| Title | 1: Single-Row ELM Suppression vs q95 | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | TBD | ITPA Joint Experiment : | No |
| Description: | Test the hypothesis that n=3 B-field from a SINGLE I-coil row should suppress ELMs over a wide range of q95. Measure the plasma response magnetic field, to learn how plasma responds to this special RMP field as a function of q. The experimental results will (1) usefully constrain theories of RMP transport and ELM suppression, and (2) maybe also show the way to simpler (single-row) ELM suppression coil arrays. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce single-row I-coil ELM suppression. Then, at I-coil n=3 current, ramp q95 slowly to find the suppression range om q95. If there is time, ramp q95 both by Ip ramp and Bt ramp. Measure plasma response with MHD magnetic pickups. | ||
| Background: | The prominent q95 resonance with a spectral peak of the RMP magnetic field is frequently considered an integral part of ELM suppression by two-row n=3 I-coil fields in DIII-D. However, ELMs were successfully suppressed in 2008 by just one row (top or bottom) of I-coils, for which SURFMN shows no resonant magnetic spectral peak at all. The 2008 experiments were brief and only demonstrated the existence of single-row suppression. The q95 dependence of the single-row I-coil suppression was not studied at that time. The single-row C-coil field has never produced ELM suppression despite repeated attempts. The magnetic spectrum of the C-coil is dominated by low-m peaks, while the single-row I-coil spectrum is broad and slowly decaying with increasing |m|. It is hypothesized that the single-row I-coil n=3 field should suppress ELMs over a wide range of q95. The unusual RMP coil geometry and the information learned from the q95 dependence of ELM suppression and plasma response will constrain theories of RMP ELM suppression. The results may also point the way to simpler (single-row) ELM suppression coil arrays. | ||
| Resource Requirements: | Standard RMP ELM suppression target plasma. 6 co-NBI, cryo pumps. Half day. | ||
| Diagnostic Requirements: | MHD magnetics, Thomson scattering, MSE, CER, reflectometry | ||
| Analysis Requirements: | SURFMN, TRIP3D, IPEC and/or MARS. Analyze plasma response B field using M. Lanctot's techniques. | ||
| Other Requirements: | -- | ||
| Title | 2: Investigate rotational screening | ||
| Name: | Mordijck smordijck@wm.edu | Affiliation: | The college of William and Mary |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Using the off-axis Neutral beam and n=3 I-coil perturbation, investigate the influence of rotation screening in L and H-mode plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use the off axis neutral beam to drive edge rotation. For L-mode plasmas we can look at the measured plasma response, the density pump-out, rotation profile and the footprint structure. For H-mode plasmas we can investigate if this changes the amount of I-coil current required to suppress/mitigate ELMs. | ||
| Background: | Up till now to investigate rotation screening, the balance of co- and counter beam injection has been altered. This changes especially the core rotation, makes the discharges prone to error-field penetration ( with lots of early disruptions). With the n=3 perturbation we are mostly interested what happens at the plasma edge, which makes the off-axis neutral beam an interesting candidate to change edge rotation. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 3: ELM suppresion/mitigation at q95=4-5 | ||
| Name: | Mordijck smordijck@wm.edu | Affiliation: | The college of William and Mary |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | By changing the ratio of coil currents in the upper versus lower I-coil (n=3 perturbation) find the best mitigation of ELMs at q95 which are relevant for steady state scenarios for ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Biggy back at the end of discharge (500ms). Use SURFMN to find best q-resonant scenario. | ||
| Background: | Can we expand the narrow q95 ELM suppression regime? | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 4: RMP in L-mode plasmas | ||
| Name: | Mordijck smordijck@wm.edu | Affiliation: | The college of William and Mary |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Continue work on RMP in lower divertor L-mode discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Investigate the influence of collisionality, q95, location of X-point on RMPs in L-mode discharges. | ||
| Background: | Last campaign during 3h of directors reserve interesting data was obtained to look at particle transport in L-mode plasma as a result of RMPs. Contrary to high collisionality discharges on MAST, turbulence levels decreased. In high collisionality discharges in H-mode on DIII-D, amplification of the RMP was observed (but density was nearly non-existent).
In L-mode a strong pump-out was observed at q95 of 3.7, but SURFMN modeling predicted pitch resonance at q95 ~ 3.2. Need experimental data to verify if response would be larger at that low value. Finally, to investigate if the creation of the tangles leads to the density pump-out, lowering the X-point closer to the divertor floor, might increase density pump-out. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 5: Controllability of pedestal and ELM characteristics by edge ECH/ECCD | ||
| Name: | Oyama oyama.naoyuki@jaea.go.jp | Affiliation: | Japan Atomic Energy Agency |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | T. Osborne, A.Leonard, Y. Kamada, H. Urano, K. Kamiya, M. Henderson, A. Loarte | ITPA Joint Experiment : | Yes |
| Description: | In order to establish new ELM control tool for ITER, the controllability of ELM and pedestal characteristics will be investigated with physics understanding of the mechanism of ELM control by edge ECH/ECCD. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This inter-machine experiment aims to confirm whether or not edge ECH/ECCD/LHCD can be a possible ELM control tool in ITER. To this end, the following experiments will be required.
(1) Reproduce the heating conditions (power density, radial profile of heating power, modulation frequency) used in AUG/JT-60U/TCV to decrease the ELM size in type I ELMy H-modes with low pedestal ļ?®e*. (2) Survey the required conditions for ELM control such as the dependence on heating power, the location of heating (HFS/LFS) and heating/CD methods (ECH/ECCD). (3) Evaluate the capabilities of ECH/ECCD/LHCD to modify ELM size (frequency), pedestal structure and plasma confinement. Based on experimental results obtained in this inter-machine experiment, we will make recommendations for the specification of the EC system in ITER for effective ELM control. | ||
| Background: | In ASDEX-Upgrade, ELM frequency was locked at ~100 Hz using modulated ECH/ECCD in H-mode plasma with fELM~150 Hz. On C-Mod a variable-phase LH launcher can be used for off-axis current drive and electron heating, with the capability of changing both the radial deposition of LH power and the relative degree of resulting LHCD and electron heating. LHCD has already been used to modify the pedestal structure in EDA H-modes, and will be used in future experiments in ELMy H-modes. In JT-60U, recent experimental results show that the ELM frequency can be increased by localized edge ECH/ECCD near the pedestal together with the reduction of the normalized ELM energy loss. It is noted that edge ECH/ECCD near the top of the plasma at high-field side is effective, while no clear effect has been observed in the case of edge ECH/ECCD near the top of the plasma at low-field side. So, we need to understand the physics mechanism of ELM mitigation by edge electron heating in order to discuss the applicability of this method to ITER and so that this joint experiment (PEP-22) has been given high priority by ITPA pedestal topical group. | ||
| Resource Requirements: | 1 day experiment
NBI+ECH(~3MW for 2-3sec) | ||
| Diagnostic Requirements: | Standard diagnostics for edge pedestal and ELM study | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 6: Rotation effect on pedestal and ELM characteristics in high beta_p plasmas | ||
| Name: | Oyama oyama.naoyuki@jaea.go.jp | Affiliation: | Japan Atomic Energy Agency |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A.Leonard, T. Osborne, Y. Kamada, H. Urano, K. Kamiya | ITPA Joint Experiment : | Yes |
| Description: | This proposal is the ITPA PEP-18. In order to improve the predictive capability for ITER H-mode operation and to reduce the ELM energy loss, the effects of toroidal rotation on the pedestal performance and ELM characteristics are investigated systematically. In addition, the effects of rotation and ripple loss on the pedestal structure and ELMs can be separated using this experiment, since DIII-D and JT-60 have quite unique capability to study plasma rotation with co and counter NBs. On the other hand, the plasma shape, thus the ELM stability, is different between DIII-D and JT-60. By combining these two conditions, this work can clarify the universal effects of rotation and dependence of rotation effects on plasma shape. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | By utilizing the unique capability of rotation control with co- and counter- NBs in DIII-D and JT-60U and by utilizing the difference in the plasma shape and edge stability between the two tokamaks, we propose to conduct inter-machine experiments on the effects of rotation on the pedestal structure and ELMs. In 2010-2011, we propose rotation scan (including a reversed Ip experiment) experiments at higher triangularity in DIII-D and will take the following data for comparison with the JT-60U data taken in 2008 to clarify the effects of rotation in different pedestal situations:
1) Frequency and energy loss (incl. ELM affected area) of ELMs, and Pedestal width and inter-ELM transport at the same ļ?¢p-ped and q95 as JT-60U; 2) Frequency and energy loss of ELMs, and pedestal width and inter-ELM transport at the same pedestal collisionality and q95 as JT-60U, and 3) Core thermal confinement of the plasmas in 1) and 2). | ||
| Background: | Recent tokamak experiments have revealed that the pedestal and core transport of the H-mode plasmas are determined under the linkage between pressure, current and rotation profiles. The goal of this research is to understand this complex system in order to improve our predictive capability for ITER, and to develop control schemes for the pedestal parameters, ELMs and core transport. Concerning the parameter linkage, plasma rotation and its radial profile seem to play critical roles. JT-60U experiments have demonstrated that a shift of toroidal plasma rotation into the co-current direction reduces the inter-ELM transport losses and increases the pedestal height and width. In addition, the ELM energy loss normalized to the pedestal stored energy (ļ??WELM/Wped) decreases with decreasing co-directed rotation. It is noted that small, so-called grassy ELMs appeared in high triangularity plasmas (ļ?¤>0.5) with zero or counter toroidal rotation. However, such clear change in the ELM characteristics has not been observed in DIII-D rotation scan experiment so far. To obtain a larger dynamic range of the counter plasma rotation, a reversed Ip experiment will be useful. As for the core confinement of H-mode plasmas, both DIII-D and JT-60U have shown improved performance with co-directed rotation compared with counter rotation. The purpose of this study is to clarify the roles of plasma rotation systematically by utilizing the unique capability of rotation control with co- and counter- NBs in DIII-D and JT-60U (existing data) and by utilizing the difference in the plasma shape and edge stability between the two tokamaks. | ||
| Resource Requirements: | 1 day Experiment. CO & Counter NBs
To increase counter rotation, reversed Ip shots will also be required. | ||
| Diagnostic Requirements: | Standard diagnostics for edge pedestal and ELM study | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 7: Divertor strike line striation as evidence for 3D boundary formation vs. plasma response | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | J.Boedo,M.S.Chu, T.E.Evans, M.E.Fenstermacher, V.Izzo, M.J.Lanctot, C-L.Lasnier, J.K.Park, H.Reimerdes, R.A.Moyer, J.Watkins | ITPA Joint Experiment : | Yes |
| Description: | Striation of divertor heat and particle fluxes is one of the most robust observations during RMP ELM suppression. This striation is caused by the perturbation of the separatrix manifolds and represents therefore a direct experimental access to the generic mechanism determining the 3D plasma boundary with RMP fields. In this experiment, the relation of the separatrix perturbation with the plasma response will be established and the relation to "stochastic boundary signatures" as particle pump out, edge temperature decrease and edge rotation spin up is investigated. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental plan foresees L-mode plasmas tailored to give access to inner and outer strike lines.
L-mode: -> repeat #142614 -> scan n=1 amplitude imposed by the C-coils to assess effect on core rotation and feedback on response (2 steps below reference n=1 value, 2 steps above) -> #5 discharges -> repeat sequence at initial rotation (does the striation come up ride away with no response phase?) -> #5 discharges -> increase power in reference shot (beta_N scan intended, still stay in L-mode) -> #5 discharges -> repeat first step with n=2 fields from I-coils at two toroidal phases (matter of preparing MARS-F analysis) Data essential for the analysis: -> obtain particle and heat flux data from both strike lines -> do in all shots OSP sweeps for LP profiles -> use fast probe for plasma edge profiles -> obtain high resolution egde Thomson data -> obtain good rotation, E_r and fluctuation data Experiment needs 25 good shots in total, i.e. this is a full day run proposal. | ||
| Background: | During the last years, divertor heat and particle fluxes were measured during application of RMP fields for ELM control. During ELM suppression a striation of heat and particle fluxes was detected and the relation to the vacuum modeled perturbed magnetic footprint in the divertor was assessed. This showed, that during L-mode plasmas the location of the lobes matched the vacuum modeled one, while in H-mode plasmas the striation width is 10-30% wider then vacuum prediction tells. As the separatrix deformation is matter of the sum of all radial fields at the separatrix, this deviation can be linked to plasma response currents. One example is #142614 were an n=3 plasma response was measured at the poloidal magnetic field sensors and particle flux striation is seen as soon as this plasma response decays (core rotation dropped potentially alowng field penetration). The plasma response measured agrees with a MARS-F prediction suggesting that this stage is determined by a resonant ideal screening response. This preliminary result motivates to assess the plasma response during RMP application and compare it to the divertor (and first wall) fluxes as well as observations of changes of the edge rotation, density and temperature profiles. The data obtained are important for ongoing strike line modeling including plasma response. | ||
| Resource Requirements: | 1.0 ITER04 patch, C-supplies on Icoils for high current capability, B_T=2.0T for edge ECE data, fresh boronization helpful | ||
| Diagnostic Requirements: | visible divertor cameras, divertor IR cameras, LLNL periscope observation, divertor (ISP) Langmuir probes, fast reciprocating probe, ECE, ECE-I, fast profile reflectometer, BES, UCLA reflectomers | ||
| Analysis Requirements: | TRIP-3D, MARS-F/IPEC for preparation and analysis, NIMROD and MARS-K as resisitive codes, EMC3-Eirene for heat and particle flux modeling including plasma response from MARS or NIMROD. | ||
| Other Requirements: | Comparison to TEXTOR will expand result over wider shape and plasma regime range. | ||
| Title | 8: Dependence of divertor heat flux striation on plasma shape and collissionality | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | J.W.Ahn, T.E.Evans, M.E.Fenstermacher, M.Jakubowski, A.Kirk, C.L.Lasnier, J.Watkins | ITPA Joint Experiment : | Yes |
| Description: | Striation of divertor heat and particle fluxes is one of the most robust observations during
RMP ELM suppression. This striation is caused by the perturbation of the separatrix manifolds and represents therefore a direct experimental access to the generic mechanism determining the 3D plasma boundary with RMP fields. In this experiment the dependence of the 3D divertor heat flux on plasma shape and collisionality is investigated to understand the heat flux pattern during RMP ELM suppression. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In this experiment, the pedestal collisionality will be scanned for different plasma shapes.
Collisionality: we suggest to do a scan of the electron pedestal collisionality by (a) changing q_95 (inside ELM suppression windows) (b) changing the pedestal density at fixed input power -> repeat RMP ELM suppressed discharge with wide q_95 ELM suppression range (e.g.#132741) #1 shot -> repeat discharge w/o RMP (!!) #1 shot -> repeat shot with 2 q_95 values, one at highest value with ELM suppression on at lowest (make sure that sufficient noRMP times are acquired before/after start of Icoils) #1 shot -> increase density at fixed input power by additional gas puffing during L-mode phase (see comments in summary of MP2008-03-07), attempt two collisionality values, repeat q_95 scan, noRMP reference and shot with two q_95 values at boundaries of ELM suppressed windows #3x2 shots Shape: this sequence shall be repeated at lower triangularity #3+6 shots Data essential for the analysis: -> obtain particle and heat flux data from both strike lines -> do in all shots OSP sweeps for LP profiles -> use fast probe for plasma edge profiles -> obtain high resolution egde Thomson data -> obtain good rotation, E_r and fluctuation data This experiment foresees 18 god shots in total with the need for several setup shots, i.e. this is a full day run proposal. | ||
| Background: | Observing divertor heat flux during RMP ELM suppression in various shapes and at various collisionalities has shown a strong dependence of (a) the location of the 3D heat flux and (b) the magnitude at the non-axisymmetric heat flux, i.e. the filling of the various lobes of the separatrix manifolds.
-> In L-mode plasmas, a match between the vacuum modeled lobe position and the measured heat (and particle) fluxes was measured. The filling of the outer lobes was found to be connected to a T_e decrease at the pedestal and agrees qualitatively with a parallel heat loss into the 3D magnetic footprint. -> In H-mode plasmas strong dependence on shape and collisionality was observed with clear deviations of the measured width from the vacuum prediction. This can be surveyed in three cases: a)At high collisionality, low triangularity, the striation width was observed to exceed the vacuum modelled width by a factor of 2-3 with evenly high heat fluxes into all lobes. b) At high triangularity, low collisionality, the width of the measured pattern exceeds the vacuum modeled footprint location by 10-30%. The heat flux into the outer lobes is considerably smaller than modeled with a plasma transport fluid and kinetic neutral transport model (EMC3-Eirene). These observations raises the question in how far (a) the plasma response determines the lobe location (see proposal ID7 on that) and (b) in how far the collisionality affecting the parallel and effective radial heat flux determines the heat flux magnitude in each lobe Both together determine in how far a thermal loss at the pedestal is connected into the 3D footprint. This is the essential aim to be determined concluding on the experimental results from ID7 and this proposal. | ||
| Resource Requirements: | 1.0ITER04 patch to get high triangularity (mean_delta ~ 0.6)
16DNRDPM12 to get low triangularity (mean_delta ~ 0.3) | ||
| Diagnostic Requirements: | visible divertor cameras, divertor IR cameras, LLNL periscope observation, divertor (ISP),
Langmuir probes, ECE, ECE-I, fast profile reflectometer, BES, UCLA reflectomers, UCLA fast CCD for tangetial HFS observation | ||
| Analysis Requirements: | TRIP-3D for preparation and diagnostic setup, MARS-F/IPEC for ideal plasma response modeling and input to field aligned grid of EMC3-Eirene, any further transport code (XCGO as kinetic reference). | ||
| Other Requirements: | Comparisons to NSTX and MAST results planed | ||
| Title | 9: Role of thermal transport for ELM suppression | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | T.E.Evans, M.E.Fenstermacher, B.Hudson, A.Kirk, M.J.Lanctot, R.A.Moyer, T.H.Osborne, J.-K.Park, P.B.Snyder | ITPA Joint Experiment : | Yes |
| Description: | A strong pedestal pressure q_95 resonance was observed during RMP ELM control experimeents in ITER similar shape [O.Schmitz et al PRL 103,165005 (2009)]. During the q_95 ramp down experiments performed three ELM suppression windows appeared which were linked to different pedestal pressure values. In order to determine the ELM stability and compare to linear P-B models like ELITE high quality edge profile data are needed. The run day proposed will be uased to acquire stationary profile data with sufficient average times during stationary ELM suppression as input for high quality EFIT fitting and subsequent ELITE analysis and use for alternative linear or non-linear stability analysis. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental plan foresees to repeat reference discharge #132741 and do a stationary shot to shot q_95 scan to obtain high quality edge profile data for EFIT equlibria suitable for ELITE analysis.
First the discharge will be repeated and then optimized for widest q_95 ELM suppression window extension. Use wall conditioning and small beta_N scans for this. -> repeat and setup #1 repeat #2 + #3 10% + 20% beta_N increase #4 + #5 10% + 20% beta_N decrease #6 pick best shot and do noRMP reference #7-17 do a shot to shot, fine q_95 scan (q_95 range of 132741 was 0.8 -> 8 steps + 2 for specifc values like edge of resonant windows etc) | ||
| Background: | A strong pedestal pressure q_95 resonance was observed during RMP ELM control experimeents in ITER similar shape [O.Schmitz et al PRL 103,165005 (2009)]. The p_e modulation observed with localized decrease spots was driven by a global (mean non q_95 dependent) particle pump out and a highly q_95 dependent electron temperature decrease. However, these data were obtained during q_95 ramping without sufficient average time for high quality edge profile data. Therefore no meaningful ELITE analysis could be performed and the understanding of the link between the resonant character of the thermal transport and the ELM stabilization is still an important open question. This gap of essential data will be closed in this experiment. | ||
| Resource Requirements: | 1.0ITER04 patch panel, C-supplies on SPAs, standard EFC, B_T=2.0T foe good ECE edge measurements | ||
| Diagnostic Requirements: | Thomson scattering at its best
ECE, ECE-I fast profile reflectometer MSE measurements and fast Lithium beam with polarimeter BES, Doppler back scattering UCLA IR and vissible cameras divertor UCSD fast camera tangential view LLNL periscope floor LP | ||
| Analysis Requirements: | kinetic EFIT
stability: ELITE transport: EMC3-Eirene (fluid) vs. XCGO (kinetic) | ||
| Other Requirements: | comparison to MAST started | ||
| Title | 10: Net-erosion and material migration with a 3D boundary | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Requested |
| Co-Author(s): | A.McLean
T.E.Evans, M.E.Fenstermacher, C.L.Lasnier, A.L.Leonard, E.A.Unterberg |
ITPA Joint Experiment : | Yes |
| Description: | Most analysis of first wall fluxes and extrapolation towards ITER assume axisymmetry. However, during RMP ELM suppression as well as during non-axisymmetric current perturbations in the plasma, a 3D plasma boundary is formed, most prominent seen in a striated divertor heat and particle flux pattern. This deformation will change the topology of the divertor (and first wall) heat and particle loads with potentially severe consequences for components designed based on axisymmetric assumptions. In this experiment a quantitative systematic assessment of the impact of the 3D deformation of the plasma boundary with an RMP field similar to that used during ELM suppression on the net-erosion and material migration properties will be performed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The aim of the experimental sequence is to scan parameters which determine the perturbed magnetic footprint as well as the effective radial transport inside of the separatrix and in the 3D SOL.
#1+#2 repeat #141905 (with OSP sweep) w/o and with PPI inserted, noRMP #3+#4 now with RMP, I-coil current I_c as in reference (4.4kA) #5-#7: I_c scan: 1.5kA, 3.0kA, 6.0kA (4.5kA is reference case) #8-#11: no OSP sweep, but q_95 ramp down for all four I_c #12-#16 repeat #1-#4 with 0 degree n=3 phasing to measure at different toroidal location in 3D boundary Together with setup shots needed and slow down due to PPI move during GDC, this is a proposal for a full day experiment | ||
| Background: | An initial assessment of the impact of a 3D boundary similar to the one observed during ELM suppression on the net-erosion and the material migration (MP2010-02-02) has shown that a significant adaptation of the divertor erosion situation situates. A 6 shot piggy-font initial experiment has shown that the chemical erosion, directly measured with the Porous Plug Injector (PPI) decreased inside if the 3D SOL with a clear dependence on the measurement location. The parallel heat and particle flux in the separatrix lobes determine the local erosion properties as well as the resulting migration of the eroded particle. Post discharge inspection of the PPI cap suggests that an enhanced local redeposition happens which indicates a dominant frction force on the eroded particle compared to the usual ExB dominated movement. This will have impact on the ability to control local redeposition as well as tritium retention.
Due to the limited capability to measure erosion quantitatively in 3D, the localized measurements as obtained from the PPI need modeling for interpretation of the resulting 3D solution. For this, effort is ongoing to couple EMC3-Eirene as a plasma background to sophisticated PSI models like ERO or MCI. The data obtained in this experiment will be essential for quantitative validation, so far only one good example was obtained. A similar experiment was performed at TEXTOR and direct comparisons will supplement the analysis. | ||
| Resource Requirements: | 1CARLSNB as patch for LSN L-mode with capability to move OSP across shelf, setup as in #141905, C-supplies on SPAs | ||
| Diagnostic Requirements: | visible divertor cameras, divertor IR cameras, LLNL periscope observation, divertor (ISP)
Langmuir probes, fast reciprocating probe, ECE, ECE-I, fast profile reflectometer, BES, UCLA reflectomers | ||
| Analysis Requirements: | TRIP-3D for preparation and analysis
EMC3-Eirene for plasma background ERO and/or MCI for erosion analysis and impurity sources | ||
| Other Requirements: | This experiment can only be performed in L-mode due to the heat flux handling capability of the PPI cap | ||
| Title | 11: 3-D Fields and ECH Density Pumpout | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | Andrea Garofalo | ITPA Joint Experiment : | No |
| Description: | The purpose of the experiment is to see if applied 3-D fields, or error fields in non-specific 3-D field experiments, play a role in the density pumpout when ECH is applied to a NBI target H-mode discharge. ECH density pumpout will be parameterized as a function of target discharge toroidal rotation at varying levels of applied 3-D field, or error field if there is an effect for small levels of perturbation. The results may also provide information regarding rotational shielding of 3-D fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The target discharge will be an ELMing H-mode with variable toroidal velocity set by co/counter NBI. Off-axis ECH will be applied to cause some density pumpout. We will try to avoid huge density pumpouts, rather looking for conditions that allow somewhat controlled experimental conditions. The effect of rotation on pumpout will be measured. Then, 3-D fields will be applied, first by compromising the error correction and then applying stronger external 3-D fields. If there is a clear 3-D effect on the ECH density pumpout, in 'break in slope' or depth of the drop, etc, then a comprehensive 3-D spectrum study should be undertaken, i.e. n=1,2.3. | ||
| Background: | * In doing DIII-D intrinsic rotation experiments in recent years the effect of large ECH density pumpout has been observed. Anecdotally, the two somewhat different conditions that were applied were 1) low toroidal rotation plasmas using balanced NBI and 2) off-axis ECH. Both were used in order to have conditions to better identify intrinsic rotation with the higher beta resulting from NBI heating in addition to ECH.
* Pumpout associated with ECH has been seen in many tokamaks, over decades. There are also cases of "pump-in". I don't know of a focussed experimental parameterization of the effect. * Theories have emerged in which the electron heating reduces the anomalous density pinch (e.g. Angioni). There are also experimental observations of ECH pumpout being associated with non-axisymmetric internal magnetic fields, due to MHD activity. * Pumpout is also seen in the RMP ELM control experiments. There it is clear that pumpout is caused by the 3-D fields applied. * In the ECH pumpout phenomenon, in DIII-D H-mode targets, as the density drops the ELM frequency typically increases, and the density decrement seems due to a reduction in the pedestal density. This increases the number of possible effects. Is it the ELMs that are reducing the density and the ECH is affecting ELMing? Even if this is the case, then it also may indicate a tie-in to the RMP ELM suppression experiments. | ||
| Resource Requirements: | 1 day experiment.
2 days if compelling results obtained. Standard DIII-D. Minimally: All beams, All gyrotrons (first 2/3 of run period). Or: CER beams + whatever gyrotrons available (last 1/3 of run period). Need I-coils and C-coils. | ||
| Diagnostic Requirements: | Standard. Need all kinetic profiles. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 12: Test edge orbit loss velocity in Helium discharges | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | Keith Burrell, Wayne Solomon, Stefan MĆ¼ller, Jose Boedo | ITPA Joint Experiment : | No |
| Description: | Measure velocity profiles of the bulk ion (He) H-modes using CER and edge Mach probes and test the thermal ion orbit loss model.
This is in the Pedestal Structure area because the indication of a loss cone distribution in the pedestal region of an H-mode will possibly bring in new effects to be considered in understanding the pedestal. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | * Measure the edge velocity with both CER and the Mach probes in ECH H-modes ( + NBI blips ) in helium discharges so that the CER and probe measurements can be cross-checked. CER of course gives the full velocity profile. Vary the discharge parameters that affect the loss cone model, primarily the location of Rx in a LSN discharge, and see if the model predictions are borne out by the measurements.
* Given time, use repeat shots to measure the CER carbon velocity profile in the same conditions. * Given time, compare USN and LSN since to lowest order this does not affect the model, but does change the effective collisionality. | ||
| Background: | * A simple ion orbit loss model has significant agreement with recent edge Mach probe flow velocity measurements near the LCFS, in H-mode conditions. The measurements show a narrow velocity layer centered near the LCFS. Subsequently, it has been found that the same layer shows up in the CER measurements of bulk ion edge toroidal velocity in intrinsic rotation experiments in ECH H-modes in helium discharges.
* By repeating these helium experiments we can cross-test the Mach probes and the CER quantitatively and vary conditions to test the loss model. | ||
| Resource Requirements: | 1+ day experiment, or part of a helium working gas campaign, since time is needed to purge the walls of deuterium. Standard DIII-D.
Deuterium diagnostic beam blips can be used. Need all gyrotrons. | ||
| Diagnostic Requirements: | Standard. CER tuned to helium. Mach probes operative. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 13: Role of Zonal Flow shear and Limit Cycles in triggering the L-H transition | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | E.J. Doyle, T.L. Rhodes, J. Hillesheim, W.A. Peebles, G. Wang, G.R. McKee, K.H. Burrell, W. Solomon | ITPA Joint Experiment : | Yes |
| Description: | Long-range flow correlations have been observed to increase around the time of the L-H transition in several experiments (TEXTOR, TJ-II, ASDEX-U) but no definitive, universal link between the evolution of time-dependent flow shear (Zonal-Flow shear) and the L-H transition has been demonstrated so far. The L-H transition is often preceded by limit-cycle oscillations of the electric field, flow shear, and turbulence level near the separatrix. These so-called "slow" or "limit-cycle" transitions potentially provide unique information on the trigger mechanism and the role of time-dependent shear flows (Zonal Flows) in the transition. The goal of this experiment is to map the evolution of Zonal Flow and turbulence level with high time resolution in a regime where Zonal Flow damping is weak (low collisionality, high temperature plasmas with a high q_95). The interaction of mean flow and Zonal flows will be investigated by changing the total neutral beam torque (using de-rated beams). Beam blips can also be used to deliberately control the trigger point/trigger the final L-H transition at different phases of the limit-cycle oscillation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | "Slow" L-H transition will be induced by ramping up neutral beam power in SN L-mode plasmas from a level below the L-H threshold. A combination of de-rated co- and counter beams will be used to control the beam power/torque input with beam modulation times down to 5 ms. A sequence of L-H transitions (and H-L back transitions) will be induced by varying the power power, with the "H" time intervals kept short to limit density increase. Doppler Backscattering and BES will be used as primary turbulence diagnostics. The L-mode separatrix density will be kept at ~ 1.5e13 cm^-3 to optimize spatial coverage for the DBS systems. In contrast to earlier experiments, recent diagnostic advances allow toroidal mapping of the mean and time-dependent shear flows (and flow correlations) on DIII-D with a time resolution down to ~10 microseconds (comparable to typical turbulence decorrelation times). This is accomplished by two Doppler Backscattering systems (an 8-channel and a 5-channel system) located 180Āŗ apart toroidally. Doppler Backscattering provides simultaneous radial profiles of the time-dependent and mean flow/ flow shear and the turbulence level. The probed poloidal wavenumber can be scanned across the ITG/TEM range in successive shots. Slow transitions with an intermediate oscillatory phase have been obtained in beam-heated low density plasmas in DIII-D recently. | ||
| Background: | Intermittent changes in edge flows and turbulence modifications before the L-H transition have been previously observed in DIII-D (Colchin, PRL 2002) as well as more recently in NSTX (using gas puff inaging, S. Zweben, APS 2010) and ASDEX-Upgrade. These experiments have been carried out at modest to high edge collisionality and in some cases with high edge radiation. In the experiment proposed here would be carried at different torque input and with the capability to reconstruct the radial profile of flow shear and turbulence levels with high time resolution (much faster than the period of limit cycle oscillations). The dual, multi-channel DBS capability on DIII-D is unique and allows reconstruction of the radial flow (and turbulence level) profile with superior time resolution. Therefore it complements and extends diagnostic capabilities on ASDEX-U, NSTX, and TJII. and a Joint ITPA Experiment investigating slow L-H transitions is proposed. | ||
| Resource Requirements: | 7 Neutral beams | ||
| Diagnostic Requirements: | DBS, BES, Core/Edge CER, Main Ion CER if available | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 14: ELM control by Vloop modulation | ||
| Name: | Cunningham geoffrey.cunningham@ccfe.ac.uk | Affiliation: | CCFE |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | C. Gimblett | ITPA Joint Experiment : | No |
| Description: | It is well known that access to H mode is strongly influenced by the loop voltage, being especially difficult while the plasma current is ramped up. The objective here is to develop a new ELM control tool using rapid modulation of the loop voltage. This has previously been done elsewhere, for example on Compass, with some success but the experiment was contaminated by plasma movement in the Z direction. Important information regarding the role of edge current in H mode barrier formation will also be obtained. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The objective is to modulate the edge current density by modulation of the loop voltage. This is most effectively done by adding a feed-forward modulation to the E coil control, adding it to the usual Ip feedback signal. Previous experiments of this type (on Compass) have suffered because of coupling between Ip and Z control, so symmetric DND operation is preferable. A likely series of experiments would thus be:
1. If a DND type 1 ELMy H mode scenario is available, apply V loop modulation, frequency dictated by the vessel penetration time. Compensating vertical field feedforward may also be needed. 2. If successful, try for higher frequency by using center-stack coils driven by SPAs. 3. If only an SND scenario is available, additional shots to determine the optimal Z control feedforward compensation will be needed. | ||
| Background: | Background: Edge current density, peeling modes and ELMs.
The peeling mode, unlike the ballooning mode, exists in simple cylindrical geometry. In this approximation, it is driven entirely by the current density Ja at the plasma/vacuum boundary at r = a (essentially it is the presence of a plasma Ja adjoining the vacuum, where of course J = 0, that provides an effective large current gradient dJ/dr). Stability in general is a function of both the edge q and edge current density. In particular if by any means the edge current density could be made negative then in general there would be no unstable peeling modes. In the peeling/relaxation ELM model [1] peeling modes are taken as the trigger for the subsequent ELM which is itself a relaxation phenomenon. The model might be thought best suited to describing Type-III ELMs; nevertheless, the model has been used to explain the ā??multi-resonanceā?? effect seen in ELM mitigation experiments using EFCC on JET [2]. Accordingly, we seek to vary ELM characteristics by varying Ja. In particular, if Ja < 0 is realised then ELMs could be stabilised according to Ref.[1]. Variation of Ja should be achieved transiently by employing ā??fastā?? current ramp scenarios. This effect has seemingly been successfully demonstrated in the past on the COMPASS tokamak [3], where a current ramp-down resulted in an ELM-free period. Transport/diffusion simulators such as TRANSP should therefore be used to examine current ramp scenarios and calculate the predicted applied loop volts V (negative V was obtained transiently in the COMPASS experiments mentioned above), Ja evolution etc. in DIII-D to re-examine this effect. References [1] Gimblett C G, Hastie R J and Helander P, Phys. Rev. Lett. 96 035006 (2006). [2] Liang Y et al., Phys. Rev. Lett. 105 065001 (2010). [3] Valovic M et al., 22nd. European Physical Society Conference on Controlled Fusion and Plasma Physics, III-125, Ed. B E Keen, P E Stott and J Winter, Bournemouth, UK (1995). | ||
| Resource Requirements: | Established type I ELMy H mode scenario | ||
| Diagnostic Requirements: | Any information on the edge current density will be very valuable. Standard temperature and density profiles to enable Transp modelling will be required. | ||
| Analysis Requirements: | Efit and Transp modelling | ||
| Other Requirements: | |||
| Title | 15: Electron Transport Stiffness Measurements - Heat Flux Scan | ||
| Name: | DeBoo debooga@att.net | Affiliation: | Retired |
| Research Area: | Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure plasma electron transport stiffness at several locations in the plasma while holding the edge plasma profiles fixed. Compare the measurements to stiffness predicted by the TGLF transport model. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Stiffness can be defined in terms of the percent change in thermal diffusivity from a power balance analysis divided by the percent change in the temperature gradient. The temperature gradient at a given plasma radius can be varied by performing a local scan in heat flux shot by shot. ECH is applied inside and outside the region of interest and the local heat flux scan is accomplished by changing the level of inside and outside ECH power while holding the total ECH power fixed. Since the total ECH power is fixed the edge plasma profiles remain fixed avoiding any concerns over edge/pedestal effects.
Stiffness can also be determined by the ratio of the heat pulse diffusivity to the power balance diffusivity. While performing the heat flux scan the heat pulse diffusivity can be determined by modulation of the power from one gyrotron. Measurement of the pulse propagation with the ECE system allows the determination of the heat pulse diffusivity. This experiment should be done in an ITER relevant plasma, H-mode and perhaps additional advanced scenarios, since one main motivation is to test the validity of the stiffness predictions of transport in ITER based on the TGLF model. This experiment can also potentially determine the existence of a critical or marginal electron temperature gradient if a low enough heat flux condition is obtained. The minimum heat flux will be set by the H-mode threshold power level. | ||
| Background: | Due to the strong stiffness of transport in ITER predicted by the TGLF model the predicted fusion power decreases almost linearly with auxiliary heating power at fixed pedestal beta (Kinsey IAEA10) potentially jeopardizing the Q=10 mission if high P_aux is required. These experiments are meant to test the validity of the TGLF model transport stiffness. Previous experiments in DIII-D L-mode plasmas with this heat flux scan technique have been employed successfully while searching for a nonlinear critical electron temperature gradient. | ||
| Resource Requirements: | All gyrotrons. | ||
| Diagnostic Requirements: | ECE and other profile diagnostics required for normal power balance analysis with ONETWO. | ||
| Analysis Requirements: | Simulation of experimental profiles and stiffness with TGLF model. | ||
| Other Requirements: | |||
| Title | 16: Electron Transport Stiffness Measurements - ECH Swing | ||
| Name: | DeBoo debooga@att.net | Affiliation: | Retired |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure plasma electron transport stiffness at several locations in the plasma while holding the edge plasma profiles and local Te fixed. Compare the measurements to stiffness predicted by the TGLF transport model. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Stiffness can be defined in terms of the percent change in thermal diffusivity from a power balance analysis divided by the percent change in the temperature gradient. To maximize the local change in temperature gradient use the so called ECH swing technique where ECH power is alternately deposited in two closely spaced regions in the plasma during an individual discharge. The local electron temperature gradient is repetitively swung back and forth this way and by phase lock averaging the ECE measurements, reasonably low errors on temperature gradients can be obtained. The region of interest can be scanned over several shots. This is similar to the heat flux scan experiment proposed but has the added advantage that Te remains roughly constant half way between the two deposition regions for equal ECH power at each deposition region. One disadvantage is that at fixed total ECH power only two steady state temperature gradient conditions are obtained so that the change in temperature gradient is determined from just two points which could result in larger uncertainties in stiffness than the heat flux approach.
This experiment should be done in an ITER relevant plasma, H-mode and perhaps additional advanced scenarios, since one main motivation is to test the validity of the stiffness predictions of transport in ITER based on the TGLF model. The maximum change in temperature gradient will be reduced if additional power is required to obtain an H-mode. | ||
| Background: | Due to the strong stiffness of transport in ITER predicted by the TGLF model the predicted fusion power decreases almost linearly with auxiliary heating power at fixed pedestal beta (Kinsey IAEA10) potentially jeopardizing the Q=10 mission if high P_aux is required. These experiments are meant to test the validity of the TGLF model transport stiffness. Previous experiments in DIII-D L-mode plasmas with this ECH swing technique have been employed successfully while searching for modulations induced in turbulence levels. | ||
| Resource Requirements: | All gyrotrons | ||
| Diagnostic Requirements: | ECE and other profile diagnostics required for normal power balance analysis with ONETWO. | ||
| Analysis Requirements: | Simulation of experimental profiles and stiffness with TGLF model. | ||
| Other Requirements: | |||
| Title | 17: Low Torque ITER-like discharges | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make an ELMing ECH H-mode with the nominal DIII-D ITER shape, q95 just above 3, betaN ~ 1.8, with no torque input, and negligible D NBI fast ions. Use D NBI blips for diagnostic purposes. Add low voltage He4 NBI injection to simulate alphas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | * Try to achieve betaN ~ 1.8 with q95 just above 3 in a LSN shape to realize what is likely the first such discharge at this betaN without significant NBI heating. Use D NBI blips for diagnostic measurements.
* Lower BT as far as possible, while keeping the ECH deposition reasonably well in the interior, rho ~ 0.3. (* The ELMing H-mode ITER confinement scaling indicates that this should be possible somewhere around a total heating power of ~ 3 MW. This should be achievable with the 6 gyrotrons now in place plus some Ohmic heating. The seventh gyrotron will help of course. An issue is whether there is any reduction in the energy confinement with ECH vice NBI.) * Then add ~ balanced NBI with He4 to simulate alpha particles, and study alpha confinement, potential instabilities, etc. If we keep the He++ poloidal ion gyroradius divided by r at the q=1 surface the same as in ITER, then the 3.5 MeV alpha scales to roughly 35 keV He beam energy, depending upon the details of the final DIII-D parameters selected. For NBI source reasons it may be better to run at a minimum of 40 keV. * If we keep the "fast" helium density to something like 4% of ne, roughly the ITER condition, then this should probably require less than 1/2 MW of total He NBI, and be a perturbation as far as heating goes. (This depends on the helium thermalization and transport times.) * The added He NBI may turn out to be what is needed to reach the final betaN ~ 1.8 summit. * This (4%) will result in a small fast helium beta by ITER standards. More He power can be injected to raise the fast ion beta somewhat. | ||
| Background: | * In recent years we have many ECH H-modes. Off axis ECH has been used to more quickly promote a steady state ELMing condition, but there may be a confinement degradation with the off axis deposition (rho > 0.5). So we want to have more central deposition.
* These conditions will allow a measurement of the intrinsic rotation at this betaN level, and will show if there is MHD activity at this relatively low rotation condition. * RMP ELM stabilization can be tested, albeit at the slightly higher q95 that has been shown to work. The low starting rotation and lower betaN will be different from more standard RMP suppression scenarios. * The plasma density is a parameter to explore, although of course the basic ITER scenario specifies a density. | ||
| Resource Requirements: | 2 Days.
All 6 gyrotrons. Helium operation in one co and one counter beamline. | ||
| Diagnostic Requirements: | Standard diagnostics. Appropriate fast ion diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 18: Electron temperature profile stiffness in TEM-dominated QH-mode plasmas | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Requested |
| Co-Author(s): | K.H. Burrell, T.L. Rhodes, C. Holland, E.D. Doyle, J. deBoo, J. Hillesheim, G. Wang, W.A. Peebles | ITPA Joint Experiment : | No |
| Description: | Improved understanding of core electron transport and profile stiffness is important with regards to alpha-particle heating in future burning plasmas, where Te ~ Ti is expected. Recent QH-mode experiments, as analyzed by TGLF, indicate a crossover from ITG- to TEM-dominated turbulence in the core plasma (r/a < 0.6) when core ECH is applied to achieve Te > Ti in the core plasma. These plasmas have very low core collisionality in the ITER-relevant range. Electron temperature gradients close to the TEM critical gradient are observed. We propose to investigate profile stiffness as function of the coupled ECH power, with the ECH deposition location scanned between 0.2 < r/a < 0.5. Profile stiffness will also be investigated at low core plasma rotation, which has shown slightly (15%) improved beta_n and global energy confinement. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Recent QH-mode experiments have achieved Te > Ti in the central plasma with injection of up to 2.7 MW of ECH power. ELM-free operation could be maintained. By scanning the ECH power successively at different radial locations, the dependence of local electron heat flux on temperature gradient scale length R/L_Te can be determined (via transport analysis). The TEM critical gradient will be evaluated by TGLF and compared with the experimentally measured gradient. QH-mode plasmas offer the benefit of a large impact of ECH heating as the absorbed power per electron is is large due to the low density. These plasmas also provide excellent core access for the DBS fluctuation diagnostics and allow measurements of the tubulence wavenumber spectra for comparison with code predictions. These plasmas are counter-injected,reversed I_p plasmas. By substituting beam power in the co-current direction from the 210L/R beams, core rotation can be substantially reduced as long as locked modes are avoided. Measurements of electron transport stiffness at low rotation are particularly valuable in extrapolating to burning plasmas and/or testing predictive capabilities and transport scaling. | ||
| Background: | A critical electron temperature gradient and indications of profile stiffness have been found in L-mode plasmas previously (for example Ryter et al., PRL 2003) but no comprehensive H-mode data have been available, in particular in low collisionality plasmas with Te ~ Ti. Suitable ECH-heated QH-mode plasmas have been achieved in the 2010 campaign. Low rotation was achieved in one shot but no documentation of fluctuation data could be made. No ECH power scan has been performed at either high or low rotation. DIII-D now has capabilities core density fluctuation measurements which enable us to measure detailed, local wavenumber spectra under conditions where we expect a transition from ITG to TEM core turbulence. | ||
| Resource Requirements: | 7 Beams, all gyrotrons | ||
| Diagnostic Requirements: | CER, DBS2, DBS5, DBS8, BES, MSE | ||
| Analysis Requirements: | TRANSP transport analysis, TGLF and GYRO analysis | ||
| Other Requirements: | -- | ||
| Title | 19: Dependence of momentum pinch and rotation peaking on beta | ||
| Name: | Tala Tuomas.Tala@vtt.fi | Affiliation: | VTT Technical Research Centre |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon | ITPA Joint Experiment : | Yes |
| Description: | This experiment is an ITPA TC-16 experiment. The objective of this experiment is to study the dependence of the momentum pinch and Prandtl numbers on beta. While the dependence of the momentum pinch on collisionality and R/Ln was studied on DIII-D in 2009, the dependence of the momentum pinch on beta remains unexplored on DIII-D and also on other tokamaks. According to the theory and very recent gyro-kinetic simulations, at high enough beta, the inward momentum pinch decreases in magnitude and eventually changes sign outwards. If experimentally verified, this could change completely the predictions of rotation peaking in future, low central torque tokamaks such as ITER when operation at high beta. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main idea of the experiment is to make a scan of beta. Standard technique to carry out the beta scan by keeping other dimensionless parameters like collisionality, rho* and q as constant as possible will be used. This technique has already been exploited for example on DIII-D, JET and ASDEX-U within the scope of TC-1 joint experiment. From this point of view, the experimental methodology should not require any development on DIII-D, except maybe the aim to keep R/Ln as constant as possible during the beta scan. Reaching a large enough variation in beta scan should be feasible. A careful planning of the experiment is needed to be sure that the possibly observed decrease in the pinch number is due to transport and not due to MHD modes.
NBI modulation technique will be the best way to induce the rotation perturbation. The main interest is in the core region at 0.2 < r/a < 0.8 which should be turbulence dominated and possibly less influenced by ELMs and sawteeth. The outcome of the experiment is to confirm (or not) the theory and gyro-kinetic predictions for the dependence of momentum pinch on beta. Carrying out the same experiment in several machines increases the confidence in the results achieved only on one machine. Combining the results from both this ITPA experiment TC-16 and the on-going TC-15 already carried out on several tokamaks will give a solid base to extrapolate the role of momentum pinch in determining the rotation profile and the peaking of the toroidal rotation in ITER. As the NBI modulation phase does not require the whole flat top phase of the shot, other experiments to study the beta dependence can be accomodated. This experiment could be combined easily for example with an experiment to study the dependence of intrinsic rotation on beta by using beam power steps and with an experiment to study the dependence of the particle pinch and density peaking on beta. Furthermore, the dependence of the overall transport properties on beta will be automatically studied during the scan. | ||
| Background: | In recent years, several tokamaks have shown that a significant inward momentum pinch exists. Now, when the existence of the inward momentum pinch has been established, in order to be able to extrapolate its significance in ITER prediction for rotation, the parametric dependencies of the pinch must be clarified. There are a few parameters believed to determine the size of the dimensionless pinch number and the Prandtl number. Experimentally, the pinch number has been found to depend strongly on R/Ln on JET. On the other hand, no significant collisionality dependence has been found on neither on DIII-D nor on JET. While the parametric dependencies (collisionality, R/Ln, q) of the momentum pinch in low-k turbulence, ITG dominated plasmas are included in TC-15 joint experiment, no studies of momentum pinch in high beta plasmas is included. TC-15 experiment was carried out on DIII-D in 2009 and published in PoP in 2010 by W. Solomon et al.
According to very recent gyro-kinetic simulations, the pinch number depends strongly on beta. At low beta, ITG completely dominates and momentum pinch depends only weakly on beta, but at higher beta, kinetic-ballooning modes become significant and momentum pinch is decreased and eventually becomes an outward convection. The principal objective of this experiment is to verify this theory and gyro-kinetic simulation result. | ||
| Resource Requirements: | 1 day experiment. Co- and counter NBI. | ||
| Diagnostic Requirements: | Standard core diagnostics, ne profile reflectometry | ||
| Analysis Requirements: | Standard, well-developed techniques will be used to analyse the NBI modulation and possible beam steps. | ||
| Other Requirements: | |||
| Title | 20: Ion Transport Stiffness Measurements - Heat Flux Scan | ||
| Name: | DeBoo debooga@att.net | Affiliation: | Retired |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure plasma ion transport stiffness while holding the edge plasma profiles fixed. Compare the measurements to stiffness predicted by the TGLF transport model. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This proposal is similar to the proposal for a heat flux scan to determine electron transport stiffness except we will use on and off-axis NBI instead of ECH to change the heat flux. By varying the level of NBI power peaked on axis and peaked off-axis at fixed total NBI power, on a shot by shot basis, the edge profiles can be held constant, avoiding any concerns over edge/pedestal effects. Changing the heat flux to both ions and electrons this way is expected to be largest near the plasma core. Thus stiffness in the plasma core will be studied with these experiments. The maximum range in variation of the heat flux obtainable will be determined by the difference in on-axis versus off-axis NB deposition profiles. If the two deposition profiles can be sufficiently separated spatially, stiffness measurements at several plasma radii may be possible by scanning the off-axis deposition. The best place to study stiffness will be where the change in heat flux is the largest. Since both ion and electron heat flux will be varied measurements of the stiffness of both species may be possible. However, at the expected Te values (Te > 1.5 keV) in the H-mode discharges to be studied, the majority of NB power will be deposited in the ions.
This experiment should be done in an ITER relevant plasma, H-mode and perhaps additional advanced scenarios, since one main motivation is to test the validity of the stiffness predictions of transport in ITER based on the TGLF model. | ||
| Background: | Due to the strong stiffness of transport in ITER predicted by the TGLF model the predicted fusion power decreases almost linearly with auxiliary heating power at fixed pedestal beta (Kinsey IAEA10) potentially jeopardizing the Q=10 mission if high P_aux is required. These experiments are meant to test the validity of the TGLF model transport stiffness in the ions and perhaps in the electrons as well if the change in electron heat flux is sufficiently large to produce clear changes in the electron temperature gradient. | ||
| Resource Requirements: | Off-axis NBI and similar level of on-axis NB power. | ||
| Diagnostic Requirements: | Full set of profile diagnostics required for normal power balance analysis with ONETWO. | ||
| Analysis Requirements: | Simulation of experimental profiles and stiffness with TGLF model. | ||
| Other Requirements: | |||
| Title | 21: ELM pacing with loop voltage blips | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | T. Evans, J. Leuer | ITPA Joint Experiment : | No |
| Description: | Investigate whether the natural ELM frequency can be overtaken and paced by short blips in the loop voltage. That is, can a low duty cycle train of blips in the loop voltage be used to entrain and increase the frequency of ELMs. And if so, how much is the per-ELM energy dump reduced? | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | *Establish an ELMing H-mode with low (enough) ELM frequency. Apply short loop voltage blips and see if there is any effect on ELMing. Two methods will be tested.
*First, apply blips with the E-coil. For example, the nominal blip length might be 8 msec at perhaps ~ 2 - 4 V / turn, applied every 30 msec (33.3 Hz). The shortest useful blip will be determined by the filtering of the vacuum vessel. There is also a lower limit on width set by the 720 pulse E-supply, but this is apparently shorter than the vessel time constant. Another restriction may be in rise time limiting circuits placed on the E-supply command. All of these issues need to be considered, together with possibly modifying the PCS command to allow the blips, while maintaining the target Ip value on the slower timescale. (We probably want to avoid having a negative loop voltage in between some hefty blips, in order to maintain Ip.) *Second, apply blips with the F-coils, say F6A and F6B. Since we want the toroidal E field applied in the pedestal region this may allow enough voltage. The vessel time constant may be shorter due to the higher poloidal mode structure of this perturbation (Jim Leuer). Using the F6 coils will also result in a modulation of the outboard flux surface, and we don't know if this will help, or hinder an ELM interaction. The squeezing of the outer surface might also promote an ELM. *The applied loop voltage from either of these methods should be amenable to simulation with the existing DIII-D circuit models. | ||
| Background: | *It is commonly observed that adding an upward ramp in Ip promotes ELMs, while a downward ramp in Ip tends to inhibit ELMs. A simple explanation could be that an increased pedestal current density pushes the edge over the current gradient limit. Can this ramp-up effect be pushed to the extreme of short, low duty cycle blips?
*Additionally, in recent years experiments have been done with modulated n=3 magnetic perturbations applied to the plasma. Some of these were done with the purpose of pacing ELMs, while, for example, others used I-coil switching to observe the temporal effect of going in and out of RMP ELM-suppressed conditions. In looking closely at such discharges there is a very rich response in the loop voltage (i.e., a can of worms), due to the high gain nature of the PCS E-coil feedback control, the time constants used, and built in time lags in the E-supply system. The standard PCS algorithm also includes derivative gain. In some cases it appears that an ELM might be triggered by the positive loop voltage response, but the similarities of all the timescales, especially in the ELM pacing experiments, makes this difficult to pin down. In one shot studied in detail (142250) it might be that the RMP modulation of the pedestal bootstrap current is generating the loop voltage response from the E-supply. These voltage swings are not small relative to the steady state voltage. They do not appear to be caused by some kind of vacuum field interaction between the I-coils and the Rogowskis, based upon reference shot pulses before t=0. | ||
| Resource Requirements: | Standard tokamak with NBI. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | *Special requirements may potentially include modifications to the PCS algorithms for Ip control, or radial position control, but perhaps going in and out of standard feedback during a blip may suffice.
*The response time of the power supply control circuits also needs to be checked. | ||
| Title | 22: Test convective cell hypothesis as explanation for the particle pump out during RMP application | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | T.E.Evans et al. | ITPA Joint Experiment : | Yes |
| Description: | RMP ELM suppression in DIII-D is strongly correlated with a reduction of the edge pressure gradient (grad_p). It was shown for ITER similar shape plasmas, that the reduction in grad_p consists out of both, a reduction in the density (density pump out) as well as the temperature in the edge [O. Schmitz et al. (2009) PRL 103, 165005]. However, while the temperature effect has very narrow resonances in q_95, the density pump out does not show a direct dependence on q95.
These recent finding suggests that for the RMP induced particle transport a dominant mechanism different than a pure stochastic layer formation is at play. We will test in this experiment the hypothesis of edge localized magnetic islands acting as convective cells driving an enhanced outward particle transport. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Addressing these physics questions require high quality profile data for different q_95 and will be obtained by a shot to shot q_95 scan, each optimized for optimal diagnostic coverage. This approach is required to obtain long periods, with steady pedestal conditions and a well established wall and SOL condition equilibrium, that will allow us to acquire sufficiently long data records needed to insure statistically significant Te, Ti, ne, ni, Zeff, v_tor and v_pol profile data that is required for calculating accurate values of Er and grad_p as a function of qa..
We aim at L-mode data as well as H-mod data, with and w/o RMP for 10 q_95 values within the ELM suppressed resonant range (3.9 > q95 > 3.1) and another 10 data points for 4.0 < q_95 < 7.0 to address the q_95 dependence of the particle pump out. The target plasma is the ITER similar shape plasma at low collisionality with even parity n=3 Icoil field at moderate RMP current capability. We will have one q_95 value per discharge, a 1s L-mode phase w/o RMP, 1 s H-mode w/o RMP, 1 s H-mode with RMP and 1s L-mode with RMP. This sequence will allow comparing to a broad range of experiments with dynamic q_95 ramp down completing the data analysis by data with good statistics. | ||
| Background: | The so-called density pump-out effect was observed in the first experiments done with RMPs in TEXT [T.E.Evans et al. (1989) Journ. Nucl. Mater] and has been observed in every RMP experiment since. In TEXT, the particle pump-out effect was attributed to an increase in radial ion convective losses caused by island localized EXB drifts. Evidence for the existence of these theoretically predicted local magnetic island convective cells has been shown by experiments in limited L-mode plasmas on TEXT, JIPP T-IIU, COMPASS, Tore Supra and TEXTOR.
The purpose of the DIII-D experiment proposed here is to find first time evidence for existence of local island convection cells in limited DIII-D L-mode plasma and to determine if their properties change across the L-H transition. Screening theory predicts that increasing poloidal rotation and total pressure at the plasma edge potentially screens the RMP fields strongly thus eliminating magnetic islands that cause the convective cells in L-mode plasmas. If both screening and island convective theory are correct then the strong density pump-out seen in DIII-D RMP experiments creates a paradox that can only be resolved by the introduction of a new particle transport mechanism in H-mode plasmas. The alternative is that screening theory is not correct and magnetic islands persist in H-mode plasmas. A similarity experiment addressing the basic mechanisms of edge transpĆ¼ort in the presence of edge localized magnetic islands is suggested at TEXTOR. The aim is to measure the local electric field and potential in and around edge localized magnetic islands. Both experimental proposals are connected to DIII-D MP-2009-02-02 (RMP transport in L-mode) and potential follow up (Boedo et al.). | ||
| Resource Requirements: | Patch panel capable to run HFS limited and ISS shaped divertor plasmas
SPAs on C-supplies for high I_coil current capability (6kA) B_T = 2T for good ECE measurements | ||
| Diagnostic Requirements: | - core, edge and divertor Thomson
- ECE, ECE-I - fast profile reflectometer - all measurements for particle balance (E.A.Unterberg) - fast reciprocating probes (midplane and X-point) - fast UCSD camera with HFS tangential view in CII, CIII, H_a (alterating) - new LLNL periscope, divertor IR and vissible cameras - fast Li beam (for density profile and polarimneter) - SXR arrays & new divertor SXR - divertor LP - filter scopes | ||
| Analysis Requirements: | - particle balance
- TRIP3D vacuum field line tracing - NIMROD resistive MHD modeling - EMC3-Eirene transport with plasma response - XCGO with self consistent plasma response | ||
| Other Requirements: | -- | ||
| Title | 23: Modulation of transport by Convective cells | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | Valerie Izzo, Todd Evans, Dmitry Orlov, Dmitry Rudakov, Oliver Schmitz, Holger Reimerdes, Zeke Unterberg, Charles Lasnier | ITPA Joint Experiment : | Yes |
| Description: | It has been for long proposed that convective cells modulate plasma transport, however, indirect and little evidence exists for this. We propose to measure the effect of convective cells on transport directly by using fast scanning probes in low power discharges while varying the island structure and plasma parameters. We will measure plasma potential and electric fields in the island structure, rms levels of density, potential and temperature fluctuations and radial transport of heat and particles. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | We will create L-mode (ECH and NBI heated), elongated (not diverted) discharges, OH discharges and add I-coil and C-coil components at various currents (i.e .island strength)such as n=1, 240 deg, 1ka C-coil where examples are shots 141919, 20, 21 and 22.These variosu configurations will position the X and O-points of the islands in front of the probes slightly inside the LCFS. Data will be taken while the island position is varied to probe various parts of the islands. | ||
| Background: | Recent DIII-D experiments have shown the suppression of ELMs by the modification of profiles to below peeling-ballooning stability thresholds. How the profiles are modified is still unknown. Data dating from the 1990's in Tore Supra and TEXT indicate the the application of external resonant fields to a tokamak discharge change the global discharge parameters, in particular the particle inventory. Data has indicated a modulation of the profiles in the vicinity of where the vacuum calculation predicts islands will be located. So there is some evidence that the islands modify the local/global transport but no precise mechanism is known at this moment. | ||
| Resource Requirements: | I and C coils, diagnostic listed below, scanning probes, NIMROD code, NBI and/or ECH. | ||
| Diagnostic Requirements: | scanning probes, IR cameras, visible cameras, visible spectroscopy, D-alpha and D_beta diode arrays, divertor probes, wall probes. | ||
| Analysis Requirements: | Need the NIMROD code to compare the data to calculations including the plasma response. | ||
| Other Requirements: | |||
| Title | 24: Snowflake Divertor | ||
| Name: | Makowski makowski@fusion.gat.com | Affiliation: | AKIMA Infrastructure Services |
| Research Area: | General PBI | Presentation time: | Not requested |
| Co-Author(s): | D Ryutov and M. Umansky | ITPA Joint Experiment : | No |
| Description: | Realize Snowflake configuration and characterize its impact on the pedestal and divertor through measurements of the pedestal and heat-flux profiles. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The snowflake divertor is characterized by a second order null at the x-point. This has the effect of approximately doubling the shear the vicinity of the x-point and substantially increasing the flux expansion there. The former is predicted (and has been qualitatively shown) to modify the pedestal and affect ELM behavior while the latter should reduce the peak heat flux and broaden the heat flux profile.
The experimental approach is to establish a reference standard single null x-point divertor configuration and, with minimal shape change (particularly with regard to triangularity), transition to a Snowflake divertor (this has been demonstrated transiently on DIII-D). The two divertor configurations will be characterized using the full suite of pedestal and edge diagnostics available on DIII-D. | ||
| Background: | The standard single null x-point divertor configuration results in a narrow strike point with correspondingly high heat flux often approaching the material limits. The Snowflake divertor configuration mitigates the high heat flux by increasing its footprint on the target plate by increasing the flux expansion (a property of the second order null). | ||
| Resource Requirements: | Development of plasma control algorithm for the the snowflake divertor. 1 or 2 2-hour evening shfits for shape development and to test the control algorithm. | ||
| Diagnostic Requirements: | Edge and core Thomson, IRTV, divertor floor probes, reciprocating probe, Tang TV, reflectometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 25: Ammonia injection through Dimes | ||
| Name: | Tabares none | Affiliation: | CIEMAT |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Not requested |
| Co-Author(s): | Dimitri Rudakov, Anthony Leonard, Clement Wong | ITPA Joint Experiment : | No |
| Description: | Ammonia (carbon radical scavenger) will be injected through the DiMES plug at a location withdrawn from the divertor floor by a few cms. The formation of carbon deposits will be investigated and compared to a reference case without ammonia injection. The possible diffusion of ammonia into the divertor plasma will be tested by OES. The effect of ammonia seeding on the optical characteristics of mirrors at DiMES will be also monitored. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The DiMES plug needs to be provided with a fueling line, a pumping system and a mass spectrometer. Ammonia will be injected during 8-10 plasma shots, then carbon deposition on the mirror samples will be measured upon DiMES plug extraction by several techniques. Simultaneous recording of N emission lines in the divertor plasma nearby will be made in order to quantify the degree of plasma perturbation by the injected species, if any. Reference experiment without ammonia injection may be required to quantify the level of carbon deposition under the selected plasma scenario if the existing reference is not close enough. | ||
| Background: | The scavenger technique for the inhibition of tritium trapping in co-deposits at remote locations has been tested in real divertors only in JET and AUG so far. Nitrogen was used then as a scavenger. Very recent results from linear plasma divertor simulators (PILOT PSI, PSI-2) indicate that ammonia is a better scavenger of film precursors (carbon radicals) than nitrogen, and can prevent the formation of deposits even when injected outside the divertor plasma, through ammonia-radical direct reactions. A 4 nm/min film deposition rate was inhibited by 1Pa.m3 flow of ammonia in PILOT PSI, downstream the plasma. However, the possible perturbation of the plasma by the injection of scavengers remains to be proved. The active, real time suppression of tritium retention in remote areas is a pre-requisite for the use of carbon facing components as an alternative to the present, tungsten-based design. | ||
| Resource Requirements: | 2 Ā½ day experiments | ||
| Diagnostic Requirements: | Dimes, Divertor spectroscopy, Mass spectrometrer | ||
| Analysis Requirements: | |||
| Other Requirements: | Gas inlet at DiMES | ||
| Title | 26: Heat flux at midplane | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Thermal Transport in the Boundry | Presentation time: | Requested |
| Co-Author(s): | Charles Lasnier, Dmitry Rudakov, Mike Makowski | ITPA Joint Experiment : | Yes |
| Description: | To identify the main transport channels and parameters responsible for the heat flux footprint in the divertor. We will measure DIRECTLY the thermal transport due to conduction and convection using scanning probes while scanning plasma parameters known to change the heat flux at the divertor significantly (such as Ip) | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Low power discharges with NBI and ECH scanning Bt and Ip while measuring midplane profiles and transport with the probe and the fllor heat deposition with the IR camera. | ||
| Background: | Previous experiments indicate strong scaling of Lambda_q with plasma parameters. We will trace those to fundamental transport processes at the midplane | ||
| Resource Requirements: | NBI, ECH. | ||
| Diagnostic Requirements: | scanning probes, IR cameras, standar DIII-D diagnostics | ||
| Analysis Requirements: | Compare measurements to turbulence codes (lodestar, LLNL, etc) | ||
| Other Requirements: | |||
| Title | 27: Super-X Divertor | ||
| Name: | Makowski makowski@fusion.gat.com | Affiliation: | AKIMA Infrastructure Services |
| Research Area: | General PBI | Presentation time: | Not requested |
| Co-Author(s): | A. Leonard | ITPA Joint Experiment : | No |
| Description: | Realize Super-X divertor configuration and characterize its impact on the pedestal and divertor through measurements of the pedestal and heat-flux profiles. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental approach is to establish a plasma analogous to those used in for validating off-axis neutral beam injection. Such plasmas only occupy about half the volume of the vessel thus allowing space for formation of a long divertor leg. The outer divertor leg will be terminated on the shelf. The major radius of the outer strike point will then be varied and the divertor properties measured. | ||
| Background: | The standard single null x-point divertor configuration results in a narrow strike point with correspondingly high heat flux often approaching the material limits. The Super-X divertor configuration mitigates the high heat flux primarily by increasing the location of the strike point to larger major radius. This has multiple effects. It increases the heat-flux footprint due to an increase in the wetted area (proportional to major radius) and an increase in the flux expansion since Bt decreases with major radius. The connection length also increases so that the electron target plate temperature should decrease and radiation along a flux should increase thus lowering the peak heat flux. | ||
| Resource Requirements: | Develop strike point control method. 1 2-hour evening shift for testing strike point control algorithm. | ||
| Diagnostic Requirements: | floor probes, Tang TV | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 28: Causality of Deuterium Flows in the Boundary | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Divertor Detachment and Plasma Flows | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov, E. Belli, J. Watkins, Lasnier, Makowski | ITPA Joint Experiment : | No |
| Description: | Boundary flows can be produced by various mechanisms: pressure asymmetries, Pfirsch-Schluter, particle momentum loss, Reynolds-stress, etc.We aim to identify these mechanisms and under which conditions they dominate | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To perform experiments where we carry:
1) pressure asymmetry at the divertor by using selective divertor detachment (inner or inner+ outer) 2) vary the edge pressure gradient by: a) comparing L and H mode flows, b) heating the edge with tangential or derated beams, c) heating the edge with ECH 2) Change radial particle loss by varying density and this increase the outboard particle/momentum loss. 3) Perform Up-Down and Bt/Ip inversion (separately) to select the thermal particle loss mechanism | ||
| Background: | Past and ongoing experiments have discovered new information about edge flows. Among them, the discovery of a thin rotating layer at the LCFS and a Pfirsch-Schluter mechanism, and the suggestion that others, such as thermal particle loss and ballooning sources, exist, but it is unclear which dominates and under which conditions. | ||
| Resource Requirements: | NBI and ECH, machine diagnostics, CER, tangential cameras, IR camera. | ||
| Diagnostic Requirements: | Scanning probes, carbon flow diagnostics, spectroscopy, CER, | ||
| Analysis Requirements: | NEO, UEDGE, SOLPS5, etc. | ||
| Other Requirements: | |||
| Title | 29: Tilted Beamline torque injection | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | Wayne Solomon | ITPA Joint Experiment : | No |
| Description: | *The purpose of this experiment is to understand NBI momentum sources and momentum confinement using the tilted beamline. A NB deposits torque in two very different ways, through the prompt fast ion radial current effect, "jXB", and through frictional drag on the bulk plasma as a fast ion loses directed momentum. These operate on vastly different timescales. The "jxB" effect deposits momentum on the timescale of the poloidal ion transit time (bounce time), while the frictional deposition is on a kinetic timescale, such as the pitch angle scattering time, or the slowing down time, or both, depending on the details of the fast ion - thermal ion, or - thermal electron, couplings.
*The tilted beam provides a mechanism to essentially eliminate the jxB torque ( reversed BT ). The experiment will compare momentum deposition (plasma acceleration) and momentum transport (modulated tilted beam) between normal and reversed-BT conditions. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | *Create target discharges (H-mode, possibly L-mode with time) with a slightly reduced volume plasma shape to allow complete elimination of tilted NBI injection into trapped orbits, based upon the line-of-sight path of the left source. Use pulsed tilted beam sources to measure the prompt velocity response (acceleration, hence momentum deposition) and also use modulated tilted sources to measure the momentum propagation and incremental confinement.
*Repeat with Btor reversed from the first day. *Additionally, by using a plasma shape that is shifted up slightly, it will be possible to avoid significant direct NBI torque deposition into the core of the discharge, allowing a measure of the timescale for diffusion, or a momentum pinch, to fill in the velocity profile. *Possibly use ECH to lengthen the slowing-down time. | ||
| Background: | *This is a straight-forward use of the new flexibility in NBI torque deposition. It could be that the jxB torque is more effectively "absorbed" because of the short timescale, whereas the slower frictional timescale may be subject to anomalous loss process.
*Experiments with counter-Ip injection of the tilted beam will also be very interesting since the prompt fast ion loss can be made much larger than for the non-tilted beam. This would also create a significant jxB torque, and become an E-field generator (negative E). It would likely be prudent to reduce the tilted source voltage, hence penetration and power, to avoid wall bombardment issues. | ||
| Resource Requirements: | 2 day experiment (minimum), normal and reversed-BT. Tilted NBI and diagnostic NBI (minimum).
ECH (desirable, not absolutely necessary) | ||
| Diagnostic Requirements: | Standard | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 30: ELM Pacing/Suppression using I-coils in axisymmetric mode | ||
| Name: | Leuer leuer@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | J. deGrassie, T. Evans, D. Humphreys,
G. Cunningham [MAST], C. Gimblett [MAST] |
ITPA Joint Experiment : | No |
| Description: | Goal of experiment is to the influence of the I-coil connected in an n=0 axisymmetric connection has on ELM frequency and size. These coils, being located inside the vacuum chamber, can produce very high frequency variations of local field and voltages in front of the coil and should produce substantial impact in the edge localized area. We expect ELM pacing as was achieved using vertical plasma oscillations in TCV and ASDEX or ELM suppression as has been achieved using n=3 in DIII-D. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish type I ELMing H-mode discharge, preferably using the ITER Scenario 2 shape and conditions. Inject constant frequency into Icoil in both Odd and Even parity at varying voltage levels. Increase frequency and determine if locking or suppression occurs. Best results may require use of the SPAs for low frequency studies and Audio Amps for high frequency studies. This probably will require testing during two different experimental days. | ||
| Background: | At least two other ELM pacing experiments are being proposed for the next campaign using axisymmetric fields from our poloidal field coils (E & F) to drive plasma loop voltage to increase/decrease plasma current [Cunningham 14, DeGrassie 21]. These proposals report that modulation of loop voltage can influence the ELMing frequency. However, the vacuum vessel greatly shields the plasma from these outside coils and limits the maximum frequency that can be effectively impressed on the plasma. However, the I-coil, when connected in an axisymmetric configuration (which I believe has never been done before) can create axisymmetric voltage which are very localized to the outer extreme of the plasma and can be injected at a much higher frequency than the outer PF coils. Although the plus/minus nature of the coils prohibits driving loop voltage (i.e. plasma current) as prescribed in the other PF/ELM pacing experiments, local gradients can be injected at a very fast time scale relative to the PF and should have some impact on the ELM behavior. Additionally, fast vertical motion of the plasma has also produced ELM triggering [Degeling, TCV, 2003; Martin, ASDEX-U, 2004) and the present proposal should have a similar influence but can be done at much higher frequency. Previous successful ELM/I-coil experiments on DIII-D have utilized primarily an n=3 connection for ELM pacing/suppression studies [Evans, DIII-D, 2004]. n==1 & 2 experiments have failed to produce ELM suppression and the loss of plasma rotation and locked modes are reportedly the issue [Schaffer, DIII-D 2009]. The advantage of axisymmetric fields is they should not overly reduce the natural rotation of the plasma (i.e. braking) but should still significantly influence the edge plasma conditions. | ||
| Resource Requirements: | ELMing H-mode with Type I elms, preferably in the ITER configuration, I-coil, SPA, Audio Amp | ||
| Diagnostic Requirements: | Edge diagnostics will be important, Fast magnetics, MSE, Thompson Scattering, Spread, Visible camera view bumper limiter, Bolometers, IR camera, CO2 Interferometer, ECE, SXR | ||
| Analysis Requirements: | EFIT/Transport | ||
| Other Requirements: | Need testing of the Audio Amp and Spa when driving the I-coil in the axisymmetric mode to determine the system characteristics, generate the optimal scenario for inducing largest voltage variations in the plasma and determine the maximum frequency attainable. | ||
| Title | 31: Magnetic perturbation plasma penetration | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | Izzo, Evans, Orlov, Rudakov, Reimerdes, Schmitz, SOlomon | ITPA Joint Experiment : | No |
| Description: | Evaluate the penetration of externally imposed magnetic field structures in the edge of DIII-D by using probe data (potentials, electric fields, Te, Ne profiles, turbulent transport, fluctuations and also magnetic field measurements from probe coils) to detect the signature of the islands (potential, temperature, density, etc) as the plasma resistivity and rotation velocity are varied. Compare to existing models. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To setup well defined islands near the LCFS where the probe can reach them (ref. 141920 and vary plasma response by changing: 1) resistivity and 2) edge rotation
Also vary perturbation strength by varying I and C coil currents. Once the perturbation is setup, introduce the probe at various times to evaluate the penetration of the perturbation vs time for a given rotation and resistivity. Compare to expectations. | ||
| Background: | It is well known that magnetic perturbation do penetrate into tokamak plasmas and that plasma rotation reduces the penetration. it is not known if the physics of the shielding are properly represented in the existing theory. A experimant | ||
| Resource Requirements: | NBI, ECH, low power ~ 1-2 MW | ||
| Diagnostic Requirements: | Scanning probes, IR cameras, Diode arrays, magnetics and DIII-D diagnostics. | ||
| Analysis Requirements: | NIMROD runs | ||
| Other Requirements: | |||
| Title | 32: ELM Suppression/reduction without density pump-out at reduced I-coil current | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T.Evans, O. Schmitz, E.J. Doyle, L. Zeng, T.L. Rhodes, G. Wang | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to demonstrate RMP ELM control (suppression or substantially reduced ELM amplitude) without substantial density pump-out in high triangularity plasmas. The proposed experiment optimizes conditions for detailed (time-resolved) turbulence/ transport measurements in the upper pedestal and outer core plasma.
Experiments during the 2008 campaign have indicated substantially reduced pump-out with lower I-coil current (<4 kA) than are typically employed for ELM suppression. With further reduction in current, small amplitude ELMs were observed. No degradation in beta_N was seen (beta_N ~2), and turbulence levels in the upper pedestal were substantially lower than during ELM suppression at high I-coil current. The proposed experiment consists of a systematic I-coil current scan (current stepping) with even and odd parity to enable measurements of the transient evolution of the density profile, upper pedestal ExB flow, and turbulence levels. ITG/TEM-scale electron temperature fluctuations can also be measured in these plasmas near the pedestal top where the optical thickness is sufficiently high. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Co-injected plasmas with are also planned to investigate differences in ELM response and density profile/turbulence modifications. Data from DBS systems located at 60 degrees and 240 degrees may help distinguish between resonant and non-resonant effects on turbulent transport. A full day with ~ 20 useful shots is requested to provide detailed turbulence measurements as a function of I-coil current/parity, with coverage of the outer core/pedestal region. Interlacing low and intermediate-k measurements in the pedestal requires 6 repeat shots for each I-coil current step (3 steps total); assuming that successive even/odd parity phases are combined in each shot. | ||
| Background: | Substantial density pump-out is often found with RMP ELM control in high triangularity plasmas is often accompanied by substantial density pump-out and core density profile modifications. In addition to the (desirable) reduction of the upper pedestal pressure gradient, increased core density gradients occur, increasing ITG or TEM turbulence growth rates and core particle loss. Experiments during the 2008 campaign have indicated substantially reduced pump-out and no degradation of beta_N with reduced I-coil current (<4 kA). Evidence for reduced upper pedestal turbulence during the reduced current phase was found in this initial experiment, however no systematic variation of I-coil current or documentation could be done in during the previous experiment. This previous experiment also suffered from the loss of key data from the CO2 interferometers, the Thomson scattering system and the profile reflectometer. | ||
| Resource Requirements: | 7 Neutral beams, I coil with SPA supplies | ||
| Diagnostic Requirements: | Profile reflectometry, all turbulence diagnostics
(DBS2,5,8) BES, CECE, FIR, MSE, CER | ||
| Analysis Requirements: | Comparisons of density/turbulence electric field profiles to TRIP-3D calculations; advanced modeling including the plasma response. | ||
| Other Requirements: | -- | ||
| Title | 33: Measure Intrinsic Rotation Size scaling in DIII-D alone -II | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | General IP | Presentation time: | Not requested |
| Co-Author(s): | Wayne Solomon, Keith Burrell, John Rice (MIT) | ITPA Joint Experiment : | No |
| Description: | *Size Scaling NEEDED to confidently extrapolate to ITER.
*Continue experiment started with 2009-51-01 *There, steady (enough) conditions were not obtained in the small and large extremes in major radius, presumably because of lack of operational time with new shapes. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | *Measure the Rice scaling slope, Rs, for three similar shapes at different size. Here, V = Rs*W/Ip.
*Focus on ECH H-modes + NBI blips to get unpolluted intrinsic rotation. *The sizes listed below have a variation in R^2 of 1.39, which we should be able to measure in the slope. *An R^2 scaling is indicated by direct comparison of the slopes between C-Mod and DIII-D, and fits with one dimensionless fit to the international database, that of MA ~ BetaN, where MA is the so-called AlfvƩn Mach number (Rice, Ince-Cushman et al). | ||
| Background: | *We obtained the three necessary shapes:
small 136868.1325 R=1.50 R/a=.35 medium 136871.1345 R=1.64 R/a=.33 large 136878.1345 R=1.77 R/a=.34 The small was plagued by a drift in the control system, shape-wise. The large had wall interaction trouble (small gapout), going in and out of ECH H-mode. Both of these issues can be solved with machine time. | ||
| Resource Requirements: | 2 day experiment (realistically)
Gyrotrons | ||
| Diagnostic Requirements: | Standard | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 34: High Collisionless NBI Torque Drive for GAMs, aka the VH-mode path? | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | George McKee, Terry Rhodes | ITPA Joint Experiment : | No |
| Description: | *Reprise of #17 (2009) and #78 (2007)
*Use high power NBI co-torque to transiently drive Geodesic Acoustic Modes and measure the plasma response and mode properties with BES. The model requires that "enough" prompt NBI radial current be injected to raise the E field "fast enough" so that the plasma rings in this fashion (see Background below). Actually, the proposed target plasma and suddenly switched-on NBI level are reminiscent of the VH-mode recipe. *Counter-Ip operation with the off axis beam will be evaluated as a possible enhanced prompt E-field driver. However, BES will hopefully be a critical diagnostic. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | *Select a target plasma with low collisionality, with q95 ~ 6. We probably want a DND biased up with normal BT to stave off the H-mode transition as long as possible. 3 NBI co-sources are turned on simultaneously and BES is deployed to look for a GAM response. Other turbulence diagnostics will be useful. If struck, perhaps the GAM response can be followed with only the one (150) beam for some time. Perhaps we will be able to do a number of measurements with various beam mixtures after the thump and ideally see if there is any correlation between the GAM response and any subsequent H-mode transition, or transport barrier formation. | ||
| Background: | *NBI torque injected by ions into promptly trapped orbits results in a radial fast ion current that delivers this torque via Jfast X B. The low collisionality plasma responds as a dielectric for times much shorter than the momentum transport timescale, that is, a return polarization current is generated in the bulk ions. This polarization is calculable for collisionless orbits, and depends upon the details of the orbit topology for an ion. For timescales much shorter than the thermal ion bounce time the gyro-orbits shift, giving the so-called classical polarizability. For timescales longer than a bounce time the banana orbits shift giving the neoclassical polarizability, about 100 times larger than the classical value. Passing-trapped ion collisions bring the plasma response to a common neoclassical value.
*So, the plasma dielectric in this regime is a function of frequency (timescale). Striking the plasma fast enough with a radial current source results in GAM generation as described in Hinton and Rosenbluth, PPCF vol 41, A653 (1999). These GAM oscillations are then collisionally damped. *We need to get the E-field to rise fast enough in a thermal ion bounce time in order to modify the orbit. An estimate shows that the prompt radial fast ion current scales with the local plasma beta, and Ip^2. So we want a low beta target (and low collisionality is important for longer GAM damping time), and low Ip, i.e. higher q95, say 5-6. The estimate indicates 3 co-sources would be enough. Hopefully, less will work to give a range to study. | ||
| Resource Requirements: | 1 day. NBI. Gyrotrons. | ||
| Diagnostic Requirements: | Standard. BES. Other fast diagnostics (turbulence, mhd, AE, ...) | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 35: Understanding the L-H Power threshold dependence on the X-Point height | ||
| Name: | Gohil gohil@fusion.gat.com | Affiliation: | GA |
| Research Area: | General IP | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Perform experiments to determine the key physics behind the dependence of the L-H power threshold on the X-point height above the divertor | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make all possible diagnostic measurements available that can reveal the key changes in the edge quantities as the X-point height is varied. In particular, examine the changes in quantities such as recycling, turbulence, SOL flows, profiles changes just inside the separatrix, etc. as the X-pint is changes and determine how these quantities relate to the changes in the L-H power threshold. | ||
| Background: | The L-H power threshold has been determined to have a strong dependence on the X-point height above the divertor surface for H, D and He plasmas. This indicates that there is common physics behind this effect, which can result in factors of 2 differences between the experimental L-H power threshold and the predictions from the present L-H power threshold scalings. Due to concurrent changes in several quantities as the X-point is varied, it is presently not possible to definitively explain the physics behind this effect. These experiments aim to separate out certain quantities such as recycling, reconnection lengths at the divertor, etc. in order to determine the important parameters that influence the power threshold. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | All available turbulence and divertor diagnostics, including all profile and SOL diagnostics. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 36: Investigate Disagreements Between Thomson Scattering and ECE Measurements in High Te Discharges | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Heating & Current Drive | Presentation time: | Requested |
| Co-Author(s): | M. E. Austin, R. Prater, B. Bray, R. Pinsker, F. Volpe | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to search for a discrepancy between Thomson scattering (TS) and ECE measurements of Te on DIII-D in discharges with high electron temperature. To carry out the experiment, L-mode discharges with core transport barriers with neutral beam injection (NBI) plus fast wave heating (FWH) are used to attain central electron temperatures of Te(0) = 9 keV (achieved last year in MP 2010-55-01). This year, we are also targeting high power, off-axis ECH discharges to reach Te(0) > 12 keV, while maintaining a Maxwellian eedf in the core plasma. Electron temperature measurements from TS and the absolutely calibrated Michelson interferometer would be compared to look for any discrepancy between the two diagnostics. With last year's data a gate timing issue with TS was discovered several weeks after the run. This limited the time resolution/available statistics for the experiment (Bray, Friday Science Meeting, May 7th , 2010) and is a main reason the experiment must be repeated. A second motivation for this new experiment this year is that based on analysis of a shot with ECH from last year, we were able to use CQL3D modeling to find a "recipe" (Prater, APS 2010) for making even higher Te plasmas (12-15 keV) of interest this year with ECH (see approach below). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Reproduce shot 140715 (beams and FW to high Te() > 9 keV). Obtain high quality TS and ECE data for detailed comparisons and modeling.
2) Apply high power ECH near r=0.25 or 0.3 starting when the current ramp is nearly complete ā??top of flux surface is best poloidal location ā??generate weak negative shear for good confinement but avoid the very strong eITB that comes with strong negative central shear ā??leave the plasma center free of ECH to maintain a Maxwellian distribution there - Use large enough k_parallel that little EC power is deposited via relativistic downshift ā??split the power equally into positive k_parallel and negative k_parallel to avoid driving excessive local current | ||
| Background: | Note that in the core of typical tokamak plasmas with Te(0) < 7 keV the TS and ECE measurements of electron temperature are in very good agreement. Also note that TS and ECE measurements of electron temperature often disagree in high Te(0) discharges that are strongly heated via ECH, but in these cases the disagreement can be explained by a well understood perturbation of the electron energy distribution function caused by the ECH [3]. In contrast to these cases, the cause of the TS/ECE discrepancy in discharges heated with only NBI and FWH where Te(0) > 7 keV is not known. Such a discrepancy has been observed in NBI discharges and discharges heated with both NBI and Ion Cyclotron Resonance Heating (ICRH) discharges in TFTR and JET [1,2]. In these cases, the central electron temperature Te(0) measured with ECE diagnostics is 10-20% higher than the TS measurement of Te(0) The discrepancy starts at Te(0) ~ 7 keV and increases approximately linearly with electron temperature. Theoretically, a non-Maxwellian electron distribution f(v) with distortion near the thermal velocity may create such a measurement discrepancy between TS and ECE measurements [4], however, no known mechanism can sustain that type of distribution with finite heating power.
As an ITPA joint experiment, either a positive or a negative result on this topic from DIII-D can significantly impact international efforts to understand the past discrepancies that have been reported on TFTR and JET [1,2]. For example, a positive result, the observation of the discrepancy between TS and ECE on DIII-D, would verify the discrepancy on an additional machine and would therefore motivate a new and detailed investigation of the phenomenon. However, a negative result, the observation of no discrepancy under a variety of conditions with high electron temperature produced with NBI and FWH, would be equally beneficial as it would show that agreement between TS and ECE can be obtained in high temperature tokamak discharges. In both cases, the experimental results and associated modeling will improve the understanding of heating and diagnostic techniques in high temperature plasmas relevant for ITER and reactor-grade tokamak experiments. [1] E. de la Luna, et al., Rev. Sci. Instrum. 74, 1414 (2003) [2] G. Taylor, PPPL report 4202 (2006) [3] C. C. Petty et al. GA Report A25804 (2007) [4] V. Krivenski et al. 29th EPS Conference on Plasma Phys. and Contr. Fusion Montreux, 17-21 June 2002 ECA Vol. 26B, O-1.03 (2002) | ||
| Resource Requirements: | All gyrotrons
FW heating systems All available NB sources | ||
| Diagnostic Requirements: | Thomson scattering, Michelson interferometer, 40-channel ECE radiometer, ECEI, CER and MSE, fast magnetics, all fast ion diagnostics, all available fluctuation diagnostics. If available, oblique ECE. | ||
| Analysis Requirements: | EFIT, gaprofiles, ECESIM, ONETWO/autoonetwo, GENRAY, TORAY,and CQL3D | ||
| Other Requirements: | |||
| Title | 37: Passive FIDA imaging | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | M. Van Zeeland | ITPA Joint Experiment : | No |
| Description: | This experiment has two objectives:
1) Diagnostic demonstration that FIDA imaging is a powerful & affordable way to measure fast-ion loss over a large poloidal angle. 2) Better diagnose the class of particles expelled by off-axis fishbones & EGAMs. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Reestablish conditions where off-axis fishbones cause large bursts of fast-ion loss.
2) Image the plasma edge with the 135 degree camera view. Vary filter angle between shots. 3) (Time permitting) Repeat measurement in plasma with EGAMs. This is a 1/2 day experiment. | ||
| Background: | Analysis of BES data from the 2010 campaign (and from NSTX) shows that large bursts of edge FIDA light occur when fast ions are expelled to the edge by instabilities. The paper on this phenomenon suggests that FIDA imaging could measure the poloidal distribution of losses with a single detector--a much simpler method than multiple scintillator detectors. It is possible that variants of this technique could measure alpha loss in ITER. The largest signals were observed during off-axis fishbones, so this is the best condition for a proof-of-principle experiment. | ||
| Resource Requirements: | 6 neutral beam sources (150 beams NOT required).
ECCD desirable | ||
| Diagnostic Requirements: | Essential: FIDA imaging camera, BES | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 38: Effect of 3D fields on fusion-product confinement | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | Yubao Zhu, Gerrit Kramer | ITPA Joint Experiment : | Yes |
| Description: | The 14 MeV neutrons produced by the triton d-d fusion product is used to assess the effect of TBM and I-coil error fields on the confinement of 1.0 MeV tritons. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a plasma condition with a large, steady, d-d rate (for example, an ELMing H-mode plasma). Add TBM error fields of ~ 500-ms duration (probably two pulses per shot). Make similar measurements for I-coil pulses. If time permits, vary plasma current to alter the triton prompt-loss boundary. | ||
| Background: | Interesting triton burnup data were obtained in 2010 but, because we used an old radiation-damaged detector, we had to average many pulses to obtain adequate counting statistics. In 2011, a factor-of-ten increase in sensitivity is anticipated, allowing for useful data in a single discharge.
The effect of 3D fields on EP confinement is the topic of an ITPA joint experiment. | ||
| Resource Requirements: | High d-d neutron rate discharges (>~2e15 n/s) of duration exceeding 1 second. Error field pulses of > 400 ms. | ||
| Diagnostic Requirements: | 2.5 & 14 MeV neutron detectors.
Plasma parameters | ||
| Analysis Requirements: | The data will be compared with calculations by several codes including SPIRAL and OFMC. | ||
| Other Requirements: | |||
| Title | 39: Expand the high li, betaN >4 operating regime through instability avoidance and higher heating power | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Steady State) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make use of ECCD stabilization of 2/1 tearing modes and modifications in the discharge evolution during the betaN ramp up in order to extend the high-performance pulse length and allow operation at lower values of q95. Take advantage of the additional neutral beam made available in 2010 and the sixth gyrotron to push betaN above 5 and test the effect of wall stabilization at high li. Make measurements of the fast ion profile in order to understand anomalous losses. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In order to operate at lower values of q95, it is necessary to avoid the early, fast-growing n = 1 mode. Stability of previous discharges is still being studied in order to understand this mode, but its occurrence is likely coupled to the current profile which has a region of negative flux surface average current just inside the H-mode pedestal which is a result of the negative surface voltage produced by the Ip feedback system because of the current overdrive by the noninductive current. One approach would be to avoid this negative current by holding the surface voltage to more positive values (a technique also used in fNI = 1 discharges in TCV). Holding the surface voltage at 0 would also allow a clear demonstration of noninductive current overdrive. The other possibility is to modify the time evolution of beta and density during the beta ramp up. ECCD would be used to avoid the 2/1 mode that terminates the high performance phase. | ||
| Background: | In 2008, high li discharges with betaN >4.5 were obtained that had fNI = 1.2 and betaN above 4 for 1 s. Bootstrap current fraction was above 80%. In the early portion of the high beta phase when li was near 1.4, even with betaN = 4.5 the discharge was operating below the no wall n = 1 ideal stability limit. BetaN was limited by available heating power. The duration of the high-performance phase was limited by onset of a 2/1 tearing mode. Best performance was obtained with q95 near 7. At lower values of q95, the high beta phase was terminated during the beta ramp up by a fast growing n = 1 instability. Comparisons with ONETWO indicate significant anomalous fast ion loss, possibly a result of semicontinuous 1/1 mode activity. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Would make use of FIDA. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 40: Maintaining high li at high betaN using RMP and near-axis current drive | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Steady State) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make high li, betaN >4 discharges more stationary by reducing the rate of decrease of li through replacement of ohmic current near the axis with ECCD and by using RMP to reduce the H-mode pedestal density in order to reduce the edge bootstrap current. Results would be used to determine what would be required to produce a true, steady-state, high li, high betaT discharge. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | There are two primary parts to this experiment. First, the deposition profile of the ECCD would be varied in order to determine its effectiveness in replacing the core ohmic current. A accompanying goal would be to determine if q(0) can be raised slightly in order to avoid 1/1 activity. The new off-axis beam injection capability may be useful for tailoring the NBCD profile near the axis. It is unlikely that, even with six gyrotrons, the ohmic current can be completely replaced in the core. However, we can obtain scaling information in order to compare with models and determine how much external current drive would be necessary to make a stationary high li discharge. In the other part of the experiment, the RMP fields would be used to reduce the H-mode pedestal pressure and, particularly, the density. Supposedly there is a window for ELM stabilization with q just above 7 which matches the best high li discharge from 2008. However, ELM stabilization isn't necessarily needed, and the RMP fields are reported to have an effect on the pedestal height even away from the resonance required to stabilize ELMs. The best discharges in 2008 were strongly overdriven with noninductive current which may have contributed to the rate of decrease of li. Some exploration of the balance between total plasma current and the noninductive current would be done in order to test the effect on the li decrease. | ||
| Background: | In the fNI >1 discharges produced in 2008, li decreases slowly as the trapped ohmic current decays in the discharge core and the bootstrap current builds in the H-mode pedestal. In the best performance discharge, because Ip was relatively low (q95 about 7), and the noninductively driven current exceeded the total programmed current, the edge surface voltage was driven negative. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 41: Beam-ion confinement of the off-axis beams (classical effects) | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | M. Van Zeeland, J.M. Park, et al. | ITPA Joint Experiment : | No |
| Description: | Goal: Use neutron & FIDA measurements to verify NUBEAM modeling of the fast-ion distribution function in plasmas with <~ 3 MW of beam power. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In all plasmas, cycle through the on-axis & off-axis Left & Right beams with two types of injection patterns:
1) Use beam blips & 2) Sources injected successively for ~ 100-ms pulses. Meanwhile, blip the FIDA beams to obtain data. Do this for as many conditions as possible, with the highest priority being +- Ip. | ||
| Background: | This proposal is based on the 2008 experiments used to test off-axis injection in vertically shifted plasmas; see PPCF 51 (2009) 125001 for details. | ||
| Resource Requirements: | All neutral beam sources except 210LT and either 30RT or 330RT. | ||
| Diagnostic Requirements: | Neutrons, FIDA spectrometers & cameras, core plasma parameters | ||
| Analysis Requirements: | NUBEAM & FIDASIM | ||
| Other Requirements: | |||
| Title | 42: Dependence of confinement and stability on toroidal rotation in high li discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Steady State) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Produce a high li discharge that runs without beta collapse at significant betaN (for example, about 4). Evaluate the dependence of confinement on toroidal rotation. Evaluate the effect of the toroidal rotation velocity on the stability limit, both the maximum attainable betaN and the no-wall limit as measured with MHD spectroscopy. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce a high betaN discharge similar to those produced in 2008. Add counter injection beams. In order to reduce the rotation to low values, it will probably be necessary to operate at less than the maximum betaN. | ||
| Background: | The normalized confinement in high li discharges seems to increase as the beam power increases, possibly indicating a dependence of confinement on toroidal rotation velocity. On the other hand, experiments in the 1990s also indicated that the enhanced confinement at higher li depends on the poloidal field strength profile. It is essential to understand the confinement at high li under low rotation conditions as might be expected in a reactor. Also, a motivation for the high li scenario is that high betaN can be obtained in the absence of wall stabilization. The stability at low rotation in DIII-D discharges should be tested to determine the role of the wall in stabilization. Discharges in 2008 had phases with betaN below the no-wall limit and phases with betaN above the limit. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 43: Dependence of confinement and stability on toroidal rotation in high li discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Produce a high li discharge that runs without beta collapse at significant betaN (for example, about 4). Evaluate the dependence of confinement on toroidal rotation. Evaluate the effect of the toroidal rotation velocity on the stability limit, both the maximum attainable betaN and the no-wall limit as measured with MHD spectroscopy. This proposal is more relevant to FNSF than it is to ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce a high betaN discharge similar to those produced in 2008. Add counter injection beams. In order to reduce the rotation to low values, it will probably be necessary to operate at less than the maximum betaN. | ||
| Background: | The normalized confinement in high li discharges seems to increase as the beam power increases, possibly indicating a dependence of confinement on toroidal rotation velocity. On the other hand, experiments in the 1990s also indicated that the enhanced confinement at higher li depends on the poloidal field strength profile. It is essential to understand the confinement at high li under low rotation conditions as might be expected in a reactor. Also, a motivation for the high li scenario is that high betaN can be obtained in the absence of wall stabilization. The stability at low rotation in DIII-D discharges should be tested to determine the role of the wall in stabilization. Discharges in 2008 had phases with betaN below the no-wall limit and phases with betaN above the limit. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 44: Effect of microturbulence on off-axis beam-ion confinement | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | D. Pace, M. Van Zeeland, J.M. Park, et al. | ITPA Joint Experiment : | Yes |
| Description: | The ratio of fast-ion energy E to plasma temperature T is varied in plasmas without MHD or fast-ion driven instabilities. The shape and beta is varied in an attempt to alter the nature of the microturbulence. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Basic plan: Use off-axis beams and ECCD to make a hot, sawtooth-free discharge. Use FIDA spectroscopy & imaging to look for inward transport of the beam ions. Document microturbulence.
Variations: 1) Lower beam voltage to lower E/T. 2) Use more EC power or beams to lower E/T. 3) Lower triangularity to (possibly) increase electromagnetic turbulence. 4) Obtain data at two values of beta with same E/T | ||
| Background: | Evidence for fast-ion transport by microturbulence was obtained in the 2008 experiments with vertically-shifted plasmas when the ratio of E/T was small; see PRL 51 (2009) 125001. With the new off-axis beams, diagnostic coverage should be better. This experiment contributes to an ITPA Energetic Particle joint experiment. | ||
| Resource Requirements: | All beam sources except 210LT and 330RT.
At least 4 gyrotrons | ||
| Diagnostic Requirements: | All EP diagnostics
Fluctuation diagnostics | ||
| Analysis Requirements: | NUBEAM & FIDASIM for initial analysis.
Gyrokinetic analysis with GYRO, GTC and (probably) GENE. | ||
| Other Requirements: | |||
| Title | 45: Direct Measurement of E_rad Corrugation at Rational Surfaces | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | M. E. Austin | ITPA Joint Experiment : | No |
| Description: | Use the combination of co and counter MSE views to directly measure the corrugation in the radial electric field at rational q surfaces that is responsible for transport barriers. The target plasmas are balanced-NBI L-modes with early heating so that q>2. The analysis will focus on MSE channels that view the radius where a rational surface, such as the q=2 surface, first enters the plasma. In the absence of E_rad effects, the co and counter viewing MSE channels will measure the same magnetic field pitch angle. Thus, a separation between the co/counter MSE signals at the time a rational q surface enters the plasma is a direct measurement of the E_rad corrugation effect. There is the option to add the off-axis beam to slow the current profile evolution, which should make the corrugation easier to observe. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish L-mode plasma with early beam heating to slow the evolution of the current profile. (2) Use 30LT and 210RT beams without modulation to collect continuous MSE and CER data. (3) May need to move the plasma location around to make sure the MSE channels are looking exactly at the location where the rational q surface (especially q=2) first enters the plasma. (4) Try adding the off-axis beam (at maximum downward angle) to slow the evolution of q_min. | ||
| Background: | Previous experiments by Max Austin found corregations in the electron temperature profile when a rational q surface entered the L-mode plasma. These corrugations were observed for both co-NBI and balanced-NBI (although only the co-NBI cases resulted in long lasting transport barriers). The GYRO turbulence simulation code predicted the existence of these corrugations by means of a equilibrium ExB shear flow driven by the zonal flows. This experimental proposal will look for direct evidence of this ExB shear flow by means of the E_rad sensitivity of the MSE diagnostic. | ||
| Resource Requirements: | NBI: 30LT and 210RT essential. 150 beamline tilted downwards is desired (but need to consider impact on BES). | ||
| Diagnostic Requirements: | MSE is critical. Fluctuation diagnostics are desirable. | ||
| Analysis Requirements: | Need GYRO simulations. | ||
| Other Requirements: | |||
| Title | 46: RMP H-mode discharges with high H-98 and mitigated/eliminated ELMs | ||
| Name: | Jakubowski marcin.jakubowski@ipp.mpg.de | Affiliation: | Max-Planck Institute for Plasma Physics |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | Todd Evans, Charles Lasnier | ITPA Joint Experiment : | No |
| Description: | Test if it is possible to achieve ELM suppression/mitigation at high triangularity, low pedestal collisionality and H-98 factor above one. Study heat loads of mitigated ELMs with RMP and non-mitigated without RMP. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The aim is to explore possibility of achieving ELM suppression/mitigation with H-98 factor significantly above 1.
-Repeat discharges #139743 & #139745 to prove reproducibility of the scenario. -Repeat discharges #139743 & #139745 without RMP for reference. -Change plasma shape to more ITER-like with high triangularity and low pedestal collisionality (e.g. like #129197) -Change q95 near resonant conditions to find best performance -Repeat the same discharges without RMP to have a reference for ELM parameters and plasma performance. Data essential for analysis: - obtain the data from IR cameras - obtain high resolution edge Thomson data - obtain fast diamagnetic energy data to calculate ELM size Goals: 1) fill in some missing data from the previous experiment 2) obtain results similar to those in 139743 & 45 but with higher energy confinement and in a shape that is closer to that in ITER 3) get higher quality pedestal profile and divertor data using improved diagnostic capabilities | ||
| Background: | As shown on DIII-D ELMs can be either completely eliminated or mitigated with RMP fields. An RMP scenario with only very small ELMs and very good confinement has been achieved in last campaign (#139743 & #139745). As the plasma shape was adjusted to be compatible with infrared view only low plasma triangularity and pedestal collisionality close to 1 have been possible. Nevertheless results were extremely interesting as complete ELM suppression or mitigation has been achieved with very good confinement (H98 ~ 1.2). This experiment would concentrate on obtaining similar performance at conditions closer to ITER plasmas. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | visible divertor cameras, divertor IR cameras, divertor (ISP) Langmuir probes, ECE, ECE-I, UCLA fast CCD for tangential HFS observation, fast diamagnetic energy measurement | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 47: Impact of Turbulent Transport on Off-axis Neutral Beam Current Drive | ||
| Name: | Pace pacedc@fusion.gat.com | Affiliation: | GA |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Requested |
| Co-Author(s): | W.W. Heidbrink, C. Holland, Z. Lin, G.M. Staebler, R.E. Waltz, A.E. White, and the Energetic Particle Working Group | ITPA Joint Experiment : | Yes |
| Description: | The goal of this proposed experiment is to obtain conclusive proof that the long-wavelength component of plasma turbulence such as ITG and TEM contributes significantly to the transport of beam ions and therefore reduces the effectiveness of off-axis neutral beam current drive. To achieve this goal it will be necessary to incorporate DIII-D's array of turbulence diagnostics and theory/simulation support into the experiment. Predictions for the level of beam ion transport due to turbulence will be made using TGYRO/TGLF and tested in a dedicated experiment employing DIII-D's off-axis neutral beam and extensive energetic ion diagnostic set. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment involves three unique research issues: core turbulence, energetic ion transport, and off-axis neutral beam injection. We will address various turbulence issues prior to the experiment by choosing discharges for which the role of turbulence is fairly well understood with respect to the thermal plasma. A series of parameter scans will be performed with TGYRO/TGLF to determine the effect of triangularity and T_e/T_i on the turbulent diffusivity of energetic ions. The dedicated experiment will simultaneously measure the turbulence and energetic ion behavior in this variety of plasmas. Emphasis will be placed on the discharges for which the largest predicted differences in energetic ion turbulent transport have been identified.
Since the off-axis neutral beam (OANB) will begin operation during the 2011 experimental campaign, it is understandable that no discharge database exists to review its impact on otherwise typical DIII-D plasmas. For this reason, it is suggested that two test discharges be obtained during commissioning of the OANB well before the scheduled experiment. One L-mode (candidate is 128913) and one H-mode (to be determined) discharge will be reproduced, but with the addition of (or replacement of another beam by) the OANB. These discharges require only basic tokamak and primary diagnostic operation as the measured plasma profiles will serve as the base case about which the parameter scans will be performed. | ||
| Background: | In experiments employing OANB injection on ASDEX-U [1] and DIII-D [2], the observed neutral beam current drive was below neoclassical values. Theoretical and simulation results [3-5] argue that long-wavelength microturbulence, such as ion temperature gradient and trapped electron modes, contribute to energetic ion transport in these regimes. The transport due to turbulence is predicted to depend on the ratio of the thermal ion temperature to the energetic ion energy, T_i/E. Higher energy beam ions feature larger gyro-orbits and average over turbulent structures more effectively than lower energy ions, an effect that has been observed experimentally in a basic plasma device [6]. Experimental validation of this effect in a tokamak plasma is of great importance to ITER due to its potential impact on OANB current drive.
[1] S. Guenter, et al., Nucl. Fusion 47, 920 (2007). [2] W.W. Heidbrink, et al., Plasma Phys. Control. Fusion 51, 125001 (2009). [3] C. Estrada-Mila, et al., Phys. Plasmas 13, 112303 (2006). [4] W. Zhang, et al., Phys. Rev. Lett. 101, 095001 (2008). [5] T. Hauff, et al., Phys. Rev. Lett. 102, 075004 (2009). [6] S. Zhou, et al., Phys. Plasmas 17, 092103 (2010). | ||
| Resource Requirements: | - availability of all neutral beams
- off-axis neutral beam at maximum injection angle | ||
| Diagnostic Requirements: | - all FIDA systems (measure the fast ion distribution)
- MSE (contribute to the determination of off-axis current drive) - turbulence diagnostic suite, including BES and DBS (characterize the turbulence across chosen parameter scan) - CER (provide T_i and rotation measurements for turbulence simulations) | ||
| Analysis Requirements: | - determination of off-axis NB driven current
- preparation of measured plasma profiles for use as input to gyrokinetic turbulence simulation codes such as GYRO and GTC | ||
| Other Requirements: | -- | ||
| Title | 48: Excitation of Alfven Eigenmodes by Fast Wave Beat Waves | ||
| Name: | Pace pacedc@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particle | Presentation time: | Requested |
| Co-Author(s): | T.A. Carter, W.W. Heidbrink, R.I. Pinsker, M.A. Van Zeeland, and the Energetic Particle Working Group | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to develop an operational scheme by which the fast wave system excites Alfven eigenmodes through a beat wave process. The successful realization of this goal will enable the study of Alfven eigenmodes, and their impact on energetic ion confinement, in DIII-D plasmas that are otherwise incapable of producing these modes (e.g., current flattop at normal magnetic field amplitude). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The fast wave 285/300 antenna will operate with an input waveform composed of two frequencies:
A) 60.0 MHz B) 60.0 - f_IF [MHz] where f_IF is the intermediate frequency desired (i.e., an Alfven eigenmode frequency in the lab frame). The value of f_IF will be repeatedly scanned between 20 and 200 kHz during the entire discharge in order to identify any Alfven eigenmode drive as the resonance condition is reached. The initial attempts to drive Alfven eigenmodes (AEs) using beat wave excitation from fast wave injection will occur in plasmas of general interest to the Energetic Particles Working Group. Neutral beam injection during the earliest stages of the current ramp produce significant Alfvenic activity in typical DIII-D shapes (122117) as well as in oval plasmas (142111). Fast wave injection of beat waves in these plasmas will be used to increase the amplitudes of the already existing AEs. A successful result in this situation serves as a proof-of-principle for this process. Other discharges will focus on exciting AEs in plasmas for which they would not otherwise be observed, but in which they are thought to be only marginally stable or weakly damped. A series of discharges that feature properties conducive to Alfven eigenmode production will be identified, including reduced magnetic fields (B_T ~ 1.0 T) in which the neutral beam ions are super-Alfvenic. Should observations indicate that the fast wave system is driving AEs in a variety of situations described above, then the experiment will transition to discharges for which the modes would never be observed (standard magnetic field, delayed neutral beam injection, etc.). | ||
| Background: | Energetic ion transport due to Alfven eigenmodes is suspected as a mechanism that may reduce the fusion power of ITER [1]. Recent work at DIII-D has quantified the transport of beam ions [2,3] in experiments that utilize early neutral beam injection during the current ramp to excite AEs. A natural progression of this work requires the ability to study wave-particle interactions in steady state plasmas that better approximate reactor conditions in which super-Alfvenic fusion alphas drive the modes. This also enables DIII-D to study fundamental features of AEs, including damping rates that are investigated elsewhere through the use of active MHD antenna systems [4]. An advantage of the fast wave excitation method compared to active MHD antennas is that it is theoretically capable of driving large amplitude and core localized modes.
Excitation of toroidicity induced AEs through beat wave interactions from ICRH systems has previously been achieved on JET [5] and more recently in AUG [6]. In the AUG work it was possible to measure the radial displacement eigenfunction of the resulting mode, which was then compared to theoretical predictions, enabling the identification of the mode and confirmation of the ICRH drive. At DIII-D, the excitation of these modes in discharges for which they were previously unavailable will enable the study of many new physics phenomena of relevance to ITER, including AE induced transport and losses of off-axis injected neutral beam ions and the resulting degradation of off-axis neutral beam current drive. [1] A. Fasoli, et al., Nucl. Fusion 47, S264 (2007) [2] M.A. Van Zeeland, et al., Phys. Plasmas, (APS 2010 invited talk, in submission) [3] W.W. Heidbrink, et al., Phys. Rev. Lett. 99, 245002 (2007) [4] A. Fasoli, et al., Plasma Phys. Control. Fusion 52, 075015 (2010) [5] A. Fasoli, et al., Nucl. Fusion 36, 258 (1996) [6] K. Sassenberg, et al., Nucl. Fusion 50, 052003 (2010) | ||
| Resource Requirements: | - fast-wave system at full power, with the 285/300 system enabled for beat-wave input
- all neutral beams available: initially used to provide early AE drive during current ramp, also used to interact with fast-wave driven modes to provide additional measurement capability | ||
| Diagnostic Requirements: | - all FILD units (midplane and R-1): measurement of any fast ion losses at frequencies coherent with the driven modes
- ECE and ECEI: detection of AEs - all FIDA systems: measurements of fast ion distribution to identify any adjustments due to the presence of AE activity | ||
| Analysis Requirements: | - possibility of using full wave solver to assess propagation of injected fast-waves: this would involve additional theory and modeling collaborators | ||
| Other Requirements: | |||
| Title | 49: Thermo-Oxidation as a Method of Reducing Dust in DIII-D | ||
| Name: | Davis jwdavis@starfire.utias.utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | ITER First Wall Issues | Presentation time: | Requested |
| Co-Author(s): | TBD | ITPA Joint Experiment : | Yes |
| Description: | The oxygen bake of DIII-D in April 2010 successfully demonstrated that the release of hydrogen from carbon-based codeposits was similar to that expected from laboratory measurements of DIII-D tiles.
Two observations made following these experiments warrant further investigation: 1) the dust level in the plasma was reduced by about a factor of two, as compared to similar clean vents; 2) there was a reduction of a factor of two in the carbon concentrations in the plasma. The second point is quite surprising given the increased oxygen concentrations. It has been observed in the past that increasing oxygen impurity levels also increases carbon concentrations due to chemical sputtering (CO formation). If these two points are confirmed, it would show that carbon dust is a significant contributor to the carbon plasma impurity, and oxygen baking may be a method to reduce plasma impurity levels. Even if carbon is not used in the DT phase of ITER, thermo-oxidation has relevance due to the use of carbon in the ITER divertor during the previous phases. It will be necessary to control carbon-based dust during the planned ITER operation with graphite at the strike points. Also, carbon deposits will need to be removed from the vessel prior to DT operation. Thermo-oxidation may prove to be one of the most effective techniques to achieve both requirements. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Proposal: Perform a thermo-oxidation experiment in DIII-D to test for dust reduction. Measurements of dust concentrations would be made in reference discharges during the campaign, prior to the O-bake. An O-bake would then be performed, followed by a return to normal plasma operation. Both dust levels and carbon impurity levels would be directly compared before and after the oxidation for reference discharges. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 50: q95 Dependence of RMP ELM Suppression | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | Test one of the hypotheses of the physics behind the q95 dependence of RMP ELM suppression, as opposed to only density pump-out or ELM mitigation, by carefully measuring net electron poloidal flow velocity profiles during q95 scans in shots with n=3 even parity I-coil fields at low collisionality typical of ELM suppression. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Setup typical LSN, ISS shape, low collisionality plasma typical of previous successful n=3 RMP ELM suppression. Use even parity I-coil and verify ELM suppression and q95 resonance window around q95=3.5 using shots with q95 down sweeps (Ip up sweeps). Then systematically run separate discharges at fixed q95 just above (mitigated ELMs) and just below (ELMs suppressed) the upper edge of the q95 resonant window for ELM suppression. For each case add all known techniques necessary to get the highest quality Thomson and CER profiles possible (radial jogs, Thomson breathing and others) so that the most accurate possible profiles of net poloidal electron flow velocity (v_e_pol = v_e_(ExB) + v_e_Gradp) in the pedestal can be generated. Look for correlation of ELM suppression with the location of the region of zero net poloidal electron flow lining up with rationale surfaces near the top of the pedestal having large resonant components from the applied n=3 RMP fields. If time allows, repeat for low q95 boundary of resonant window for ELM suppression. If more time, repeat with BT ramps/variation for q95 variation. If more time, repeat with odd parity n=3 fields, Ip up sweeps and q95 resonant window about q95=7. | ||
| Background: | One of the hypotheses of RMP ELM suppression (Snyder RMP brainstorming https://diii-d.gat.com/diii-d/Click_here_to_access_PDF_presentation_files) is that some mode must form near the top of the pedestal in such a way that the pedestal width and/or height can not continue to evolve toward the instability boundary. For rotating H-mode plasmas some theories and code simulations (eg. Becoulet PEP ITPA April 2010 Naka and Oct 2010 Seoul) predict that the RMP fields will be screened by the plasma response everywhere except at the very edge of the plasma and at locations further in for which the net electron poloidal flow velocity goes to zero. This experiment attempts to look for an experimental correlation between the suppression of ELMs (as distinct from particle pump-out due to the RMP, or mitigation of ELMs size) and the alignment of the zero crossing of the net poloidal electron flow with strong harmonics of the applied RMP spectrum at the top of an ELM-stable pedestal. Because the net poloidal electron flow is the difference of two large quantities (poloidal Grad-p and ExB flows) each with experimental uncertainties, the experiment requires very detailed methodical effort to produce the best Thomson and CER profile data possible on DIII-D. | ||
| Resource Requirements: | Same resources as in previous reference RMP ELM suppression shots with q95 sweeps showing q95 resonant window for ELM suppression, eg. 136221 (Ip ramp) or 136240 (BT ramp) | ||
| Diagnostic Requirements: | All pedestal profile and turbulence diagnostics, in particular the best possible Thomson and CER profiles in the pedestal with best channel-to-channel calibrations, jogs and breathing techniques etc. | ||
| Analysis Requirements: | If possible, control room analysis of pedestal profiles by Osborne python analysis tools to generate profile of net electron poloidal flow velocity. Will eventually require full kinetic EFIT analysis and CERFIT/E-radial analysis for accurate v_e_pol | ||
| Other Requirements: | Data mining of existing RMP ELM suppression shots with q95 variation to determine the quality of present TS and CER profiles for calculation v_e_pol profile and whether any correlation exists between zero crossing at the top of the pedestal and ELM suppression. | ||
| Title | 51: Role of ECRF on toroidal rotation profile | ||
| Name: | Yoshida none | Affiliation: | JAEA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, J. deGrassie, P. Gohil | ITPA Joint Experiment : | No |
| Description: | Intrinsic rotation and momentum transport coefficients (diffusivity and pinch) with ECRF will be studied, in order to understand the physical mechanisms determining the toroidal rotation velocity profile with ECRF. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A scan of ECH power deposition profile (for example r/a~0.3, 0.45, 0.6) will be carried out in the parameter regime similar to that in JT-60U (especially collsionality, eta).
Momentum diffusivity and pinch velocity will be measured using NB modulation or magnetic perturbative experiments in both EC injected and non-injected plasmas. Effects of the change in momentum transport and the change in the intrinsic rotation will be separately evaluated through a scan of the toroidal rotation velocity (CO, BAL, CTR-rotation) in target plasmas. We will focus on L-mode plasmas in order to minimize other effects such as ELMs and pressure gradient-driven intrinsic rotation. | ||
| Background: | In DIII-D, the correlation of the intrinsic rotation profile with the ECH power deposition profile was investigated in H-mode plasmas [deGrassie, PoP2007]. Although the intrinsic rotation profile varies with the ECH power deposition profile, the toroidal rotation is almost in the co-direction both on-axis and off-axis deposition.
On the other hand, the direction of the EC-driven intrinsic rotation changes at the EC deposition radius in JT-60U L- and H-mode plasmas. ECRF drives the CO-intrinsic rotation inside the EC deposition and drives the CTR-intrinsic rotation outside the EC deposition. One difference between the condition in the two devices is heating system for target plasmas; ECH or NBI. Thus, the direction of the intrinsic rotation might correlate with the parameter regime such as density, temperature and so on. Also theoretical model predicts the direction of the intrinsic rotation changes by changing collisionality or eta. Measurement of the momentum diffusivity and pinch velocity is needed to separately evaluate the diffusive term, pinch term and residual term in the momentum flux. | ||
| Resource Requirements: | 1 day experiment. CO & Counter NBs, ECH (~3 MW, ~3 sec) | ||
| Diagnostic Requirements: | Standard diagnostics, especially fast CER for toroidal and poloidal rotation and ion temperature, Thomson for electron density and temperature, ECE for sawtooth and electron temperature, high k FIR scattering, BES. | ||
| Analysis Requirements: | Transient momentum transport analysis using modulated NBs or magnetic perturbations | ||
| Other Requirements: | |||
| Title | 52: Role of ECRF on toroidal rotation profile | ||
| Name: | Yoshida none | Affiliation: | JAEA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, J. deGrassie, P. Gohil | ITPA Joint Experiment : | No |
| Description: | ntrinsic rotation and momentum transport coefficients (diffusivity and pinch) with ECRF will be studied, in order to understand the physical mechanisms determining the toroidal rotation velocity profile with ECRF. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A scan of ECH power deposition profile (for example r/a~0.3, 0.45, 0.6) will be carried out in the parameter regime similar to that in JT-60U (especially collsionality, eta).
Momentum diffusivity and pinch velocity will be measured using NB modulation or magnetic perturbative experiments in both EC injected and non-injected plasmas. Effects of the change in momentum transport and the change in the intrinsic rotation will be separately evaluated through a scan of the toroidal rotation velocity (CO, BAL, CTR-rotation) in target plasmas. We will focus on L-mode plasmas in order to minimize other effects such as ELMs and pressure gradient-driven intrinsic rotation. | ||
| Background: | In DIII-D, the correlation of the intrinsic rotation profile with the ECH power deposition profile was investigated in H-mode plasmas [deGrassie, PoP2007]. Although the intrinsic rotation profile varies with the ECH power deposition profile, the toroidal rotation is almost in the co-direction both on-axis and off-axis deposition.
On the other hand, the direction of the EC-driven intrinsic rotation changes at the EC deposition radius in JT-60U L- and H-mode plasmas. ECRF drives the CO-intrinsic rotation inside the EC deposition and drives the CTR-intrinsic rotation outside the EC deposition. One difference between the condition in the two devices is heating system for target plasmas; ECH or NBI. Thus, the direction of the intrinsic rotation might correlate with the parameter regime such as density, temperature and so on. Also theoretical model predicts the direction of the intrinsic rotation changes by changing collisionality or eta. Measurement of the momentum diffusivity and pinch velocity is needed to separately evaluate the diffusive term, pinch term and residual term in the momentum flux. | ||
| Resource Requirements: | 1 day experiment. CO & Counter NBs, ECH (~3 MW, ~3 sec) | ||
| Diagnostic Requirements: | Standard diagnostics, especially fast CER for toroidal and poloidal rotation and ion temperature, Thomson for electron density and temperature, ECE for sawtooth and electron temperature, high k FIR scattering, BES | ||
| Analysis Requirements: | Transient momentum transport analysis using modulated NBs or magnetic perturbations | ||
| Other Requirements: | |||
| Title | 53: RMP ELM suppression in dRsep=0 DN Plasma | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Revisit attempt to obtain RMP ELM suppression in a dRsep=0 true symmetric DN shape for applicability of many existing analysis codes from the stellarator community that work best in up/down symmetry. Also part of ITPA PEP-23 joint experiment with MAST CDN plasmas. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Re-establish setup from previous attempts at RMP ELM suppression in dRsep=0 DN shape (April 1, 2010, shots 142582-607) and reproduce results of one of the shots without I-coil (142585) and with large effect (excessive pump-out) due to the I-coil (eg. current steps in 142587 or 589 etc.). Then try several techniques to minimize the excessive density pump-out and allow larger I-coil currents to try for ELM suppression, vis: 1) reduce pumping by increasing gap to both outer pumps, 2) dRsep scan from LSN to DN to determine dRsep range giving excessive particle pump-out, 3) in dRsep=0 DN warm upper outer pump, 4) also warm upper inner pump. Increase I-coil current as much as possible consistent with pump-out not being large enough to lock the plasma. Vary q95 to check for ELM suppression resonance. If ELM suppression still not obtained vary betaN. | ||
| Background: | In 2010 there was one day of dedicated attempts to obtain low collisionality RMP ELM suppression in a pure DN (dRsep=0) plasma. ELM suppression was not obtained although there was a strong effect of the RMP application on the plasma. The primary difficulty was that the RMP produced very large density pump-out, even for I-coil currents much lower than typically needed for ELM suppression in LSN ISS plasmas. One hypothesis for the large pump-out is that the upper outer pump was significantly more effective in these experiments than the lower pump is in typical LSN RMP ELM suppressed plasmas. This experiment focuses on techniques to vary (reduce) the particle pump-out in dRsep=0 shapes and allow larger RMP magnetic perturbation strength to be applied to DN plasmas to attempt ELM suppression. | ||
| Resource Requirements: | Same resources as in previous dRsep=0 DN RMP ELM suppression experiments (eg. 142589) including best possible dRsep control near DN. | ||
| Diagnostic Requirements: | All pedestal profile and turbulence diagnostics, upper and lower divertor diagnostics (filterscopes, visible TVs, IRTVs) | ||
| Analysis Requirements: | Pumping speed of various cryopumps and overall particle balance to determine contributions to density pump-out during RMP | ||
| Other Requirements: | |||
| Title | 54: Dependence of momentum transport on Te/Ti | ||
| Name: | Yoshida none | Affiliation: | JAEA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, P. Gohil | ITPA Joint Experiment : | No |
| Description: | Dependence of momentum transport diffusivity and momentum pinch velocity on the ratio of the electron temperature to ion temperature (Te/Ti) with low or no external torque input will be investigated in order to increase an understanding of the momentum transport in ITER.
DIII-D has advantages in studying momentum transport in ITER-like plasmas: Te/Ti>1, low or no external torque input using CO & Counter NBs and EC, and ITER-like plasma shape. In JT-60U momentum transport diffusivity and pinch velocity were measured at Te/Ti<1. By comparing the measurements in DIII-D to those in JT-60U, properties of momentum transport in the plasmas dominated by electron heating and ion heating will be clarified. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Momentum transport diffusivity and momentum pinch velocity will be evaluated using NB modulation or magnetic perturbative experiments.
Dependency of the momentum transport diffusivity and momentum pinch velocity on Te/Ti with low or no external torque input will be investigated through a scan of the power ratio of BAL-NB to EC at a totally constant absorbed power. Relation between the changes of momentum transport and of turbulence properties will be investigated using turbulence measurements including long and short wavelength. The target plasma will be a type-I ELMy H-mode with ITER-like plasma shape. | ||
| Background: | In the momentum database for ITPA Transport and Confinement group, we have recently found the ratio of momentum diffusivity to ion heat diffusivity tends to be smaller with a large Te/Ti, and the ratio of pinch velocity to momentum diffusivity tends to increase with decreasing Te/Ti.
However these trends depend on JT-60U and JET data. In addition, data at Te/Ti<1 dominate the database. | ||
| Resource Requirements: | 1 day experiment. CO & Counter NBs, ECH (~3 MW, ~3 sec) | ||
| Diagnostic Requirements: | Standard diagnostics, especially fast CER for toroidal and poloidal rotation and ion temperature, Thomson for electron density and temperature, ECE for sawtooth and electron temperature, high k FIR scattering, BES | ||
| Analysis Requirements: | Transient momentum transport analysis using modulated NBs or magnetic perturbations | ||
| Other Requirements: | |||
| Title | 55: Dependence of Pedestal Turbulence on RMP Strength During ELM Suppression | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Try to determine if magnetic perturbation due to the I-coils directly affects pedestal turbulence (and possibly thereby turbulent particle flux) by repeating RMP ELM suppression experiment with I-coil current modulated (eg. 4kA to 2 kA) at constant beam power, and measuring all changes to pedestal turbulence characteristics (BES, profile reflectometer, DBS, ECE?, CECE?, others) with maximum time resolution. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Re-establish conditions used in very successful experiments measuring effect of RMP strength (I-coil current) on density turbulence (from BES) from 2010 (eg. shot 142250).
Repeat with absolutely constant beam power to remove difficulty of small beam modulations aligned with I-coil modulations in 2010 experiments. Do detailed scan of turbulence diagnostics across pedestal and far core regions (0.85 < PsiN < 1.0) to measure radial dependence of response of turbulence to magnetic perturbation strength. | ||
| Background: | Very interesting experiments from 2010 indicated rapid changes of density fluctuations from BES on the timescale of changes in the magnetic perturbation fields during ELM suppression with modulated I-coil current. The timescale of the fluctuation response to the I-coil current change increased at positions further into the core plasma. However data were not obtained throughout the pedestal out to the very edge of the plasma. Also, small beam power modulations (700 kW out of 6 MW) in phase with the I-coil modulation complicated the analysis. Extending this experiment with constant beam power and greater spatial coverage of multiple turbulence diagnostics in the edge should contribute to understanding the effect of 3D fields on turbulent particle transport in the pedestal during ELM suppression. | ||
| Resource Requirements: | Same resources as in previous successful RMP ELM suppressed shots with modulated I-coil current , eg. 142250. | ||
| Diagnostic Requirements: | All pedestal and turbulence diagnostics, especially BES, profile reflectometer, DBS etc. | ||
| Analysis Requirements: | Control room analysis of turbulence spectra changes vs. I-coil current for different locations in core, pedestal and very edge. | ||
| Other Requirements: | |||
| Title | 56: Zero Input Torque QH-mode Without Internal Coils | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Attempt to retain QH-mode at near zero input NB torque using NTV torque from predominantly non-resonant magnetic fields generated from coils external to the vacuum vessel alone. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce successful sustainment of QH-mode at zero input NB torque using NTV torque from NRMF generated by combination of I-coil and C-coil n=3 fields (shot 141439). Then try to reduce I-coil contribution while still maintaining QH-mode. For each combination of NRMFs perform input torque scan to determine minimum required input torque to maintain QH-mode. | ||
| Background: | Experiments in 2010 showed that QH-mode could be maintained as the input NB torque was reduced to zero by application of NTV torque from non-resonant magnetic perturbation fields generated by a combination of the I-coil and C-coil. This experiment attempts to map out the operating space of low input torque QH-mode as a function of the mix of NRMFs from coils internal to the vacuum vessel vs NRMFs from coils external to the vessel. This is an important question for future applications of QH-mode at low input torque such as in ITER since the engineering of coils external to the vessel could be considerably easier than for internal coils. | ||
| Resource Requirements: | Same resources as in previous successful experiment, eg. 141439 including both co-and counter-NBI and reversed Ip. Will want maximum capability to vary spectrum of magnetic perturbations from the C-Coil to test effect of different NRMF spectra on NTV torque and QH-mode. | ||
| Diagnostic Requirements: | All pedestal and turbulence diagnostics, magnetics for EHO. | ||
| Analysis Requirements: | Control room calculation of input NBI torque, pre-experiment calculations of anticipated NTV torque from various combinations of I-coil and C-coil configurations. | ||
| Other Requirements: | |||
| Title | 57: Higher Freq Pellet ELM Pacing Physics | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate higher frequency pacing of ELMs by triggering pellets than obtained with 2010 pellet hardware. Determine difference in ELM pacing performance with pellets launched from various poloidal locations. Understand the physics and scaling of pellet triggered and additional ELMs (if any) between pacing pellets. Measure and understand the scaling of the peak divertor target heat flux vs injection frequency of the pacing pellets. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Re-establish the low ELM frequency ITER DEMO discharge used in successful pellet ELM pacing experiments in 2010 (vis. shot 141132). Inject high frequency pacing pellets with half the mass of 2010 pellets from LFS X-point region using new injector hardware. Measure ELM frequency and divertor peak heat flux during ELMs as function of pellet injection frequency. Measure all necessary pedestal quantities to understand triggering physics of ELMs at pellet times and any ELMs triggered between pellets. Determine differences in ELM pacing response to pellets injected from multiple poloidal locations. | ||
| Background: | Experiments in 2010 with 14 Hz pellet injection into a low natural ELM frequency (5 Hz) ITER DEMO discharge showed triggering of up to 25 HZ ELMs. ELM amplitude and divertor target peak heat flux were reduced during the paced ELMs compared with the natural ELMs, although by factors somewhat smaller than the increase in ELM frequency. ELM pacing by pellets is the primary technique for ELM Control in the ITER baseline. New hardware on DIII-D for 2011 will allow approximately 2x increase in pellet injection frequency (to 30 Hz) for ELM pacing, and new poloidal launch locations including the LFS near the X-point similar to the planned launch in ITER. This experiment focuses both on the empirical goal of establishing higher frequency pellet ELM pacing on DIII-D and also on understanding the physics of both ELMs triggered at the time of pellet injection, and also any additional ELMs triggered between pellets. | ||
| Resource Requirements: | Same resources as used in previous successful pellet ELM pacing experiments from 2010 (vis shot 141132-33) plus new higher frequency, smaller pellet injector systems. Also all fast IRTV systems to measure peak heat flux during ELMs. | ||
| Diagnostic Requirements: | All pedestal and edge fluctuation diagnostics with fast time resolution, fast midplane imaging systems for pellet measurements. | ||
| Analysis Requirements: | Control room analysis of pellet trajectories and precision timing of pellet injection times and ELM times (filterscopes), analysis of peak divertor heat flux during ELMs from IRTV. | ||
| Other Requirements: | |||
| Title | 58: Retracted | ||
| Name: | James jamesan@fusion.gat.com | Affiliation: | |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | -- | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 59: Exploration and Characterization of I-mode | ||
| Name: | Whyte none | Affiliation: | MIT |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | D. Whyte, A. Hubbard, J. Hughes, P. Gohil, J. Rice, B. Lipschultz, E. Marmar, M. Greenwald | ITPA Joint Experiment : | Yes |
| Description: | Search for the stationary "I-mode" transport regime on DIII-D to determine parameters necessary access to I-mode, avoid H-mode and characterize the pedestal conditions. These results will be compared with the C-Mod I-mode results [Whyte, et al. Nucl Fusion (2010)]. I-mode is possibly a highly favorable regime: it features a stationary edge temperature barrier without an edge particle barrier. This leads to stationary discharges with high T pedestal, H-mode energy confinement, and no core impurity accumulation (L-mode particle confinement) without the requirement for ELMs. Almost all I-mode discharge on C-Mod are stationary and ELM-free.
In addition to its attractiveness as an ELM-free regime, I-mode has important scientific contributions to make in transport studies since the regime clearly separates the energy and particle/impurity channels from one another. C-Mod reports correlations between I-mode and changes in edge fluctuation changes including a weakly coherent EM mode existing in the pedestal region. It is speculated that this mode is responsible for enhanced particle transport in the edge. Detection of this mode and comparison to C-Mod results, as well as the EHO, is desirable. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental approach is to use the operational and physics insights gained from C-Mod. The I-mode regime is most readily observed with unfavorable grad-B drift topology and medium to high triangularity. As q95 is decreased towards ~3 sudden transitions are found from L to I-mode, with a rapid increase of edge T (within < sawtooth pulse). In addition the required power to access I-mode in this condition increases with Ip (contrary to the LH threshold in favorable grad-B). This tends to lead to high absolute performance at ITER's q95~3 but without ELMs: a very favorable result for ITER.
We will use the identification parameters as defined by C-Mod 1) Formation of a T pedestal without a n pedestal 2) Observation of reduced broadband edge fluctuations and/or presence of a high frequency (>100 kHz) weakly coherent EM mode 3) No standard H-mode transition signatures of D-alpha drop and/or increasing density Based on the C-Mod experience we will explore for I-mode by scanning within a shot the heating power in small steps, and then change other global parameters of interest on a shot to shot basis namely: Ip, Bt (or q95) and density. A clear result from C-Mod was that the I and H power threshold has different dependences in Ip, Bt (or q95) than favorable direction. Namely the power threshold clearly increases with Ip. Also the threshold increased with lower B, the opposite trend of standard LH scaling. Shaping also played a role in avoiding H-mode access, e.g. x-point spacing, triangularity. In a way this is not news for D3D which used a high delta, upper null shape to avoid H-mode transitions in high fusion gain experiments of 90's (Lazarus et al) Another key feature in accessing I-mode is density. C-Mod found optimal density windows using cryopumping. There is an intriguing possibility that the heating gap between I and H could be made much larger by going below the "low-density" LH threshold. This has not been possible yet on C-Mod due to L-mode impurity limitations with ICRF. However it may be possible to explore this limit with NBI heating. Experimental plan: unfavorable grad-B (e.g. USN with forward BT) (starting) Plasma shape as from C-Mod In each shot perform a stepped power scan near projected L-I (or L-H) threshold shot to shot scans (in approx. order of priority): - density (4-5 values) - Ip (to scan q95 ~3, 3.5, 4) - shape (X-point clearance) - Bt (2 values) | ||
| Background: | Excerpts from abstract of C-Mod paper on I-mode (Whyte, et al. Nucl. Fusion 2010)
"An improved energy confinement regime, I-mode is studied in Alcator C-Mod, a compact high-field divertor tokamak using Ion Cyclotron Range of Frequencies (ICRF) auxiliary heating. I-mode features an edge energy transport barrier without an accompanying particle barrier, leading to several performance benefits. H-mode energy confinement is obtained without core impurity accumulation, resulting in reduced impurity radiation with a high-Z metal wall and ICRF heating. I-mode has a stationary temperature pedestal with Edge Localized Modes (ELMs) typically absent, while plasma density is controlled using divertor cryopumping. I-mode is a confinement regime that appears distinct from both L-mode and H-mode, combining the most favorable elements of both. The I-mode regime is investigated predominately with ion ā??B drift away from the active X-point. The transition from L-mode to I-mode is primarily identified by the formation of a high temperature edge pedestal, while the edge density profile remains nearly identical to L-mode. Laser blowoff injection shows that I-mode core impurity confinement times are nearly identical with those in L-mode, despite the enhanced energy confinement. In addition a weakly coherent edge MHD mode is apparent at high frequency ~ 100-300 kHz which appears to increase particle transport in the edge. The I-mode regime has been obtained over a wide parameter space (BT=3-6 T, Ip=0.7-1.3 MA, q95=2.5-5). In general the I-mode exhibits the strongest edge T pedestal and normalized energy confinement (H98>1) at low q95 (<3.5) and high heating power (Pheat > 4 MW). I-mode significantly expands the operational space of ELM-free, stationary pedestals in C-Mod to Tped~1 keV and low collisionality Ī½*ped~0.1, as compared to EDA H-mode with Tped< 0.6 keV, Ī½*ped>1. " "Previously, an ā??improved L-modeā?? regime was found and briefly described in ASDEX-Upgrade [Ryter], which focused on the mode as an intermediate step in the transition to H-mode in the unfavorable ā??B configuration....Threshold studies on these tokamaks, as well as DIII-D, showed clearly that edge temperatures, T gradients and required heating power are substantially higher at the L-H transition in the unfavorable vs. favorable ā??B configuration [11-14]. " So there is a good chance that plasmas that are "I-mode-like" have been previously obtained on DIII-D. It seems reasonable to explore this further. Based on the Cmod results, the "new" parts of these studies for DIII-D is that a) The T pedestal can have a sudden, clear bifurcation from L-mode rather than just being a continuous slow modification from L-mode profiles b) The confinement mode can be made stationary, rather than intermediate of L to H. c) The sudden changes in edge fluctuations associated with the T pedestal without a density pedestal have been identified. | ||
| Resource Requirements: | NBI heating
Cryopumping Unfavorable grad-B drift topology | ||
| Diagnostic Requirements: | Full suite of pedestal diagnostics
Full suite of edge fluctuation diagnostics - BES - reflectometry - magnetics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 60: PCS control of runaway electrons using the Yoshida equilibrium | ||
| Name: | James jamesan@fusion.gat.com | Affiliation: | |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Requested |
| Co-Author(s): | P. Parks, D. Humphreys, N. Eidietis | ITPA Joint Experiment : | No |
| Description: | Control of runaway electron plateaus is a topic of recent interest since these plateaus may be inevitable in future machines due to the continuous presence of high energy electrons from inverse compton scattering. Similar plateaus can be generated in DIII-D, but the PCS presently uses the Grad-Shafranov equilibrium for feedback which is not a valid model during a runaway plateau.
The Yoshida equilibrium replaces the pressure gradient term in the Grad-Shafranov equilibrium with a relativistic centrifugal pressure term, which should be more valid during runaway plateau since the actual plasma pressure is almost zero during the plateau, making any gradients necessarily small as well. By incorporating measurements of the average runaway electron energy using the recently developed bismuth germanate scintillator array into the PCS, this equilibrium can be used to more correctly feed back on the runaway plateau equilibrium, possibly enabling improved control. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Initiate a low elongation inner wall limited ECH heated discharge
2) Inject an argon killer-pellet to induce a rapid shutdown and generate a runaway electron plateau 3) Switch the PCS algorithm to a Yoshida equilibrium after the plateau has settled 4) Attempt to demonstrate radial control of the runaway plateau | ||
| Background: | In 1989 Zensho Yoshida published a paper on the toroidal equilibrium of plasma with a concentrated relativistic electron beam, similar to the situation after generation of a runaway electron plateau. Experiments last year focusing on runaway electron plateau control succeeded in establishing vertical control, but failed at radial control, the direction of the runaway electron inertial pressure term introduced in Yoshida's equilibrium. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | Scintillator array average energy measurement. Also since runaway confinement routinely lasts in excess of 100ms now, ECE michaelson interferometer scans can be used to investigate the spectrum of microwave emission from runaway current. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | Specialized programming of the PCS to switch to the Yoshida equilibrium during the runaway electron plateau phase. | ||
| Title | 61: Studying runaway electron generation and confinement using profile switches | ||
| Name: | James jamesan@fusion.gat.com | Affiliation: | |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys, V. Izzo, M. Kornbluth | ITPA Joint Experiment : | No |
| Description: | Using the subtle profile differences between diverted discharges which do and do not produce runaway plateaus as identified with GATO and NIMROD, we propose experiments using various profile actuators (ECCD, NBCD, pellet fueling, etc) to establish similar current and pressure profiles before a killer pellet shutdown.
By adjusting the profiles between these configurations, it should be possible to switch the subsequent runaway electron plateau on and off. The implication for future machines would be tremendous: essentially introducing an operating criterion under which runaway free shutdowns are possible, hence preventing the damage associated with them. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Use available profile actuators to establish a stable, well diagnosed plasma configuration matching one which previously generated runaways or not
2) Inject argon killer pellet to shutdown the plasma 3) Observe runaway generation and transport | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | MSE, Thompson scattering | ||
| Analysis Requirements: | Need to identify specific profile targets matched with actuators based on prior analysis. | ||
| Other Requirements: | -- | ||
| Title | 62: Dependence of ion transport stiffness on rotation | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Requested |
| Co-Author(s): | T.L. Rhodes. E.J. Doyle, J. deBoo, G.R. McKee, J. Hillesheim, W. Solomon, L. Zeng,
W.A. Peebles |
ITPA Joint Experiment : | No |
| Description: | Knowledge of the critical ITG gradient and the degree of ion transport stiffness is crucial for predicting performance in future burning plasmas. The degree of transport stiffness directly determines the achievable temperature profile, and the response of the radial temperature profile to changes in power deposition. In this experiment we propose to vary toroidal rotation and rotational shear independently to investigate their effects on the critical ITG gradient and ion transport stiffness. Low rotation H-mode plasmas are chosen as they are of particular interest for extrapolating to burning plasma scenarios. One goal of the experiment is to extrapolate the critical gradient from the dependence of normalized ion heat flux on temperature gradient (at fixed radius). A comparison to linear (TGLF) and nonlinear (GYRO) code predictions can then be made, in particular with the predicted nonlinear upshift of the critical gradient (Dimits shift). Core fluctuation measurements with BES, DBS will allow us to correlate density fluctuation levels with measured (via transport analysis) and predicted (via nonlinear Gyro simulations) diffusivities. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose to use ECH H-mode target plasmas and vary core plasma ion power deposition by injecting neutral beam power at constant beam torque. ECH H-modes have lower core plasma rotation and rotational shear compared to NBI H-modes. As ECH power, power deposition radius, and neutral beam torque can be varied independently, we expect that toroidal rotation and rotational shear can be varied independently. For constant ECH power deposition profile, the ion heating power is varied by stepping the beam power at constant beam torque to measure profile stiffness. The power scan is repeated for three different values of net beam torque. During this shot sequence, a transition transition to lower ECH power can be made later in the shot (duration ~ 1s) to modify ExB core shear. The dependence of normalized ion heat flux (or Gyro-Bohm normalized diffusivity) on the ion temperature gradient (at a given radius) is extracted using transport analysis. This analysis will focus on radii r/a < 0.6 where electron-ion coupling is weak. The stiffness is given by the slope, and the critical gradient length is then determined by extrapolating to zero ion heating power. The goal of this analysis is to determine the dependence of the critical gradient and the stiffness on toroidal rotation and ExB shear. | ||
| Background: | Earlier experiments on profile stiffness have mostly focused on electron heating (for example Ryter et al, PRL 2005). The ion temperature critical gradient in DIII-D has been investigated (Baker et al, PoP2003) but no direct local measurement of profile stiffness has been made. Recent experiments on JET have suggested a rotation dependence of stiffness but toroidal rotation and rotational shear have not been varied independently. This experiment builds on recent observations that the core ExB shear for given NBI torque depends sensitively on ECH power deposition. | ||
| Resource Requirements: | All beams, 7 gyrotrons | ||
| Diagnostic Requirements: | All fluctuation diagnostics, MSE, CER | ||
| Analysis Requirements: | TRANSP, linear gyrofluid (TGLF) and nonlinear GYRO analysis | ||
| Other Requirements: | -- | ||
| Title | 63: Real-time Control of RE Beam | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys | ITPA Joint Experiment : | No |
| Description: | Demonstrate and characterize robust real-time position control of an RE beam while examining the minimal actuator set needed to maintain such control. Specific goals:
1. Establish reliable R control using RZ control, Isoflux, or combination of both. 2. Avoid elongating/diverting plasma in order to maintain passive vertical stability as li rises. 3. Determine if z position can be maintained throughout RE lifetime without fast vertical stabilization. 4. Determine if RE beam can be held steady state with Ip cont. 5. Establish control of RE beams with initial current < 200kA |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce RE beams using a limited plasma + Ar killer pellet formula. Tailor inner PF coils current to minimize their applied field after transient vessel current disappear in order to avoid significantly elongating or diverting RE beam. Apply rtEFIT/Isoflux as soon as possible using new snap file developed by A. Welander in order to obtain robust measurement of plasma position. Demonstrate reliable isoflux control, focusing on the ability to control R. Use isoflux to maintain a low-elongation RE beam, possibly using the PF5-8 coil to squish RE beam. Determine if passive vertical stability can be established, negating need for fast vertical stabilization. Using Ip control, determine if RE beam can be held in "steady-state" up to the Ecoil flux limit. Determine factors that can increase the chance of low current RE survival. | ||
| Background: | In 2010, D3D produced its first demonstration of reliable Z&Ip control of an RE beam for up to 250ms. However, this control was not particularly refined and several significant challenges remain.
1. Reliable R control was not demonstrated. The linear estimator used for control was erratic and the slew rate limit of the outer PF coils renders them incapable of any real control for tens of ms. 2. Long-lived RE plasmas often elongated and sometimes diverted. In order to keep the early RE beam from crushing into the inner wall, the inner coils must push hard to buck the attractive force of the vessel eddy currents. This leads to elongation and increased vertical instability. 3. All RE discharges ended in a VDE. It is critical to establish passive vertical stability, or at least minimize elongation, in order to delay/avoid this occurrence as li rises. Also, there is some evidence that Ip control delayed (avoided?) VDE. 4. Initial RE current < 200kA often crushed into the center stack, as coils are unable to respond fast enough to resist vessel currents. 5. rtEFIT would not start converging until 15-20 ms after CQ, making early isoflux control impossible. Only 1 (unsuccessful) attempt at applying isoflux to a steady late-time RE beam was made. | ||
| Resource Requirements: | 4x ECH, Ar killer pellets, 30L (MSE), 1 DAY | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, scintillators, Fast camera | ||
| Analysis Requirements: | EFIT, Toksys/Simserver tests to establish run PCS setting before run. | ||
| Other Requirements: | -- | ||
| Title | 64: ECCD within magnetic islands | ||
| Name: | Prater prater@fusion.gat.com | Affiliation: | GA |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | C. Petty, F. Volpe, R. La Haye, T. Strait | ITPA Joint Experiment : | No |
| Description: | This experiment is designed to measure the Electron Cyclotron Current Drive within a slowly rotating magnetic island. The experiment will show whether the ECCD within an island is correctly evaluated by a code like TORAY, which assumes poloidal uniformity. The result could reduce the calculated ECCD power requirement for NTM stabilization in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Entrainment of a saturated 2/1 NTM in a rotating externally-applied magnetic field can produce a slowly rotating mode, say at 50 Hz, which allows the ECCD to be modulated synchronously with the O-point. If the magnitude of the ECCD is significant but too small to stabilize the island, the current driven in the island will be measureable through the modulation of the MSE pitch angles at 50 Hz. This will support determination of the driven current density. | ||
| Background: | he control of neoclassical tearing modes by ECCD within the islands has been shown to be highly robust, maybe more so than expected from the Rutherford equation. One reason may be that the poloidal bunching of the ECCD within the islands provides an un-included improvement in the ECCD efficiency compared to the poloidally uniform case. Alternatively, if the current-carrying electrons diffuse out of the islands before they isotropize through collisions, the ECCD will be diminished. These effects should be understood to project correctly to the requirements for ITER. | ||
| Resource Requirements: | At least 5 gyrotrons. The I-coil is an essential element. | ||
| Diagnostic Requirements: | MSE is essential. | ||
| Analysis Requirements: | Analysis will be challenging. | ||
| Other Requirements: | |||
| Title | 65: Edge localized ECH to increase ELM frequency | ||
| Name: | Prater prater@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | T.Osborne, R. Groebner, J. Lohr | ITPA Joint Experiment : | No |
| Description: | ECH in the pedestal region may be capable of affecting the ELM frequency. In one sequence of recent shots, the ELM frequency doubled when modest ECH power was applied near the top of the pedestal. However, a scan of the position of the power relative to the pedestal top was not carried out, so the effect on the ELM frequency may have been due to the additional global power rather than the local power. This experiment would clarify the effect. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Duplicate shot 139780 which had 3rd harmonic central heating and 2nd harmonic heating on the high-field side pedestal. Apply ECH at slightly different values of Bt so that the location of the ECH is scanned across the pedestal. If the effect is due to the localized heating in the pedestal, the change in the ELM frequency should be sensitive to the exact radial location of the heating. | ||
| Background: | ECH heating near the pedestal top was found to have an effect on ELM frequency when the 60 GHz inside launch system was used on Doublet III. In that experiment, small changes in the location of the resonance made a significant different in the ELM frequency, with heating outside the pedestal causing a decrease in the ELM frequency and heating inside the pedestal causing an increase. | ||
| Resource Requirements: | 5 gyrotrons | ||
| Diagnostic Requirements: | Edge Thomson would be very useful. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 66: Vertical control of RE from ITER-like targets | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys, V. Izzo | ITPA Joint Experiment : | No |
| Description: | Demonstrate robust vertical control of RE beams formed from vertically unstable elongated targets, particularly an ITER-like LSN. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with ITER-like LSN (or at least elongated limited) target. A suitable target may come from the RE deconfinement via enhanced MHD proposal by Humphreys. Attempt to minimize li using off-axis current drive, particularly off-axis beam, if available, in order to enhance RE production. If this results in RE beam production, catch and control the resulting beam using the established methods from the 2010 campaign. | ||
| Background: | Vertical control of RE beams from low elongation, limited target plasmas was demonstrated on D3D in 2010. However, it is interest for ITER to demonstrate vertical control of an RE beam derived from an ITER-like LSN. Modeling studies (see S. Konovalov, 2010 IAEA FES) have indicated that ITER cannot vertically control RE beams after a major disruption. However, it is the opinion of this author (Eidietis) that those studies assume an extremely pessimistic, and perhaps un-realistic, cross-section for the RE beam. They appear to assume that the RE beam essentially maintains the elongated, vertically unstable cross section of the target plasma. Experience on D3D indicates that RE beams tend to initially form as small, circular, high-aspect ratio plasmas independent of the shape of the target plasma, making them much more vertically stable. Hence, it is of interest to experimentally test if an RE beam from an ITER-like target can be vertically controlled. This has not been attempted up until now because RE production in LSN plasmas only occurs in 10-20% of shots, making run days very inefficient. However, NIMROD modeling indicates that lower li plasmas may be more likely to produce reliable RE beams, allowing for more productive run days. | ||
| Resource Requirements: | 6x ECH, Ar killer pellets, 30L (MSE), off-axis NBI 0.5 DAY | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, scintillators, Fast camera | ||
| Analysis Requirements: | NIMROD, current-drive codes? | ||
| Other Requirements: | -- | ||
| Title | 67: Controlled RE deconfinement | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys, P. Parks | ITPA Joint Experiment : | No |
| Description: | Apply active RE deconfinement/dissipation techniques upon controlled RE beam. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with limited, low-kappa target + Ar killer pellet to form RE beam. 3 techniques will be tested to observe their effect upon the RE beam evolution:
1. Attempt to scrape off RE by pushing RE beam against inner wall 2. Holding RE beam in steady state using RZIp control, heat the outer (thermal) plasma to reduce resistivity and gradually increase density do enhance RE dissipation. Given Wesley 2010 IAEA observation that RE seems to a be high resistance, it is hoped that Ip will gradually transfer from RE current to entirely thermalized current. 3. Attempt RMP deconfinement again, but limit plasma outboard to enhance Icoil field strength in plasma. | ||
| Background: | -- | ||
| Resource Requirements: | 6x ECH, Ar killer pellets, 30L (MSE), 1 DAY | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, scintillators, Fast camera | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | RE beam R control must have been established in previous experiment. | ||
| Title | 68: Direct measurements of Zonal Flows versus rational q-surface spacing | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | J. Hillesheim, T.L. Rhodes, G. Wang, E.J. Doyle, L. Zeng | ITPA Joint Experiment : | No |
| Description: | Mean ExB flows and Zonal Flows are predicted to peak at or near rational q-surfaces (R.E. Waltz et al, PoP 2005). ZF's can regulate ITG turbulence levels and, hence, indirectly, the critical ITG gradient. Zonal Flows near the q=2 surface are also a possible trigger mechanism for Internal Transport Barriers. We propose to directly measure the mean ExB flow and Zonal Flow corrugations near the q=2 rational surface under conditions were the rational surface spacing can be varied, with the goal to provide data for a comparison with the Mean Flow and ZF structure predicted by gyrokinetic simulations. Closely spaced low order rational surfaces are expected to show different behavior as Zonal Flow layers can overlap with possible flow shear cancellation, and the turbulent eddy size may span adjacent rationals.Accordingly, changes in ZF and mean flow shear and in turbulence saturation (or density fluctuation levels) may be detectable under these conditions. The goal of the experiment is to compare experimentally Zonal Flow shearing rates with simulations for rarefied and closely spaced rational surfaces. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The proposed experiment utilizes L-mode plasmas ( | ||
| Background: | Initial measurements of a localized E_r shear layer near the q=2 surface have been previously made by Doppler Backscattering. The shear layer has been shown to locally reduce density fluctuations (L. Schmitz, APS invited talk, 2008). Mean flow and time-dependent flow shearing rates on the order of 1-2x10^5 s^-1 were observed from the differential Doppler shift across four DBS channels probing within a 4 cm region around the q=2 surface. These shearing rates are comparable to those found in weak pedestal shear layers. | ||
| Resource Requirements: | All neutral beams, possibly ECH
A dedicated day is requested for this experiment. Several shots will be required to line up the DBS probing radii with the resonant surfaces locations for each condition as the current profile evolves. time is required to perfectly align | ||
| Diagnostic Requirements: | All core fluctuation diagnostics, MSE, CER | ||
| Analysis Requirements: | Low-k global GYRO runs to resolve radial ZF structure near rational surfaces. | ||
| Other Requirements: | |||
| Title | 69: Characterization of I-mode on DIII-D with rf (FW and/or EC) heating | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | J.S. deGrassie, M. Porkolab, J. Hosea, A. Nagy, P.M. Ryan | ITPA Joint Experiment : | No |
| Description: | This experiment is a follow-on to ROF #59, "Exploration and Characterization of I-mode", where it is proposed to obtain C-Mod's I-mode on DIII-D with neutral beam heating. On C-mod, I-mode has been obtained with ICRF heating as the sole auxiliary heating. The principal motivation for the study of I-mode is an ELM-free confinement mode with H-mode level confinement, and as such I-mode is potentially a game-changer for ITER. However, I-mode also could also solve another problem for ITER, which is coupling high levels of ICRF power without excessive rf electric fields near the antenna. Those rf electric fields cause impurity problems (particularly in high-Z first-wall environments) and also limit the power density that can be obtained without electric breakdown of the antennas. The rf electric field level needed to couple a given power level scales with the antenna load resistance RL as RL**-2, and the L-mode-like edge plasma in I-mode yields antenna loading that is about twice what is observed after an H-mode transition. Hence, at a given fixed maximum antenna voltage, about four times as much ICRF power can be coupled in I-mode (or L-mode) than in an equivalent plasma with a conventional H-mode edge.
Furthermore, studies of HHFW heating on NSTX indicate that ELMs cause a significant degradation of HHFW heating efficiency, compared with ELM-free regimes {Hosea, et al., 2010). This result is consistent with previously published DIII-D FW results (Petty, et al., NF 1999, Vol. 39, 1421), which showed a dependence of FWCD efficiency on ELM frequency and character - at the highest ELM frequencies, the FW edge losses became much more important than in ELM-free regimes such as VH-mode, or with infrequent Type I ELMs. Hence, we expect that FW edge losses could be significantly reduced in I-mode, compared with ELMing H-modes. A crucial aspect of I-mode accessibility for use with ICRF is whether the accessibility has a dependence on the plasma/wall gap, i.e. the outer gap. To maximize the antenna loading, the outer gap is reduced to the minimum compatible with outer wall heating, acceptable rotation/confinement and accessibility to the desired confinement mode. Hence it is important to extend the I-mode accessibility study of ROF #66 to include the affect of the outer gap on I-mode threshold. Another aspect of I-mode accessibility that will be addressed partly by ROF #66 is the possible role of rotation on I-mode access, in that the applied torque in C-Mod is very much lower with ICRF heating only than will be obtained with co-injection NBI. The scans listed in ROF #59 might be extended by adding NB torque to the possible controlling parameters, scanning from all co-injection to balanced injection; replacing NBI with non-torqueing FW heating in steps is another approach to this important aspect. Along similar lines, the possible effect of the mix of ion and electron heating will be studied by changing the mix of NBI and FW (and/or ECH.) |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The proposal would be to follow the NBI-only I-mode characterization experiment and start with the best (i.e. most robust I-mode access) case from that experiment. The NB power would be replaced with FW power from shot to shot, maintaining the total heating power level at that which maintains I-mode for as long a period as possible. If the BT for optimal I-mode turns out to be high enough to permit reasonably central 110 GHz X2 heating, and the maximum FW power level is reached (should be ~3.5 MW or higher), further replacement of NBI power with EC power is a possible continuation.
If the outer gap dependence of I-mode access has not yet been addressed by the time of this experiment, it should be studied as part of this work. The I-mode accessibility with NBI only could be extended to include a scan of outer gap and the FW loading measured non-perturbatively during that scan. The high-power FW portion of the experiment should be carried out at the minimum outer gap at which satisfactory I-modes are sustained. | ||
| Background: | Experiments on DIII-D with an L-mode edge but high heating power have a long history, as alluded to in ROF #59. One particular example of this kind of work in the 1990s was the fact that to measure the non-inductive current profile with MSE, as in the FWCD work, it was necessary to maintain a longish sawtooth-free period and an L-mode edge, to keep the density low and the power-per-particle as high as possible. When all three FW systems were added to a sufficiently high NB power to keep q0>1 for a long enough period to make the measurement, it was necessary to bias the shape upwards and reduce the inner gap to maximize the L-H transition power in order to maintain L-mode (see 19960311 for a typical example of a session of this kind.) A more recent example of something similar is 140715, from the TS/ECE discrepancy experiment, where an H-factor relative to 89P scaling of about 1.9, or an H-factor relative to H98y2 of about 0.86, is obtained with an L-mode edge. Again, unwanted H-mode transitions limited the beam power in this recent experiment. It is not being claimed that these discharges are I-modes, only that good confinement with an L-mode edge has been obtained with FW plus NB heating in DIII-D in regimes limited by unwanted H-mode transitions, even after having raised the L-H transition power as much as possible.
The combination of ROF #59 with this proposal will constitute a good first look at I-mode in DIII-D with NBI, FW and possibly EC heating. | ||
| Resource Requirements: | NBI, up to 6 sources (off-axis not needed, but both co- and counter- sources needed)
All three FW systems At least 4 gyrotrons Upper and lower cryopumps | ||
| Diagnostic Requirements: | Usual core diagnostics, plus edge fluctuation and profile (e.g., reflectometers) diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 70: Beta limit and bootstrap current fraction in ITER steady-state scenario discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | Study the performance of steady-state scenario discharges in the ITER discharge shape in order to establish the physics basis and optimum operating scenario for the ITER steady-state mission. Determine the beta limit and bootstrap current density as a function of q_min. Make comparisons between performance in the single null ITER shape and the double null DIII-D AT shape in order to establish the physics basis for the evolution between ITER and DEMO end for optimization of steady-state scenario discharges in DIII-D. A portion of this experiment addresses ITPA IOS group high-priority experiment 3.1. Increase the plasma current over what has been used previously to push q95 down to 5 in order to reach conditions that project to Q = 5 in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce the fNI = 1 discharges produced in 2008 and vary q_min, beta and density gradient in order to test the effect on the achieved bootstrap current and beta limit. Use the ECCD to better advantage to avoid 2/1 tearing modes in order to either raise the achievable betaN or establish the maximum betaN value as determined by ideal stability. Do this in a discharge shape that better matches the ITER scaled shape in the outer, lower squareness. Using shape adjustments, modify the density by taking advantage of the divertor cryopump. As conditions are varied, test the effect of the outside gap on the betaN limit. Make use of the off-axis beams to improve the capability to reach elevated values of q_min. | ||
| Background: | During 2008 the first attempts were made at making a fNI = 1 discharge in a scaled ITER shape in DIII-D. FNI = 1 was successfully obtained at relatively low betaN = 3.1 with fBS = 0.7. The beta limiting instability was a 2/1 NTM and the outside gap seemed to have a moderate effect on the achievable beta. This contrasts with the double null shape steady-state scenario discharges which had less density gradient and correspondingly less bootstrap current but which have been operated at betaN = 3.7 without a 2/1 NTM. The discharge shape that was used doesn't quite match the intended ITER scaled shape. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 71: Improve the ability to produce the exact, scaled ITER discharge shape in DIII-D | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Find a new patch panel configuration that will allow the lower, outer squareness in the DIII-D "scaled ITER shape" to better match the actual squareness in the ITER shape. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Try a different patch panel configuration, probably with return current in the F4B coil. This will probably require more than just a couple of shots because it is hard to predict the effect of changes on the patch panel on the currents during the full discharge evolution. Typically there are unexpected excursions in current on various coils that need to be dealt with. During 2008, a fair amount of time was spent trying to overcome this type of problem. If possible, modeling using TokSys would be useful before the experiment in order to test the effect of various patch panel configurations. | ||
| Background: | During 2008, the ITER demonstration discharge task force attempted to produce a discharge shape in DIII-D that was an exact match to the ITER shape scaled by a factor of 3.7. The shapes produced were a good match except in the lower, outer squareness. The shape mismatch was a consequence of the VFI constraint on the DIII-D patch panel configuration. Two different patch panel configurations were tried. Both had too much current in the F7B coil, resulting in lower outer squareness which was too small. At least one alternative patch panel configuration exists which might improve the squareness match. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 72: Long duration, 100% noninductive fraction discharges in the ITER shape | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make use of increased pulse length capability in DIII-D (longer toroidal field duration, more neutral beam energy) and increased ECCD power to push the duration of fNI = 1 discharges in the ITER shape toward 2 tauR. Make use of off-axis beam injection to increase the stationary value of q_min. Make use of a long duration, stationary phase of the discharge to optimize the NVLOOP determination of the noninductive current profile. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In 2011, in comparison to 2008 when steady-state scenario discharges were last run in the ITER shape, DIII-D will have an additional neutral beam available, some of the beams will be able to operate with longer pulse length, the toroidal field pulse length capability will be increased, there will be six gyrotrons available, and there will be off-axis beam injection available. This provides increased hardware capability that should allow the extension of the pulse length duration of discharges with moderate values of betaN and elevated q_min. Take advantage of this capability and the good performance of the 2008 ITER steady-state scenario discharges to study longer duration fNI = 1. | ||
| Background: | In the steady-state scenario ITER demonstration discharges produced in 2008, fNI = 1 was obtained with relatively low betaN = 3.1. These discharges had very broad Te profiles and a relatively large core density gradient which resulted in fBS of about 0.7. This discharge is a good candidate for the study of longer duration with fNI = 1 because the relatively low neutral beam power required will allow long beam pulse duration. In addition, the high noninductive current fraction expected (based on previous results) increases the likelihood that a steady-state, high q_min discharge can be obtained. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 73: Delaying TQ onset during disruption mitigation using ECH | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use ECH heating of the q=2 surface during pellet injection to delay the onset of the thermal quench and enable impurity deposition in the core. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Standard disruption LSN target. Turn on ECH at q=2 surface shortly before injection of Argon killer pellet. Measure the time from the pellet launch to TQ with and without ECH. Vary applied ECH power and timing relative to pellet launch. | ||
| Background: | A basic problem for almost all massive impurity injection techniques is that the TQ often occurs before the impurity payload can be delivered to the plasma core (< q=2). This results in greater heat conduction to the divertor and inhibits density coalitional suppression of runaway electrons. Delaying the TQ onset time even 1-2 ms would greatly enhance the ability of the various mitigation methods to penetrate to the core. ECH on the q=2 has been shown to delay/avoid numerous types of disruptions on numerous tokamaks. | ||
| Resource Requirements: | 6x ECH, Ar killer pellets, 30L (MSE), 0.5 DAY | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, scintillators, fast camera | ||
| Analysis Requirements: | |||
| Other Requirements: | Is this the correct group to put this in? | ||
| Title | 74: Operations development time | ||
| Name: | Walker walker@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Requested |
| Co-Author(s): | Al Hyatt, Dave Humphreys, Jim Leuer | ITPA Joint Experiment : | No |
| Description: | Establish regular 2 hour experimental slots for initial testing of experimental use of new operational capabilities. Possible uses include shape development, diagnostic calibrations, control development, tests of ICRF or ECH systems, tests of off-axis beam, etc. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Define a method of proposing 2 hour development experiments a week or more in advance. Define a process for prioritizing and scheduling competing tests. We recommend that priorities be based on demonstration of readiness to proceed, need for capability in upcoming physics experiments, relevance to goals of DIII-D program, and available manpower and budget constraints (an option should be available to not conduct a particular 2 hour experimental session if constraints are too severe). | ||
| Background: | Presently, any experiment that uses a new or modified capability must take on the substantial risk associated with the lack of experimental testing of that capability. In the past, this has sometimes resulted in significant portions of an experimental day or even the whole experimental day being consumed by difficulties associated with a lack of readiness of the new capability. In addition, the time it takes for new systems to come on-line can be increased significantly when scheduled experiments are not willing to take on that risk. In recent years, the use of regularly scheduled 2 hour experimental sessions on Thursday 5-7PM for operational development has significantly accelerated the development new capabilities. | ||
| Resource Requirements: | Machine Time: 2 hours experimental time each week.
Other requirements will vary with the tests being performed. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 75: Correlation between ELM activities and plasma edge voltage with respect to divertors | ||
| Name: | Zheng none | Affiliation: | IFS, Univ. of Texas-Austin |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | TBD. This work has colloborated with H. Takahashi and E. Fredrickson with DIII-D experimental data used. | ITPA Joint Experiment : | Yes |
| Description: | Both theoretical analyses and experimental observations seem to point out that ELM activities and electric voltage between plasma edge and divertors are correlated. We propose to further clarify this issue experimentally and to explore the method of ELM mitigation through reducing the negative edge electric voltage with respect to divertors.
The theoretical support is described in the background section. Our prediction that the reduction of edge negative voltage can be tapped for ELM mitigation is consistent with the existing experimental observations. Some examples are briefly discussed as follows: 1. I-Modes in C-Mod (A. Hubbard, APS-DPP Talk PI2:6, Nov 2010): Most C-Mod I-Modes have been obtained with unfavorable BĆ?ā??B drift, which leads Er to be only weakly negative. This exactly corresponds to a reduced electric voltage between plasma edge and divertors. 2. QH modes in DIII-D (K. H. Burrell. et al, PHYSICS OF PLASMAS 12, 056121 (2005)): To date, QH-mode operation in DIII-D requires neutral beam injection opposite to the plasma current direction (counter-injection). Counter-injection create a negative Er, which pulls ions toward the core plasma. Consequently, the edge ion orbit loss is reduced and so does the voltage between plasma edge and divertors. 3. RMP suppression of ELMs in DIII-D (T. Evans et al): The RMP causes the edge field line stochasticity. Under this circumstance, electron accumulation cannot be built up at plasma edge due to the parallel mobility of electrons. Our theory can also explain the ELM dependence on pedestal collisionality, as well as ELM mitigation by the pallet injection, etc. (see for example, L. J. Zheng, et al. TTF09 presentation, http://ttf2009.ucsd.edu/working/ab_storage/after_presentation/zheng_ttf09_elm0.pdf). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Adjusting the edge plasma voltage with respect to divertors can be achieved by various ways: such as applying electric bias directly by probes, injecting ions parallel to the magnetic field lines from divertor area, controlling neutral beam injection direction and radial deposition position, etc. They are just suggestions as I am not an experimental expert. | ||
| Background: | To explain the physics, we can imagine tokamak as a heat ā??reservoirā??. If overheated, the tokamak ā??reservoirā?? would overflow and eventually might cause its ā??damā?? to collapse. ELMs are an phenomenon of tokamak energy/particle overflow. Nevertheless, it is not a big overflow, since they occur at marginal stability. In most cases, they need a trigger and nonlinear amplification. Possible triggers can be plasma instabilities, e.g., the so-called peeling-ballooning modes. Note that interchange-type of modes (e.g., peeling-ballooning) interchange not only plasma and magnetic energies, but also the current.
Consequently, current sheets are created on the mode rational surfaces and interchange-type of modes are converted to the newly discovered current interchange tearing modes (CITMs, L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)). As this type of tearing modes (current-interchange peeling-ballooning-tearing modes) develops, edge magnetic surfaces are destroyed and the plasma edge is connected to the scrape-off-layer (SOL) by magnetic islands and stochastic field lines. This reconnection induces the radial particle and heat transport from pedestal to SOL and results in the burst of the SOL current; In return, the SOL current further drives the tearing modes. This positive feedback process explains the nonlinear development of ELMs [L. J. Zheng, H. Takahashi, and E. D. Fredrickson, Phys. Rev. Lett. 100, 115001 (2008)]. However, there is an important ingredient for this positive feedback process to happen: the electric voltage between plasma edge and divertors. In usual DIII-D H-mode confinement, plasma edge is charge-negative, i.e., there is non-neutralized electron accumulation, because of ion orbit loss and co-current rotation drive by neutral beam injection. Meanwhile, the divertor sheaths are charge-positive, i.e., there is non-neutralized ion accumulation, because more electrons are collected by the divertors due to their larger parallel thermal speed. As a consequence, a negative ā??biasā?? voltage exists between plasma edge and divertors. The reconnection between plasma edge (i.e., pedestal region) and SOL can cause a ā??short-circuitā?? and results in a current seed. This current seed is further enhanced by the neutralization of sheath ion charges due to the arrival of pedestal electrons. Consequently, the burst of SOL current occurs. The strength of SOL current burst depends on the voltage between plasma edge and divertors. Large burst results in ELMs, while small one can only create the so-called harmonic oscillation (EHO) as observed in C-Mod. Therefore, the reduction of edge negative voltage with respect to divertors may mitigate ELMs. | ||
| Resource Requirements: | Within normal DIII-D experimental setup. | ||
| Diagnostic Requirements: | measurement of edge plasma voltage with respect to divertors or wall, as well as usual ELM diagnostics. | ||
| Analysis Requirements: | Nonlinear simulation codes are not ready for current interchange tearing modes. We have developed a linear AEGIS-R code, that can investigate the stability condition of the current-interchange peeling-ballooning-tearing modes. | ||
| Other Requirements: | |||
| Title | 76: Off-axis NBCD Measurement | ||
| Name: | Park parkjm@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Requested |
| Co-Author(s): | M. Murakami, C. Petty, W. Heidbrink, M. Van Zeeland, and, T. Suzuki | ITPA Joint Experiment : | No |
| Description: | Measure local profile of off-axis NBCD using the new off-axis beams over range of operation conditions (+-BT, +- IP, injection power, and energy) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Establish a target L-mode plasma for off-axis NBCD measurement: high current at full-field, reverse BT; stationary as much as possible; delay sawteeth start time as much as possible; lower density for higher CD efficiency but until Alfven instabilities.
2) Start with 150LT beam at normal operating voltage and maximum tilting angle with 30LT and 210RT beam blips every 50 msec for co- and counter- MSE measurement. 3) Obtain a fiducial discharge with balanced injection for CD analysis. Use 30LT and 210RT and adjust duty factor and/or add ECH to match both Te and beta. 4) Reduce 150LT beam power (50% duty factor) 5) Reduce injection energy (50 keV) 6) Repeat with 150RT and 150LT+150RT 7) Repeat with normal BT 8) Repeat with reverse Ip | ||
| Background: | The eventual goal of this experiment is to establish physics basis for off-axis NB current drive and develop predictive modeling capabilities to use the new off-axis beams in many area of advanced tokamak development. ITER physics, and fusion science. In 2008, a prototype experiment for off-axis NBCD has been carried out using vertically shifted small plasmas (Park PoP 2009, Murakami NF 2009, Heidbrink PRL 2009, Heidbrink PPCF 2009). The off-axis NBCD profiles measured from the magnetic field pitch angles by the motional Stark effect diagnostics agree well with calculation using the orbit-following MC code, NUBEAM. The magnitude of off-axis NBCD turned out sensitive to the toroidal magnetic field direction due to change of the beam injection alignment relative to the local helical pitch of the magnetic field lines. The measured off-axis NBCD increases approximately linearly with the injection power, although a modest amount of fast ion diffusion is needed to explain an observed difference in the NBCD profile between the measurement and the calculation at high injection power PNB > 7 MW. This proposal focuses on the classical aspects of off-axis NBCD using the new off-axis beams. | ||
| Resource Requirements: | All neutral beam sources except either 30RT or 330RT with 150 beams at maximum tilting angle
Minimum 3 Gyrotron | ||
| Diagnostic Requirements: | MSE, Neutrons, FIDA spectrometers & cameras, Core reflectometer | ||
| Analysis Requirements: | NUBEAM, ONETWO, TRANSP, NVLOOP, FIDASIM analysis | ||
| Other Requirements: | |||
| Title | 77: Access to AI scenario with direct stabilization of tearing modes | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Requested |
| Co-Author(s): | T.Luce, M.Wade | ITPA Joint Experiment : | No |
| Description: | We propose to preemptively stabilize the n=1 mode that appears during the betaN ramp-up and limits access to the AI scenarios. This can be done by applying ECCD at the q=2 surface during the betaN ramp-up, using a feedback method based on real-time equilibrium reconstruction that tracks the position of the rational surface in time. If the mode is stabilized, different betaN trajectories can be used, demonstrating that a broad range of parameters exist, in the access to this scenario. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will start by reproducing discharge #142567 with the same ramp-up rate and path that were originally used. No ECCD will be applied for the first shot. If an n=1 mode appears, interrupting the betaN ramp-up, we will use the discharge to track the q=2 surface before and during the mode with real-time efits with and without MSE data, and double-check its position with respect to the rotation profile and the BES data. Using the information gathered in this test shot, the PCS algorithm that tracks the location of the q=2 surface and controls the radial position of the plasma (Rp) will be used to apply ECCD at the rational surface to preemptively stabilize the mode and get the discharge through to the high-betaN phase. Since the fastest betaN ramp-up is performed in ~200 ms, the algorithm cannot use Bt as an actuator to move the ECCD deposition, because it has a longer time scale. We plan to use ECCD for ~500 ms into this phase, then turn it off to maintain a high fusion gain Q. If the mode reappears after the EC power has been shut off, we will leave the gyrotrons on for as long as the system allows, to check whether the instability can be avoided. Keeping the n=1 NTM stabilized, we will explore a range of different betaN ramp rates, and different ramp timings (from 1 s to 2.5 s into the discharge). | ||
| Background: | he hybrid q95~4 scenario is of interest of ITER operations, but the present experiments have shown that the access to it is limited by the appearance of an n=1 NTM that is triggered during the N ramp-up phase. A very narrow path has been found in the Asdex-U tokamak to access this scenario from a sawtoothing ohmic plasma, and attempts have been made in DIII-D, which led to a similar conclusion, although only one discharge survived the ramp-up phase. It is vital to demonstrate that this scenario can be obtained reliably and reasonably independently from the precise betaN trajectory chosen in the front end of the shot and the initial q profile. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 78: Where are the antenna arcs occuring? | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | General IP | Presentation time: | Not requested |
| Co-Author(s): | F.W. Baity, A. Nagy, J. Yu, J.C. Hosea, A.G. Kellman, D. Rasmussen, P.M. Ryan, J. Caughman | ITPA Joint Experiment : | No |
| Description: | Essentially all experiments that use high-power ICRF heating are limited in the power that can be coupled to a given plasma by high-voltage antenna arcs/breakdown. Yet in no experiment has the actual location in the antenna structure where the breakdown occurs been unambiguously determined, except in pathological cases. This data is absolutely needed to inform any reasonable design for a significantly improved antenna. In the 2008 campaign, DIII-D commissioned a data acquisition system of unprecedentedly high time resolution (up to 1 GHz sampling rate) to study the electric signatures of antenna breakdown at the rf timescale. The UCSD fast camera was also used to look at the 0 deg FWCD antenna for the first time. What is needed is to combine these diagnostics to attempt to image an arc on the 0 deg antenna while at the same time acquiring fast electrical data on the same event.
During the present LTO, we are installing diagnostics on the 285/300 antenna, including probes in the antenna backplane and fibers and collimating lenses to bring out light from the antenna box. We also propose to use these diagnostics in an attempt to locate arcs, both in vacuum operation (no plasma) and possibly with plasma, in the 285/300 antenna. Also, we are installing an entirely new arc diagnostic, called GUIDAR, on the 285/300 system. This is intended to allow measurement of the location of abrupt changes in impedance with a spatial resolution of perhaps a meter or so. This system will be commissioned first in vacuum operation. It is possible in the best-case scenario that GUIDAR will allow localization of arcs even in the presence of tokamak plasmas. If so, this would be of very great international interest, as such a system has never been tried in a fusion experiment to date. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | What needs to be done is to trigger both the camera running at its maximum frame rate and the fast digitizer system with the same event trigger that tells the transmitter to blank the rf power, and to acquire both pre-trigger and post-trigger data. It may prove to be necessary to intentionally slow down the response of the blanking system to allow enough energy to be deposited in the arc to produce a visible image on the camera. The usual system shuts off the rf within about 20 microseconds of the arc detection. Perhaps an L-mode plasma with a reasonably large outer gap, without NBI, would be the simplest target plasma to start with. | ||
| Background: | The ITER ICRF antenna system will be required to operate at higher rf voltage levels than have ever been obtained in a reliable, long-pulse way in any previous experiment. The limitation has always been electrical breakdown in the antenna system, yet nobody has been able to actually determine the location of the arcs in existing antennas, presumably due to the successful fast shut-off of the rf power when an arc is detected. This is clearly important for learning how to achieve the needed reliability for ITER. | ||
| Resource Requirements: | Fast framing camera looking at an optimized view of the 0 deg antenna, fast digitizer system connected to raw rf samples on the same antenna system, triggering set up specially as required. | ||
| Diagnostic Requirements: | Special diagnostics set up as listed above in resource requirements. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 79: FW-only L/H Transition Power Study | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | General IP | Presentation time: | Not requested |
| Co-Author(s): | A. Nagy, M. Porkolab, P.M. Ryan, J. Hosea, G. Wang | ITPA Joint Experiment : | No |
| Description: | ITER may not have enough auxiliary heating power to exceed the L/H transition power in the Day 1 configuration (hydrogen ops). For this reason, several machines have remeasured the L/H transition power in hydrogen with hydrogen or helium beams and compared those results with deuterium. Furthermore, recent work has shown that the L/H transition power has a dependence on plasma toroidal rotation speed, with lower rotation speeds being associated with lower L/H transition power levels. Even 20 percent-level effects may be important in this context. With this in mind, the fact that the Fast-Wave only H-mode observed in 1991 had a distinctly lower power threshold than with NBI heating in the same discharge may be of importance. The fact that H-mode transitions were observed with fast wave (FW) power as the only auxiliary heating source, under conditions of rather low single-pass absorption was an important piece of evidence that multiple-pass absorption of the FW power can be efficient. By expanding the range of FW frequencies, densities (and hence target electron temperatures), and using 3rd harmonic ECH, we can get a more quantitative measurement of the edge losses by determining the L-H transition threshold power under varying single-pass absorption conditions. This is important to ITER, both from the point of view of improving knowledge of access to H-mode in plasmas with only intrinsic rotation (no torque) and also to improve understanding of FW edge losses under varying edge conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment consists of scans of target density, rf power (at two different frequencies: 60 MHz and 90 MHz), toroidal field, and whether 3rd harmonic ECH is added (at the appropriate toroidal field), and comparison of co-, counter-current, and push-pull phasing. A beam is used for comparison, later in the shot. Minimal beam blips are used for CER, MSE diagnostics. At each condition, the power threshold for L-H transition is observed for FW, for the comparison beam, and for ECH (at the appropriate fields). | ||
| Background: | H-modes with fast wave heating by direct electron absorption as the only form of auxiliary heating were discovered at DIII-D in July 1991, and have not been studied since. In particular, the fast waves in that experiment were launched with the shortest available parallel wavelength ("Pi phasing") at 60 MHz at around 1 T, and we have never studied H-modes with current drive phasing, either co- or counter-current, or at higher frequency than 60 MHz. Furthermore, the great interest in the dependence of L/H transition power levels on rotation and/or applied torque in recent years has provided a new motivation for this experiment, as mentioned in the description section above. Finally, insofar as this study provides further data on FW edge losses, the emphasis on this area on NSTX in the past several years has increased the need to obtain data at higher toroidal fields than can be run on NSTX to see how these effects scale with BT, to provide data both for possible future STs and for ITER FW heating.
In a piggyback experiment in 2010, it was found that H-mode transitions could be rather easily obtained with a very low NBI power plus 2-3 MW of FW power in directional (counter-current) phasing. This reduces the uncertainty that FW-only H-modes can be obtained in the lower-k-parallel 90 deg phasing. | ||
| Resource Requirements: | Machine Time: 1 day Experiment
Number of gyrotrons: 4 Number of neutral beam sources: 4 Three FW systems, one at 60 MHz and the others at 90 MHz. | ||
| Diagnostic Requirements: | Edge reflectometry with the antennas adjacent to the 285-300 FW antenna, along with the UCLA profile reflectometers, would be a very helpful addition to the usual diagnostic set for this experiment. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 80: FW coupling and electron heating in ELM-stabilized H-modes with RMPs - continued | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | C.C. Petty, T.E. Evans, M. Porkolab, P.M. Ryan, A. Nagy, J.C. Hosea, G. Hanson | ITPA Joint Experiment : | No |
| Description: | It is arguable that since uncontrolled large ELMs are probably not acceptable for ITER and beyond and therefore ELM control is an absolute requirement, the most relevant regime for FW coupling is one in which ELMs have been suppressed with Resonant Magnetic Perturbations (RMPs). In this experiment, we would continue the study of high-power FW coupling and central electron heating in ELM-stabilized discharges with RMPs that we began in FY07. In a single day's exploratory experiment, we showed that ~2 MW of FW power could be coupled to an ELM-stabilized discharge, and obtained the first signs of central electron heating due to the FW. Much of the day was occupied with establishing an ELM-suppressed case with much smaller outer gap than had previously been used. Subsequent experiments in the ELM-control area have produced more robust ELM-stabilized cases at smaller outer gaps, so that we have a better starting point. Poor B-supply regulation in 2007 led to significant difficulty in staying in the narrow q-resonant window. We did not get a good no-rf comparison shot in either of the two cases which we studied (different outer gaps). No significant power was coupled at 60 MHz from the 285/300 antenna, due to a problem in the antenna which was remedied in the Fall 2007 vent. We need to continue this experiment, which is the world's first such attempt to couple fast wave power to an RMP-stabilized edge, with all of these issues addressed to obtain a publishable result on this important topic. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Continuation of the experiment on this topic from 2007, in which the beam power (=programmed beta) and outer gap are minimized while maintaining the ELM suppression with RMPs and acceptable outer wall heating, the FW power added and documenting the resulting electron heating. Comparison of different antenna phasings (both co-, both counter, push-pull), measurement of electron heating profile with modulation of the FW power. | ||
| Background: | See description. | ||
| Resource Requirements: | Machine time: one day experiment, Number of beam sources: 6, three FW systems, one at 60 MHz and two at ~90 MHz, 7 kA operation of the I-coil in the configuration used for RMP experiments. | ||
| Diagnostic Requirements: | The addition of the edge reflectometer adjacent to the 285-300 FW antenna, along with the UCLA profile reflectometers, would be a significant plus to these studies. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 81: High central fast wave current drive efficiency at high electon beta with ECH preheating | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | C.C. Petty, P.M. Ryan, J.C. Hosea, M. Porkolab, A. Nagy, J.M Lohr, R. Prater | ITPA Joint Experiment : | No |
| Description: | Combine 6 gyrotrons-worth of central 110 GHz ECH (all launchers aimed at or near the center of the discharge without driving toroidal current) with the combined 60 MHz and 90 MHz fast wave power. We would use the minimum neutral power necessary to create the sawtooth-free discharge in which the driven currents can be accurately measured, and to make the (MSE) measurement. Both L-mode and H-mode target discharges would be tried, although at full rf power, one would expect difficulty in keeping the discharge in L-mode. The basic scan would be a density scan, at each case obtaining at least a matching pair of discharges with co- and counter-current FW phasing. The object of the exercise would be to extend the range of central electron beta values, and hence of single-pass absorption of the FW power, considerably beyond what was possible without high power ECH. If time permits, comparison of the current drive efficiency of the 60 MHz and 90 MHz systems could be performed - as the single-pass absorption increases, we expect at some point to observe more efficient current drive at higher launched parallel phase velocity (the higher frequency case). The experiment would seem to be the logical precursor to full utilization of the combined rf systems for AT work involving tailoring of the current profile. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | See description. | ||
| Background: | This experiment was tried on two days in the 2004-2005 campaign. However, technical problems prevented any useful data to be obtained. The experiment was also tried for a fraction of a day in 2010, but again technical problems precluded any advances from being made. This experiment is a natural one to split time with the TS/ECE discrepancy experiment (see A. White's contribution to the ROF), as the setup is almost identical. The DIII-D FWCD system was designed to be most efficient in a plasma with central electron temperature of about 10 keV, but the maximum electron temperature at which we have measured the FWCD efficiency to date is about 6 keV. The theoretical prediction is that the FWCD efficiency scales roughly linearly with central electron temperature, and all experiments to date have conformed with this prediction. | ||
| Resource Requirements: | Machine time: 1 day experiment. 5 gyrotrons minimum,4 NB sources, all three FWCD system | ||
| Diagnostic Requirements: | All usual profile diagnostics, with MSE being especially important. | ||
| Analysis Requirements: | Analysis of current drive with MSE tools, NVLOOP. | ||
| Other Requirements: | |||
| Title | 82: Investigate parameter range for zero-NBI-torque QH-mode with NRMF | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | A. Garofalo, J.K. Park | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the parameter range over which we can achieve QH-mode with zero net NBI torque. This will involve scans in toroidal field, plasma current, density, beam power and plasma shape. The scans will be guided by peeling-ballooning theory (to investigate stability boundaries) and IPEC calculations (to optimize NTV torque). Issues to be investigated include 1) Can we achieve low rotation QH-mode in the upper single null and balanced double null shapes used in previous QH-mode experiment? 2) Does the NTV torque vary with shape in the way predicted by IPEC? 3) What is the collisionality range where QH-mode exists and how does this change with plasma shape (LSN versus USN versus DND)? 4) What is the range of q for low rotation QH-mode operation? 5) What are the limits on input power or beta_N? |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Complete investigation of the low rotation QH-mode parameter space will require several experimental days. Exactly how we order the various scans will require detailed discussion within the working group. One possible scenario would be to start with the LSN plasma used previously and perform density (e.g. edge collisionality), q and power scans to establish the ranges for this shape. We would then move to the USN shape in order to make contact with the previous, higher rotation QH-mode work and repeat the density, q and power scans. Finally, we would try the DND shape; previously, this shape has shown the best peeling-ballooning stability and, hence, the highest density limit. Investigation of the NTV torque under the various conditions would require taking shots both with and without the nonresonant magnetic fields. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. The toroidal rotation in these plasma is much lower than in previous QH-modes with unbalanced NBI and is similar to what one might expect in future devices. Having established the existence of QH-mode with low rotation, we now need to investigate the parameter range over which we can reliably achieve QH-mode with low rotation. | ||
| Resource Requirements: | Reverse Ip operation | ||
| Diagnostic Requirements: | All profile and fluctuation diagnostics, especially edge BES and ECE-I for EHO studies. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 83: Determine how much co-torque can be used in QH-mode with NRMF | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | A. Garofalo, J.K Park | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to determine now much co-NBI torque can be used while still maintaining a QH-mode plasma using nonresonant magnetic fields (NRMF) to provide counter torque. In order to compare with peeling-ballooning theory of the EHO, we will make careful measurements of the edge rotation profiles at the various torques. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment will be an NBI torque scan. We will start by reproducing the zero NBI torque QH-mode using the NRMF that we have studied previously. Plasma conditions will be chosen as indicated in the background section. Unlike previous QH-mode work, this starting condition will be produced with forward Ip in order to have more co-NBI torque at high power. This may require some development work. Once we have the balanced beam case operational, we will gradually add co-NBI torque until we find the point where the QH-mode can no longer be sustained. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices,since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. The toroidal rotation in these plasma is much lower than in previous QH-modes with unbalanced NBI and is similar to what one might expect in future devices. A key question for compatibility with machines like ITER is whether we can maintain the QH-mode with NRMF in the small amounts of NBI torque in the co-direction. This work will build on previously proposed experiments that investigate the low rotation QH-mode parameter space and which compare the NTV torque from the NRMF to theoretical calculations. We will choose the shape, beta_N value and the I and C-coil configurations which optimize the NTV torque and minimize the intrinsic plasma torque. Previous work by Solomon (IAEA 2010) has shown low intrinsic torque in QH-modes in the DND shape. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | All profile and fluctuation diagnostics, especially edge BES and ECE-I for EHO studies. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 84: Test of Neoclassical Toroidal Viscosity theory using modulated I-coil currents | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | Use modulated I-coil currents to investigate the theory of braking of plasma toroidal rotation by non-resonant error fields | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate the I-coil currents to modulate the non-resonant drag on the plasma. Investigate the effects as a function of modulation frequency, background plasma rotation, collisionality and I-coil parity. | ||
| Background: | This experiment was given 1/2 day in 2008. Unfortunately, there were issues of machine cleanliness since it was run after an experiment with significant gas puffing. Accordingly, the QH-modes were poor. Attempts to perform this experiment in ELMing H-mode lead to locking of the ELM frequency to the I-coil modulation. Although this locking was a significant discovery, the ELM effects on the rotation masked the direct I-coil effects. We need to perform this experiment in high quality QH-mode plasmas, since this avoids the ELM problem while still allowing us to probe H-mode plasmas. | ||
| Resource Requirements: | I-coil system connected to do both error field correction and n=3 braking. QH-mode will require reversed plasma current. | ||
| Diagnostic Requirements: | All profile diagnostics. CER at high enough speed to have 10 samples per I-coil modulation period. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 85: Further development of co-NBI QH-mode | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | P. Gohil, T.H. Osborne, P.B. Snyder, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | Broaden operating range for the co-NBI QH-mode discovered in 2008 through systematic, theory-guided parameter scans | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | he set of experiments listed here are designed to 1) optimize QH-mode operation under the conditions used in the 2008 experiments and to 2) broaden the QH-mode operating space. Optimization of existing conditions: 1) Find minimum possible target density by lowering gas injection rate early in the shot and moving beam start time as early as possible. 2) Extend QH-mode duration by operating at higher input power and torque Expand parameter space: 1) Scan Drsep and upper triangularity. 2) Vary safety factor by changing current and toroidal field. 3) Vary outer gap to see the effect on the EHO. In order to properly carry out all these scans, three experimental days are requested. | ||
| Background: | QH-mode with all co-injection was discovered during serendipitously during the 2008 campaign and a dedicated experiment was performed for one day. We have just barely begun the investigation of the co-NBI QH-mode. The goal of the present proposal is to use our knowledge of counter-NBI QH-mode to find ways to broaden the co-NBI QH-mode operating space so that co-NBI QH-mode can be used more routinely. The parameter scans listed in the experimental approach are based on empirical results from counter-NBI QH-mode combined with theoretical understanding of the QH-mode operating boundaries based on peeling-ballooning mode stability analysis. All QH-mode experiments to date indicate that lowering the target density is beneficial for QH-mode. Theory tells us that more strongly shaped plasmas and increased rotational shear are both beneficial for QH-mode. In addition, edge stability depends on safety factor. Finally, the theory of the EHO says there is a range of outer gaps over which the EHO will exist and modify the particle transport. | ||
| Resource Requirements: | Reverse Ip operation. 6 NBI sources. | ||
| Diagnostic Requirements: | All profile and edge fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 86: ELM dynamics Study & theory comparison | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Divertor Detachment and Plasma Flows | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov, Z Unterberg, A Leonard, C Lasnier, J Watkins, R. Moyer | ITPA Joint Experiment : | No |
| Description: | Study ELM motion once it leaves the plasma, compare to analytical and numerical work. Setup various ELM sizes and use all fast diagnostics to evaluate velocity scaling, size scaling (means Te and Ne) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | All fast boundary diagnostis | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 87: Investigating turbulent transport of fast-ions in off-axis NBCD | ||
| Name: | SUZUKI none | Affiliation: | Japan Atomic Energy Agency |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | J.M. Park, M. Murakami, T. Luce, C. Petty, W. Heidbrink | ITPA Joint Experiment : | No |
| Description: | To investigate dominant mechanism of turbulent transport of fast-ions for a high heating power in DIII-D by scanning Te (in Te/Eb) and beta independently, where Te and beta were coupled in the previous experiments in 2008. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment puts emphasis on decoupling Te and beta of the plasma (described in ā??Backgroundā?? section) during off-axis NBCD, where Te and beta were coupled in the previous experiments in 2008. For this purpose, a pair of discharges having different Te but the same beta is produced through changing gas-puffing rate (and ne). The beta is scanned as well from ~1.2% (5.6MW in 2008) to higher values via NB heating power. As higher beta is preferable as far as off-axis NB driven current can be measured, since beta is estimated to be ~4% in the ITER SS scenario #4(type-I). Using a set of the density and heating power scans, dominant mechanism of turbulent transport of fast-ions in DIII-D is investigated and off-axis NBCD capability in ITER is to be discussed as well. In these experiment, plasma configuration will be different to that in 2008 where vertical jog of small-volume plasma was employed for realization of off-axis NBCD. In this experiment, measurement of turbulence intensity during the off-axis NBCD is essential in order to understand the difference in off-axis NBCD as fast-ion transport by background turbulence. NBCD location is to be discussed. For physics study, NBCD (r/a~0.5) will be better for larger effect on NBCD, while for ITER similarity, r/a~0.3 will be better. | ||
| Background: | Under the ITPA joint experiment IOS-5.1, validation experiment of off-axis NBCD was performed in DIII-D in 2008, where nice agreement of beam driven current between measurement and calculation assuming Db=0 was obtained for heating/CD power below 5.6MW. The Db is fast-ion diffusion coefficient in TRANSP. In a higher power (7.2MW), measured beam driven current was smaller than the calculation (Db=0) and best agreed with Db=0.3m2/s. It was considered that turbulent transport of fast-ions is a cause of the larger Db for the higher heating power. Theoretically proposed two models of turbulent transport of fast-ions predict that the fast-ion diffusion depends on Te/Eb for electrostatic turbulence or beta for electromagnetic turbulence, respectively. The larger the parameters, the larger the effect of turbulence becomes. In the 2008 experiments, Te (and probably beta) increased with increase in the heating power for almost the same density so that Te and beta were presumably coupled (or not clearly separated); form fig.22 in J.M. Park et al, PoP16 (2009) 092508. | ||
| Resource Requirements: | off-axis NBI | ||
| Diagnostic Requirements: | MSE, FIDA, BES, and standard diagnostics for transport study in order to evaluate bootstrap current and neo-classical resistivity in loop-voltage-profile analysis. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 88: Fast-Ion Losses and Associated Heat Load from Edge Perturbations (ELMs and RMPs) | ||
| Name: | Garcia-Munoz manuel.garcia-munoz@ipp.mpg.de | Affiliation: | U of Seville |
| Research Area: | Energetic Particle | Presentation time: | Requested |
| Co-Author(s): | R. Fisher, W. W. Heidbrink, R. Nazikian, D. Pace, M. A. Van Zeeland, V. Kiptily, K. Toi and the DIIID and ITPA Energetic Particle Groups. | ITPA Joint Experiment : | Yes |
| Description: | The role that fast-ions play on ELM stabilization by RMPs through an eventual fast-ion redistribution/loss will be investigated. Time-resolved energy and pitch angle measurements of fast-ion losses induced by ELMs and RMPs will provide a more comprehensive picture of the interplay between fast ions and edge perturbations, improving our ability to make reliable predictions for future devices. The DIII-D expertise and advanced heating schemes (counter-current, on- and off-axis NBI and ECRH), diagnostics (to precisely measure mode structure and fast-ion redistribution and loss) and external coils (to produce a rich variety of RMPs) will be crucial for the success of these joint experiments. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The aim of this work is to systematically compare the measured and predicted loss of fast ions arising from known RMPs and current ELM MHD models in order to assess the role that fast-ions play in ELM stabilization by RMPs.
It is proposed to create as simple as possible experimental conditions in which an identified perturbation gives rise to measurable fast-ion losses; exploiting recent developments in diagnostics (of internal fluctuations and fast-ions) and machine capabilities (especially off-axis NBI and external coils). The experimental programme has three well defined steps: 1) ELMy H-mode in relatively low density plasmas with large off-axis fast-ion population and plasma shapes optimized for fast-ion loss detection will be performed to study the sole effect of ELMs on edge fast-ion population. The workhorse of these experiments will be large, radially extended, type-I ELMs. A scan of the NBI injection energy, namely gyroradius and orbit width, will be useful to scan the pedestal and SOL region with lost ions. 2) L-mode experiments consistent with optimal external coil performance and large edge fast-ion population will be performed to understand the effect of RMPs and associated magnetic stochastic layers on fast-ion transport. Given this target scenario, we plan a scan in q_edge shear, coil current and phasing, as well as RMP helicity. Low toroidal magnetic field would be beneficial to increase the relative strength of the perturbation field vs. vacuum toroidal field. An NBI injection energy scan would also be very useful. 3) Finally, based on the experimental results obtained previously, we plan to study the fast-ion role on ELM stabilization by RMPs. We plan type-I ELM stabilization in H-mode with external heating given separately by on-axis, off-axis NBI and ECRH. The different heating schemes and external coil features available at AUG, DIII-D, JET and LHD, combined with their comprehensive suite of diagnostics, makes their participation in these Joint Experiments essential in order to get a throughout understanding of the basic physical mechanisms. | ||
| Background: | Inherent to high-confinement modes are steep edge pressure gradients which suffer from spontaneous energy release in the form of edge MHD perturbations, ELMs. In present-day fusion devices, ELM-induced heat loads are still acceptable but somewhat uncertain extrapolations to ITER indicate causes for concern. This has motivated the search for plasma regimes with small ELMs, as well as for methods, such as pellets or RMPs, to control the occurrence and size of the ELMs.
While the ELM impact on the global stored energy is empirically rather well established, less is known about how their MHD structures influence energetic particle confinement. Even less is known about the interplay between fast ions and RMPs and its impact on the final ELM stabilization. In fact, in H-mode discharges with ECRH only, so far ELM stabilization by RMPs has not been achieved. Recent breakthroughs in fluctuation and fast-ion diagnosis techniques may shed some light on the physical mechanisms responsible for the fast-ion transport induced by ELMs and their possible stabilisation/mitigation techniques. In particular, time-resolved energy and pitch angle measurements of ELM-induced fast-ion losses from AUG and LHD have revealed that ELMs might cause severe fast-ion losses. ELM-induced fast-ion losses were found to show different time scales correlated with SOL density changes and ELM MHD sub-structures. | ||
| Resource Requirements: | All neutral beams and I-coils. | ||
| Diagnostic Requirements: | FILDs, NPA, FIDA spectrometers and fast-camera, ECE, neutron and BES. IR-camera? | ||
| Analysis Requirements: | In the framework of this new ITPA proposal for joint experiments (AUG, DIII-D, JET and LHD), the collected data will be compared with calculations by several codes including SPIRAL, OFMC and ASCOT. | ||
| Other Requirements: | -- | ||
| Title | 89: Commissioning of new pellet injection port and pellet size for ELM triggering | ||
| Name: | Baylor baylorlr@ornl.gov | Affiliation: | ORNL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | T. Jernigan, N. Commaux | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 90: Optimizationf of Plasma Performance with Pellet ELM pacing at 30 Hz | ||
| Name: | Baylor baylorlr@ornl.gov | Affiliation: | ORNL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | ELM Control Group, A. Loarte | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 91: Determination of Pellet ELM Pacing Synergy with HFS Pellet Fueling | ||
| Name: | Baylor baylorlr@ornl.gov | Affiliation: | ORNL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | ELM Control Group | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 92: Measurement of ECE resonance layer width using Correlation ECE | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | General IM | Presentation time: | Not requested |
| Co-Author(s): | M. E. Austin, R. Prater, T. Rhodes, G. Wang | ITPA Joint Experiment : | No |
| Description: | Validate fundamental EC emission and absorption physics by comparing the predicted width of the EC resonance layer against measurements of radial correlation length of electron temperature fluctuations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce simple plasmas (L-mode) with a variety of emission layer widths. This can be done by varying the electron temperature and density using different combinations of heating methods: ECH, FW and beams. In these plasmas the CECE system will use two tunable narrowband IF filters to make measurements of correlated Te fluctuations under two very different conditions: 1) the IF filters are within the same emission layer and 2) the IF filters are in different emission layers. Make measurements of the correlation length of electron temperature fluctuations in many emission layer width conditions. With this data set, an integrated modeling effort that includes EC emission and absorption calculations and GYRO turbulence calculations with synthetic diagnostic modeling can be undertaken determine if the measured correlation lengths are limited by the width of the emission layer, as has been suggested by observations in the past [Watts, White], or if the measurement is truly a measure of the turbulence correlation length. | ||
| Background: | The emission layer width presents the fundamental limitation on radial resolution of Te profiles measured with ECE diagnostics. On ITER, predictions for the emission layer width have been crucial for determining whether or not to use O-mode or X-mode ECE systems. Yet in a tokamak the emission layer width has not been directly measured. Interestingly, measurements of electron temperature turbulence with CECE diagnostics have shown that the measured radial correlation length of the turbulence tracks very closely the calculated emission layer width instead of the actual radial correlation length of turbulence [Watts]. These observations have not been confirmed with a carefully controlled experiment. Specifically, to test this one needs at least two extreme values of the emission layer (one narrow and one wide) and a calibrated CECE diagnostic and careful measurements of the radial correlation length of turbulence. If indeed the measured radial correlation length of temperature fluctuations reflects the emission layer width rather than the true underlying radial correlation length predicted by GYRO+syn. diag., then the CECE diagnostic is actually making a measurement of the emission layer width rather than the correlation length of the turbulence. The measured emission layer width can be used to validate fundamental EC emission and absorption theory, thus providing more confidence in calculations of the EC resonance layer. This information will also be crucial for developing more sophisticated CECE synthetic diagnostics that are used to validate gyrokinetic codes because this experiment will provide new data useful for understanding the response of the CECE measurements to kr, which has a strong impact on the predicted fluctuation level (Holland)
Refs: 1) Watts, FUSION SCIENCE AND TECHNOLOGY VOL. 52 AUG. 2007 2) White, PhD thesis, UCLA, 2008 3) Holland, ļ»æļ»æļ»æPHYSICS OF PLASMAS, 16, 052301 (2009) | ||
| Resource Requirements: | ECH, FW, NBI: 330L, 30L, 210R | ||
| Diagnostic Requirements: | Standard diagnostics (magnetics, TS, 32 channel ECE radiometer, etc. )
CECE (calibrated against Michelson/ECE radiometer) Michelson, oblique ECE (if available), ECEI, DBS and reflectometer for radial correlation length measurements of ne-tilde, BES(if 150L used) | ||
| Analysis Requirements: | gaprofiles, GENRAY, CQL3D, ECESIM, ONETWO, TGLF, GYRO, synthetic CECE diagnostics. | ||
| Other Requirements: | -- | ||
| Title | 93: Effect of mitigating gas density on RE decay rate | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | N. Commaux | ITPA Joint Experiment : | No |
| Description: | Assess the effect of varying mitigating high-Z gas density upon RE beam decay rate. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will begin with the limited seed plasma that reliably produced RE beams during the 2010 campaign. This plasma will be terminated with varying levels of Ar gas puff. The subsequent decay rate of the resulting RE beam current will be measured and recorded as a function of injected gas density. | ||
| Background: | Initial evidence from the 2010 run campaign indicates that high-Z gases can increase the decay rate of an RE beam current. However, our ability to vary the density of the mitigating gas has been very limited, as only large Ar killer pellets have historically been capable of producing RE beams. This is not universal to all devices. In particular, Tore Supra can reliably create RE beams in limited plasma with minimal high-Z gas input. D3D found in 2010 that a limited plasma produces RE beams much more reliably than an diverted plasma, but no attempt was made to induce RE beams with anything other than large Ar pellets. We propose to produce RE beams using minimal impurity gas puffing, similar to Tore Supra, in order to enable a wide variation in the high-Z gas density and view its effect on the RE current decay rate. | ||
| Resource Requirements: | 4x ECH, MGI, 30L (MSE), Pellet injector, MGI, Ar regular valve, 0.5 DAY | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, scintillators, Fast camera | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 94: Compatibility between pellet fueling and RMP ELM suppression | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, T. E. Evans | ITPA Joint Experiment : | Yes |
| Description: | The heat load from ELMs on the plasma facing components is an important issue for the design of ITER. A new technique using non axisymetric resonant magnetic perturbations to suppress the ELMs has been successfully tested on DIII-D. But some experiments combining pellet fueling and RMP on DIII-D have shown that pellet injection can temporarily disturb the ELM suppression. Some cases of individual pellets show no sign of immediate ELM triggering. Pellet injection is planned to be the main fueling method on ITER. Therefore the combination of RMP ELM suppression and pellet fueling is very likely in ITER plasmas. It is then important to understand the physics of the interaction in order to maintain reliably the ELM suppression during the pellet fueled high density plasmas. This is a proposed ITPA joint experiment (PEP-35) for RMP ELM suppression. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Apply the n=3 ā??usualā?? ELM suppression RMP field on ELMing H modes to obtain high quality ELM suppression. Add individual HFS pellets injection (low frequency ~1Hz) and try to change the plasma parameters to suppress the ELM bursts that pellet injection can trigger (injected torque, shaping, strike point position to increase the pumping efficiencyā?¦). Then increase the pellet injection frequency to test the influence of the pellet injection frequency. One thing that has to be thoroughly reproduced is the front end of the scenario to get a reproducible flattop initial density. One possibility is modifying the injected torque since the rotation damping induced by the pellet injection could be an effect explaining the loss of the ELM suppression in some cases. The influence of the q95 should be tested (through a q95 scan) also inside the resonant window. It will be also interesting to change the pellet parameters to test the influence of the pellet penetration and injection location on the occurrence of ELM bursts after a pellet injection. The new upgrade of the pellet injector allowing shallower injections (slower and smaller pellets) could provide new data on shallow pellet fueling. It is very important during these experiments to maintain as much as possible the initial electron density constant during the day in order to allow a reliable comparison between shots. | ||
| Background: | Several shots with non-optimized RMP ELM suppression have been proven sensitive to pellet fueling when the density is significantly increased. In these shots (for example 131466, 131467) the first pellet injected triggered a bifurcation back to ELMing H mode. Results from a preliminary experiment conducted last year show possible significant differences between LFS injections and HFS injections and a possible partial recovery from the density pump-out triggered by the application of the n=3 perturbation without a bifurcation to ELMing H mode. But due to the occurrence of lock-modes at the onset of the perturbation, the density levels between the different shots are very different making a direct comparison unreliable. It is important to understand what triggers this bifurcation and a recipe to avoid it. Cases exist with single pellet injection where ELMs are not triggered and the plasma maintains an ELM suppressed state. One of the possible explanations is maybe that the density pump-out is in fact necessary to the ELM suppression and thus recovering from it by injecting pellets force the plasma back to ELMing H mode. | ||
| Resource Requirements: | 1 day experiment with Co/Cn NBI, Bt at 2.15T (for ECE), pellet injector, I coils with current capacity up to 7kA | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 95: Study of the pellet injection particle deposition profile | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | General IP | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, P. B. Parks | ITPA Joint Experiment : | No |
| Description: | Pellet injection is planned to be the main fueling method on ITER. The very high density and temperature of the plasma will not allow a deep penetration of the pellet. But this shallow penetration is expected to be compensated a strong curvature and ļ??B drift which will deposit the particles much deeper in the plasma for the pellets injected from the inner wall. A relation between the pellet deposition profile and the q profile has been observed on several machines. This effect could have important consequences on the particle deposition profile on ITER. In order to evaluate the consequences and the efficiency of the fueling on ITER, a good understanding of this drift effect is important. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Inject HFS pellets in ELMing H modes and changing slowly the penetration of the pellet with respect to integer q surfaces (penetration inside or outside of q=3). This can be done by changing the pellet speed (from 80 m/s up to 200 m/s and changing the q profile using Ip, Bt and NBI current drive to change the position of q=3 in the plasma. The analysis of the pellet deposition profile will be used to determine how the q profile affects the fast drift of the pellet mass. The new upgrade of the pellet injector allowing shallower injections (slower and smaller pellets) could provide new data on shallow pellet fueling. | ||
| Background: | Using the pellet injection data from DIII-D, a relation has been found between the pellet deposition profile and the integer q surfaces. But there is a strong variation of plasma and pellet injection conditions for these shots (NBI power, Ip, Bt, H or L modeā?¦). A systematic study with controlled changes in the plasma parameters is still required to prove this effect and to evaluate its consequences on the fueling of ITER. | ||
| Resource Requirements: | 1 day experiment with Co/Cn NBI, fast reflectometry, 30 NBI line (for MSE), pellet injector, Thomson scattering, UCSD fast camera | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 96: ECH modulation experiment to search for radial variation of profile stiffness | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | J. DeBoo, C. C. Petty | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to explore the role of turbulent transport in determining observed stiffness of plasma profiles across the core-edge transition region ( 0.65 < rho < 0.85). ECH heat pulse modulation experiments can be used to study changes in electron heat diffusivity versus local electron temperature gradient in great detail at DIII-D, and have been used in the past to look for critical gradient behavior in L-mode (DeBoo 2005). Analysis of these type of ECH modulation experiments can be done with a critical gradient transport model (Garbet 2004, Asp 2007) to determine over what radial range the profiles exhibit stiffness. By comparing changes in turbulence with changes in inferred diffusivity across the core-edge transition region (0.65 < rho < 0.85), it will be possible to determine 1) Where, if at all, in this region are the profiles stiff? 2) Does the stiffness vary dramatically across this region, i.e. are profiles stiff in the core but not the edge 3) To what degree do measured changes in turbulence (used as markers for turbulent-transport) track the changes in the gradients and the diffusivities? | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Approach will be similar to ECH modulation experiments done by DeBoo, NF, 2005, but in contrast to that work that looked in the core plasma for critical gradient behavior, this new experiment will target the transition region nearer the edge.
The following recipe will be explored for the experimental approach: 1) inner wall limited and diverted plasmas 2) USN to avoid H-mode 3) early beam heating to delay sawteeth 4) plasmas that are shifted up/down to aid in varying the radial deposition of ECH as greatly as possible 5) two types of ECH modulation (DeBoo 2005) to monitor changes in gradients, fluxes, and turbulence across radial transition region. 6) Te scan using FW independent of NBI and the ECH modulation to reduce v_eff, increase Te/Ti and drive the TEM more strongly. The ITG/TEM balance may be key to achieving stiffness and in determining how stiffness varies with radius. With FW, it may be good to revisit search for stiffness at core radii from DeBoo 2005 (rho = 0.35, rho = 0.6). 7) using the 210R beam it may be possible to also vary rotation and track changes in stiffness (Mantica PRL) The radial variation in profile stiffness can be experimentally explored by 1) Measuring changes in the Te profiles during the ECH modulation experiment in the core-edge transition region (0.65 < rho < 0.85) 2) Using a simple critical gradient transport model and heat pulse propagation measurements to quantify the "stiffness" of the transport by plotting diffusivity versus local gradient 3) Measuring in great detail the changes in the equilibrium profiles and the turbulence across this region. | ||
| Background: | Past work (DIII-D, JET, Asdex Upgrade, Tore Supra) have studied profiles and transport analysis results in great detail. As reported by DeBoo 2005, no evidence in the core (rho < 0.6) of DIII-D L-mode plasmas for the existence of a critical gradient was found. But as reported by Garbet in 2004 there is a transition region between the core and the edge where the stiffness may come into play and there is an observed discontinuity in the experimentally inferred heat pulse diffusivity. Past fluctuation measurements from DIII-D have shown a break in slope of the fluctuation amplitudes near rho = 0.75 (White 2008), even though the local normalized density and temperature gradients are smoothly varying across this region. This indicates that possibly there is profile stiffness across that region as reported by Garbet. This motivates a resumed search for critical gradient behavior in L-mode plasmas beyond the core. Determining the boundaries between core and edge and the range of the transition region is critical for mapping out over what radial range the profiles are expected to be stiff in ITER, and therefore over what range of radii a model such as TGLF (which has been shown to be quite stiff (Kinsey IAEA10 )) can be used to predict transport. | ||
| Resource Requirements: | NB (330L/30L, 150 L, 210R), ECH, FW | ||
| Diagnostic Requirements: | standard diagnostics (TS, ECE, magnetics, etc., additionally as many turbulence diagnostics as possible, core and edge. | ||
| Analysis Requirements: | power balance, critical gradient models, TRANSP, ONETWO, fourier mode analysis (ECH modulation analysis), TGLF, GYRO, TGYRO, NEO | ||
| Other Requirements: | -- | ||
| Title | 97: Cross-diagnostic comparisons to verify turbulence amplitudes and cross-phase measurements | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | T. Rhodes, G. Wang, G. R. McKee, M. Austin, D. Rudakov | ITPA Joint Experiment : | No |
| Description: | By cross-correlating fluctuations measured with a variety of diagnostics (reflectometer, BES, CECE, 32-channel radiometer, Langmuir probes) it will be possible to verify measurement of fluctuation amplitudes and the cross-phase angle between temperature and density fluctuations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 0) CECE, BES< ECE radiometer, Langmuir probes all on the same digitizer
1) Compare ECE and probe measurements of electron temperature fluctuations in low-power density, Ohmic plasmas. Optical depth is expecetd to be marginal. Cross-correlate with reflectometer. 2) In NBI, sawtooth free L-Mode plasmas perform q-scan to align BES and CECE sample volumes on the same flux surface to cross-correlate BES/CECE and reflectometer/CECE to verify measurements of the turbulent density-temperature cross-phase angle 3) in L-mode or H-mode (hybrid?) plasmas measure the n-T cross-phase angle with BES/CECE and reflectometer/CECE associated with neoclassical tearing modes. q scan may be used to align BES and CECE | ||
| Background: | For comparing theory/experiment, detailed cross-diagnostics comparisons can be used to aid in the interpretation of fluctuation measurements. | ||
| Resource Requirements: | all available neutral beams | ||
| Diagnostic Requirements: | standard profile measurments, BES, CECE, reflectometer, Langmuir probes, MSE | ||
| Analysis Requirements: | fluctuation data analysis | ||
| Other Requirements: | BES, CECE, Langmuir probes and 32-channel ECE radiometer need to have at least some channels all on the same time base (same digitizer). | ||
| Title | 98: Low rotation, elevated q_min steady-state scenario discharges at high betaN | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Attempt to operate discharges with q_min >1.5, betaN >3.5 and q95 = 5-7 at less than the maximum attainable toroidal rotation. Do this by adding counter-injection beams to the standard recipe for steady-state scenario discharges. Assess the effect on the discharge stability, transport and noninductive current fraction of the addition of counter-injection beams. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce a steady-state scenario discharge following the standard procedure and gradually add the two available counter-injection beams. Explore various methods and timing for adding the beams in order to avoid instability. It is likely that at reduced rotation, tearing modes will be a significant problem. If possible, use direct stabilization with ECCD to preemptively suppress tearing modes (possibly developed in a separately proposed experiment). | ||
| Background: | Steady-state scenario discharges operate at the highest betaN possible in order to maximize fBS. This requires all of the co-injection beams available at DIII-D and results in large toroidal rotation velocities. At these high values of betaN, n = 1 tearing modes are often observed, providing a limit to discharge performance. In the past, when an attempt has been made to add even a small fraction of counter-injection beams in order to obtain additional MSE data, the counter injection beams were almost always removed after only a few shots because of the worry that they make the discharge more susceptible to tearing modes. A serious attempt to add significant counter-injection beam power has never been made. If the goal of the experiment is to maintain high betaN while reducing the rotation, it is likely that all of the co-injection beams will still be needed. That means that the extent to which the rotation can be reduced will be limited. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 99: Use the off-axis beams in the high q_min steady-state scenario | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make use of off-axis aiming of the 150 degree beams to broaden the NBCD profile and the fast ion pressure profile in high betaN, elevated q_min steady-state scenario discharges in order to obtain performance improved over the discharges produced in 2008-2010. The overall goals are a) stationary, q_min >1.5 with fNI = 1 for duration greater than tauR; b) increased bootstrap current fraction obtained by increasing betaN. A more detailed list of goals is the following: 1) provide the additional off-axis current drive that is required to match the noninductive current profile to the current profile for a weak shear q profile with q_min about 1.5 and fNI = 1; 2) broaden the profile of fast ions so that the total pressure profile is broader and the betaN limit is increased; 3) evaluate whether a stationary q profile with q_min >1.5 can be maintained; 4) test the ability to maintain q_min >2 for longer duration and at higher beam injection power than is possible with all on-axis injection; 5) evaluate fast ion transport, thermal energy transport and compare with on-axis injection; 6) test the high betaN stability to tearing modes and fast ion driven modes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The most off-axis current drive will be present in reverse Bt (a.k.a. positive Bt). So, first any issues with reproducing the high-performance steady-state scenario discharges (such as 133103) in reverse Bt will need to be addressed. Then, with the 150 beams aimed as far off-axis as possible, scan the value of q_min at the beginning of the high betaN phase between 1.5 and 2.5. Evaluate the ability to maintain a stationary value of q_min. Adjust q95 and Bt in order to tune the discharge for fNI = 1. Adjusting q95 changes the total noninductive current required, and adjusting Bt changes the beam power required to obtain a fixed value of betaN with a scaling such that fNI increases with Bt as a result of increasing NBCD. Maximize betaN. Evaluate the relation between the achievable betaN and the total pressure peaking factor. Make detailed FIDA measurements in order to look for differences between the predicted and observed fast ion pressure profiles. | ||
| Background: | A careful study of the dependence of the noninductive current densities on the q profile performed in 2009 established that q95 in the range 6-6.8 is required in order to make access to fNI = 1 feasible at the values of betaN (3.5-3.8) obtained thus far in steady-state scenario discharges. In those discharges there was typically noninductive current overdrive on the axis that pushed q_min to lower than desired values. Off-axis, there was a noninductive current deficit of about 20 A/cm2 for a discharge with q_min about 1.5. By moving enough of the beam current drive off-axis, these problems can be corrected. Also, the pressure peaking factor in these discharges was about 3.3-3.5, consistent with the calculated betaN stability limits near 4. The thermal pressure peaking factor, however, was only 2.6. If the fast ion pressure peaking can be reduced by depositing enough of the fast ions off-axis, then the total pressure peaking factor could be reduced and the betaN limit increased. With off-axis beam current drive and total noninductive driven current similar to what was produced in 2009, it should be possible to create a discharge with a stationary q profile with q_min near 1.5 at q95 at in the range 6.5-7. If betaN can be increased or density and temperature profiles otherwise modified to produce higher bootstrap current density, fNI = 1 at q95 near 6 should be possible. Depending on the profiles of noninductive current and the beam power required to produce the target betaN, the stationary q_min could be well above 1.5. At the least, it should be possible with off-axis beam injection to reduce the rate of decay of q_min in discharges which start the high betaN phase with q_min >2. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | A well verified, reverse Bt MSE calibration is essential. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 100: Fast ion diffusion in high betaN, steady-state scenario discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Steady State) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Produce dedicated discharges for a comparison of predicted and measured fast ion density profiles in the steady-state scenario at high betaN. Optimize the discharges for all of the fast ion diagnostics including FIDA and the fast ion loss detectors. Compare losses from on-axis and off-axis beams. Make use of techniques such as short beam pulses to look at the rise and fall times of the neutron signals. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | See the "Description" paragraph. | ||
| Background: | Analysis of the high betaN steady-state scenario discharges requires calculation of the noninductive current density profiles from models. In order to calculate the neutral beam current density profile, the beam deposition profile must be calculated and this calculated profile must be assumed to be correct. However, typically if the measured thermal pressure and the calculated fast ion pressure profiles are summed, the on-axis pressure and the total stored energy are inconsistent with equilibrium reconstructions using EFIT with magnetics and MSE data. It is necessary to assume an anomalous fast ion diffusion profile in order to obtain agreement. It is highly desirable to evaluate whether this technique of using an anomalous diffusion profile actually produces fast ion density profiles from the model that match the experiment. Otherwise, analysis of the steady-state scenario discharges and calculation of the noninductive current fraction involves significant uncertainty because the total neutral beam driven current is not well known.
The problem is particularly significant at high betaN. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 101: Maximize betaN and fBS with direct tearing mode stabilization in steady-state scenario discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The primary goal is to produce high betaN, fNI = 1 discharges for long duration. Part of the approach to achieving this goal would be to develop the capability to preemptively stabilize tearing modes in steady-state scenario discharges using ECCD narrowly deposited at the resonant surface. With this capability activated, push the betaN to the maximum attainable in order to study the limit determined by modes other than resistive tearing modes and in order to maximize the bootstrap current fraction. With tearing modes stabilized, maximize the duration of fNI = 1 at high betaN. With some or all of the gyrotrons dedicated to narrowly deposited ECCD, there will be reduced or zero current density in the broadly deposited noninductive current density profile normally provided by ECCD. Use the off-axis neutral beams to replace the broad ECCD profile normally used. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dedicate the first portion of the experiment to development of the techniques for direct tearing mode stabilization using ECCD. Apply the standard methods of search and suppress and active tracking. Preferably, use real-time aiming of the gyrotron steering mirrors rather than plasma motion or toroidal field changes in order to adjust the ECCD deposition profile to lie on the resonant surface. Standard steady-state scenario discharges at high betaN would be used as the targets rather than a custom discharge with a reproducible tearing mode. It is highly likely that tearing modes will appear often in the high betaN discharges. Once there is confidence that tearing modes can be avoided, the primary goal of the experiment would be to produce fNI = 1 discharges for long duration. The approach would be to maximize fBS by increasing betaN. The direct stabilization would be counted on to maintain tearing mode stability both as betaN is increased and as the current profile evolves during the increased duration of the discharge. | ||
| Background: | Tearing modes (2/1 and/or 3/1) impose the betaN stability limit and limit to discharge duration in steady-state scenario discharges. Progress toward developing fNI = 1 discharges has been very slow because most discharges are interrupted by the growth of a tearing mode. It is essential that we do something to eliminate these modes. The general approach has been to either assume that the resonant surfaces will not be present if q_min is maintained at a high enough value or that tearing modes can be avoided using broadly deposited ECCD. Tearing modes have not been reproducibly avoided using these techniques, though. A proven technique for avoiding tearing modes is direct stabilization with ECCD narrowly deposited at the resonant surface. Very precise aiming is required but searching techniques to find the correct aiming have been developed. This technique may be very valuable in steady-state scenario discharges at high betaN, but the work to test this has never been done. The hesitation to use the direct stabilization technique has stemmed from the need to use the available gyrotron power to provide a broad profile of off-axis noninductive current density in order to produce a total noninductive current profile compatible with stationary, fNI = 1 operation. However, now that the off-axis beams are available, it may be possible to divert some or all of the gyrotrons to tearing mode stabilization and still have enough off-axis noninductive current drive. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 102: Pellet fueling for L mode phase density build-up | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | General IP | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, E. A. Unterberg, G. L. Jackson, D. A. Humphreys | ITPA Joint Experiment : | No |
| Description: | The startup scenario in ITER has been studied thoroughly during the last few years. Because of the high heat flux on PFCs, it is not possible to maintain the plasma limited on the PFC for more than a few seconds. The control of the li during the initial phase is also very important. But something that has been studied is a way to reduce the particle inventory during this initial phase. Previous experiments have shown that most of the particle inventory originates from the pre-fill and L mode density ramp-up phase. The problem is that on ITER, the particle confinement time will be so high that controlling precisely the wall recycling is very important to keep the density under control. It has also been observed that a high wall recycling has a detrimental effect on the global performance of the discharge (confinement, MHD stabilityā?¦) Using a higher fueling efficiency scheme like pellet injection as soon as possible in the discharge could allow lowering globally the particle inventory. It could also provide another way to control the li during the startup by avoiding the edge cooling by the gas puff thus keeping the li at a lower and more controllable level that could also be used for advanced or non inductive scenarios. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using the ITER startup scenario developed at DIII-D during the last years by Gary Jackson, the main idea would be to apply low penetration pellet injection as soon as possible during the startup. The challenge would to inject the pellets very early in the discharge to shutdown the gas puff but not too early so that the plasma is too cold to ablate the pellet or to trigger an MHD instability due to the additional cooling from the pellet. It would be interesting to adjust the pellet parameters to test which has the best efficiency and the additional heating scheme to help the plasma survive the injection. One of the possible issues is to run into a Pellet Enhanced Performance mode (PEP) that would create an internal transport barrier. To avoid that problem the upgrade of the pellet injector allowing smaller pellets will help because only deep penetrations usually trigger this kind of mode. The main data gathered during these experiments would be a measure of the wall retention during these experiments, global discharge performance changes, current density profile,ā?¦ | ||
| Background: | The 2008-2009 campaign was successful at demonstrating a viable ITER startup scenario on DIII-D. It was possible to reproduce the different requirements (early diversion, big bore startup, current ramp-up feedback controlled to obtain a target liā?¦) but the important particle inventory in this early phase can generate problems later on in the discharge on ITER. It is therefore important to extend these studies to obtain a low particle inventory startup scenario to be able to reduce as much as possible the particle inventory thus the recycling during the rest of the discharge. | ||
| Resource Requirements: | 1 day experiment with Co/Cn NBI, pellet injector, fast wave heating, ECH heating, baking of the vessel to reduce the particle inventory, particle balance experiments setup for the cryo-pumpsā?¦ | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 103: Pellet fueling in high Greenwald fraction discharges | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | General IP | Presentation time: | Requested |
| Co-Author(s): | T. C. Jernigan, L. R. Baylor, E. A. Unterberg | ITPA Joint Experiment : | No |
| Description: | The fusion performance expected on ITER has been calculated assuming an energy confinement in agreement with the H98 empirical confinement scaling. ITER plasmas are planned to be operated at significant Greenwald fraction (0.85 for the 400 MW inductive baseline scenario and 0.95 for the 500MW inductive scenario) with H98=1 in order to achieve high Q. But it has been proven on several tokamaks that such high density conditions can be difficult to obtain using gas puffing and can show confinement degradation when compared to the H98 scaling. DIII-D is uniquely capable of producing high performance discharges with the ITER shape. It is therefore important to test the consequences of such high density operation on the plasma performance of the ITER scenario using this unique characteristic. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Obtain the ITER baseline scenario plasmas. Compare the behavior of the plasma when using gas puffing or pellet injection (HFS shallow injection) in order to increase the density to a Greenwald fraction of ~0.9. If possible, try to improve the pumping efficiency by small adjustments of the strike points. Use the pellets also as a probe to determine the particle transport coefficients. Use a combination of NBI and RF heating to maintain a strong central heat deposition at high density. NBI alone at high density will lead to off axis heating, which will result in reduced stored energy and lower confinement than the H98 scaling. USing a lower plasma current and magnetic field would allow lowering the Greenwald limit thus operating with better NBI efficiency. Using the ITER baseline scenario could also improve the knowledge concerning how well the different parameters of an ITER discharge can be reproduced simultaneously but this experiment can also be carried out using less challenging scanarios. | ||
| Background: | The 2008 campaign was successful at demonstrating several ITER operational scenarios on DIII-D. It was possible to reproduce the shaping, the beta, the collisionality... but the density was not at the Greenwald fraction expected on ITER (it is difficult to achieve both ITER relevant collisionality and Greenwald fraction on DIII-D). It is therefore important to extend these studies to the high density regime to understand the consequences of high density on the confinement. | ||
| Resource Requirements: | 1 day experiment with Co/Cn NBI, pellet injector, fast wave heating, ECH heating | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 104: Experimental identification of the plasma response to off-axis NBI | ||
| Name: | Moreau didier.moreau@cea.fr | Affiliation: | CEA Cadarache |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University), D. Mazon (CEA) | ITPA Joint Experiment : | Yes |
| Description: | The objective of this experiment is to experimentally characterize the plasma response to off-axis NBI on DIII-D. It requires only a small number of shots (4-5) with off-axis NBI and is a simple extension to the experimental model identification that was performed during the last experimental campaign. These 2009 experiments allowed very good control-oriented models to be obtained to describe the response of the poloidal flux and toroidal rotation profiles to 5 actuators, namely co-current NBI, counter-current NBI, balanced NBI, ECCD and loop voltage (see reference [1]).
This new experiment will be very useful for the evaluation of off-axis NBI physics as they will provide an experimentally measured space-time plasma response that can be compared with theory-based modeling. The off-axis current drive capability of the DIII-D NBI system will also provide additional flexibility for controlling the current profile during ramp-up and/or during the high performance phase of AT steady state discharges, and also for the combined control of the current profile, Ī²N and/or the toroidal rotation profile. The proposed experiment will provide essential data to complete the 2009 model in order to use off-axis NBI as an extra actuator for model-based control applications. This would therefore allow future real-time control experiments to be done while making use of the best heating and current drive mix available on DIII-D (see other proposals, e.g. #122, on "Integrated and Model-Based Control "). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental plan will be the same as in 2009, but with modulations of the off-axis beams only (4-5 shots).
First we shall reproduce shot #140090, the reference shot without power modulations (2 shots). Then, up to t=2.5 s, all subsequent discharges will be similar to the reference one (1.8 Tesla, Ī²N-controlled AT scenario, at a central plasma density, ne0 = 3.5 x 1019 m-3 and plasma current, Ip = 0.9 MA). At t=2.5 s (i.e. after a 1 s current flat top), in all discharges, the Vloop control mode (i.e. the use of Vloop as a control actuator) is enabled and the Ip and Ī²N controls are disabled in order to avoid feedback in the response data. Between t=2.5 s and t=7 s, modulations of the off-axis beams will be applied, first 150L only, then 150R and both. The modulation waveforms are determined in advance and uploaded into "futureshot" files that can be readily used during the experiment. | ||
| Background: | The algorithms that are used to numerically identify the various elements of a semi-empirical plasma response model using experimental data have been developed for model-based control purposes and used successfully on JET, JT-60U and DIII-D [1-2]. The models relate a set of (machine-dependent) input parameters or actuators (e.g. H&CD powers) to measured output profiles for which control will be needed, namely, the current density (or safety factor) profile, which characterizes the magnetic state of the plasma, and one or several fluid/kinetic parameters and profiles (Ī²N, plasma rotation velocity, ion and/or electron temperature, etc ...).
A model-based controller can then use all the available heating and current drive (H&CD) systems in an optimal way to regulate the evolution of the plasma profiles [2]. The development of such integrated control-oriented models is requested by the ITPA under the IOS group (Joint Experiment IOS-6.1). References: [1] D. Moreau et al., "Plasma Models for Real-Time Control of Advanced Tokamak Scenarios", submitted to Nuclear Fusion (see 2010 IAEA FEC, paper EXW/P2-07). [2] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. | ||
| Resource Requirements: | NBI with various power waveforms is needed including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. This experiment requires that commissioning of the off-axis beams has been completed (full power modulations). Other additional heating and current drive systems from 6 gyrotrons will also be required although not essential. The discharges are to be run partly in the loop voltage control mode (PCS). The system identification requires the availability of the profile data from MATLAB. | ||
| Diagnostic Requirements: | Magnetic measurements, MSE, CER, Thomson. Equilibrium reconstruction including the q-profile (RTEFIT2) are essential, and measurements of the density profile, toroidal rotation profile, as well as ion and electron temperature profiles, are required. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 105: Sorting out effect of elongation vs limited/diverted configuration | ||
| Name: | Granetz granetz@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | Eric Hollmann, Nick Eidietis, Val Izzo, John Wesley, Dave Humphreys, Alex James | ITPA Joint Experiment : | No |
| Description: | Determine which shaping effect is more important for having runaways in disruptions: low elongation or limited configuration. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop a diverted equilibrium with as low an elongation as possible, preferably close to circular. Trigger disruptions with Ar killer pellets and determine statistics of runaway presence in the current quench. Then develop a limited configuration with a high elongation, preferably as high as a typical diverted equilibrium, then trigger disruptions with Ar pellets and determine statistics of runaway presence. C-Mod has plans to do a similar experiment (with LH instead of Ar killer pellets). | ||
| Background: | Recent experiments on DIII-D and C-Mod have confirmed the hypothesis that low-elongation limited plasmas have a much higher probability of obtaining runaways in the current quench compared to normal-elongation diverted plasmas. This hypothesis was based on comparisons among many tokamaks around the world. But these two shaping effects (elongation and limited/diverted configuration) have been carried out together so far. We propose trying to determine whether one of the two effects is more important. | ||
| Resource Requirements: | Ar pellet injector, development of unusual magnetic configurations | ||
| Diagnostic Requirements: | HXR/scintillators, photo-neutron detector, synchrotron imaging | ||
| Analysis Requirements: | NIMROD w/RE physics included | ||
| Other Requirements: | We could probably turn this into a formal ITPA joint experiment (MDC-16) w/C-Mod and others | ||
| Title | 106: 150 Off-axis Beam Characterization and Check-out | ||
| Name: | Van Zeeland vanzeeland@fusion.gat.com | Affiliation: | GA |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | W.W. Heidbrink, J.M. Park, B. Grierson, D. Pace, J.H. Yu | ITPA Joint Experiment : | No |
| Description: | The primary goal of this experiment is to carry out initial operation of the 150 off-axis beam, characterize it, make sure we have the necessary data for modeling it in NUBEAM, and make an initial comparison with modeling of the beam deposition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The beam divergence, vertical profile, and actual tilted trajectory will be obtained by injecting it into helium gas and imaging the resulting beam emission with the fast framing camera, as was done for the 30L beam in previous experiments (3 shots ea. voltage). This will rely on installation of the re-entrant view on the 225 midplane flange. Additionally, the temporal evolution of the species mix and response of the beam voltage to modulation will be measured with a spectrometer viewing the 150 beam if one can be installed in time. Injection into an L-mode plasma with a density ramp will be used to vary the beam attenuation and obtain a check on the deposition for comparison to modeling. 3 discharges will be carried out at each voltage: 1 with 150L only, 1 with 150R only, and 1 with both 150 beams (3 shots ea. Voltage). Measurements of the beam emission will be made with the fast camera as for beam into gas, the confined fast ions will be measured with the various fida diagnostics, and the promptly lost ions will be measured by BILD and FILD. The end of these discharges will have short blips of the 150 beams to measure the neutron emission (which is sensitive to the largest energy ions) for a comparison of the full energy beam mix at the two different voltages as well as the neutron decay rate.
These shots should be repeated at maximum and minimum tilt as well as an intermediate angle. | ||
| Background: | The tilted 150 beam is a major upgrade to the device with the purpose of driving off axis current. Key to being able to predict and assess the achieved performance is having the beam properly represented in our modeling codes. The primary focus of the proposed experiment is obtaining accurate measurements of the beam profile, divergence, species mix, and voltage dependencies of these parameters. The beam profile is not expected to be the same as previous campaigns due to a new mask, different focusing, and new steering that may cause clipping of the beam. Errors in estimates of these parameters will cause errors in our predictions of the beam ion profile, current drive, etc. The secondary major focus of this experiment is to make an initial assessment of the beam ion deposition. If carried out early in the campaign, this experiment will also provide us an opportunity to make initial measurements with several beam related diagnostics and optimize these measurements before dedicated experiments to assess current drive are executed. | ||
| Resource Requirements: | 150, 330, 30 beams | ||
| Diagnostic Requirements: | Imaging view of 150 beam and ideally a d-alpha spectrometer with Doppler shifted view and ~ 1 ms time response viewing the 150 beam. | ||
| Analysis Requirements: | TRANSP and ONETWO | ||
| Other Requirements: | |||
| Title | 107: Controlled loss of superthermal ions and the H-mode edge (pedestal, ELMs, transitions) | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | Pedestal Structure | Presentation time: | Requested |
| Co-Author(s): | J. deGrassie, B. Grierson | ITPA Joint Experiment : | No |
| Description: | Understand the physics connection between loss of superthermal ions at the plasma edge and the H-mode edge barrier/pedestal, and develop methods for controlling the edge physics by regulating this ion loss. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. The connection between edge ion loss and the H-mode edge is supported by several observations. Among these are: the ubiquitous presence of the Er well, which is a manifestation of an ion-poor region at the edge; the recent observation that the profile of the velocity (Mach number) near the separatrix is similar to the computed profile of ion orbit losses; and the well-known dependence of H-mode threshold on x-point height.
b. With the off-axis beams it will be possible to modify and modulate the superthermal ion edge loss. Operation at the lowest accessible energy (~30-40 keV) with both current and field reversed will maximize the out-going trapped ion orbits. The loss can be modulated by changing the plasma shape, outer gap, and x-point height. Particularly important is that changes in the superthermal ion population at the edge can be monitored using edge CER. c. Determination of a link between ion edge loss and the H-mode edge provides a prospect for -- control of the Er well -- control of L-H and H-L transitions -- modification and/or pacing of ELMs -- modification of pedestal parameters -- control of the edge rotation/torque. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | also submitted to Alternative Techniques for ELM Control | ||
| Title | 108: Model-based control of the current profile for steady state scenarios | ||
| Name: | Moreau didier.moreau@cea.fr | Affiliation: | CEA Cadarache |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University), D. Mazon (CEA) | ITPA Joint Experiment : | Yes |
| Description: | This proposal aims at demonstrating model-based current profile control in AT operation scenarios on DIII-D. The goal is to apply the control as early as possible during the ramp-up phase in order to obtain, in a reproducible manner, various requested target q-profiles for the high-Ī²N phase of the advanced scenarios. The ability to vary the requested current profile will also be very interesting for physics studies in which it plays an essential role, because it takes a long and precious experimental time to adjust by trial and error the various actuator waveforms in order to reach a particular goal.
The control algorithm has been previously developed and validated on JET [1] (see background). In this DIII-D experiment, the actuators are (i) co-current NBI power, (ii) counter-current NBI power, (iii) balanced NBI power, (iv) total ECCD power from all gyrotrons in a fixed off-axis current drive configuration, and (v) loop voltage. Off-axis NBI could possibly be included as an additional actuator for better flexibility in the profiles, depending on the execution of the experimental proposal # 118 ("Experimental identification of the plasma response to off-axis NBI"). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The demonstration of adequate current profile control will require half a day to possibly one day, after some dedicated tests of the algorithm in the PCS have been conducted in a couple of short (2 hours) preliminary sessions.
First we shall reproduce shot #140090, the reference shot of the system identification series (1.8 Tesla, Ī²N-controlled AT scenario, at a central plasma density, ne0 = 3.5 x 1019 m-3 and plasma current, Ip = 0.9 MA). As in the identification shots, the Ip and Ī²N controls will be disabled and Vloop control enabled at t=2.5 s (i.e. after a 1 s current flat top). On this reference shot, the same actuator waveforms as for #140090, including Vloop, will be run in open-loop from t=2.5 s. Then we shall try to obtain the same current profile at t=2.5 s but replacing the Ip / Ī²N controls by our profile controller earlier in time (e.g. 2 s, 1.5 s, 1 s, ...), thus gradually extending the duration of closed-loop operation forward, until the earliest time for reaching the target profile (end of ramp-up) is found. The same experiments will then be repeated while varying the requested target profile and, finally, closed-loop operation will be extended beyond t=2.5 s in the aim of slowing down the evolution of the current profile or maintaining it quasi-steady for longer periods. | ||
| Background: | Real-time control of the plasma current profile is important to achieve stable and reproducible operation of tokamaks in the advanced steady state regime. A multi-variable approach based on a semi-empirical dynamical plasma model has been proposed in which the controller uses a combination of the available heating and current drive systems, including the external loop voltage, in an optimal way to control the evolution of the plasma parameters and profiles [1]. The controller design uses singular perturbation techniques and is quite generic (i.e. independent of the tokamak).
The control-oriented (semi-empirical) model to be used to determine the controller matrices has been obtained from system identification experiments performed on DIII-D in 2009. The model was shown to provide excellent fits to the experimental data [2], not only during the phase when the system identification was performed (t > 2.5 s) but also during current ramp-up (from t = 0.3 s). This includes shots that were not used for model identification, in particular those in which an n=1 mode was present, and therefore shows the robustness of the model for control applications. The development and experimental tests of such control methods is requested by the ITPA under the IOS group. References: [1] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. [2] D. Moreau et al., "Plasma Models for Real-Time Control of Advanced Tokamak Scenarios", submitted to Nuclear Fusion (see 2010 IAEA FEC, paper EXW/P2-07). | ||
| Resource Requirements: | NBI at full power is needed and with waveforms generated in real-time by the PCS, including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. Full power ECCD from 6 gyrotrons will also be required. Note: If the experiment proposal #118 has not been executed ("Experimental identification of the plasma response to off-axis NBI"), the 150Ā° beams should aim on-axis and should be able to provide the same input characteristics (geometry, voltage, etc ...) as in November 2009. | ||
| Diagnostic Requirements: | Real-time magnetic measurements, MSE and equilibrium reconstruction including the poloidal flux and the q-profile (RTEFIT2) are essential. Measurements of the density profile, toroidal rotation profile, as well as ion and electron temperature profiles are also required for analysis, not necessarily in real time. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 109: Controlled loss of superthermal ions and the H-mode edge (pedestal, ELMs, transitions) [DUP 107] | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | J. deGrassie, B. Grierson | ITPA Joint Experiment : | No |
| Description: | Understand the physics connection between loss of superthermal ions at the plasma edge and the H-mode edge barrier/pedestal, and develop methods for controlling the edge physics by regulating this ion loss. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. The connection between edge ion loss and the H-mode edge is supported by several observations. Among these are: the ubiquitous presence of the Er well, which is a manifestation of an ion-poor region at the edge; the recent observation that the profile of the velocity (Mach number) near the separatrix is similar to the computed profile of ion orbit losses; and the well-known dependence of H-mode threshold on x-point height.
b. With the off-axis beams it will be possible to modify and modulate the superthermal ion edge loss. Operation at the lowest accessible energy (~30-40 keV) with both current and field reversed will maximize the out-going trapped ion orbits. The loss can be modulated by changing the plasma shape, outer gap, and x-point height. Particularly important is that changes in the superthermal ion population at the edge can be monitored using edge CER. c. Determination of a link between ion edge loss and the H-mode edge provides a prospect for -- control of the Er well -- control of L-H and H-L transitions -- modification and/or pacing of ELMs -- modification of pedestal parameters -- control of the edge rotation/torque. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | also submitted to Pedestal Structure | ||
| Title | 110: Model-based control of the toroidal rotation profile in AT scenarios | ||
| Name: | Moreau didier.moreau@cea.fr | Affiliation: | CEA Cadarache |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University), D. Mazon (CEA) | ITPA Joint Experiment : | Yes |
| Description: | This proposal aims at providing a tool to study the physics of low rotation AT discharges on DIII-D. The goal is to control the full radial profile of the plasma rotation in advanced steady state discharges, and to achieve, in a reproducible manner, various requested rotation profiles for the high-Ī²N phase of the advanced scenarios.
The control actuators are (i) co-current NBI power, (ii) counter-current NBI power, (iii) balanced NBI power, and (iv) total ECCD power that was shown to have a significant effect on rotation [1]. Off-axis NBI could possibly be included as an additional actuator for better flexibility in the profiles, depending on the execution of the experimental proposal # 118 ("Experimental identification of the plasma response to off-axis NBI"). The controller takes into account the coupling between the rotation profile and the measured current profile (even if the latter is not controlled), as well as the response to the various actuators. If the experimental proposal # 122 ("Model-based control of the current profile for steady state scenarios") has been executed successfully and there is enough time, combined control of the rotation and the current profile, with the same group of actuators and controller, will be attempted. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The demonstration of adequate control will require half a day to possibly one day, after some dedicated tests of the algorithm in the PCS have been conducted in a couple of short (2 hours) preliminary sessions.
First we shall reproduce the reference shot from the system identification series (1.8 Tesla, Ī²N-controlled AT scenario, at a central plasma density, ne0 = 3.5 x 1019 m-3 and plasma current, Ip = 0.9 MA). In subsequent shots, Ī²N control will be disabled at t=2.5 s (i.e. after a 1 s current flat top) when rotation profile control will start for periods of time that will be short in the first tests, and will increase if the closed-loop response of the plasma is favourable, possibly up to the end of the flat-top phase. Then we shall request different target rotation profiles, including low and, possibly, zero rotation, and repeat the procedure. | ||
| Background: | A multi-variable profile control approach based on experimentally determined, semi-empirical plasma models has been proposed in which the controller uses a combination of the available heating and current drive systems in an optimal way to control simultaneously the evolution of magnetic and fluid/kinetic plasma parameters and profiles [2]. The controller design uses singular perturbation techniques and is quite generic (i.e. independent of the tokamak).
The control-oriented (semi-empirical) model to be used to determine the controller matrices has been obtained from system identification experiments performed on DIII-D in 2009. The model was shown to provide excellent fits to the experimental data [2]. The development and experimental tests of such control methods is requested by the ITPA under the IOS group. References: [1] D. Moreau et al., "Plasma Models for Real-Time Control of Advanced Tokamak Scenarios", submitted to Nuclear Fusion (see 2010 IAEA FEC, paper EXW/P2-07). [2] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. | ||
| Resource Requirements: | NBI at full power is needed and with waveforms generated in real-time by the PCS, including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. Full power ECCD from 6 gyrotrons will also be required. Note: If the experiment proposal #118 has not been executed ("Experimental identification of the plasma response to off-axis NBI"), the 150Ā° beams should aim on-axis and should be able to provide the same input characteristics (geometry, voltage, etc ...) as in November 2009. | ||
| Diagnostic Requirements: | Real-time magnetic measurements, MSE, equilibrium reconstruction (RTEFIT2) and toroidal rotation profiles (CERQUICK) are essential. Measurements of the density profile as well as ion and electron temperature profiles are also required for analysis, not necessarily in real time. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 111: Does low li enhance runaway population? | ||
| Name: | Granetz granetz@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Not requested |
| Co-Author(s): | Val Izzo, Nick Eidietis, Eric Hollmann, John Wesley, Dave Humphreys, Alex James | ITPA Joint Experiment : | No |
| Description: | Determine whether current profile plays a role in the observation of runaways in the current quench. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using ECCD and/or off-axis NBI to create plasmas with varying li, and disrupt with Ar killer pellets. Determine statistics of runaway presence in the current quench. | ||
| Background: | NIMROD modeling w/inclusion of RE physics by Val Izzo has shown that the MHD generated by Ar cooling in low-li vs high-li disruptions is different enough to drastically effect the amount of runaways in the current quench. | ||
| Resource Requirements: | ECCD, off-axis beam, Ar pellet injector | ||
| Diagnostic Requirements: | HXR/scintillators, photo-neutron detectors, synchrotron imaging | ||
| Analysis Requirements: | NIMROD w/RE physics included | ||
| Other Requirements: | -- | ||
| Title | 112: Effect of gas puffing from different poloidal location on ELMs | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A. Nagy, F. Chamberlain | ITPA Joint Experiment : | No |
| Description: | During studies of gas puffing from outboard midplane locations to enhance FW antenna loading, it was observed that the ELM character and frequency was strongly affected by the gas puffing. This observation, along with the search for actuators to control ELMs, leads to the idea of studying the effect of ELMs caused by D2 puffing from different poloidal locations. In 2011, there will be new outboard midplane puffing orifices near the 0 deg and 180 deg FW antennas, along with a comb injector near the 285/300 FW antenna, in addition to puffing locations at the top and bottom of the cross section. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The basic experiment is to create a regularly ELMing H-mode plasma and compare the effect of puffing at a given, calibrated rate from different poloidal locations on the ELM character and frequency. The best-case result would be to find that a gas puff from one poloidal location has a positive effect on the ELMs - increases their frequency and reduces their size - without a strongly deleterious effect on confinement. | ||
| Background: | |||
| Resource Requirements: | Puffing locations near the 0 deg, 180 deg and 285/300 deg midplane in addition to the 'usual' puffing locations. | ||
| Diagnostic Requirements: | All available filterscopes and IR camera would be used to characterize the effect of the puffing on the ELMs. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 113: Combined real-time control of the current profile and Ī²N in AT scenarios | ||
| Name: | Moreau didier.moreau@cea.fr | Affiliation: | CEA Cadarache |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University), D. Mazon (CEA) | ITPA Joint Experiment : | Yes |
| Description: | Ī²N control is routinely used in advanced steady state operation on DIII-D. The goal of this experiment is to test a new controller which integrates current profile and Ī²N control, as early as possible during the ramp-up phase of the discharge. The ability to vary these coupled parameters in a controlled fashion will be very interesting for steady state integration studies. This proposal complements proposal #122 ("Model-based control of the current profile for steady state scenarios") and assumes that the latter has been executed successfully.
The model-based control algorithm has been already developed as well as the control-oriented model to be used in the controller [1-2]. The actuators are (i) co-current NBI power, (ii) counter-current NBI power, (iii) balanced NBI power, (iv) total ECCD power from all gyrotrons in a fixed off-axis current drive configuration, and (v) loop voltage. Off-axis NBI could possibly be included as an additional actuator for better flexibility in the current profile, depending on the execution of the experimental proposal #118 ("Experimental identification of the plasma response to off-axis NBI"). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The demonstration of combined control will require half a day to possibly one day after some dedicated tests of the PCS have been conducted in a couple of short (2 hours) preliminary sessions.
First we shall reproduce the same reference shot as in proposal #122 (1.8 Tesla, Ī²N-controlled AT scenario, at a central plasma density, ne0 = 3.5 x 1019 m-3 and plasma current, Ip = 0.9 MA). In subsequent shots, the usual Ī²N control will be disabled at t=2.5 s (i.e. after a 1 s current flat top) and the new control algorithm will be used, first for Ī²N control only (i.e. with zero weight on the current profile target), and with the same Ī²N target value as in the reference shot. Note that the controller takes into account the coupling between Ī²N and the measured current profile (even if the latter is not controlled). Control will be applied from short periods of time to longer periods and also start earlier, during the ramp-up phase. If this is successful, the weight of the current profile target in the control algorithm will be increased gradually so that combined control can be achieved. The experiments will then be repeated while varying independently the requested Ī²N and current profile targets. | ||
| Background: | Combined real-time control of the plasma current profile and Ī²N is important to achieve stable and reproducible high-Ī²N operation of tokamaks in the advanced steady state regime. A multi-variable approach based on a semi-empirical dynamical plasma model has been proposed in which the controller uses a combination of the available heating and current drive systems, including the external loop voltage, in an optimal way to control the evolution of the plasma parameters and profiles [1]. The 2-time-scale controller design for magnetic and kinetic control uses singular perturbation techniques and is quite generic (i.e. independent of the tokamak).
The control-oriented (semi-empirical) model to be used to determine the controller matrices has been obtained from system identification experiments performed on DIII-D in 2009. The model was shown to provide excellent fits to the experimental data [2]. This experiment will be carried out only when proposal #122 (current profile control) has been executed and successful. References: [1] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. [2] D. Moreau et al., "Plasma Models for Real-Time Control of Advanced Tokamak Scenarios", submitted to Nuclear Fusion (see 2010 IAEA FEC, paper EXW/P2-07). | ||
| Resource Requirements: | NBI at full power is needed and with waveforms generated in real-time by the PCS, including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. Full power ECCD from 6 gyrotrons will also be required. Note: If the experiment proposal #118 ("Experimental identification of the plasma response to off-axis NBI") has not been executed, the 150Ā° beams should aim on-axis and should be able to provide the same input characteristics (geometry, voltage, etc ...) as in November 2009. | ||
| Diagnostic Requirements: | Real-time magnetic measurements, MSE and equilibrium reconstruction including the poloidal flux and the q-profile (RTEFIT2) are essential. Measurements of the density profile, toroidal rotation profile, as well as ion and electron temperature profiles are also required for analysis, not necessarily in real time. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 114: Toroidal Alfven Eigenmodes Driven by ECRH Fast Electrons | ||
| Name: | Pace pacedc@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | W.W. Heidbrink, R. Prater, R.S. Granetz, M.A. Van Zeeland, G. Wallace, A.E. White, and the ECH Team | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to develop an operational scheme in which the ECRH system drives a large anisotropy in the electron distribution that leads to the excitation of toroidal Alfven eigenmodes (TAEs). The successful realization of this goal will enable the study of TAEs and their impact on energetic ion confinement in DIII-D plasmas that are otherwise incapable of producing these modes (e.g., current flattop at normal magnetic field amplitude). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A survey of discharges that are amenable to maximum ECRH power will be identified. Through collaboration with the ECH group, we will purposely drive a large anisotropy in the electron population by way of oblique injection. | ||
| Background: | Recent work at DIII-D has quantified the transport of beam ions [1,2] in experiments that utilize early neutral beam injection during the current ramp to excite AEs. A natural progression of this work requires the ability to study wave-particle interactions in steady state plasmas that better approximate reactor conditions in which super-Alfvenic fusion alphas drive the modes. This also enables DIII-D to study fundamental features of AEs, including damping rates that are investigated elsewhere through the use of active MHD antenna systems [3]. An advantage of the fast electron excitation method compared to active MHD antennas is that it is theoretically capable of driving large amplitude and core localized modes.
Fast electron driven fishbones were observed during early experiments employing ECCD on DIII-D [4], suggesting that the increased power available from the present system is capable of extending this behavior to other modes. More recently, fast electron driven TAEs have been observed on Alcator C-Mod during lower-hybrid current drive experiments [5]. The successful generation of TAEs (modes for which there is a fundamental understanding [6]) through ECRH fast electrons will provide additional constraints on the physics of EC wave damping and serve as an additional constraint to models and codes that attempt to calculate the resulting electron distribution. [1] M.A. Van Zeeland, et al., Phys. Plasmas, (APS 2010 invited talk, in submission) [2] W.W. Heidbrink, et al., Phys. Rev. Lett. 99, 245002 (2007) [3] A. Fasoli, et al., Plasma Phys. Control. Fusion 52, 075015 (2010) [4] K.L. Wong, et al., Phys. Rev. Lett. 85, 996 (2000) [5] J.A. Snipes, et al., Nucl. Fusion 48, 072001 (2008) [6] W.W. Heidbrink, Phys. Plasmas 15, 055501 (2008) | ||
| Resource Requirements: | - all gyrotrons available | ||
| Diagnostic Requirements: | To measure TAE fluctuations: ECE, ECEI, interferometer, PCI, and BES
To measure fast electrons: HXR/scintillators, photo-neutron detectors, and synchrotron imaging | ||
| Analysis Requirements: | - Electron distribution due to EC injection calculated by CQL3D
- TAE stability analysis from NOVA-K or similar | ||
| Other Requirements: | |||
| Title | 115: Transient divertor reattachment | ||
| Name: | Pitts richard.pitts@iter.org | Affiliation: | ITER Organization |
| Research Area: | Divertor Detachment and Plasma Flows | Presentation time: | Not requested |
| Co-Author(s): | A.W. Leonard, C. Lasnier, J. G. Watkins | ITPA Joint Experiment : | Yes |
| Description: | Characterise divertor response to sudden loss of impurity seeding or fuelling and across confinement transitions in the presence or absence of seeding | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Experiments are sought in which impurity seeding for divertor power flux control in a high power ELMing H-mode is abruptly removed, simulating failure of divertor gas injection in ITER. The divertor parameters in response to the seeding loss are to be carefully monitored in the steady state phases before impurities are removed and during the phase directly following removal. The key parameters are power fluxes to the target and the divertor radiation distribution. The essential information required is the timescale for increase in the divertor power flux density from a previously semi-detached state to full attached. The timescale will in fact be largely dominated by the pumping speed (itself a function of magnetic geometry) and by the radiating species in question. Candidate radiating species for ITER are nitrogen, neon and argon, with nitrogen or neon currently favoured. Since N2 is partially recycling and neon full recycling, both should be tried in these experiments, hopefully in very similar plasmas. The experiments should be performed in SND with drsep larger than the midplane power e-folding length. Priority should be on outer target power flux measurements since this will be where the highest stationary power fluxes will be experienced on ITER.
In addition to seeded experiments, measurements are required of the divertor response to H-L back transitions. Such measurements should already be available in abundance, but any additional data would be welcome. The issue here is the modification of divertor target parameters as a consequence of the back transition, when a heat pulse might be expected at the divertor. | ||
| Background: | The ITER divertor high heat flux components cannot tolerate the elevated power fluxes that will occur during fusion burn if the divertor plasma falls out of the partially detached state in which the machine must operate at full power. Thermal load calculations show that the time to react if plasma reattachment occurs must be on the order of 1-3 seconds at most. In this sense, reattachment is defined as a case in which divertor radiated fraction falls from the high (~60%) required in the partially detached regime to values in the region of 20%. Even somewhat higher values (e.g. 30%) extend the time to react, but not significantly. In the case of a full W divertor, foreseen for the DT phase of ITER, this radiation will have to be almost entirely supplied by extrinsic seeding impurities, most likely N2, Ne and or Ar. However, there is not yet a quantitative physics basis on which to assess how fast and under which circumstances the reattachment might occur. The issue of divertor reattachment is the subject of an ITPA Divsol task DSOL-20, being led by the main author of the current proposal | ||
| Resource Requirements: | In principle, provided previously run seeded discharges which fit the approximate experimental requirements can be re-executed, this proposal should not require many discharges. It is important, if possible, to attempt with both N2 and Ne. Similar discharges were requested in ROF 2009, but not obtained. | ||
| Diagnostic Requirements: | Divertor target IR, distribution of total radiation, particularly in the divertor region, target Langmuir probes, divertor neutral pressures, divertor spectroscopy | ||
| Analysis Requirements: | Basic analysis of divertor radiation and target power flux profiles is the most important aspect here - very fast time resolution is not required. Simplified modelling of the system response, accounting for pumping speed etc will be required to link the observations to the ITER case. Some modelling of the data obtained will be possible at the IO with plasma boundary codes. | ||
| Other Requirements: | |||
| Title | 116: Experimental identification of the plasma response to off-axis NBI | ||
| Name: | Moreau didier.moreau@cea.fr | Affiliation: | CEA Cadarache |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University), D. Mazon (CEA) | ITPA Joint Experiment : | Yes |
| Description: | The objective of this experiment is to experimentally characterize the plasma response to off-axis NBI on DIII-D. It requires only a small number of shots (4-5) with off-axis NBI and is a simple extension to the experimental model identification that was performed during the last experimental campaign. These 2009 experiments allowed very good control-oriented models to be obtained to describe the response of the poloidal flux and toroidal rotation profiles to 5 actuators, namely co-current NBI, counter-current NBI, balanced NBI, ECCD and loop voltage (see reference [1]).
This new experiment will be very useful for the evaluation of off-axis NBI physics as they will provide an experimentally measured space-time plasma response that can be compared with theory-based modeling. The off-axis current drive capability of the DIII-D NBI system will also provide additional flexibility for controlling the current profile during ramp-up and/or during the high performance phase of AT steady state discharges, and also for the combined control of the current profile, Ī²N and/or the toroidal rotation profile. The proposed experiment will provide essential data to complete the 2009 model in order to use off-axis NBI as an extra actuator for model-based control applications. This would therefore allow future real-time control experiments to be done while making use of the best heating and current drive mix available on DIII-D (see other proposals on "Integrated and Model-Based Control "). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental plan will be the same as in 2009, but with modulations of the off-axis beams only (4-5 shots).
First we shall reproduce shot #140090, the reference shot without power modulations (2 shots). Then, up to t=2.5 s, all subsequent discharges will be similar to the reference one (1.8 Tesla, Ī²N-controlled AT scenario, at a central plasma density, ne0 = 3.5 x 1019 m-3 and plasma current, Ip = 0.9 MA). At t=2.5 s (i.e. after a 1 s current flat top), in all discharges, the Vloop control mode (i.e. the use of Vloop as a control actuator) is enabled and the Ip and Ī²N controls are disabled in order to avoid feedback in the response data. Between t=2.5 s and t=7 s, modulations of the off-axis beams will be applied, first 150L only, then 150R and both. The modulation waveforms are determined in advance and uploaded into "futureshot" files that can be readily used during the experiment. | ||
| Background: | The algorithms that are used to numerically identify the various elements of a semi-empirical plasma response model using experimental data have been developed for model-based control purposes and used successfully on JET, JT-60U and DIII-D [1-2]. The models relate a set of (machine-dependent) input parameters or actuators (e.g. H&CD powers) to measured output profiles for which control will be needed, namely, the current density (or safety factor) profile, which characterizes the magnetic state of the plasma, and one or several fluid/kinetic parameters and profiles (Ī²N, plasma rotation velocity, ion and/or electron temperature, etc ...).
A model-based controller can then use all the available heating and current drive (H&CD) systems in an optimal way to regulate the evolution of the plasma profiles [2]. The development of such integrated control-oriented models is requested by the ITPA under the IOS group (Joint Experiment IOS-6.1). References: [1] D. Moreau et al., "Plasma Models for Real-Time Control of Advanced Tokamak Scenarios", submitted to Nuclear Fusion (see 2010 IAEA FEC, paper EXW/P2-07). [2] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. | ||
| Resource Requirements: | NBI with various power waveforms is needed including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. This experiment requires that commissioning of the off-axis beams has been completed (full power modulations). Other additional heating and current drive systems from 6 gyrotrons will also be required although not essential. The discharges are to be run partly in the loop voltage control mode (PCS). The system identification requires the availability of the profile data from MATLAB. | ||
| Diagnostic Requirements: | Magnetic measurements, MSE, CER, Thomson. Equilibrium reconstruction including the q-profile (RTEFIT2) are essential, and measurements of the density profile, toroidal rotation profile, as well as ion and electron temperature profiles, are required. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 117: Transport control via realtime NTM width modulation | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | A. Welander | ITPA Joint Experiment : | No |
| Description: | We propose to control plasma beta by realtime modulation of the width of 3/2 or 2/1 NTMs using ECCD. The islands will be grown or suppressed in order to obtain and maintain a beta target by a reduction or improvement in confinement. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will begin with a well-established NTM control scenario in order to establish a nominal beta for the scenario. The target beta will then be lowered and raised in steps from that nominal value in order to exercise the full spectrum of NTM growth and reduction control. That sequence will be followed by a series of over-heated shots in which the beams are turned on at successively higher powers without feedback (a very crude approximation of a nuclear burn) in order to explore the limits of the beta control and to examine the differences (if any) between 3/2 and 2/1 modulation. | ||
| Background: | In present-day tokamaks, beta control is primarily accomplished by modulating the auxiliary heating sources, particularly NBI. However, in burning plasmas the effect of auxiliary heating will be merely a perturbation relative to the much greater nuclear heating. Beta control will have to migrate from the present "turning the hose on and off" approach to more subtle methods. Some candidates include fuel mixture control and total density control. The present experiment proposes the alternative (or complementary) method of deliberately modifying the confinement properties of the plasma via NTM width control, effectively placing a well-controlled pressure "release valve" on the plasma.
The suppression of NTMs has been well established in DIII-D and other devices. This experiment expands NTM control from a binary control problem (suppression or no suppression) to the continuous problem of matching a target width. This control capability has been available for a while but has remained untested and un-utilized. | ||
| Resource Requirements: | All NBI & Gyros, PCS algorithm development, testing of NTM width control algorithm
TIME: 0.5 + 1.0 day (initial test of new algorithms) +(full experiment) | ||
| Diagnostic Requirements: | Magnetic (fast and slow),CER on 30L and 330L,MSE, Thomson,CO2 interferometers, ECE radiometer | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 118: Dependence of ITG critical gradient on rotation in H-mode with ITER shape | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This experiment must be run after #80 Ferron has been successful.
Investigate the impact of the ITER shape and rotation on the ITG critical gradient in H-mode plasmas. The data collected in an ITER shape H-mode plasma during a rotation scan (Ti profile (R/LTi), heat flux, qi, profile, BES, FIR, high-k backscattering) can be used to test and validate the TGLF transport model. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Carefully document the changes in qi vs R/LTi as a function of radius as rotation is scanned in a standard DIII-D H-mode using the ITER shape plasma. Repeat the measurements in a non-ITER shape (DIIID standard, but compatible with ITER shape patch panel). Make many repeat measurments of the Ti profile using CER and main ion CER if available to obtain as accurately as possible changes in R/LTi versus radius. Document as extensively as possible changes in turbulence using BES, FIR and high-backscattering. Compare the ONETWO/TRANSP qi vs measured R/Lti profiles and the profiles of turbulent amplitude across low-k and high-k to predictions from TGLF and from ITG critical gradient models. | ||
| Background: | It has been demonstrated theoretically and experimentally that transport depends strongly on shape (Holland, Kinsey) and recent work at JET showed that the ITG critical gradient depends on rotation (Mantica). It is not clear exactly how sensitive the ITG critical gradient is to the combination of plasma shape and rotation, and whether the ITER shape and increased rotation will conspire together to increase the critical gradient. It is critical to obtain data in the ITER scaled shape at DIII-D, tracking the changes in the ion temperature profile (R/LTi) versus radius as rotation is scanned. By comparing this to a rotation scan in a standard DIII-D shape, and comparing both cases to TGLF predictions, insight may be gained into how favorable/unfavorable the ITER shape is for increasing/decreasing the ITG critical gradient in a low-roatation reactor relevant regime. Additionally, as shape and rotation are scanned it will be imperative to attain any information possible on high-k (ETG) turbulence activity to understand what impact this has on electron and ion heat transport, since the interplay between the low and the high-k turbulence in determining the ITG critical gradient is not well understood. | ||
| Resource Requirements: | Patch panel to make ITER shape. All beam sources: Co and Counter NBI. | ||
| Diagnostic Requirements: | CER and MSE. Standard profile diagnostics.
BES, FIR, High-k backscattering | ||
| Analysis Requirements: | ONETWO, TRANSP, TGLF+TGYRO transport model, TRINITY+GS2 transport model | ||
| Other Requirements: | |||
| Title | 119: The impact of current evolution in the ITER baseline scenario | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | T. Luce | ITPA Joint Experiment : | No |
| Description: | We plan to investigate the role of the current profile, and of the li measurement of the current density peaking, in the tearing stability of DIII-D ITER demonstration discharges. From recent studies, it seems that the central factor leading to the appearance of the n=1 instabilities that limit these discharges is the evolution of the J profile, and not the betaN level as previously believed. The analysis of the database shows that all the discharges end (with a tearing mode or at the end of the NBI power) when the li trace reaches a precise range of values, all included between 0.85-0.95. For this reason it is important to explore different starting li levels, at different betaN values, to understand whether any ITER discharge of this type will inevitably be terminated due to the natural evolution of the current density profile. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce an ITER demo discharge, with q95~3, and evolve the current profile ina way that the starting li value at the beginning of the high-betaN phase is progressively increased to the limit of the PF coils capabilities. Keep the betaN flattop value fixed, and check whether the tearing modes appear at different times, and at what li values. | ||
| Background: | In DIII-D discharges the target scenario in the ITER shape
at q95ā?¼3 and betaNā?¼1.8 is often limited by the appearance of a tearing mode with toroidal mode number n=1, which significantly reduces the confinement and often ends the discharge with a disruption. The standard approach to define the operational limits due to these instabilities characterizes this stability limit as a beta limit. In the general approach, it is inferred that the plasmas become unstable above a marginal beta limit where the NTMs are metastable and can be triggered by a seed island, provided usually by sawteeth, ELMs or fast particles instabilities like fishbones. The classical stability term is not taken into account for the stability limit in this approach. In the presence of ELMing and sawtoothing activity during the discharge, it is implied that the instability will be triggered if the discharge is run above a certain beta value, and the classical delta-prime term is considered constant and negative. This approach is also implied in discussion of DIII-D results. However, it has been shown that in the experimental data of the ITER-like DIII-D discharges, the betaN level is not a discriminator for the stability of the discharge, and that the destabilizing bootstrap term is not varying as the plasma evolves towards the triggering of the instability. Moreover, the li traces for all the discharges is still decreasing when the modes start. For these reasons we believe that the change in the classical stability term related to the evolution of the current profile plays the main role in the destabilization of the modes. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 120: Avoiding locked mode disruptions by triggered ECH | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | We wish to characterize the ability of ECRH to avoid locked mode disruptions. The ECH will be triggered by disruption precursor signals and aimed at the 2/1 or 3/2 surface, | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Generate locked modes in high-beta discharges. Trigger ECH (aimed near q=2 surface) asynchronously using Vloop or locked-mode detector. Vary ECRH power to determine threshold for disruption avoidance. Assess effect of varying ECH timing (by changing trigger threshold) upon ability to avoid a disruption. Repeat with co/counter ECCD to compare ECH to ECCD. | ||
| Background: | ECH applied to the low-order rational surfaces (particularly q=2) has been found to be very effective in avoiding density limit & q-limit disruptions in various limited/L-mode tokamaks (T-10,JFT-2MFTU,RTP,AUG). Recently, AUG reported the avoidance of NTM locked modes in high-beta discharges using off-axis ECH. This experiment aims to implement real active disruption avoidance on DIII-D. In particular, by re-creating the AUG results and then expanding them by assessing the power threshold, timing, and ECCD/ECH comparison. | ||
| Resource Requirements: | 6x ECH, MGI, NBI, Pellet injector, 1 day
Vloop dud trip resettable dud trips for multiple instances | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, | ||
| Analysis Requirements: | toray | ||
| Other Requirements: | -- | ||
| Title | 121: Measurements with new CER Chords Inside the Magnetic-Axis | ||
| Name: | Chrystal chrystal@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | Keith Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to test new tangential CER chords whose points of intersection with the neutral beam are inside the magnetic-axis. Combining their measurements of toroidal rotation with the same measurements from existing tangential chords allows the poloidal rotation to be calculated. These results need to be compared to direct measurements of the poloidal rotation from vertical CER chords. Scans of ion temperature and toroidal field will help determine how corrections to the charge-exchange cross-section affect these measurements. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure tangential and poloidal rotation of carbon impurities in plasmas with an ion temperature scan and plasmas with a toroidal field scan. The rate the temperature and field can be scanned is increased if ELM's are infrequent (as in a QH mode for instance) since this allows for greater time resolution of quality CER data points. | ||
| Background: | Upgrades to the CER system will introduce eight new tangential chords which intersect the neutral beam inside the magnetic axis. These views, coupled with existing views outside the magnetic axis, can be used to calculate poloidal rotation values based on neoclassical theory. These calculations compliment direct measurements of the poloidal rotation made with current vertical CER views because these direct measurements depend on a calculated cross-section correction whereas the new calculations are able to directly measure the cross-section correction (by viewing co and counter injected neutral beams). This could prove advantageous for measuring poloidal rotation and allow for better tests of neoclassical theory.
Initial work will involve simply comparing results for poloidal rotation based on these two measurement techniques. Any disagreement between the two will be important for investigating the validity of neoclassical theory, but it is expected that discrepancies could be a result of the different methods used to correct the charge-exchange cross-section. Since this correction depends strongly on ion temperature and toroidal field, scans of these parameters could be useful in determining the extent this cross-section correction influences these measurements of poloidal rotation. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER tangential and vertical chords. Special beam modulation is required to obtain quality CER measurements for all the required chords. All other profile diagnostics. | ||
| Analysis Requirements: | Comparison with neoclassical codes (NCLASS etc.) | ||
| Other Requirements: | |||
| Title | 122: Test - Please Ignore | ||
| Name: | Leblanc leblanca@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the parameter range over which we can achieve QH-mode with zero net NBI torque. This will involve scans in toroidal field, plasma current, density, beam power and plasma shape. The scans will be guided by peeling-ballooning theory (to investigate stability boundaries) and IPEC calculations (to optimize NTV torque). Issues to be investigated include 1) Can we achieve low rotation QH-mode in the upper single null and balanced double null shapes used in previous QH-mode experiment? 2) Does the NTV torque vary with shape in the way predicted by IPEC? 3) What is the collisionality range where QH-mode exists and how does this change with plasma shape (LSN versus USN versus DND)? 4) What is the range of q for low rotation QH-mode operation? 5) What are the limits on input power or beta_N? |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 123: Investigate Poloidal Variation of Electrostatic Potential | ||
| Name: | Chrystal chrystal@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Keith Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure the poloidal variation of the electrostatic potential within a flux surface for high rotation plasmas. This will be accomplished with CER measurements of impurity ions inside and outside the magnetic axis. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create a plasma that has high carbon toroidal rotation and use CER tangential views to measure parameters of the carbon impurities. | ||
| Background: | Neoclassical theory predicts that when ion toroidal rotation is comparable to the ion thermal speed, density will no longer be a constant on flux surfaces. In this scenario, comparing ion density at different positions on a flux surface can be used to determine the change in electrostatic potential between the different positions.
Upgrades to the CER system will introduce eight new tangential chords which intersect the neutral beams inside the magnetic axis. Using these chords in combination with existing tangential chords (which intersect the neutral beams outside the magnetic axis), the density of carbon impurities can be measured on the low-radius and high-radius side of a flux surface. Combining these measurements with knowledge of other plasma parameters allows the change in electrostatic potential around a flux surface to be calculated. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER tangential views require special beam modulation. All other profile diagnostics. | ||
| Analysis Requirements: | Comparison with neoclassical codes (NCLASS etc.) | ||
| Other Requirements: | |||
| Title | 124: Investigate Neoclassical Theory of Poloidal Rotation | ||
| Name: | Chrystal chrystal@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | Keith Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to test neoclassical predictions of poloidal rotation using CER chords that measure poloidal rotation. This will be done by isolating certain plasma parameters for adjustment and comparing their effect on the measurements with the theory. Also, plasmas that should have low agreement with neoclassical theory will be investigated. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure impurity poloidal rotation in several different scenarios: a plasma where rho* is scanned, a plasma where nu* is scanned, H-mode and L-mode plasmas with similar parameters, and a plasma with an internal transport barrier. Investigating all of these plasma scenarios in one day would be difficult. | ||
| Background: | Neoclassical theory for a plasma with one ion impurity species predicts a linear relationship between poloidal rotation and rho*, and a more complicated relationship between poloidal rotation and nu* based on transport coefficients. Measuring impurity poloidal rotation with the CER diagnostic while scanning these parameters will provide a test of this portion of neoclassical theory.
A different way to test neoclassical theory is to purposely try to violate its assumptions and see how well measurements and theory coincide. To this end, being able to analyze two plasmas which are largely the same with one having increased turbulence (an effect not accounted for in neoclassical theory) could be informative. This could be accomplished by looking at poloidal rotation in similar H-mode (lower turbulence) and L-mode (higher turbulence) plasmas. Another method of violating neoclassical theory assumptions would be to create an internal transport barrier while increasing the gyro-radius as much as possible. This would decrease the validity of the assumption of smallness of rho when compared to the gradient scale length in a region readily viewed by CER chords. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER tangential and vertical chords. Special beam modulation is required to obtain high quality CER measurements for all the required chords. All other profile diagnostics. | ||
| Analysis Requirements: | Comparison with neoclassical codes (NCLASS etc.) | ||
| Other Requirements: | |||
| Title | 125: Beta Induced Alfven Acoustic Eigenmode (BAAE) Studies | ||
| Name: | Van Zeeland vanzeeland@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | B. Tobias, R.K. Fisher, N.N. Gorelenkov, W. W. Heidbrink, D.C. Pace | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to document the 2D structure of the Beta Induced Alfven Acoustic Eigenmodes (BAAE) using ECE imaging as well as investigate the stability of BAAEs driven by off-axis beam injection. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will begin this half day experiment with discharge 132710 in which a spectrum of BAAEs up to n=20 have been observed. The discharge will be repeated with varying levels neutral beam power, both on and off-axis to investigate the effects on the eigenmode stability as well as on mode structure. These discharges will have 1-2 gyrotrons deposited on-axis to increase the Te gradient and create a measurable temperature perturbation. The eigenmode structures will be measured with ECEI, BES, and ECE and the impact on the fast ion population will be measured with FIDA, FILD, neutrons, equilibrium reconstructions, and FIDA imaging. | ||
| Background: | The BAAE is a relatively new type of Alfven eigenmode found in the lowest Alfvenic gap formed by the coupling of the shear and acoustic continuum. It was recently observed in NSTX, JET, ASDEX, and DIII-D. In all devices except JET, the BAAE has been associated with enhanced fast ion transport. Due to its low frequency and compressional component, it can also interact strongly with the thermal ion population and possibly turbulence.
The radial BAAE mode structure has been measured previously using the linear ECE array. In order to make a measurable temperature perturbation, however, on-axis ECH is required. In 2010 BAAE experiments, ECE imaging was a newly installed diagnostic and to avoid damage to the diagnostic, ECH was not allowed into the vessel while ECE imaging was in operation. As a result, no 2D mode structures were obtained with ECEI. 2D mode structures are a very important component of modeling BAAEs since these modes do not appear on magnetics, making measurements of their toroidal mode number (with the existing diagnostics) impossible. The off-axis neutral beam will provide an inverted radial fast ion pressure gradient that should significantly impact BAAE stability. This is in contrast to the typical centrally peaked fast ion pressure profiles obtained with on-axis injection and will be an excellent opportunity for stress testing models of the BAAE. | ||
| Resource Requirements: | 0.5 days
150R,L 30L, 330L at least 2 gyrotrons | ||
| Diagnostic Requirements: | ECEI, FIDA, ECE | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 126: FILD2 calibration using prompt losses | ||
| Name: | Fisher fisherr@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | W. Heidbrink, M. Van Zeeland, D.Pace and EP Working Group | ITPA Joint Experiment : | No |
| Description: | The goal is to checkout the performance of the new FILD2 diagnostic at 165R0. Prompt losses from beam blips will be measured during current ramp and compared to the results of reverse orbit modeling | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 10 ms beam pulses from each neutral beam source will be injected consecutively during a plasma current ramp shot similar to that done on shot 141223 for the 225R-1 FILD. Due to the larger heat loads expected at the midplane, the new FILD2 diagnostic must be inserted slowly into the bumper limiter shadow while monitoring the FILD2 temperature using TC's and views from any available DIII-D cameras viewing the detector head. | ||
| Background: | These beam blip current ramp shots should be done for different plasma shapes including standard single null divertor shots and also for the oval shots used for TAE studies | ||
| Resource Requirements: | All NB sources | ||
| Diagnostic Requirements: | FILD2 at 165R0 and FILD1 at 225R-1, DIII-D cameras viewing detector head at 165R0 if available | ||
| Analysis Requirements: | reverse orbit modeling based on observed losses | ||
| Other Requirements: | Note that some portions of this experiment can be combined with Proposal 130 | ||
| Title | 127: Alfven Eigenmodes and Fast Ion Transport with the Off-Axis Beam | ||
| Name: | Van Zeeland vanzeeland@fusion.gat.com | Affiliation: | GA |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | W.W. Heidbrink, R.K. Fisher, M. Garcia-Munoz, D.C. Pace, B.J. Tobias, and Y. Zhu | ITPA Joint Experiment : | No |
| Description: | The primary goal of this experiment is to use the new off-axis beam to further our understanding of Alfven eigenmode (AE) stability and the impact of these modes on the fast ion population. This experiment will directly contribute to our understanding of the off-axis neutral beam performance as well as our ability to predict its current drive and fast ion deposition during the current ramp phase of DIII-D discharges by mapping out the impact of AEs in a range of typical operating conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment will use 122117 (-Bt) as a target discharge with well-documented AE activity and fast ion transport. The operating space from this typical 2 beam on-axis startup (with injection beginning at t=300 ms) will be spanned in a series of discharges by stepping from 2 equivalent on-axis beams to 2 off-axis beams while maintaining constant power. The progression will be as follows:
2 on-a, 0 off-a 1 on-a, 1 off-a 0 on-a, 2 off-a 1.5 on-a, 0.5 off-a 0.5 on-a, 1.5 off-a 2 off-a, 1 on-a Following this, a power scan of the 150 off-axis beam will be carried out going from 0.25 to 2 equivalent beams (previous experiments showed AE instability beginning near 0.5 equiv. beams). To further map out the impact of these modes in a range of parameter space, 2 equivalent off-axis beams will be injected with the injection delayed to as late as 600 ms in the current ramp. Previous experiments in 2010 showed this can severely impact AE stability. Since a major prediction of the off-axis current drive is its dependence on the sign of Bt, these discharges will be repeated with +Bt. The impact on the fast ion population will be measured with FIDA, FILD, neutrons, equilibrium reconstructions, and FIDA imaging. | ||
| Background: | Virtually all beam heated DIII-D discharges contain some level of identifiable Alfvenic activity. It is well documented that this activity is responsible for a large central depletion of the fast ion profile. Not surprisingly, it is difficult to find discharges in which simulations with TRANSP or ONETWO accurately reproduce the volume averaged neutron emission, fast ion pressure profile, and NBCD when classical beam ion diffusion is assumed. This discrepancy is particularly large during the current ramp phase when the NB beta is a significant fraction of the total. As a result, we cannot make reliable predictions of the off-axis neutral beam performance during the current ramp without first measuring the AEs it drives in a range of conditions. | ||
| Resource Requirements: | 1 Day
330, 30, 150 beams | ||
| Diagnostic Requirements: | FIDA, FILD, ECE, ECEI | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 128: Work towards 3.6 MW coupled FW power to AI discharges - 'existence proof' | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Requested |
| Co-Author(s): | T.C. Luce, M. Porkolab, A. Nagy, P.M. Ryan, P. Politzer, R. Goulding, D. Rasmussen, E. Fredd, N. Greenough, A. Eguizabal, J. Hosea, A. Horton | ITPA Joint Experiment : | No |
| Description: | This is the continuation of the 'main line' of experiments aiming at the 5-year plan goal of robust coupling to the plasma core of 3.6 MW of FW power in advanced discharges, here defined as having at least betaN=2.4 and H98y2=1. This information is needed to inform the Fast Wave assessment at the end of FY11. Since these discharges are always rapidly ELMing H-modes, to date, the achievement of this goal essentially comes down to finding regimes with adequate FW antenna resistive loading RL. To couple about 1.33 MW per 4-strap antenna with acceptable peak voltage, minimum RL (between ELMs) of about an ohm is needed. In FY10 experiments, we found that marginally sufficient antenna loading can be found in this regime by changing the balance between ELM-induced fueling of the far SOL and cryompumping, which is done by adjusting DRSEP. We found that it is crucial to have a comparison of the FW power with an equal level of EC power, to allow separation of the effects on confinement of pure electron heating from possible inefficiencies of core FW absorption. In the best cases, it appears that the FW absorption in the core is similar to that of the EC, which is expected to be ~100%. This is consistent with ray-tracing studies that show that the 90 MHz FW undergoes strong upshift of the n-parallel spectrum such that the single-pass absorption on electrons in the core can be ~75%. Under such high first-pass absorption conditions, we expect that the only edge losses that could be important are the 'prompt' losses that would occur as part of the initial wave launching (not multiple-bounce losses).
So far we have coupled up to 1.5 MW of FW power and compared it with the incremental effect of 1.5 MW of EC power. Without change, this case may allow up to ~2.5 MW of FW power with acceptable antenna voltages. To get the next ~20%, we are moving all three FW antennas radially inwards by ~1 cm during the present LTO. This should increase RL at a fixed plasma condition and position by about 20%, and hence allow that much increase in FW coupled power at the same antenna voltage. Furthermore, we are making improvements to the transmission lines on the 285/300 antenna and 0 deg antennas, and improvements to the ELM/arc discrimination on the 0 and 180 deg antenna systems that should allow a further increase in coupled power. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We expect that time for this experiment is probably best allocated in half-day increments, preferably the morning, so that if we find that a given condition is conducive to high FW power, we could hope to extend the day to include the afternoon, while if technical problems crop up that preclude higher power levels, we can cede the remainder of the day to another related, but non-FW, experiment. | ||
| Background: | See description. | ||
| Resource Requirements: | All three FW systems, up to 6 gyrotrons, full NBI power. | ||
| Diagnostic Requirements: | Profile diagnostics, neutron measurements, possibly visible and/or IR views of as many of the three FW antennas as possible, new 285/300 antenna diagnostics including GUIDAR (to be installed during present LTO) | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 129: TBM Error Field Compensation in H-mode | ||
| Name: | Park jpark@pppl.gov | Affiliation: | PPPL |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | M. J. Schaffer, J. E. Menard, TBD | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to test the n=1 compensation for TBM mockup error fields, and to investigate the effects on rotation braking and confinement degradation in H-mode. It has been predicted that the dominant component in TBM mockup error fields is the n=1 perturbation rather than many other higher n perturbations, even for braking in H-mode, and therefore can be compensated by I-coil corrections. Results will be very useful for the future ITER error field correction study. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Reproduce LSN target shots in the early 2009 TBM campaign (such as #140033). Start with IP=1MA, moderate co-injected NBI (2~4MW).
(2) Turn on TBM 1kA and check the rotation braking and confinement degradation. (3) Apply the n=1 compensation using I-coils. (4) Optimize correction if the compensation is favorable. Theoretical values are 240 phasing, 440A peak current for each. (5) Repeat (perhaps without (4)) different targets with higher IP and NBI. | ||
| Background: | The 2009 TBM campaign found that TBM mockup error fields can induce plasma locking in Ohmic plasmas, due to the amplified n=1 component, and thus can be compensated using the conventional I-coil corrections. The results were consistent with ideal perturbed equilibrium calculations. The calculations coupled with NTV theory also found that the n=1 component in TBM is most important even for magnetic braking in H-mode and thus the n=1 compensation using I-coils can largely mitigate magnetic braking effects. This is counter-intuitive since TBM mockup error fields are highly localized but I-coil correction fields are widely distributed in space. The prediction therefore should be validated in experiments. Results will provide important guidance on error field correction in the future devices. | ||
| Resource Requirements: | LSN plasmas, all 5 co-beams, full capability for I-coil corrections. | ||
| Diagnostic Requirements: | magnetic measurements, MSE, CER, MPTS, IR camera | ||
| Analysis Requirements: | EFIT, SURFMN, IPEC, and NTV calculations | ||
| Other Requirements: | -- | ||
| Title | 130: Beam Ion Heating of Bumper limiters | ||
| Name: | Fisher fisherr@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Not requested |
| Co-Author(s): | David Pace, Arnie Kellman | ITPA Joint Experiment : | No |
| Description: | The goal is to determine if the heat loads on the existing bumper limiters are due to neutral beam prompt losses. By measuring the prompt losses using the new mid-plane FILD2 diagnostic, we hope to check the results of the reverse orbit modeling to determine if prompt losses at the outer midplane are consistent with the observed bumper limiter heating. If prompt losses are responsible,the proposed long pulse plasma operation with neutral beam heating can potentially be improved by moving the toroidal location of the bumper limiters to better accommodate the prompt losses. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Monitor bumper limiter heating and FILD2 data on prompt losses at midplane during neutral beam injection. NOTE this experiment can be combined with Proposal 126.....longer beam pulses from beam sources predicted to be responsible for heating BL's will then likely be needed for BL TC measurements | ||
| Background: | The present bumper limiters observe variations in heat load and damage that are not clearly understood, and and are thought to possibly be due to neutral beam prompt losses. Analysis of existing data is underway to determine if the observed bumper heat loads correlate with the expected prompt losses based on reverse orbit modeling. If the results of this analysis look promising, data from the new midplane FILD would be an important addition to confirm the conclusions on the contribution of the prompt losses to the heat loads on the BL's | ||
| Resource Requirements: | All NB sources | ||
| Diagnostic Requirements: | new midplane FILD2 at 165R0 | ||
| Analysis Requirements: | reverse orbit modeling of prompt losses | ||
| Other Requirements: | bumper limiter TC data?
IR camera data if views of BL's are available? | ||
| Title | 131: Oblique-ECE-assisted MECCD suppression of 2/1 NTM | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | M. Austin, R. La Haye, R. Prater, D. Truong, A.Welander | ITPA Joint Experiment : | No |
| Description: | Use oblique ECE to radially align and correctly modulate ECCD in 2/1 island. Stagger poloidally the launching directions of the various gyrotrons to reproduce ITER-like broad deposition and further enhance the relative benefits of modulation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- Dial up a discharge with a 2/1 NTM rotating at f<5kHz, for example #135861.
-- Interface oblique ECE and ECCD as in #132113 -- Add new oblique ECE capabilities developed at UW-Madison: - new ultra-low-noise video-amplifiers - new broader-band phase-shifter, to compensate for the oblique ECE being collected and ECCD being injected at different locations - (if ready) new radiometer with lower noise and more channels (16 instead of 2) - (if ready) new hardware ECE-Mirnov correlator. Mirnov will provide the correct frequency, oblique ECE the correct phase. -- Compare: - narrow vs. broad ECCD - modulated vs. Continuous - deposition in the O-point, X-point and in between. | ||
| Background: | The system has been already applied with success to the alignment and modulation of narrow ECCD to a rotating 3/2 island, resulting in its complete stabilization and in saving 30% of average power compared to continuous ECCD. Further improvements, recognizable also in a reduced demand of peak power, are expected for broad ECCD.
So far, the more malicious 2/1 mode has been successfully "tracked" at TEXTOR but has never been stabilized by modulated ECCD, neither using Mirnov drive, nor oblique ECE. It will be important to do this for the first time, on the way to ITER. | ||
| Resource Requirements: | 5 gyrotrons | ||
| Diagnostic Requirements: | Oblique ECE | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 132: Compare co-/ctr-ECCD in O-/X-point (4 cases) | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Drive current in the O-point or X-point of a neoclassical island in the co- or ctr-direction (4 cases). Compare NTM stabilization efficiency and characterize (de)stabilization mechanisms. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate co-ECCD. Decide whether to modulate after oblique ECE (analogically) or Mirnov signals (analogically or digitally) depending on whatā??s available and most reliable at the moment of experiment. Compare O- and X-point phasing. Tilt launchers toroidally and repeat for ctr-ECCD. | ||
| Background: | It has already been shown, both at AUG and DIII-D, that co-ECCD in the O-point is more effective than in the X-point (although X-point deposition is still better than doing nothing).
The aim of this proposal is to add two more cases, namely ctr-CD in the O- and in the X-point. These are similar but not identical to co-CD in the X- and O-point, respectively. One possible difference is that ctr-CD might give rise to a 4/2 component that would make the 2/1 island narrower and thus easier to stabilize. Similar considerations apply to a 6/4 distortion of the 3/2 island. Further differences between co- and ctr-CD stabilization might be unveiled by the experiment, in particular by magnetic probe and ECE measurements of the island width evolution and of the poloidal and toroidal mode numbers. MSE measurements of the total local current and ONETWO and TORAY calculations of the Bootstrap and EC-driven currents will also be essential in the analysis. | ||
| Resource Requirements: | 5 Gyrotrons | ||
| Diagnostic Requirements: | Oblique ECE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 133: "Catching" rotating modes with rotating fields before locking | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Apply rotating magnetic perturbations (MPs) to the rotating precursor of a locked mode to "catch it" and entrain it while it slows down. After the mode locks to the rotating MP, the rotation can either be kept constant, at a safely high level to avoid locking to the static error field or to the walls, which is one of the main causes of disruptions, or accelerated. As a result the mode would accelerate too, and be stabilized by rotational shielding and rotation shear. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | As soon as mode rotation frequency f<1kHz (dud), apply intense (I-coil current 1.5kA) n=1 travelling wave at frequency <=f, to account for dwell time, then slow down at the same rate as the mode. Can be pre-programmed or in f/back (rtnewspec). Alternatively, use magnetic feedback, for the I-coils to feed back on Mirnov. While the mode slows down, reduce its amplitude accordingly, to account for reduced rotational shielding.
If rotating mode locks to rotating field at time t, repeat with pre-programmed changes in MP rotation after t, for example keep the rotation steady, or accelerate it again. | ||
| Background: | So far, MPs successfully controlled initially locked modes at DIII-D. Here we propose to pre-emptively apply rotating MPs and avoid locking altogether. For entrainment, it will be crucial for the travelling wave to match the mode frequency and phase, to avoid differential rotation (thus, rotational shielding) between the MP and the mode. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | PCS changes involving rtnewspec and/or adaptation of magnetic feedback to NTMs | ||
| Title | 134: Forced rotation of initially locked islands, with diagnostic applications | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye | ITPA Joint Experiment : | No |
| Description: | Apply rotating n=1 error field to unlock and spin up to 200-1000Hz an initially locked mode. Try to reproduce rotational mitigation at 10Hz. Resolve (forcefully) rotated islands with diagnostics such as CER and MSE which normally have too little temporal resolution. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use dud detectors to trigger an I-coil travelling wave right after mode locking. Pre-program concomitant ramps of frequency and amplitude of travelling wave. Begin with 1-300Hz, 1-2kA, 2s linear ramps.
Experimentally (and with help from calculus of variations?) make ramps slow enough to avoid slipping but fast enough to avoid that the mode grows too much and disrupts. For fixed frequency ramp, improve amplitude ramp: make it as steep as necessary to defeat shielding at high frequencies by image currents in the wall, but not too steep, as excessive amplitude results in excessive error-field penetration and possible disruption. For higher current, the SPAs will be adopted as I-coil power supplies. Even so, the amplitude limit will be hit before the frequency limit. At that time, amplitude will be kept to the maximum or slowly ramped down, according to SPA capabilities, while the frequency will be further increased. The intention is to maintain the coupling between the travelling wave and the mode as long as possible, even if the travelling wave is not as intense as desired. Non-linear ramps could help in this respect. We will also try to reproduce and understand the unexpected mode mitigation obtained during the frequency ramp at about 10Hz, for which various interpretations have been formulated. If successful, we will compare post-locking intervention (present proposal) with the more difficult and ambitious pre-locking intervention (proposal 133), i.e. with the attempt to "catch the mode" while it slows down and sustain its rotation without letting it lock at all. | ||
| Background: | Sustained rotation (a.k.a. entrainment) has been already demonstrated, but at relatively modest frequencies of up to 60Hz, possibly 180Hz (unclear, waiting for confirmation). Sustained rotation at higher velocities, of the order of 200-1000Hz has a number of advantages, ranging from locking avoidance, rotational mitigation, an easier ECCD modulation, accessibility to diagnostics which are normally too slow to resolve NTMs. The frequencies considered here are accessible to most diagnostics, including MSE and CER. Much higher frequencies are impossible for the SPAs, or are possible but at too low amplitudes. When compensation from image currents in the wall is taken into account, the effect in the plasma would be far too weak. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER, MSE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 135: Easier modulated ECCD on forcefully rotated mode | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye, A. Welander | ITPA Joint Experiment : | No |
| Description: | Demonstrate a new NTM stabilization technique in which, instead of adapting the ECCD modulation to the natural, non-uniform mode rotation, the mode is magnetically forced to rotate at a known frequency and with a known phase, and the ECCD is modulated accordingly. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Magnetically force a tearing mode to rotate at a prescribed frequency and with known phase. At the same time, inject ECCD and modulate it at that same frequency and phase. This is easier than adapting the ECCD modulation to the spontaneous, non-uniform mode rotation. Note that everything is pre-programmed. No feedback required.
It can be attempted either on a mode which is initially locked to the wall or error-field, or on an NTM which is initially rotating at its spontaneous frequency. In the first case it would be a follow-up of proposal 134, in the latter of proposal 133. Any frequency in the 30Hz-5kHz range is acceptable. The following reference shots will be needed for comparison: with continuous ECCD, with no ECCD, with deliberately misaligned ECCD (e.g. 2-3 gyrotrons too far in, 2-3 too far out) and with X-point phasing. | ||
| Background: | Sustained mode rotation (a.k.a. entrainment) at up to 60Hz (and possibly 180Hz, unclear, waiting for confirmation) have been already demonstrated in the absence of ECCD. Rotational mitigation has been observed, but ECCD will be necessary for full stabilization. Furthermore, because the frequency and phase of rotation will be known and controllable, the control by modulated ECCD is expected to be easier. | ||
| Resource Requirements: | SPAs, 4-6 gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 136: Benefits of ECCD modulation as functions of phase, deposition width, misalignment & duty-cycle | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye, A. Welander | ITPA Joint Experiment : | No |
| Description: | Confirm and quantify benefits of proper phase, broad deposition and good alignment in the stabilization of NTMs by modulated ECCD. Investigate duty-cycles alternative to the conventional 50/50%. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This can be an extension of any successful NTM stabilization by modulated ECCD experiment in the next campaign, regardless of the diagnostic (Mirnov, oblique ECE, horizontal ECE) and interface (digital, analogue).
Simply repeat the experiment for various phasings (O-point, X-point and intermediate) and for various deposition widths delta_ECCD, by staggering the gyrotron launches more and more in the poloidal direction. Note that the ratio of delta_ECCD to the island width w is scanned automatically and for free in every shot, dynamically, as delta_ECCD remains fixed and w shrinks. However, various other plasma parameters change in the process, not to mention that at the end of the stabilization process w becomes small, delta_ECCD/w large, hence the modulation more beneficial and the stabilization more rapid. Unfortunately other effects leave the same signature, of making the stabilization quicker (e.g., when w becomes comparable with the marginal island width). For this reason, it is preferred to scan delta_ECCD instead, in a dedicated series of 3-5 shots. 2-4 misaligned shots for each case should suffice to study the resilience of modulated ECCD to misalignment in 4 cases (broad/narrow deposition, O/X point phasing) and compare with formulas by Rip Perkins. A preliminary DIII-D result suggest that ECCD in the X-point might be more resilient to misalignment, but needs confirmation. Finally, in addition to the conventional 50/50% duty cycle, we intend to keep the ECCD on for 10, 30, 70 and 90% of the time. | ||
| Background: | An extensive phase scan and a qualitative (2-3 point) width scan were carried out at AUG, for a 3/2 mode. The goal at DIII-D is to repeat, confirm and improve those scans for the 3/2 mode, especially as far as delta_ECCD is concerned, and to perform them also for the 2/1 mode, for the first time. It is still unclear whether ITER really needs the ECCD to be modulated, and this scan can help answering this question.
As a by-product, the phase-scan can help separate the ECCD effect on Delta', which does not depend on the phase, from the replenishment of the missing BS current, which evidently does. Should modulation be recommended for ITER, the proposed misalignment scan is of obvious importance in setting the requirements for the real-time scan of the launchers. Finally, the scan of the duty cycle is unique in its kind. Duty cycles of 50% on, 50% off time are customarily considered both in theory and experiment. However, stabilization is the resultant of the two mechanisms mentioned above, one of which does not depend on the phase and, as a matter of fact, does not improve with modulation. It would rather benefit from a longer duty cycle, or from the ECCD being on all the time. For all these considerations, it is speculated that the optimum might be other than 50/50%, and we propose to find it experimentally. Note the optimum would be machine- and scenario-dependent. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 137: ECCD modulated by horizontal ECE | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | M. Austin, R. La Haye, A. Welander | ITPA Joint Experiment : | No |
| Description: | Show that horizontal ECE can replace oblique ECE as a driver for ECCD modulation in phase with NTM O-point. Coincidentally, no phase correction is required for the 2/1 mode, because the ECE and the ECCD happen to be about 180deg out-of-phase. Some correction ā??to optimize experimentally- is required for the 3/2 mode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The approach will be similar to shots where ECCD was modulated by oblique ECE, e.g. #132113, except that the analogue interface in the annex ("the box") will read the horizontal ECE signals.
The horizontal ECE is located at phi=60deg, i.e. approximately 180deg apart from the gyrotrons (phi=240-270deg) and thus the ECCD deposition regions (which differ from the launching positions by <10deg). Hence, the horizontal ECE is measured in a ideal position from which it can directly modulate the ECCD in phase with a rotating 2/1 mode, with no need for phase correction. To stabilize a 3/2 mode, instead, we will need the phase-shifter already used with success in the past, or the new shifter with flatter response which we recently developed. | ||
| Background: | Horizontal and oblique ECE have the advantage, over Mirnov probes, of being local, internal diagnostics of NTMs. This simplifies the phase correction when they are used to modulate the ECCD in phase with a rotating island. Oblique ECE simplifies this phase correction even more than horizontal ECE, if collected along the ECCD launch direction, or an equivalent one.
In the last campaign, the new oblique ECE radiometer was successfully interfaced, by an analogue circuit in the annex, to the gyrotron power supplies, and ECCD was modulated in synch and in phase with the O-point of a rotating 3/2 NTM. Complete stabilization was obtained. The analogue interface worked very reliably and correctly manipulated the signals that it was receiving from the radiometer. Those signals, however, were not perfect NTM measurements, partly because of intrinsic reasons (ECE is sensitive to all Te fluctuations, not just to NTMs) and partly because of the signal-to-noise ratio of the radiometer, which needs to be improved. While these improvements are under way (a new oblique ECE radiometer is under construction at UW-Madison), we suggest to connect the analogue interface to the horizontal ECE. Its superior signal-to-noise ratio will make up for the slightly more difficult phase correction. Should it work, it would be a more robust, reliable and easy-to-use driver for ECCD modulation: more robust and reliable because horizontal ECE has been tested for years, is one of the main DIII-D diagnostics, regularly maintained, easier to use because it is always available and does not require various shutters and an optical switch to be opened or closed, as in the case of oblique ECE, which shares a transmission line and a launcher (receiver) with one of the gyrotrons. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | Cables or optical links between horizontal ECE radiometer and annex. | ||
| Title | 138: Analogue approach to Mirnov modulation of ECCD, with phase scan | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | A. Welander, R. La Haye, E.J.Strait | ITPA Joint Experiment : | No |
| Description: | Use Bp probes at various locations, connected to the analogue interface originally developed for oblique ECE, for modulated ECCD at various phases relative to NTM. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Similar if not identical to shots where ECCD was modulated by oblique ECE, but with the analogue interface in the annex ("the box") connected to a Mirnov probe (or a combination of probes, to isolate odd/even modes or specific n numbers). Then repeat for different probes around the torus, to scan the phase of ECCD relative to the island O-point and X-point. | ||
| Background: | At DIII-D, both Mirnov signals and oblique ECE signals are being explored as drivers for ECCD modulation in synch and in phase with the O-point of rotating NTMs.
Mirnov signals are digitally interfaced to the gyrotrons, whereas a fully analogue approach has been adopted for the oblique ECE. Here it is proposed to combine the best aspects of the two approaches. In the last campaign, the analogue electronics worked very reliably: it delivered to the gyrotrons the expected modulation waveforms, on the basis of the received oblique ECE signals. These signals, however, were not perfect NTM measurements, partly because of intrinsic reasons (ECE is sensitive to all Te fluctuations, not just to NTMs) and partly because of the signal-to-noise ratio of the radiometer, which needs to be improved. Mirnov signals, on the other hand, are with no doubt the best, cleanest measurements of NTMs. Their digital interface with ECCD, however, is still being commissioned. Here we propose to use Mirnov signals in combination with the analogue interface. The only drawback of this approach is the uncertainty in the phase extrapolation from the Mirnov sensor to the ECCD actuator, lying in different positions. A simple way around the problem is to utilize different sensors (from the same toroidal array) in different discharges. In this way, the phase-delay would automatically be scanned from shot to shot, with the twofold result that a complete toroidal scan would be performed and the sensor yielding the best phase would be found. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 139: NTMs "on demand", by ECH | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ECRH slightly outside q=3/2 to provoke a local flattening of pressure, thus a Bootstrap deficit and therefore a 3/2 NTM when desired, for example in hybrid discharges. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Very simple, with ECRH, perpendicular launch and deposition slightly outside q=3/2. A toroidal field scan (within the shot, or from shot to shot) will allow to find the best location. | ||
| Background: | NTMs are not always undesired instabilities: small 3/2 NTMs, for example, are desirable in 'hybrid' discharges, where they help to prevent sawteeth. Occasional difficulties were encountered, for example in March 2007, in reproducing hybrid scenarios with "natural" 3/2 NTMs. The purpose of this proposal is to develop a tool to trigger these modes on demand, when required. It might also shed light on the physics of seeding, and permit accurate measurement of the NTM growth rate, although under "artificial" conditions, which can be contrasted with predictions from the Rutherford equation. In particular, the flattening of the pressure profile is expected to affect the Bootstrap term in the Rutherford equation and, to a higher order, to make Delta' less negative. The presence of other NTMs in the plasma (typically 4/3, in the absence of 3/2) is not expected to constitute a problem, as ECRH will be deposited outside the 3/2 surface and thus even farther from the 4/3 one. | ||
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| Title | 140: NTMs "on demand", by modulated ECCD | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Drive current filament on the q=3/2 surface (around which island will form), by means of ECCD modulated at twice its CER rotation frequency. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure with CER the toroidal rotation velocity of the 3/2 surface (identified/localized through MSE and EFIT). Repeat the shot with modulated ECCD at twice that frequency (because n=2). In case of excessive variation from-shot-to-shot or within-the-shot, some work on the PCS might allow to respond in real-time to the changes of CER frequency. Compare co- and ctr-CD, expected to drive classical and neoclassical TMs, respectively. | ||
| Background: | NTMs are not always undesired instabilities: small 3/2 NTMs are desirable in ā??hybridā?? discharges, where they help to prevent sawteeth. Despite the existence of consolidated experimental recipes, difficulties are often encountered (see for example in March 2007) in developing hybrid scenarios with ā??naturalā?? 3/2 NTMs. The purpose of this proposal and of #371 is to develop a tool to trigger these modes on demand, when required. The basic idea is that, in the absence of mode and therefore of filamentation, the 3/2 surface is ā??smoothā?? and consists of identical 3/2 current filaments, all carrying the same current. Artificially increasing or decreasing the current in one of them by means of modulated co- or ctr-CD would break the axisymmetry and introduce the helical current perturbation around which an island would form. If sufficiently big (wider than ā??marginalā??), this island would evolve, grow and saturate. Tailoring the island to the hybrid discharge needs would then be a question of setting the plasma beta or the ECCD power ā??or otherwise modifying the Rutherford equation- in such a way that the saturated width is tolerable. | ||
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| Title | 141: Test of causality: mode rotation vs. plasma rotation | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Does the plasma "drag" the mode or the mode drags the plasma? Can electromagnetic torques imparted to the mode spin the plasma? Can viscous torques imparted to the plasma spin the mode? When the mode locks, does it always and exactly arrest the plasma rotation? These are the questions that we intend to answer. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use NBI to control plasma rotation and I-coils to control mode rotation. Use CER to measure plasma rotation and Mirnov (newspec) to measure mode rotation. Then use an approach similar to K.Burrell's H-mode studies (PoP 1999): modulate the mode (plasma) rotation and measure the delay of the plasma (mode) response. Hysteresis curves will be obtained. | ||
| Background: | NTMs are approximately "frozen" in the plasma and spinning the latter (by momentum injection with NBI) also spins the former. Various proposals and some experimental evidence exists, suggesting that the opposite is also true, i.e. that torque imparted to the mode also spins the plasma. If confirmed, plasma rotation (and not just mode rotation) would represent one more application for internal coils in devices with low NBI momentum injection like ITER. Similar considerations apply to plasma and mode braking, and how they affect each other. For example, rotating modes are known to lock even in rotating plasmas which keep rotating. On the other hand, slow plasma rotation encourages mode locking. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | CER | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 142: Effect of impurities and wall conditioning on NTMs | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure beta threshold for NTM onset for different wall conditions and under controlled core/edge cleanliness or impurity seeding. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop a small database of 10-20 discharges with NTM onset during NBI (hence, beta) ramp. A 2/1 mode is preferred, as it is closer to the wall and possibly more dependent on impurities and wall conditioning. The otherwise identical shots should differ only by wall conditions (e.g. be taken after a boronization, disruption, or glow discharge) and/or deliberate impurity contamination (impurity pellet, puffing, laser blow-off). There are various ways of building the database.
Approach 1: modify the plasma test shot run every morning at DIII-D (the same used for long term monitoring of wall conditioning and impurities) by adding an NBI ramp at the end. Make additional, small changes, if needed for NTMs but then leave them fixed for the duration of the campaign. Approach 2: try building the database in a single day or half-day of experiment. However, this would only tackle the dependence on impurities and the short-term dependence on wall conditions, but fail to do any monitoring in the long term. | ||
| Background: | Dedicated experiments at NSTX (F. Volpe, L. Delgado-Aparicio, S. Gerhardt, S. Sabbagh et al.) showed that the onset of NTMs is delayed by Lithiumization and anticipated by Neon puffing. Post-disruption data at DIII-D also suggest an effect of wall conditioning and impurities on NTMs.
Here we propose to systematically characterize these effects at DIII-D, by operating otherwise NTM-unstable discharges under controlled conditions of core/edge cleanliness or impurity seeding. We also propose to compare with the NSTX results. This is important in view of ITER, where it might be preferable to wait for good wall conditioning before trying high-beta operation, if this poses the risk of a major NTM, possibly resulting in locking and disruption. | ||
| Resource Requirements: | pellet, impurity pellet, laser blow-off? | ||
| Diagnostic Requirements: | all edge diagnostics | ||
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| Title | 143: Offset velocity of NTMs | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye | ITPA Joint Experiment : | No |
| Description: | Establish at least the direction in which the NTM rotates, in the plasma frame of reference: whether it is the ion or the electron diamagnetic direction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The NTM offset velocity is the small difference between two large velocities: of the fluid and of the mode, measured respectively by CER and by magnetics. We need to minimize the error on these measurements, or on their difference. For this, we will adopt two approaches. In the first one, we will keep the rotation as constant as possible for as long as possible (e.g. by means of J. Ferronā??s algorithm), average the two velocities over that long interval, and take the difference. The second approach is a bit more risky in terms of scenario: the idea is to keep the total NBI power constant (or, even better, beta constant) but alternatively inject the beams in the co and counter direction. The consequent modulation of co/ctr-torque injection should take place on a time-scale of 2-3 momentum confinement times (200-300ms), to give the plasma the time to respond. The plasma will alternatively rotate in the co- and ctr-direction. If done carefully, the CER fluid rotation will average to zero, on a long time-scale. As a result, the time-averaged Mirnov measure of the mode rotation in the lab frame will, de facto, represent the mode rotation in the plasma frame.
Both approaches should be repeated for Ohmic and NBI plasmas, where electron and ion diamagnetic effects, respectively, are expected to dominate. | ||
| Background: | The Neoclassical toroidal Velocity (NTV) theory has successfully predicted the rotation frequency of Resistive Wall Modes (RWMs) in the plasma frame of reference. This natural, ā??offsetā?? rotation frequency is a fraction of the Alfven frequency. Works by H.Wilson et al. suggest that Neoclassical Tearing Modes (NTMs) obey a different physics, related with the diamagnetic frequencies. However, there is still little agreement and experimental validation on whether and under which conditions NTMs should rotate in the ion or the electron diamagnetic direction, and at which fraction of the corresponding diamagnetic frequency. | ||
| Resource Requirements: | 210 NBI | ||
| Diagnostic Requirements: | CER | ||
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| Title | 144: Imaging island formation for various NTM triggers | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Study NTM triggering by resolving ā??e.g. by ECE- the formation of a neoclassical island triggered by an ELM, sawtooth, fishbone or other trigger. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Consider three types of discharges: ELMing, sawtoothing and fishboning. Slowly ramp up NBI power. NTMs are expected to develop at the first event (ELM, sawtooth or fishbone) occurring after beta has exceeded the threshold for NTM meta-stability, provided such event is intense enough to exceed the marginal stability condition. The scope is to image the formation of the island with ECE and, at the same time, acquire as many data as possible to reconstruct the dļ?·/dt vs. ļ?· stability curve. This serves to determine when the plasma enters in an NTM-unstable regime, according to Rutherfordā??s theory, and compare the measured width with the calculated marginal and saturated width. | ||
| Background: | The modified Rutherford equation satisfactorily predicts under which conditions a plasma is NTM-unstable, i.e. neoclassical islands can form. However, very little is known on when and how NTMs do form. We only know that NTMs need to be triggered by other MHD instabilities such as ELMs, sawteeth and fishbones. | ||
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| Title | 145: Beam Ion Losses Due to AE's and EGAMs | ||
| Name: | Fisher fisherr@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | W. Heidbrink, M. Van Zeeland, D. Pace and EP Working group | ITPA Joint Experiment : | No |
| Description: | The goal is to better understand the beam ion losses due to Alfven eigenmodes and EGAMs using data from the new midplane FILD2 detector not available during the March 2010 dedicated experiments | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat experiments on beam ion losses due to EGAMs, TAEs and RSAEs in oval plasmas (~ shots 142111 to 142116) with the important addition of the new midplane FILD2 diagnostic. Then run similar experiments using single null divertor plasmas and compare results. | ||
| Background: | Reverse orbit modeling is being used to better understand the physics of the interactions between beam ions and AEs and EGAMs. This modeling predicts that the results for new midplane FILD2 detector should be significantly different from the existing FILD1 detector at 225R-1. Data using both detectors should allow us to better understand the loss mechanisms. The losses at the new midplane FILD2 should be larger for standard divertor plasmas. Large losses to the FILD1 diagnostic were only observed for oval plasmas, but this is based on a very limited number of shots and FILD insertion distances. | ||
| Resource Requirements: | ~ full day experiment | ||
| Diagnostic Requirements: | both FILD diagnostics, FIDA, neutrons, core plasma diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 146: Error field correction for DIII-D and ITER (1): Locked mode phase | ||
| Name: | Strait strait@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Requested |
| Co-Author(s): | M. Chu, A. Garofalo, R. La Haye, H. Reimerdes, M. Schaffer, F. Volpe | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to develop the physics basis for an efficient method of determining error field correction in ohmic or low beta plasmas. The capability to determine the error field correction continuously during the evolution of a single discharge would be beneficial to DIII-D and to ITER.
Such a method would be a significant advantage over the usual 4-quadrant, locked mode onset method used for low beta plasmas, which typically requires four repeated discharges in order to obtain a measurement at a single time during the discharge evolution. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The proposed approach is based on the addition of a rotating n=1 perturbation to the static error field, in the presence of a locked mode. The resulting toroidal phase of the locked mode depends on the vector sum of the two non-axisymmetric fields, as well as a possible viscous torque between the stationary island and the rest of the plasma. The error field and viscous torque can be determined by fitting the measurements to an appropriate model.
A key element of the experiment is to make the measurement under conditions of weak co- and counter-neutral beam injection as well as ohmic plasmas, in order to demonstrate that an unknown viscous torque can be successfully fitted in the model. | ||
| Background: | The phase of a locked mode has been observed to vary nonlinearly with the phase of a rotating applied n=1 field [F. Volpe, Phys. Plasmas 16, 102502 (2009)], as expected in the presence of an additional static error field. Preliminary analysis of a single ohmic case [E. Strait, APS-DPP talk XO4.002 (2010)] shows that the resulting error field estimate is reasonably consistent with that of the 4-quadrant method. Additional data and variation of the viscous torque are needed to establish the physics basis for this approach. | ||
| Resource Requirements: | I-coil with SPAs, for application of a slowly rotating n=1 field.
One co- and one counter-NBI source, for variation of plasma rotation. | ||
| Diagnostic Requirements: | Magnetic diagnostics, including RWM pairs for locked mode measurements. CER measurements desirable but not essential. | ||
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| Title | 147: Role of MHD in Disruption Runaway Suppression | ||
| Name: | Humphreys humphrys@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | N. Commaux, N. Eidietis, T. Evans, E. Hollmann, V. Izzo, A. James, T. Jernigan, P. Parks, J. Wesley, J. Yu | ITPA Joint Experiment : | No |
| Description: | The goals of this 1-day experiment are to study the role MHD activity plays in post-thermal quench RE suppression/deconfinement, and to identify scenarios for producing effective levels of MHD for rapid plasma shutdown. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Using standard low-kappa limited or LSN target plasmas with good RE seed confinement (latter developed in other proposed experiments: low li?), the nature of unstable MHD eigenmodes will be varied by applying co- or ctr-ECCD, varying edge q95, beta, and/or rotation. Ar pellets will be injected to trigger disruption and observe the impact on conversion of thermal current to RE. Specific choices may be guided by GATO and/or NIMROD studies predicting varying effectiveness of RE seed (de)confinement, making this experiment an important illustration of scenario development through theoretical prediction. | ||
| Background: | Recent analyses with GATO (Kornbluth) and NIMROD (Izzo) have suggested that the nature of the MHD instabilities triggered during a disruption may determine the size of the RE seed population that survives the thermal quench. Modes with large amplitude penetrating to the core (consistent with deconfinement of seeds even at the core) are correlated with low post-thermal quench RE current. Modes with zero amplitude at the core are correlated with high post-thermal quench RE current. Thus, the amplitude of MHD activity and the spatial form of the unstable eigenmodes may provide effective knobs to enhance the effectiveness of RE current mitigation. | ||
| Resource Requirements: | Cryogenic Ar killer pellet injection to generate RE | ||
| Diagnostic Requirements: | Usual disruption diagnostics: 5 kHz magnetics, DISRAD,
Thomson, new IR camera strongly desirable, fast cameras (LLNL and UCSD), UCSD scintillators for RE detection, FPLASTIC | ||
| Analysis Requirements: | JFIT; RE simulation codes (Parks/Humphreys); KPRAD, other impurity/radiation codesā?¦ | ||
| Other Requirements: | |||
| Title | 148: Error field correction for DIII-D and ITER (2): Multiple coil arrays | ||
| Name: | Strait strait@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Requested |
| Co-Author(s): | M. Chu, A. Garofalo, R. La Haye, H. Reimerdes, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to understand the importance of multiple modes in the plasmaā??s response to error fields, and to improve error field correc
A successful experiment would: (1) Yield improved error correction for DIII-D (2) Provide guidance on how best to use ITERā??s arrays of error field correction coils (3) Test IPEC predictions of optimum error field correction (4) Provide direct evidence of multiple-mode coupling to the error field |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiment will apply n=1 error field correction with three independent toroidal arrays of coils: upper I-coils, C-coils, and lower I-coils.
The number of free parameters for the coil currents is reduced from 6 to 2 by fixing the relative amplitude and phase of the three arrays in order to match either (a) the observed locked mode structure, or (b) IPEC calculations of the ā??most dangerousā?? error field. The free parameters now consist of a single amplitude and phase, which can be optimized empirically using standard methods. Error field correction is evaluated by the ohmic plasma low-density limit. Comparison to the standard EFC algorithm will show whether optimization of the spectrum is important. Comparison of methods (a) and (b) above will validate IPEC predictions. Observation of a locked mode that has a measurably different structure than a locked mode without error correction will be direct evidence of participation by additional modes. | ||
| Background: | A similar optimization approach was used several years ago to determine that a 240-degree phase difference between the upper and lower I-coils was best. This led to better error correction than with the C-coil. Here we extend the approach to the optimization of three coil sets, and expect an improvement over error correction with the I-coil alone.
Previous work on multi-mode error field correction showed improvement through optimized use of the old n=1 coil and the C-coil [J.T. Scoville and R.J. La Haye, Nucl. Fusion 43, 250 (2003)]. To date, there has been no attempt to optimize the combined use of the I-coil and C-coil for error field correction. | ||
| Resource Requirements: | I-coils and C-coils with SPAs. The coils must be connected in 9 independent anti-symmetric pairs, in order to allow arbitrary amplitude and toroidal phase of n=1 for each of the three arrays of coils.
It may be desirable to add waveforms to the PCS that will allow scaling of the amplitude and rotation of the toroidal phase for a given current distribution in the three rows of coils. | ||
| Diagnostic Requirements: | Magnetic diagnostics, including RWM pairs for locked mode measurements. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 149: Runaway Electron Suppression by Applied Nonaxisymmetric Fields | ||
| Name: | Humphreys humphrys@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | N. Commaux, N. Eidietis, T. Evans, E. Hollmann, V. Izzo, A. James, T. Jernigan, P. Parks, J. Wesley, J. Yu | ITPA Joint Experiment : | No |
| Description: | Continue study of RE suppression using nonaxisymmetric fields applied by the I-coil. Study effects of I-coil fields on thermal quench MHD eigenmodes as a possible mechanism for the previously-observed deconfinement. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Using LSN target plasmas with good RE seed confinement (latter developed in other proposed experiments: low li?), applied n=1 or n=3 I-coil fields will be alternated with no applied field to study effects on MHD spectra and post-thermal quench RE current amplitude. Ar pellets will be injected to trigger disruption and observe the impact on conversion of thermal current to RE. | ||
| Background: | Experiments over the last two years have shown apparent suppression of RE with n=3 at full I-coil current in several discharges, but attempts to reproduce the effect have been inconsistent. Low kappa limited plasmas were developed last year with very high reliability of RE current conversion, but these plasmas showed no response to applied I-coil fields. A new LSN target is expected to be developed (with low li?) based on theoretical predictions of good seed confinement, with a plasma boundary very close to the I-coils. Such a target may show a strong contrast in post-thermal quench RE current between no I-coil field and applied I-coil field. | ||
| Resource Requirements: | Cryogenic Ar killer pellet injection to generate RE | ||
| Diagnostic Requirements: | Usual disruption diagnostics: 5 kHz magnetics, DISRAD,
Thomson, new IR camera strongly desirable, fast cameras (LLNL and UCSD), UCSD scintillators for RE detection, FPLASTIC | ||
| Analysis Requirements: | JFIT; RE simulation codes (Parks/Humphreys); KPRAD, other impurity/radiation codes | ||
| Other Requirements: | |||
| Title | 150: Shear and Beta Dependence of Turbulence and Heat Flux to test Profile Stiffness | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | C. Holland, C. Petty, T. Rhodes, L. Schmitz, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Examine the variation of electron and ion heat flux with local (ion/electron) normalized temperature gradient scale length as a function of rotation, and simultaneously measure profile of turbulence characteristics across the mid-radial region. Perform a power scan at each rotation to test predictions of profile stiffness. Study 2D turbulence structure as a function of rotationally varied ExB shear in long-duration, medium to high beta hybrid H-mode plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform a power (beta) scan at constant (low) rotation in long-pulse low-current Hybrid discharges. Hybrids will be utilized to obtain quasi-steady, sawtooth free H-mode plasma conditions. Low-current will maximize the turbulence amplitudes in these relatively low-fluctuation H-modes (successful in Te/Ti scan in 2010). Start with as low a rotation in which a hybrid H-mode can be sustained (typically, a finite co-current toroidal rotation velocity is required to avoid NTMs). Increase power at constant PCS feedback-controlled rotation to scan heat flux and examine flux vs. normalized temperature gradient for ions and electrons. Increase rotation (all co-current injection) and perform similar power scan up to the maximum sustainable beta. Obtain turbulence measurements with the 8x8 BES array to examine structure, amplitude and eddy shear as heat flux is varied. Use DBS, FIR and other fluctuation diagnostics to fully examine turbulence characteristics for comparison with model predictions. Presumably, turbulence amplitude will increase significantly with heat flux once above the critical gradients. If possible, perform a similar scan at intermediate rotation values. | ||
| Background: | Transport in Hybrid scenario discharges has been shown to depend strongly on the toroidal Mach number (M = v_tor /c_s). Profile stiffness (function of gradient of heat flux vs. normalized temperature gradient) has also been shown on JET to vary with shear, becoming less stiff at higher rotation (higher ExB shear), [Mantica-IAEA-2010]. By varying the injected neutral beam torque into hybrid plasmas and simultaneously maintaining beta constant via feedback control, the "H-factor" decreases by approximately 20% as the Mach number is reduced from about M=0.5 to M=0.1 [Poitzer, Nuclear Fusion-2008]. This has been shown to be consistent with the a reduction in ExB shearing at lower Mach number from GLF23 modeling.
Previous measurements of turbulence characteristics in hybrid discharges (McKee, APS-2005) with the upgraded BES diagnostic, showed that turbulent eddies exhibit a strongly tilted structure. This is in sharp contrast to the more radially-poloidally symmetric eddy structure typically observed in the core of L-mode discharges. The direction of this tilted eddy structure is consistent with the ExB shear flow in these plasmas. The combination of detailed turbulence and transport measurements will allow for comprehensive tests of the profile stiffness predictions from TGLF and other models. This experiment will contribute to stiffness/critical gradient studies, transport model validation, and fundamental studies of turbulent transport. | ||
| Resource Requirements: | 7 sources NBI | ||
| Diagnostic Requirements: | BES, DBS, FIR, PCI, CECE, UF-CHERS, full profile diagnostics | ||
| Analysis Requirements: | Transport models (TGLF and others) | ||
| Other Requirements: | |||
| Title | 151: Bifurcated helical core equilibrium at DIII-D | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | W.A.Cooper, R. Buttery, A. Cole, T.E.Evans, J.R.Ferron, E.A.Lazarus, H. Reimerdes, M.A.Van Zeeland, J.Yu | ITPA Joint Experiment : | No |
| Description: | Evolution of helical cores with improved confinement and stability properties was reported recently from RFX [R. Lorenzini et al. Nature Physics 5 (2009) 570]. These regimes are an attractive option to operate fusion plasmas concerning improved stability and confinement properties motivating exploration in tokamaks. A bifurcation into a helical equilibrium state with axisymmetric boundary is predicted [W.A.Cooper et al PRL 105, 035003 (2010)] for plasmas with flat core q profiles with low q_min. The trigger for the bifurcation event is a small displacement of the magnetic axis as perturbed initial condition in the modeling. This can now potentially be realized experimentally utilizing the unique DIII-D capability of off-axis beam heating in combination with the real time control of the current profile evolution achieved earlier [J.R.Ferron et al Nucl. Fusion 46 (2006) L13-L17]. The aim of this experiment is the demonstration of a controlled and sustained generation of a plasma state with a helical core and - if obtained - study the interaction with an edge resonant magnetic perturbation field. This is an exploratory experiment and the potential result would be transformational as this helical core shall represent a stage of minimized free energy being more stable against internally and externally induced instabilities. We therefore submit this proposal as a Torkil Jensen Award application. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The key ingredients identified by VMEC like stability modeling with the ANIEC code [W.A.Cooper et al PRL 105, 035003 (2010)] are a flat or hollow q-profile with q_min close to unity. In addition there is evidence for a Exact specification of the experimental sequence needs preparatory modeling. A task ordered sketch of the experimental plan looks as follows: - establish discharge with suitable q-profile Test how magnetic axis could be inductively displaced: - use off-axis beam heating during current profile evolution - use off-axis beam heating at start of current flat top Test how magnetic axis could be non-inductively displaced - use off-axis ECCD during current profile evolution - use off-axis ECCD at start of current flat top Both attempts will be run with feedback on the q-profile [J.R.Ferron et al Nucl. Fusion 46 (2006) L13-L17] to maintain the low to inverse share q-profile susceptible to the bifurcation This scan will be repeated for various plasma stages, i.e. L-mode, low power to high power H-mode to explore dependence on Follow up studies once this stage is achieved will allow addressing topics as stability and core confinement with a helical core, NTV driven transport and in particular combination with edge RMP fields to shape the boundary in presence of a helical core. | ||
| Background: | Confining high energy plasmas in tokamaks with a strong, axisymmetric magnetic cage often stimulates instabilities which in most cases will result in non-axisymmetric structures. The 3D filaments of ELMs, sawtooth crashes in the plasma center and tearing modes forming helical islands as state of minimized free energy are examples. Often the plasma confinement is degraded by these events with significant impact for the intimate contact of the expelled fluxes with the wall elements. Observations form the RFX reversed field pinch with a self organized helical core [R. Lorenzini et al. Nature Physics 5 (2009) 570] resulting in a stable, enhanced confinement plasma motivates assessing the feasibility of this regimes at tokamaks.
Employing an energy minimization scheme within ideal MHD revealed that plasma with low shear core q-profiles and qmin~1 can bifurcate from an axisymmetric core solution into a helical core with slightly reduced free energy [W.A.Cooper et al PRL 105, 035003 (2010)]. Evidence for transient transitions into such a regime were found, e.g. the "snake structure" at JET [J.A.Wesson et al., PPCF 37, A337 (1995)], sawtooth free discharges at TCV [H.Reimerdes et al., PPCF 48, 1621 (2006), Y. Camenen et al., NF 47, 586 (2007)]. The rational for this proposal comes from the fact that [W.A.Cooper et al PRL 105, 035003 (2010)] presents the potential to trigger formation of a helical core in tokamaks by introducing small perturbations of the initial condition. In this study the magnetic axis was displaced in stable equilibria solutions for plasmas with a flat or hollow core q-profile, qmin~1 and finite As such a helical core will potentially exhibit improved stability and confinement properties, assessment of this stage could open a way to new operational regimes and will enable studying 3D effects on transport and confinement in detail. In combination with the edge resonant magnetic perturbation field coil set at DIII-D (I- and C-coils), a combination of helical (3D) cores with a 3D shape boundary can be studied in a unique fashion at DIII-D. The technical requirements involve control of the q profile to establish a discharge with a flat core q-profile and qmin~1 in a range of <ļ?¢>. At DIII-D, discharges with feedback controlled q-profiles were achieved [J.R.Ferron et al Nucl. Fusion 46 (2006) L13-L17]. The new capability of off-axis NBI heating enables to displace inductively the magnetic axis during start up and/or at the beginning of the flat top. In addition, on-axis or off-axis ECCD heating could be used to tailor the current profile and by that the position of the | ||
| Resource Requirements: | Most of the detailed - plasmas with high triangularity interesting (ISS shape) due to relevance for ELM suppression, oval shaped plasmas as contrast, maybe intermediate shapes and also DN plasmas, matter of preparatory modelling - scan from L-mode to H-mode with various heating levels (<ļ?¢> and ļ?¢N dependence) - need off-axis beam heating - need ECCD for current profile control - need q-profile feedback a lĆ” [J.R.Ferron et al Nucl. Fusion 46 (2006) L13-L17] | ||
| Diagnostic Requirements: | All diagnostics able to detect a helical deformation of the plasma core and help to judge on the extension of this helicity towards the edge are useful.
- MSE measurements for q-profile feedback - Polarimeter on fast Li beam for edge q-profile measurements - SXR poloidal arrays - Divertor SXR (new) - Magnetic sensors - Profile measurements: TS, DBS, BES, fast reflectometer, ECE, ECE-I - Fast UCSD camera with FFT analysis (VanZeeland/Yu) (maybe combined with pellet injection) - New LLNL periscope - Divertor IR and visible cameras - Filter scopes | ||
| Analysis Requirements: | kinetic EFITS and high quality profile data will be essential
ANIMEC/VMEC modelling of plasmas with indications for helical core and control shots w/o bifurcation evidence | ||
| Other Requirements: | Explore bifurcation phenomena for DIII-D reference equilibria with ANIMEC code [W.A.Cooper et al PRL 105, 035003 (2010)] to define exploration space (started for discharges suggested by E.A.Lazarus) | ||
| Title | 152: Destabilization of sawteeth by local ECCD in the presence of energetic ions | ||
| Name: | Chapman ian.chapman@ccfe.ac.uk | Affiliation: | CCFE |
| Research Area: | General IP | Presentation time: | Requested |
| Co-Author(s): | RJ Buttery, R La Haye, B Pinsker, V Igochine (IPP), O Sauter (CRPP) | ITPA Joint Experiment : | Yes |
| Description: | ITER will need to deploy ECCD sawtooth destabilisation to avoid large sawteeth triggering the onset of low betan NTMs with potentially large size at mode onset. Demonstrations to date have not been in relevant ITER-like baseline scenarios with significant heating power and fast ion beta. Therefore a demonstration of the technique proposed for ITER (with ECCD to change local magnetic shear) is required, to ascertain whether other strategies (eg ICRH) will be required. This demonstration should start with testing the principle of whether the sawtooth can be controlled in the right regime, and extend to tracking the q=1 radius to provide a viable demonstration for ITER and to ascertain impact on NTM threshold. DIII-D can make a major impact in this area, which is of high priority for ITER baseline scenario. Key part of ITPA MDC-5, comparison experiment with AUG | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop ELMy H-mode low-density scenario with long period sawteeth in the presence of NBI in the core. Sweep co-ECCD from outside q=1 to inside q=1. Repeat scan with counter-ECCD. Heating ramps should be applied to measure 3/2 and 2/1 mode onset thresholds comparing ECCD and no ECCD cases. Ideally, real time systems can be used to keep ECCD tracking q=1 surface. | ||
| Background: | Long period sawteeth have been observed to result in the low-beta triggering of neoclassical tearing modes [Chapman, NF 2010], which can significantly degrade plasma confinement. In ITER this problem is likely to be exacerbated by the strongly stabilising effect of the energetic alpha particles, which are predicted to lead to long sawtooth periods. Consequently it is imperative to develop a robust control actuator which can deliberately destabilise the sawteeth, even in the presence of very energetic ions. Initial results from Tore Supra [Lennholm et al, PRL, 2009] and AUG [Igochine, accepted PPCF, 2011] suggest that ECCD can destabilise sawteeth in the presence of ICRH fast ions. However, sawtooth control by tailoring the magnetic shear remains to be demonstrated in high performance plasmas with a significant fast ion beta. See Chapman PPCF 2011 for topical review on this subject. | ||
| Resource Requirements: | Full power from on-axis co-beams (plus off-axis counter beams if possible?), Maximum ECRH power from gyrotrons with mirrors for sweeping desposition, ICRH may be advantageous for creation of fast ion population, real time mode targeting systems | ||
| Diagnostic Requirements: | CER, MSE, Magnetics, TS, SXR, FIDA | ||
| Analysis Requirements: | Transport simulations (eg ASTRA) to model ECCD and changes in magnetic shear (aided by MSE equilibrium reconstruction). TRANSP simulations for fast ion population. Drift kinetic simulations to show effect of fast ions on kink stability | ||
| Other Requirements: | |||
| Title | 153: Heat loads from different beams on the TBM protective tiles. | ||
| Name: | Kramer gkramer@pppl.gov | Affiliation: | PPPL |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | Mike Schaffer, Raffi Nazikian, Mike van Zeeland | ITPA Joint Experiment : | No |
| Description: | Heat loads on the plasma facing surfaces of the TBMs are of a concern in ITER. In the 2009 TBM mock-up experiments in DIIID heat loads on the TBM tiles were measured that were caused by beam-ion losses from co-going beams. The co-going beams, however, span only a small range of pitch angles while in ITER a uniform distribution of pitch angles is expected from the alpha particles. We propose to use various combinations of co, counter, and off-axis beams to probe different parts the pitch angle distribution and compare the results with simulations from beam-ion loss codes to validate those codes. The validated codes can then be used to make more accurate head load predictions for ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create discharges similar to 140144, 140153, and 140156 and inject in separate discharges the following combination of beams:
1. co-going beams 330L and 330R, 2. counter-going beams 210L and 210R, 3. off-axis beams 150L and 150R For each beam combination create discharges with and without the TBM fields engaged at different gaps between the separatrix and TBM surface. | ||
| Background: | In the 2009 TBM experiments it was found that the two middle protective tiles of the TBM mock-up were heated significantly when the TBM fields were present and NBI was injected. This heating increased with a decreasing gap between separatrix and TBM surface. Both the location and the trend in the heat load could be reproduced by beam-ion loss calculations with various codes. In those simulations it was found that beam-ions deposited near the edge were lost to the TBM surface in the presence of the TBM fields. In those experiments only co-going beams were used, spanning only a small range of pitch angles. By replacing the co-going beams with counter-going ones, we can increase the range of pitch angles and study the effects of TBM fields on counter-going beams. Predictions with the full orbit following code indicate that the hot spots on the TBM tiles generated by the counter-beams are 50% higher than with the co-beams. | ||
| Resource Requirements: | TBM mock-up, co-, counter-, and off-axis NBI. | ||
| Diagnostic Requirements: | Thermocouples array on the TBM tiles, fast-ion loss diagnostics, ... | ||
| Analysis Requirements: | |||
| Other Requirements: | Main analysis codes: ASCOT, OFMC, SPIRAL, DELTA5D, TRANSP. | ||
| Title | 154: Experimental prove of fast-ion heating of the TBM tiles. | ||
| Name: | Kramer gkramer@pppl.gov | Affiliation: | PPPL |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | Mike Schaffer, Raffi Nazikian, Mike van Zeeland | ITPA Joint Experiment : | No |
| Description: | Heat loads on the plasma facing surfaces of the TBMs are of a concern in ITER. In the 2009 TBM mock-up experiments in DIIID heat loads on the TBM tiles were measured. There are strong indications that those heat loads were caused by TBM-induced beam ion losses. In order to prove unambiguously that those losses are caused by fast ions and not by thermal plasma touching the tiles we would like to compare discharges one heated with NBI and the other with ECCH in which the TBM fields are present. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create a discharge that is suitable to be heated with ECCD and with NBI. Minimize the gap between the separatrix and TBM surface for the largest effects on the TBM tiles and measure the heat load in both cases. | ||
| Background: | In the 2009 TBM experiments it was found that the two middle protective tiles of the TBM mock-up were heated significantly when the TBM fields were present and NBI was injected. In beam-ion loss calculations it was found that the heating could be caused by the deposition of lost beam ions. In order to prove unambiguously that the hot spots are created by fast ions rather than by thermal plasma pulled to the TBM surface, two similar discharges should be created, one with NBI heating and another where the NBI power is replaced by the same amount of ECCD power. | ||
| Resource Requirements: | TBM mock-up, NBI and ECCH at comparable power levels. | ||
| Diagnostic Requirements: | Thermocouples array on the TBM tiles, fast-ion loss diagnostics, ... | ||
| Analysis Requirements: | Main analysis codes: ASCOT, OFMC, SPIRAL, DELTA5D, TRANSP. | ||
| Other Requirements: | |||
| Title | 155: Effect of TBM/RMP on the confinement of trapped and passing fast ions | ||
| Name: | kurki-suonio none | Affiliation: | Aalto U |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | William Heidbrink | ITPA Joint Experiment : | No |
| Description: | Investigate the effect of local magnetic perturbation on the transport and confinement of fast ions on qualitatively different orbits. Also looking at the beta dependence would be very useful. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Diagnose the confinement properities of both on-axis and off-axis NB-ions with and without the TBM Mockup. The cleanest results can naturally be obtained using only one beam at a time. Different perturbation strengths (at least two) would be useful to get a more quantitative grip on the issue. Time and resources permitting, studying the confinement with TBM mockup at two very different beta-values would be extremely useful. | ||
| Background: | Continuation of the 2009 TBM mockup experiments and ITER fast ion wall load studies using ASCOT (see NF 49 (2009) 095001). | ||
| Resource Requirements: | Most perpendicular and most off-axis beams | ||
| Diagnostic Requirements: | FIDA spectrometers & cameras, edge plasma parameters, thermocouples | ||
| Analysis Requirements: | NUBEAM & FIDASIM | ||
| Other Requirements: | |||
| Title | 156: Sensitivity of TGLF predicted transport to high-k modes in Hybrid plasmas | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | T. Rhodes, G. R. McKee, C. C. Petty, J. Kinsey, G. Staebler | ITPA Joint Experiment : | No |
| Description: | Explore the sensitivity of TGLF predicted transport to high-k (ETG) modes in hybrid plasmas via a rotation scan to eliminate low-k contributions to transport, leaving only high-k. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using low q95 Hybrid discharges with strong rotation variations, while keeping beta fixed by varying input beam energies, find an experimental condition where the ExB shear is expected to result in strong reduction of ion transport (Chi_i approaching neoclassical (Kinsey APS 2007)). In these cases, the Te profile predicted by TGLF in hybrids depends sensitively on high-k contributions. Vary effectiveness of electron heating by scanning density, try to produce at least three different values of Te(0). Measure high-k density fluctuations with high-k backscattering. Compare TGLF predicted profiles and fluxes to measured profiles and fluxes, compare changes in ETG growth rate to changes in measured density fluctuations. Monitor low-k with BES. Can we make a Hybrid plasma that is compatible with high-k backscattering measurements? | ||
| Background: | Balance between low-k and high-k transport remains unresolved in L-mode, H-mode and Hybrid plasmas, however, there is very strong evidence from past experiments and theory that high-k modes can become dominant in hybrid plasmas as EXB shear is increased as CHi_i approaches neoclassical values (Politzer, Kinsey APS 2007). | ||
| Resource Requirements: | co and counter NBI sources | ||
| Diagnostic Requirements: | full profile diagnostics, CER, MSE,
BES and high-k backscattering | ||
| Analysis Requirements: | ONETWO, TRANSP, TGLF linear stability analysis and TGYRO wrapped around TGLF flux matched profile predictions | ||
| Other Requirements: | |||
| Title | 157: Improvement of advanced scenarios by counter ECCD | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | General IP | Presentation time: | Requested |
| Co-Author(s): | M. Henderson | ITPA Joint Experiment : | No |
| Description: | Modify Advanced Scenarioļæ½??s plasma current profile using a combination of co-ECCD off-axis and counter or pure ECH inside of barrier. The counter-ECCD is used as a tool in controlling the shear from weak to strong, resulting in an increase bootstrap current and stronger ITB. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Repeat the plasma discharge with 100% non-inductively driven plasma current (133103), additional EC power is added inside of the ITB with first pure heating. Then, on a shot-to-shot basis, the toroidal injection angle is added for negative current drive inside the ITB. | ||
| Background: | Use of Counter ECCD in the fully non-inductive ITB scenarios. Recently DIII-D used co-ECCD off axis to enhance the ITB resulting in an increased bootstrap current and a slightly high off axis peaked plasma current profile. The co-ECCD contribution (and resulting BS current) was crucial in achieving the non-inductive discharge. Previous work from TCV has demonstrated that a small contribution of counter ECCD in the center has a significant impact on the ITB. The TCV results demonstrated in some cases that the net current from I_counter (0.5MW) + I_co(1.0MW) + I_BS exceeds the net current with co-ECCD off axis alone I_co(1.5MW) + I_BS. A similar scenario development using a combination of co and counter ECCD could be practical for optimizing the EC system on ITER and justify or invalidate the recently accepted PCR-098 | ||
| Resource Requirements: | 1-2 run days. 5 to 6 gyrotrons (depending on availability. | ||
| Diagnostic Requirements: | Equivalent to shot 133103 | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 158: Search for ETG critical gradient in ITB plasmas | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | M. E. Austin, B. Pinsker, C. C. Petty, T. Rhodes, G. R. McKee, J. Kinsey, G. Staebler | ITPA Joint Experiment : | No |
| Description: | Search for the existence of an ETG critical gradient across the radius in ITB plasmas using ECH and FW to make extreme variations in the electron temperature profile. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using the q=2 surface ITB trigger recipe (Austin), produce ITB plasmas at fixed rotation (higher rotation with all co-NBI leads to longer lived ITBs, so this is best target).
Use preheating with ECH to improve FW coupling in core. FW will couple well if we boost core Te to 6 o 7 keV with ECH. We should attempt to make ITB plasmas with low Te(0) = 5-6 keV with beams only, then high Te(0) = 9-12 keV with beams and FW and some ECH, then extremely high Te(0) > 12 keV with beams, FW and all ECH available. Examine changes in Te profiles, R/LTe and high-k density fluctuations as Te is scanned using combination of ECH and FW. Make the base ITB with as little co-NBI power as possible so that we can increase rotation if ECH slows down the plasma rotation. It may also be useful to match the density in all cases, but if extremely high Te is only attainable with lower ne, then we should reduce ne. Compare observed changed in transport and fluctuations with the ETG critical gradient model. Need careful monitoring of Te profile across the radius with 32-channel ECE and TS. | ||
| Background: | The ETG mode is predicted to have a critical gradient (Jenko) that depends on Te/Ti Other critical gradient models also vary with Te/Ti. If Te = Ti the critical gradient for ITG equals that of the ETG and the Te / Ti dependence is strong (Horton). For ITG, the threshold is R/LTi=1.88 |s/q| (1 + Ti / Te). If we make extremely different Te and Ti profiles, as is possible with ITB plasmas, we can explore the ETG turbulence (high-k) response and predictions for the changes in critical gradient | ||
| Resource Requirements: | co and counter beams, FW and ECH | ||
| Diagnostic Requirements: | standard profile diagnostics, CER, MSE, high-k backscattering, BES | ||
| Analysis Requirements: | ONETWO, TRANSP, TGLF, ECESIM, CQL3D | ||
| Other Requirements: | -- | ||
| Title | 159: Toroidal rotation studies with TBM | ||
| Name: | Tala Tuomas.Tala@vtt.fi | Affiliation: | VTT Technical Research Centre |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, M. Schaffer, A. Salmi, T. Koskela, T. Kurki-Suonio | ITPA Joint Experiment : | No |
| Description: | A series of experiments was performed on DIII-D to mock-up the field that will be induced in a pair of ferromagnetic Test Blanket Modules (TBMs) in ITER to determine the effects of such error fields on plasma operation and performance. In the TBM experiments, the following effects were studied: plasma rotation and locking, confinement, L-H transition, edge localized mode (ELM) suppression by resonant magnetic perturbations, ELMs and the H-mode pedestal and energetic particle losses. The largest effect was slowed plasma toroidal rotation v across the entire radial profile by as much as ~50% decrease due to TBM. A decrease in global density, beta and confinement were typically ~3 times smaller. Non-resonant braking by NTV has been identified as the most probably candidate to explain the dramatic decrease of toroidal rotation caused by the TBM. Theoretical NTV braking by the TBM mock-up field was evaluated numerically using the code IPEC, and the computed global NTV torque was ~3 times larger than the braking torque inferred from the experiment. However, in the previous TBM experiments, no detailed rotation and momentum transport studies were carried out to understand experimentally the physics on how the TBM affects toroidal rotation. The detailed rotation and momentum transport study is the scope of this experiment. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | There are several experimental methods, not exploited in the last campaign, to understand the mechanisms on how the TBM affects plasma rotation. The first part of the experiment is to separate the NTV type of torque and the torque due to fast ion losses. This requires similar pulses with co-NBI and counter-NBI. In the case of co-NBI, both NTV and fast ion losses induced torque are in counter-current direction,a nd thus difficult to identify their impact on rotation separately. However, in the counter-NBI case, NTV increases the rotation and fast ion induced torque decreases the rotation, and then by comparing the rotation in the co- and counter-current pulses, one can separate the two effects.
The torque due to fast ion losses can be compared with the ASCOT code calculation of the same quantity. The second experimental method is based on NBI modulation where it is possible to obtain both the momentum transport coefficients, diffusivity and pinch, separately. By comparing the diffusion coefficients between similar plasmas with TBM and without TBM, it will be possible to quantify the effect of TBM via chances in momentum transport, rather via the torque source(like NTV), on plasma toroidal rotation. NBI modulation technique has been used on DIII-D in the past without TBM tsuccessfully yield the momentum diffusivity and pinch. The third experimental method is based on NBI steps, also widely exploited on DIII-D to study the intrinsic rotation in beam heated or ECRH heated plasmas. Here, the idea is to make similar plasmas with and without TBM and with NBI steps, to quantify the intrinsic (non-NBI) torque. The resulting intrinsic torque will include all the torque terms besides the beams, i.e. it will contain all the terms like NTV, torque due to fast ion losses and the 'usual' intrinsic rotation torque. Subtracting the intrinsic torque from the plasma without TBM from the plasma with the TBM will yield the torque profile due to TBM. This can be further compared with NTV torque calculation, for example with IPEC code. | ||
| Background: | |||
| Resource Requirements: | TBM mock-up, counter and co NBI, ECH, divertor cryo pumps | ||
| Diagnostic Requirements: | Thomson scattering, CER, MSE, fast ion diagnostics | ||
| Analysis Requirements: | Detailed calculation of fast ion losses with ASCOT orbit-following code. ASCOT calculations for the contribution of fast ions to torque. NTV torque calculations, for example with IPEC. Detailed TRANSP and ASCOT analyses of the NBI torque. | ||
| Other Requirements: | |||
| Title | 160: Neon Shattered Pellet into RE beam | ||
| Name: | Jernigan jernigan@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | Baylor, Commaux | ITPA Joint Experiment : | No |
| Description: | Fire large, shattered, neon pellets into post disruption, RE discharges to dissipate the RE energy | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | See description | ||
| Background: | Massive neon injection into the RE discharges ha increased energy dissipation from the RE beam. The massive, rapid pulse of atoms from the shattered pellet may improve the dissipation. | ||
| Resource Requirements: | Noting unusual | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 161: Low Impurity RE Generation | ||
| Name: | Jernigan jernigan@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | Baylor, Commaux | ITPA Joint Experiment : | No |
| Description: | Extend the existing DIII-D RE generation prescription to lower impurity densities by using a small gas puff of neon or argon instead of the 2.7 mm argon pellet as a trigger. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use existing RE prescription (low elongation, limited) discharge but try a small impurity gas puff just large enough to trigger a radiation induced disruption. Start with neon (can be regenerated from cryopumps between shots), then argon (tends to stick to cryopumps between shots), then a 1.8 mm neon pellet, and finally a 2.7 mm neon pellet. Stop when a reliable RE beam is generated. | ||
| Background: | Following a suggestion by Robert Granetz (MIT), DIII-D has developed a recipe for reliable generation of RE beams during disruptions. It consists of using a low elongation, limited discharge with a 2.7 mm argon pellet as the disruption trigger. While we can reliably generate the RE's during the Ip decay, we still have a faster decay of the RE beam than other tokamaks like Tore Supra or Textor which may be due the large amount of argon in the pellet (7.2 Torr-liters). In addition the machine performance degrades during the day presumably due to the buildup of argon in the machine. Using small amounts of neon or argon may ameliorate these problems. | ||
| Resource Requirements: | Nothing unusual (except machine time) | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 162: Neon Shattered Pellet For Disruption Mitigation | ||
| Name: | Jernigan jernigan@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | Baylor, Commaux | ITPA Joint Experiment : | No |
| Description: | Use large, shattered neon pellets to try to reach the Rosenbluth density required to prevent RE generation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use the ORNL Shattered Pellet Injector (SPI) with neon to try to reach the Rosenbluth density. The existing pellet size (15.4 mm barrel) with neon has 135% of the electrons (free plus bound) required to reach the Rosenbluth density in DIII-D. | ||
| Background: | Previous experiments with the SPI using neon were unsuccessful due to entrained neon with the helium propellant blow-by. Several techniques are being explored to reduce or eliminate the blow-by. | ||
| Resource Requirements: | Successful modifications of the SPI to reduce of eliminate the blow-by. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 163: Small, Probing Pellets into RE Beam | ||
| Name: | Jernigan jernigan@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | Baylor, Commaux, Van Zeeland, Parks | ITPA Joint Experiment : | No |
| Description: | Fire small pellets into the decaying discharge for diagnostic and control purposes. Neon pellets would be an excellent trigger for the Putvinski scheme for RE control and D2 pellet could make a diagnostic probe for charge exchange analysis (Van Zeeland) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Stream a series of small pellets into a decaying disruption. D2 pellets could be used for diagnostic purposes, and neon pellets could be used the test the Putvinski proposal to prevent RE generation. | ||
| Background: | Putvinski has suggested a series of gas puffs during the Ip rampdown to trigger MHD events that would dump the RE's before a full RE beam was generated. A stream of small (1.8 mm), neon pellets would provide a reliable trigger source to test the sufficiency of late MHD events during the Ip decay. | ||
| Resource Requirements: | Requires a different disruption trigger than the argon pellet. See ROF proposal 161. Otherwise, nothing unusual. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 164: RE Prevention Below the Rosenbluth Density | ||
| Name: | Jernigan jernigan@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | Baylor, Commaux | ITPA Joint Experiment : | No |
| Description: | Try massive, impurity injection from the Medusa array to see if we can prevent the RE beam generation from the low elongation, limited discharges with argon pellets. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create RE discharges with the low elongation, limited configuration with impurity trigger. Use these discharges to see if MGI with the Medusa array can prevent the formation of the RE beam. | ||
| Background: | Theory indicates that a very high density ("the Rosenbluth density") will prevent RE formation during the current decay following a disruption. No runaways have been observed in MGI mitigated gas with full flow even though the densities are only about 10% of the Rosenbluth density. The objective is see if the obtainable densities are enough to prevent RE formation in an optimal RE generation regime. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 165: Study of stability of 'off-axis fishbone mode (OFM)' | ||
| Name: | Matsunaga matsunaga.go@qst.go.jp | Affiliation: | QST |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | M. Okabayashi(PPPL), M. Takechi(JAEA), A. Isayama(JAEA), Y. In (FAR-TECH) | ITPA Joint Experiment : | No |
| Description: | To clarify mechanism that 'Off-axis Fishbone Mode (OFM)' can induce RWM onset, investigation of OFM behavior will be continued on DIII-D. In particular, dependence of OFM stability on plasma rotation and beta will be investigated. Moreover, identification of original mode of OFM by MHD spectroscopy by using I-coil in the kHz range will be done. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Beta and plasma rotation scans will be done. In particular, to investigate resonance condition of OFM and EP, plasma rotation will be scanned so that OFM frequency will miss the precession frequency of trapped energetic particles including the counter direction. MHD spectroscopy with I-coils and synchronous detectors to pick up faint response signal in the kHz range will be done with frequency sweeping of I-coil current up to 10 kHz. If marginal stable mode exists in the range, a response can be measured as eigenvalue of transfer function. | ||
| Background: | In the wall-stabilized high-beta plasma on JT-60U, energetic particle driven mode named as 'Energetic particle driven Wall Mode (EWM)' has been observed. Since it is found that the EWM can induce RWM onset and trigger ELM, understanding of EWM physics is important. On DIII-D, similar mode is observed in the high-beta plasma; this mode was named as 'Off-axis Fishbone Mode (OFM).' The OFM can also induce RWM onset but not interact with ELM. EWM/OFM is predicted to appear in future reactor with high-beta operation, and plays an important role to determine achievable beta. At previous campaign, we have found the OFM mode frequency has dependence of plasma rotation. As one of possibilities, stable kink-ballooning mode (IKBM) is original mode of OFM. The mode frequency of stable IKBM is sensitive to plasma rotation and beta. To validate the hypothesis that OFM originates in IKBM, the dependence of plasma rotation and beta should be clarified. | ||
| Resource Requirements: | 1 day experiment with NBs to exceed no-wall beta limit and ECCD for NTM suppression | ||
| Diagnostic Requirements: | Standard for RWM and EP experiments | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 166: Quiescent H-modes with an externally driven EHO | ||
| Name: | Reimerdes reimerdes@fusion.gat.com | Affiliation: | CRPP-EPFL |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | A. Fasoli, A.M. Garofalo, J.M. Hanson, M.J. Lanctot, P.B. Snyder, W.M. Solomon, D. Testa | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to extend the parameter regime of quiescent H-mode discharges by driving a perturbation similar to the EHO in discharges where the edge transport usually results in ELMs. Even if an MHD mode is stable, it can be driven to a finite amplitude by applying a suitable non-axisymmetric magnetic perturbation with external coils (antennas) [A. Fasoli, et al., Phys. Rev. Lett. 75, 645 (1995), A.M. Garofalo, et al., Phys. Plasmas 10, 4776 (2003)]. Since the EHO has a dominant low n structure and rotates in the 5-10kHz frequency range, the DIII-D I-coil could apply a suitable external field. Driving a stable perturbation has the advantage that the perturbation amplitude can be controlled by the amplitude of the driving field and, therefore, does not only dependent on plasma parameters. In addition any resonant magnetic braking torque generated by the external field would pull the plasma towards the rotation frequency of the external field rather than zero, thereby avoiding the locking in the case the external field is too large. While an extended parameter regime could possibly result in an attractive ELM suppression technique for ITER, the response to the external field also yields information about the stability of the EHO. An external control of the EHO would also enable studies of the transport enhancement as a function of the mode amplitude at otherwise similar plasma parameters. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In this experiment the I-coil is used to apply a rotating n=1 field with a kink mode helicity (240-300Deg I-coil phasing with the exact phasing to be determined by modeling) at 5-10kHz. In order to access this frequency range and maximize the external field amplitude the I-coil will be connected in toroidally opposed anti-symmetrical pairs and powered by parallel audio-amplifier pairs. It is estimated that this configuration will result in approximately 200A of current in the I-coil. Possible target plasmas are:
1) Counter-rotating H-modes that are close to the QH-mode operating regime. 2) ELM-free H-modes, where a small enhancement of the particle transport should have a large effect on the density evolution. 3) Standard ELMing H-mode. Since the EHO stability is strongly affected by plasma rotation, the experiment includes frequency sweeps in order to find a frequency, where the external field couples best to an edge mode. Measurements will include magnetic measurements of the plasma response and measurements of the density and temperature profiles. Of interest are the modification of the time averaged kinetic profiles, which indicates transport changes, as well as the component that oscillates at the applied frequency, which indicates the perturbation structure. Since the rotation period of the external field is short compared to confinement time scales the oscillating component can be interpreted as a displacement of flux surfaces. | ||
| Background: | Quiescent H-modes, i.e ELM-free discharge at constant density and radiated power, but with improved energy confinement given by an edge pedestal, have been observed in various machines, such as the QH-mode in DIII-D [C.M. Greenfield, et al, Phys. Rev. Lett. 86, 4544 (2001)] and the Enhanced D-alpha H-modes in C-Mod [Y. Takese, et al., Phys Plasmas 4, 1647 (1997)]. In these operating regimes MHD fluctuations, namely the edge harmonic oscillation (EHO) in DIII-D and the quasi-coherent mode (QCM) in C-mod, are thought to be responsible for an enhanced particle transport that avoids the onset of ELMs. In DIII-D the EHO is observed in a limited parameter regime. The discharges typically exhibit a relatively low pedestal density and high pedestal temperature as well as a large edge rotation shear [K.H. Burrell, et al., Phys. Rev. Lett. 102, 155003 (2009), A.M. Garofalo, et al., 23rd IAEA FEC, EXS/1-2]. It is thought that the EHO is a low n peeling mode that is driven unstable by rotation shear at edge conditions slightly below the ELM stability limit and which saturates due to a change of rotation shear at finite mode amplitude [P.B. Snyder, et al., Nucl. Fusion 47, 961 (2007)]. This interpretation is consistent with the observed strong dependence of the EHO amplitude and the ensuing transport enhancement on plasma rotation. Limits of the QH-mode operating regime are encountered when the transport enhancement is too weak and cannot avoid the onset of ELMs or when the EHO amplitude is too large and causes locking. The operating regime could be greatly extended, if the drive of the mode by rotation shear could be replaced with an external field and the resulting transport controlled by the amplitude of the external field. First attempts to drive MHD instabilities in order to control transport have been carried out on JET [D. Testa, et al., 28th EPS conference on Controlled Fusion and Plasma Physics (2001)], where internal saddle coils were used to drive global Alfven waves in the range from 30-70kHz. | ||
| Resource Requirements: | I-coils with audio-amplifiers, preferably with two parallel amplifiers powering pairs of I-coils. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 167: Interaction between EP-driven mode and ELM in high-beta plasmas | ||
| Name: | Matsunaga matsunaga.go@qst.go.jp | Affiliation: | QST |
| Research Area: | General SSI | Presentation time: | Not requested |
| Co-Author(s): | M. Okabayashi(PPPL), M. Takechi(JAEA), A. Isayama(JAEA), N. Oyama(JAEA) | ITPA Joint Experiment : | No |
| Description: | Survey parameters so that OFM can trigger ELM. In particular, edge density scan will be done to change the pedestal stability similar to JT-60U where the EP-driven mode can trigger ELM. To clarify the mechanism EP-driven mode can trigger ELM, EP profiles will be measured. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In high beta region with OFM, pedestal density scan will be done by gas puffing so that ELM stability becomes close to ballooning limit, that is similar to JT-60U ELM stability. And maximize OFM amplitude, NB combination will be scanned. These results will be compared with JT-60U results. | ||
| Background: | In the wall-stabilized high-beta plasma on JT-60U, energetic particle driven mode named as 'Energetic particle driven Wall Mode (EWM)' has been observed. The EWM can often trigger ELM. The energy release due to ELM becomes smaller and ELM frequency becomes higher during EWM appearance. On DIII-D, similar mode is observed in the high-beta plasma; this mode was named as 'Off-axis Fishbone Mode (OFM).' The OFM can also induce RWM onset but not interact with ELM. The EWM-triggered ELM is interesting in the aspect of ELM control. If this phenomenon will be observed on DIII-D, it becomes interesting target as universal physics in tokamak. One of possibilities for the EWM-triggered ELM is enhanced EP-transport by EWM act as an additional pressure at pedestal region. This can be studied in detail with many EP diagnostics on DIII-D. Finally, obtained results will be compared with JT-60U data. | ||
| Resource Requirements: | 1 day experiment with NBs to exceed no-wall beta limit | ||
| Diagnostic Requirements: | Standard for RWM, EP and ELM experiments | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 168: Measurement of chemical erosion products via FTIR Spectroscopy | ||
| Name: | Umstadter karl@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Not requested |
| Co-Author(s): | Steve Allen, Chris Chrobak, Ron Ellis | ITPA Joint Experiment : | No |
| Description: | Carbon PFCs in tokamaks are subject to chemical erosion due to hydrocarbon (e.g. CD4, C2T2) formation during plasma operation, when deuterium and tritium ions and atoms are present. Understanding the formation, transport, breakup, redeposition and removal of these molecules, both during plasma operation as well as during bake cycles, is important for full accounting of fuel particle balance. Measurements of hydrocarbon content in the exhaust of DIII-D will improve the accuracy of during shot particle balance measurements. The monitoring of hydrocarbons, and oxides of carbon and hydrogen during wall conditioning procedures, such as helium glow discharge cleaning, baking, and baking in oxygen, is key to long-term fuel particle accounting. These measurements also contribute to a better understanding of the net erosion rate of plasma facing graphite. Such measurements on DIII-D will be important to developing and benchmarking codes for predicting the amount of tritium, which will accumulate in ITER during its lifetime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A Fourier transform infrared (FTIR) spectrometer will be utilized to monitor exhaust gas during DIII-D shots. The FTIR spectrometer will be coupled to the vacuum system between the turbo pumps and roughing pumps in the pump room below. Measurements will be made inside a gas cell that increases the sensitivity of measurements. Analysis of the gases present in the exhaust will be the guide for future experiments. | ||
| Background: | Traditionally, CD4 release from carbon surfaces is estimated by using passive spectroscopy of CD-band emission. Calibration of these measurements requires either complex modeling of the CD4 to CD breakup chain in the plasma or injection of a calibrated flow of CD4 from the material surface. Ideally, one would like a direct, non-perturbing measurement of the methane production from the plasma-facing surface under a variety of plasma conditions. By simultaneously measuring the absorption of infrared radiation by CD4 molecules and emission of CD-band photons, we can determine the range of plasma conditions where CD emission provides an accurate measure of CD4 release. This work can then be extended to evaluate other hydrocarbon gases such as deuterated acetylene and ethane. Additionally, tritium products if present may be detected if concentrations are above background.
D2O, CO, CO2 and others are created when carbon plasma facing components that have been exposed to plasma, undergo an oxygen bake cycle. This technique has proven to be applicable to carbon and hydrogen oxides that are produced during cleaning cycles in tokamaks. The efficiency of the bake can was successfully monitored by measuring the respective gas concentrations versus the oxygen bake conditions (i.e. bake temperature, oxygen pressure, etc.). Traditionally, residual gas analyzers (RGAs) have been used to measure the off-gas of tokamaks during operation and bake cycles. The mass number spectrum will be very complex and confusing when D and T are present in roughly equal amounts because nearly all mass numbers will have significant signal. This happens in the RGA because molecular fragments produced in the ionization chamber contain arbitrary ratios of C:D:T. Use of cracking pattern analysis is not reliable under these conditions. FTIR Spectroscopy does not suffer from this handicap as it is specific to the absorption band of the molecule and overlap is avoidable. | ||
| Resource Requirements: | The system will be connected in the pumping room and should not affect operation in any way. Access to the pump room and room for the FTIR spectrometer is required. A vacuum connection to from the gas cell to the pump duct should be established. Remote USB access to the system via fiber optic extenders will be established for remote monitoring. Dedicated machine time during operations is not required but is desirable. Dedicated time outside ops is required for calibration. | ||
| Diagnostic Requirements: | MDS of CD-band, C i-iii emission; Fast Filterscopes - CD-band, C i-iii emission; Divertor Thomson measurements of ne and Te in the divertor | ||
| Analysis Requirements: | Calibration of gas cell with CD4, C2D2 and other gases observed in exhaust will allow qualitative analysis of gases. This can be accomplished without connection to DIII-D. A measured flow of calibration gas in DIII-D is required to quantify the amount absolutely. | ||
| Other Requirements: | |||
| Title | 169: Measurement of chemical erosion products near DiMES via FTIR Spectroscopy | ||
| Name: | Umstadter karl@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Not requested |
| Co-Author(s): | Steve Allen, Chris Chrobak, Ron Ellis, Dmitry Rudakov, Clement Wong | ITPA Joint Experiment : | No |
| Description: | Carbon PFCs in tokamaks are subject to chemical erosion due to hydrocarbon (e.g. CD4, C2T2) formation during plasma operation, when deuterium and tritium ions and atoms are present. Understanding the formation, transport, breakup, redeposition and removal of these molecules, both during plasma operation as well as during bake cycles, is important for full accounting of fuel particle balance. Precise measurements of hydrocarbon content as close as possible to the divertor of DIII-D will improve the accuracy of during shot particle balance measurements. Such measurements on DIII-D will be important to developing and benchmarking codes for predicting the amount of tritium, which will accumulate in ITER during its lifetime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A Fourier transform infrared (FTIR) spectrometer will be utilized to monitor erosion products near DiMES during DIII-D shots. The FTIR spectrometer will be coupled to the vacuum system of DiMES. Measurements will be made inside a gas cell that increases the sensitivity of measurements. | ||
| Background: | Traditionally, CD4 release from carbon surfaces is estimated by using passive spectroscopy of CD-band emission. Calibration of these measurements requires either complex modeling of the CD4 to CD breakup chain in the plasma or injection of a calibrated flow of CD4 from the material surface. Ideally, one would like a direct, non-perturbing measurement of the methane production from the plasma-facing surface under a variety of plasma conditions. By simultaneously measuring the absorption of infrared radiation by CD4 molecules and emission of CD-band photons, we can determine the range of plasma conditions where CD emission provides an accurate measure of CD4 release. This work can then be extended to evaluate other hydrocarbon gases such as deuterated acetylene and ethane. Additionally, tritium products if present may be detected if concentrations are above background.
D2O, CO, CO2 and others are created when carbon plasma facing components that have been exposed to plasma, undergo an oxygen bake cycle. This technique has proven to be applicable to carbon and hydrogen oxides that are produced during cleaning cycles in tokamaks. The efficiency of the bake can was successfully monitored by measuring the respective gas concentrations versus the oxygen bake conditions (i.e. bake temperature, oxygen pressure, etc.). Traditionally, residual gas analyzers (RGAs) have been used to measure the off-gas of tokamaks during operation and bake cycles. The mass number spectrum will be very complex and confusing when D and T are present in roughly equal amounts because nearly all mass numbers will have significant signal. This happens in the RGA because molecular fragments produced in the ionization chamber contain arbitrary ratios of C:D:T. Use of cracking pattern analysis is not reliable under these conditions. FTIR Spectroscopy does not suffer from this handicap as it is specific to the absorption band of the molecule and overlap is avoidable. | ||
| Resource Requirements: | he system will be connected to the vacuum system of DiMES. Access to the DiMES manipulator and room for the FTIR spectrometer is required. A vacuum connection to from the gas cell to the pump duct should be established. Remote USB access to the system via fiber optic extenders will be established for remote monitoring. Dedicated machine time during operations is not required but is desirable. Dedicated time outside ops is required for calibration. | ||
| Diagnostic Requirements: | MDS of CD-band, C i-iii emission; Fast Filterscopes - CD-band, C i-iii emission; Divertor Thomson measurements of ne and Te in the divertor | ||
| Analysis Requirements: | Calibration of gas cell with CD4, C2D2 and other gases observed in exhaust will allow qualitative analysis of gases. This can be accomplished without connection to DIII-D. A flow of calibration gas in DIII-D is required to quantify the amount absolutely. | ||
| Other Requirements: | |||
| Title | 170: Time evolution of halo current width and temperature in disruptions/VDEs for model development | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Requested |
| Co-Author(s): | M. Sugihara | ITPA Joint Experiment : | No |
| Description: | This proposal is to measure the time evolution of the halo current width and temperature (if possible their profiles) for the development of more relevant halo model with robust physics basis to be used in the 2D predictive disruption codes. Various disruptions (MD) and VDE conditions should be investigated but, particularly interesting discharges are VDEs with fast current quench which generate large halo current as well as MD which generates large halo current. From these measurements, relevancy of the existing models can firstly be examined. As a next step, relevant physics model to represent the halo width should be developed. Relevancy of the developed model needs to be confirmed by at least 2D disruption codes, since the developed model is to be used in the 2D codes, and thus, the modelling of DIII-D experiments by DINA and/or TSC codes is highly desirable. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Poloidal distribution of the poloidal halo current should be measured by Rogowski coil or shunt during whole phase of current decay of MD/VDEs. Electron temperature (if possible its profile) in the halo and core region should also be measured simultaneously during the current decay phase. VDEs should be triggered intentionally and large halo current discharges for different current quench time should be selected. For MD case, any of the triggering mechanisms, like density limit, locked mode, which can generate large halo current, can be used. | ||
| Background: | Robustness of the machine against electromagnetic (EM) loads under various disruption (MD) and VDE conditions expected is essential for ITER. In particular, design of the blanket modules (BM) will be finalized within 1-2 years, and thus, relevant physics guidelines for the comprehensive MD and VDE scenarios must be provided for appropriate design of BMs to withstand the possible most severe load conditions.
Preparation of the MD and VDE scenarios have been done using 2D disruption codes, DINA and TSC, based on international disruption database (IDD) compiled by ITPA group. Of particular importance for the design are the scenarios, which generate large halo current and/or eddy current. So far, 2D codes predict that large halo current is generated by VDEs with slow current quench, since large plasma current remains in the unstable magnetic field pattern region. For this reason, large eddy current, which is generated by fast current quench, is not overlapped with large halo current. Very strong supporting structure for the BMs is required if these two currents (forces) are actually overlapped. Recent new IDD, however, include the corresponding current quench time for the halo current data, and it is found that largest halo current is observed in fast current quench discharges. It is also found that there is no essential difference between MD and VDEs, i.e., large halo current is also observed in MD. In parallel to the database activity, model validation of the 2D codes using ASDEX-U, NSTX, JT-60U is also on-going under ITER and EFDA tasks, and common understanding has been gradually recognized that the halo current model is still very insufficient to reproduce the experimental results. Presently, the width is simply fixed or modelled by simple analytic formula related to the instantaneous plasma current during its decay. It is also identified that the halo current width and temperature are likely to be most influential parameters for the halo model, and it has been suggested that such a large halo current could be associated with fast current quench by choosing different halo width model, e.g., rather narrow halo width during VDEs. According to this background, it is of primary importance to develop a relevant halo current model for the preparation of representative MD and VDE scenarios in ITER. | ||
| Resource Requirements: | 1-2 run days | ||
| Diagnostic Requirements: | Measurement of poloidal halo current and its poloidal distribution by Rogowski coil or shunt.
Electron temperature measurement in the halo current region as well as the core plasma region during current decay by, e.g., line ratio of He I and II. | ||
| Analysis Requirements: | Modelling of time evolution for halo current and its poloidal distribution by 2D codes, DINA or TSC, is highly desirable to reproduce the experimental time evolution of halo currents and to derive the relevant model and to confirm. | ||
| Other Requirements: | -- | ||
| Title | 171: "Spiralling field" EFC | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Requested |
| Co-Author(s): | H. Reimerdes, E.J. Strait | ITPA Joint Experiment : | No |
| Description: | Let a mode lock to the resultant of the unknown static machine EF and a known, uniformly rotating, growing Magnetic Perturbation (MP). Infer the EF amplitude and orientation from the measured non-uniform mode rotation and amplitude modulation, and from behaviours that can only be observed for a spiralling, not simply rotating, MP: the mode rotation will initially be intermittent, then complete but non-uniform, then more and more uniform. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Generate a non-disruptive locked mode by ramping the beams and thus beta in a low-rotation (balanced injection) plasma. Tweak beta, q95 and post-locking NBI to ensure no disruption.
If necessary, add ECH or ECCD at q=2 location: the former is expected to keep the mode small (thus, less disruptive) regardless of being deposited in the island O- or X-point; the latter will introduce a modulation in the mode-amplitude and will make the rotation non-uniform in correspondence of O- and X-point deposition. This modulation and non-uniformity allow an even better characterization of the locked mode and, thus, of the EF. Apply growing, rotating (spiralling) MPs. Infer the EF from the non-uniform response of the mode response (amplitude and phase) measured via internal saddle loops. | ||
| Background: | At DIII-D, modes initially locked to the wall or machine EF were forced to rotate by applied rotating MPs. More accurately, the mode locked to the resultant of the static EF and the rotating MP. This was a nuisance from the standpoint of driven rotation in that it caused the island to rotate non-uniformly, to the point that it "slipped" or changed direction [F. Volpe et al., Phys.Plasmas 16, 102502 (2009)].
This, however, might turn useful from an EFC perspective, to find an unknown EF as the difference between a measured EF+MP and a known (pre-programmed) MP. A standard EFC method consists in fixing the MP phase phi_MP and ramping the density down until locking. Then phi_MP is scanned shot-by-shot. The EF is inferred from the values of A_MP (or n_e) at locking. Note that ramps are pre-programmed, they do not stop at locking and often terminate with disruptions. Moreover, 3-4 discharges of this kind are needed for every new EFC, i.e. in principle for every new scenario. The approach proposed here, by contrast, consists in scanning both phi_MP and, more slowly, A_MP, within the same shot. The resulting MP rotates and grows, i.e. it spirals out. In doing so, it scans the A_MP/A_EF vs. phi_MP-phi_EF, where A_MP and phi_MP are known, and the EF amplitude and phase, A_EF and phi_EF, are the unknown. The behaviour of a mode locked to and dragged by the EF+MP resultant changes as different regions of the A_MP/A_EF, phi_MP-phi_EF plane are explored and A_EF and phi_EF can be indirectly inferred. For example A_EF coincides with the smallest A_MP for which the mode is successfully dragged for a complete toroidal rotation. Before then, rotation will be incomplete, and phi_EF will be the mid-phase of the arcs. All this requires the presence of a non-disruptive mode in the plasma. This mode can either be pre-existing, seeded by EF-penetration, e.g. by an earlier density ramp-down, or, inevitably, it will automatically be generated during the MP "spiral", as soon as the total amplitude becomes high enough. It is understood that the MP of choice has the same n as the dominant EF component. However, once a certain n-component has been identified and corrected, one can repeat the process for the second most important component. This method might represent a non-disruptive, ITER-relevant, quicker (as it requires one shot or a fraction of it, instead of four shots) alternative to the conventional technique mentioned above. Note that conventional EFC is restricted to low-density locked modes and thus, inevitably, low beta. This method, instead, also works at high density, high beta and low q95. Finally, there are prospects of generalization to multi-mode EFC, by identifying other features in the A_MP/A_EF vs. Phi_MP-phi_EF plane, and there are prospects of generalization to spontaneously rotating NTMs and Quasi-Stationary Modes (QSMs) rather than to forcefully rotating locked modes. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 172: "Spiralling field" EFC | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | H. Reimerdes (CRPP-EPFL), E.J. Strait | ITPA Joint Experiment : | No |
| Description: | Let a mode lock to the resultant of the unknown static machine EF and a known, uniformly rotating, growing Magnetic Perturbation (MP). Infer the EF amplitude and orientation from the measured non-uniform mode rotation and amplitude modulation, and from behaviours that can only be observed for a spiralling, not simply rotating, MP: the mode rotation will initially be intermittent, then complete but non-uniform, then more and more uniform. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Generate a non-disruptive locked mode by ramping the beams and thus beta in a low-rotation (balanced injection) plasma. Tweak beta, q95 and post-locking NBI to ensure no disruption.
If necessary, add ECH or ECCD at q=2 location: the former is expected to keep the mode small (thus, less disruptive) regardless of being deposited in the island O- or X-point; the latter will introduce a modulation in the mode-amplitude and will make the rotation non-uniform in correspondence of O- and X-point deposition. This modulation and non-uniformity allow an even better characterization of the locked mode and, thus, of the EF. Apply growing, rotating (spiralling) MPs. Infer the EF from the non-uniform response of the mode response (amplitude and phase) measured via internal saddle loops. | ||
| Background: | At DIII-D, modes initially locked to the wall or machine EF were forced to rotate by applied rotating MPs. More accurately, the mode locked to the resultant of the static EF and the rotating MP. This was a nuisance from the standpoint of driven rotation in that it caused the island to rotate non-uniformly, to the point that it "slipped" or changed direction [F. Volpe et al., Phys.Plasmas 16, 102502 (2009)].
This, however, might turn useful from an EFC perspective, to find an unknown EF as the difference between a measured EF+MP and a known (pre-programmed) MP. A standard EFC method consists in fixing the MP phase phi_MP and ramping the density down until locking. Then phi_MP is scanned shot-by-shot. The EF is inferred from the values of A_MP (or n_e) at locking. Note that ramps are pre-programmed, they do not stop at locking and often terminate with disruptions. Moreover, 3-4 discharges of this kind are needed for every new EFC, i.e. in principle for every new scenario. The approach proposed here, by contrast, consists in scanning both phi_MP and, more slowly, A_MP, within the same shot. The resulting MP rotates and grows, i.e. it spirals out. In doing so, it scans the A_MP/A_EF vs. phi_MP-phi_EF, where A_MP and phi_MP are known, and the EF amplitude and phase, A_EF and phi_EF, are the unknown. The behaviour of a mode locked to and dragged by the EF+MP resultant changes as different regions of the A_MP/A_EF, phi_MP-phi_EF plane are explored and A_EF and phi_EF can be indirectly inferred. For example A_EF coincides with the smallest A_MP for which the mode is successfully dragged for a complete toroidal rotation. Before then, rotation will be incomplete, and phi_EF will be the mid-phase of the arcs. All this requires the presence of a non-disruptive mode in the plasma. This mode can either be pre-existing, seeded by EF-penetration, e.g. by an earlier density ramp-down, or, inevitably, it will automatically be generated during the MP "spiral", as soon as the total amplitude becomes high enough. It is understood that the MP of choice has the same n as the dominant EF component. However, once a certain n-component has been identified and corrected, one can repeat the process for the second most important component. This method might represent a non-disruptive, ITER-relevant, quicker (as it requires one shot or a fraction of it, instead of four shots) alternative to the conventional technique mentioned above. Note that conventional EFC is restricted to low-density locked modes and thus, inevitably, low beta. This method, instead, also works at high density, high beta and low q95. Finally, there are prospects of generalization to multi-mode EFC, by identifying other features in the A_MP/A_EF vs. Phi_MP-phi_EF plane, and there are prospects of generalization to spontaneously rotating NTMs and Quasi-Stationary Modes (QSMs) rather than to forcefully rotating locked modes. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 173: Controlling a helical tokamak equilibrium with a small external 3D field | ||
| Name: | Piovesan paolo.piovesan@igi.cnr.it | Affiliation: | Consorzio RFX |
| Research Area: | General SSI | Presentation time: | Not requested |
| Co-Author(s): | L. Piron, M. Gobbin, L. Marrelli, P. Martin, D. Terranova | ITPA Joint Experiment : | No |
| Description: | Helical equilibria with good confinement properties are spontaneously obtained in RFX-mod at high plasma current [R. Lorenzini et al. Nature Physics 5, 570 (2009)]. The safety factor profile in these cases has a reversed shear and internal transport barriers develop where the magnetic shear vanishes. In addition, recent RFX-mod experiments have been shown that such equilibria can be induced and controlled by applying helical boundary conditions at the edge with magnetic feedback [P. Piovesan et al., presented at the 2010 MHD workshop].
Simulations with the ANIMEC code support the idea that similar helical equilibria can also be obtained in tokamaks with an axi-symmetric boundary, by slightly perturbing the magnetic axis [W.A. Cooper et al PRL 105, 035003 (2010)], for example with an off-axis neutral beam, as proposed for DIII-D by O. Schmitz et al [proposal n. 151]. As found in RFX-mod, small 3D external perturbations could be sufficient to stimulate such a state or in any case to control its phase and rotation frequency once it is achieved. This can be realized in DIII-D through magnetic feedback with the I/C coils. Such an approach may be also useful to control helical equilibria obtained with the method proposed by Schmitz et al [proposal n. 151]. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The approach introduced above may be tested in plasmas with a q profile close to that predicted in [W.A. Cooper et al PRL 105, 035003 (2010)] to favor a bifurcation to a helical equilibrium, i.e. a flat or hollow q profile with qmin of about 1. The path to reach such a configuration may be similar to that proposed by O. Schmitz et al in proposal n. 151, e.g. off-axis beam heating at the start-up or off-axis ECCD assisted by q profile feedback control.
As far as the magnetic feedback is concerned, it may be interesting to vary the amplitude and phase of the applied perturbation, even within a single discharge, to study its effect on the helical equilibrium. | ||
| Background: | A helical tokamak equilibrium may have potential advantages, for example in terms of stability, as described in [W.A. Cooper et al PRL 105, 035003 (2010)]. In addition, the hollow or flat q profiles typical of such equilibria would favor the development of internal transport barriers, as observed in many tokamaks and similarly to what also observed in RFX-mod. Long-lived n=1, m=1 kink modes are also found to replace sawtooth activity in MAST and TCV, which may turn out beneficial with respect to NTM avoidance. For these reasons such helical equilibria may be worth being investigated in tokamaks. | ||
| Resource Requirements: | The proposed experiments require to impose at the edge a finite radial field with some prescribed n=1 amplitude and phase to be determined by feedback, so that the external field sustains the n=1 field produced by the plasma. This requires a finite reference value for the n=1 Bp field to be implemented in the DIII-D feedback controller, different from the present zero-reference value normally used. | ||
| Diagnostic Requirements: | All diagnostics needed to image the core topology (soft x ray, ece, ece-i, ...) and to determine internal magnetic field profiles (mse, polarimeter, ...) | ||
| Analysis Requirements: | kinetic EFITS and high quality profile data
ANIMEC/VMEC simulations | ||
| Other Requirements: | |||
| Title | 174: Behavior of PFCs in ELMing deuterium plasma following significant exposure to helium & deuterium | ||
| Name: | Umstadter karl@ucsd.edu | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | Clement Wong, Dmitry Rudakov, William Wampler | ITPA Joint Experiment : | No |
| Description: | Studies of helium-seeded deuterium plasmas indicates that D will remain trapped near the surface for exposure temperatures up to 725K. The retention of gas in PFCs may lead to enhanced erosion of material that will behave differently if ionized near the surface. All prior heat pulse testing of PFCs have been completed in vacuum environments without the presence of background plasma. ELMs will not be this kind of isolated event and one should know the effect of a plasma background during these transients. Current models may underestimate the damage caused by ELMs due to these phenomena. This experiment is a continuation of successful experiments in the last campaign. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | DiMES samples of graphite and tungsten will be bombarded with helium and deuterium ions in the PISCES-A device. Saturated and unsaturated samples will be loaded into the DiMES system and exposed for as many shots as possible during experiments that have the strike point on or near. Disruptions, if they should occur will not adversely affect the experiment. The DiMES holder should be lowered between shots so that it is not exposed to He glow discharges. | ||
| Background: | Heat-pulse experiments have begun in the PISCES-A device utilizing laser heating in a divertor-like plasma background. Initial results indicate that the erosion of PFCs is enhanced as compared to heat pulse or plasma only tests. This enhanced erosion may be caused by trapped gases released during the heat pulse. Also self-sputtering of material that is ejected during the transient, ionized by the plasma near the surface and subsequently driven back to the surface may occur. Gas retention in PFCs and ELM energy are currently indicated as the cause. Current machines around the world don't see the damage witnessed in the lab because the sample (divertor) has a lower fluence before seeing an ELM, disruption, or glow cleaning. This will not be the case for ITER as one can expect >1E25 D/m2 between ELMs (2 Hz). These experiments will be compared to experiments performed on PISCES-A. | ||
| Resource Requirements: | The system should be operating for as many shots as possible during experiments that have the strike point on or near DiMES. DiMES will raise/lower before/after shot to avoid He glow. Dedicated machine time during operations is not required but is desirable for several shots. | ||
| Diagnostic Requirements: | Fast/IR Cameras looking at DiMES; MDS ā?? looking at Hg and W i & ii: center at 4320Ć? for W i 4294.6Ć?, W i 4302.1Ć?, H-gam 4340.5Ć?, W ii 4348.1Ć?; Fast Filterscopes with filters as in MDS; Divertor Thomson measurements of ne and Te in the divertor | ||
| Analysis Requirements: | NRA to be completed by SNL, SEM imaging, TDS and mass loss analysis will be completed at UCSD PISCES lab. | ||
| Other Requirements: | |||
| Title | 175: Core Transport Stiffness in H-mode Plasmas | ||
| Name: | Kinsey jon.kinsey@comxco.com | Affiliation: | CompX |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Assess core transport stiffness while holding the H-mode pedestal conditions fixed. Compare the experimentally analyzed stiffness to the stiffness found in TGLF predictive simulations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform a power scan at fixed pedestal beta in non-sawtoothing H-mode discharges. The emphasis will be on discharges with low toroidal rotation but the mix of on-axis NBI and off-axis NBI should
be varied in order to assess if the core stiffness changes with rotation. Weak shaping will be necessary to keep the pedestal height from changing with auxiliary power. | ||
| Background: | Predictive transport simulations using TGLF for ITER have shown that the fusion Q scales like P_aux^(-0.8) at fixed pedestal beta as a result of stiff core turbulence. Since stiffness is such an important ingredient in our predictions for ITER, it is essential to validate this aspect
of TGLF experimentally. | ||
| Resource Requirements: | Comparable levels of off-axis and on-axis NB power, ECH power. | ||
| Diagnostic Requirements: | Full set of profile diagnostics required for normal power balance analysis with ONETWO or TRANSP. | ||
| Analysis Requirements: | Simulation of experimental profiles and assess core stiffness with TGLF transport model using XPTOR. | ||
| Other Requirements: | |||
| Title | 176: Effects of various techniques for ELM mitigation on average pedestal pressure and plasma confinement | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, W. Solomon | ITPA Joint Experiment : | No |
| Description: | Compare effects on ELMs and plasma confinement of various approaches to ELM control by frequency enhancement | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Two plasma conditions would be explored at a level of P_net/P_th ~ 1.3 : one with very low natural f_ELM*tau_E ~ 1 (ITER demonstration discharges) and one with high frequency f_ELM*tau_E ~ 4 (lower delta). In these plasmas the ELM frequency would be increased in several steps to the highest level possible with the three schemes before Type IIII ELMy H-mode transition and the effects on ELM size, pedestal energy and density changes at ELM, average plasma pressure, plasma confinement and divertor ELM power fluxes will be documented. The three methods for ELM control to be applied are pellet pacing, oscillating RMPs and gas puffing. | ||
| Background: | Control of ELM frequency by pellet pacing is one of the two baseline techniques for ELM control in ITER. Several major issues remain open with regards to the application of this technique : the effects on plasma confinement of ELM control by frequency increase, the effect on ELM power fluxes and the additional losses caused by the use of pellets to trigger the ELMs. | ||
| Resource Requirements: | 2 run days. Pellet injector and oscillating RMPs | ||
| Diagnostic Requirements: | Pellet diagnostics, Pedestal diagnostics, ELM power flux diagnostics | ||
| Analysis Requirements: | None | ||
| Other Requirements: | None | ||
| Title | 177: L-H and H-L Power Thresholds in He Plasma Diluted by Hydrogen | ||
| Name: | POLEVOI none | Affiliation: | ITER Organization |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | Punit GOHIL | ITPA Joint Experiment : | Yes |
| Description: | Study of dependence of L-H and H-L power threshold on helium contamination by hydrogen in ITER-like conditions | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Make a scan in the range ne ~ 2-5 1019m-3 with EC heating only and He fuelling by gas puffing (no He NBI). For H-L threshold studies: For each density of this scan create an H-mode type-I plasma in a pure He. Then increase fraction of hydrogen gradually by LHS hydrogen pellet injection keeping the electron density constant. Determine the critical hydrogen fraction for Type-I-III and for H-L transitions. For L-H threshold studies: For each density of this scan in the L-mode vary hydrogen fraction by LFS pellet injection keeping ne constant. Determine PL-H and P(Type-III-I).
Secondary mission of experiment is the assessment of the speed of He residual from hydrogen-LFS pellets in L- and H-modes. In general using of hydrogen puffing for hydrogen fuelling instead of pellets is also possible. But then the secondary mission will be lost. Using of H-NBI for fuelling is undesirable because of additional torque and CX effects. Another secondary mission is the assessment of particle transport and energy confinement properties of mixed He-H plasmas | ||
| Background: | In ITER pre-DT phase H-mode operation is more likely in He plasmas with dominant electron heating, low torque input, He fuelling by gas puffing in the range ne ~ 2-5 1019m-3. The LFS hydrogen pellet injection required for ELM pace making will cause He contamination by hydrogen. Power threshold for the H- mode operation has different dependence on density in low density range for hydrogen and He. Thus, compound of the mix can be critical for H-mode operation. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 178: ICR Heating in Helium Plasma Diluted by Hydrogen | ||
| Name: | POLEVOI none | Affiliation: | ITER Organization |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | Punit GOHIL, J.M. PARK, Ron PRATER | ITPA Joint Experiment : | Yes |
| Description: | Study of dependence of ICRH with hydrogen minority scheme on the fraction of hydrogen in ITER-like conditions | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This study can be combined with task 1(L-H and H-L Power Thresholds in He Plasma Diluted by Hydrogen). Make a scan in the range ne ~ 2-5 1019m-3 with EC heating only and He fuelling by gas puffing (no He NBI). For H-L threshold studies: For each density of this scan create an H-mode type-I plasma in a pure He. Then increase fraction of hydrogen gradually by LHS hydrogen pellet injection keeping the electron density constant. For L-H threshold studies: For each density of this scan in the L-mode vary hydrogen fraction by LFS pellet injection keeping ne constant. For each of the scans choose a few points in a hydrogen fraction scan and heat plasma by short FW pulse, sufficient to assess power absorbed by plasma. Keep density and minority fraction constant during this heating pulse.
Assess power, absorbed by plasma as a function of hydrogen fraction when minority becomes a majority and the single pass absorption is lost. The results are required for validation of the ICRH code prediction for ITER. | ||
| Background: | In ITER pre-DT phase H-mode operation is more likely in He plasmas with dominant electron heating, low torque input, He fuelling by gas puffing in the range ne ~ 2-5 1019m-3. In ITER condition the single path absorption for ICRH is expected for nH/ne ā?¤ 5% The LFS hydrogen pellet injection required for ELM pace making will cause He contamination by hydrogen. Hydrogen fraction in a scale of a second can become high nH/ne ~50%, which strongly affects the power absorption | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Analyses the results of FW absorption to compare with the ICRH code predictions | ||
| Other Requirements: | |||
| Title | 179: Determination of minimum pellet size for ELM triggering by plasma current and shaping scans | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, W. Solomon, G. Huysmans, S. Futatani | ITPA Joint Experiment : | No |
| Description: | To determine the minimum pellet size for ELM triggerring in DIII-D by utilizing the smallest pellet and increasin the edge pressure limit by plasma sahping and high current operation | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiment attempts to determine the minimum pellet size for ELM triggering by employing the smallest available pellet size in ITER and use of plasma shape to increase edge stability. The experiments would consist on injection of smallest size pellets in discharges at the highest possible currents with high Pinput (for highest Tped) with increasing shaping (up to double null) to improve edge stability to find out if pellet triggering of ELMs ceases at a given value of plasma shaping (and edge pressure). If this is achieved, finer scans of shaping and input powers around the thershold observed will be performed. | ||
| Background: | Determining the minimum pellet size for ELM triggering in ITER is required to demonstrate the viability of the scheme (in terms of the associated plasma thorughput required) and for its optimisation | ||
| Resource Requirements: | 1 run day. Pellet injection | ||
| Diagnostic Requirements: | Pellet diagnostics, Pedestal diagnostics | ||
| Analysis Requirements: | None | ||
| Other Requirements: | None | ||
| Title | 180: Assessment of limits to ELM frequency enhancement by pellet pacing | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, W. Solomon, G. Huysmans, S. Futatani | ITPA Joint Experiment : | No |
| Description: | Determination of possible limits to ELM frequency control by pellet pacing at high repetition frequencies when the after-ELM pedestal plasma may be well away from edge stability limits | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Starting from an ITER demonstration discharge increase the ELM frequency by a factor of 2, 4, 6 (and 8 in if possible) with oscillation RMPs. Inject pellets at 2, 4, and 6 (and 8 in possible) times the natural ELM frequency at the outer midplane and then at the X-point and vary the timing of pellet injection versus the RMP ELM triggering time. If successful perform a power scan to understand the effect on maximum ELM-triggered pellet frequency of pedestal recovery speed. | ||
| Background: | ELM control by pellet pacing is one of the Baseline ELM control technique in ITER. For ITER the required increase of the ELM frequency is ~ 30 which is well above the natural ELM frequency. As pellets lead to the triggering of ELM when the pedestal plasma is away from MHD stability limits it is not clear if such high frequency enhancement is possible as the pedestal will not have recovered from the previous ELM | ||
| Resource Requirements: | 1 run day. Pellet pacing, oscillation RMPs | ||
| Diagnostic Requirements: | Pellet diagnostics, Pedestal diagnostics, ELM power Flux diagnostics | ||
| Analysis Requirements: | None | ||
| Other Requirements: | None | ||
| Title | 181: Test bed for validation of transport models and controllers for ITER | ||
| Name: | POLEVOI none | Affiliation: | ITER Organization |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Requested |
| Co-Author(s): | Dave HUMPHREYS, Tim LUCE, Didier MOREAU | ITPA Joint Experiment : | No |
| Description: | Creation of the framework which can be used as the test bed for validation of transport models and controllers for ITER on DIIID experiments | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Creation of the framework/code combined with emulation of the real actuators and diagnostics of DIII-D with possibility of implementation of the transport models and models of actuators proposed for ITER simulations.
Build the model, required for plasma control for each of the integrated models proposed for ITER simulation with DIIID actuators. Then use this built model in the close loop control scheme for real DIII-D experiments to judge whether it is adequate. And vice versa, test whether the algorithm developed and used in real DIII-D experiments provide an adequate control with integrated plasma model used for ITER to judge whether the transport/actuator models used for ITER simulations are adequate. | ||
| Background: | Creation of adequate models for control design purpose including model of plasma, actuators, diagnostics and plasma reconstruction of ITER is set as the first task in the R&D Needs in PCS Requirements for Plasma Kinetic Control in ITER. Adequacy of plasma and actuator model and approach to control should be tested and justified in real present day machines. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 182: Effectiveness of pellet fuelling with simultaneous ELM control by pellet pacing | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, W. Solomon | ITPA Joint Experiment : | No |
| Description: | Determination of the effective fuelling efficiency of combined pellet fuelling and ELM control by pellet pacing in ITER-like plasma conditions | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Establish an ITER demonstration discharge with HFS pellet fuelling aiming at similar particle deposition profiles as in ITER by adjusting the level of input power (and possibly, plasma current) and a given ratio of | ||
| Background: | ITER is expected to be fuelled by HFS pellet pacing with LFS pellet pacing for ELM control. Understanding of plasma fuelling in these conditions is needed to evaluate the adequacy of the installed systems and their compatibility with the total particle throughput. | ||
| Resource Requirements: | 1-2 run days depending on feasible scans. Pellets for fuelling from HFS fuelling and for pacing from LFS | ||
| Diagnostic Requirements: | Pellet diagnostics, Pedestal and core plasma diagnostics, ELM power flux diagnostics | ||
| Analysis Requirements: | No | ||
| Other Requirements: | No | ||
| Title | 183: Triggering ELMs with Injected Li Granules | ||
| Name: | Mansfield none | Affiliation: | PPPL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of the experiment is to investigate whether or not injected spherical Li granules (d ~ 1mm) can trigger and pace ELMs in a manner similar to deuterium pellets. Further goals are to allow a quantitative comparison of the relative efficacies of Li and deuterium in the ELM triggering process and thereby to test the current models of pellet-induced ELM triggering. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish an H-mode discharge that exhibits robust type 1 ELMs. This would best be identical to those discharges for which deuterium pellet injection has previously succeeded in triggering ELMs. The NSTX Li granule injector would then be used in a similar attempt to trigger ELMs. The injection speed of the Li granules could be swept during the course of a single discharge to test the hypothesis that there are both threshold sizes and speeds in the ELM-triggering process. The size of the Li granules can be changed overnight. | ||
| Background: | Recent ablation calculations (P Parks and Wen Wu) suggest that Li granules can complete favorably with deuterium pellets in role of ELM initiation. We (NSTX) have recently built and tested a simple and compact prototype device that can inject small (d ~ 1mm) spherical objects horizontally at speeds that are consistent with those needed to trigger ELMs by deuterium pellet injection (up to ~ 100 m/s). This device employs a simple impeller to redirect a falling stream of granules horizontally. Such a device can be used to inject Li granules near the horizontal midplane in an effort to trigger ELMs. The speed of injection is controlled by an air motor turning the shaft of the impeller while the frequency of injection is controlled independently by a resonating piezoelectric disk in a dropping device. Hence we have been able to achieve average injection frequencies of 50 - 100 Hz at speeds approaching 100 m/s. Both the speed and the instantaneous injection frequency can, in principle, be swept in a linear fashion during a single discharge.
On NSTX Li coatings on plasma facing components have eliminated ELMs in essentially all H-mode plasma configurations. While this has led to enhanced confinement and plasma performance, it has also led to the accumulation of impurities in the core. It is hoped that eventually these impurities can be flushed from the core with appropriate ELM triggering techniques. | ||
| Resource Requirements: | ELMing H-mode with type 1 ELMs. Similar to target plasma for low frequency ELM triggering with D pellet injection. | ||
| Diagnostic Requirements: | Pedestal and edge fluctuation diagnostics with fast time resolution as well as fast imaging systems for pellet measurements. | ||
| Analysis Requirements: | Analysis will be done in a fashion similar to D pellet injection experiments | ||
| Other Requirements: | Access to a midplane port with a gate valve having at least 1.5 inch aperture. | ||
| Title | 184: Testing predictions for ITB formation from a gyrokinetics-based flux model | ||
| Name: | Barnes none | Affiliation: | U of Oxford |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Felix I. Parra, Anne E. White, Max E. Austin | ITPA Joint Experiment : | No |
| Description: | The goal of the proposed experiments is to develop and test predictions from a new theoretical model of ITB formation. Plots of the Gyro-Bohm normalized heat flux versus R/LT (constructed from our model by calculating the heat flux, temperature, and temperature gradient at each radial location in steady-state) show a striking similarity to those given for ASDEX Upgrade discharges (we have compared our results with Fig. 3 of [Wolff, PPCF 45 1757 (2003)]). We would like to construct these heat flux versus R/LT plots at different time slices of DIII-D discharges with varying deposition profiles. Predictions of the presence and width of ITBs from our theoretical model will be compared with the experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Previous experiments by M. E. Austin observed ITBs in both co-NBI and balanced-NBI plasmas at DIII-D. Prior to the experiments we propose, the empirical recipe for the formation of ITBs in past discharges will be compared with the results obtained using our simplified turbulent flux model. We will conduct a detailed study of the characteristics of ONETWO/TRANSP experimental heat flux versus measured R/LTi at various time slices for discharges with and without ITBs. This will allow us to map out the regions in parameter space (beam energy, beam power, and current profile) where ITBs form.
In the proposed experiments, we will target conditions for ITB formation identified using our simplified model. This will require discharges with a combination of co-NBI and balanced-NBI. We will also target plasmas where the ITB is not predicted to form. 1) Begin with standard co-NBI plasma with early beam heating to slow evolution of q-profile and get formation of ITB when q=2 surface appears in plasmas. 2) Vary mix of co- and counter-NBI according to model predictions to prevent the ITB from forming. The beam energies can also be varied from maximum values (75-80 keV) down to 60 keV on a shot to shot basis to scan energy and momentum input. 3) Develop discharge with co-NBI and counter-NBI plasmas where reduced flux model predicts an ITB should form without using the q=2 recipe. 4) If we are successful in developing an ITB without the q=2 trigger, we will vary the mix of co- and counter-NBi and energy input to prevent the ITB formation. | ||
| Background: | Internal transport barriers (ITBs) have been observed on a wide range of experimental devices. While the detailed conditions necessary for their formation varies, it is clear that both the current and rotation profiles play a significant role. As a community, we have a basic theoretical understanding of how the shear in these profiles affects turbulent transport, and there have been a number of numerical studies of turbulence suppression using gyrofluid and gyrokinetic simulations. However, we have yet to move beyond this simple picture to explain abrupt transitions to improved confinement and to predict transport barrier locations and widths.
Recent gyrokinetic flux tube simulations have demonstrated that transitions from low to high flow and temperature gradients are possible for a limited range of input powers and neutral beam energies, with the range depending on the current profile [Highcock, arXiv:1008.2305; Parra, arXiv:1009.0733]. By fitting the turbulent fluxes from these simulations with a simplified model, we have developed a basic understanding of the circumstances under which these transitions occur at a given radius in a tokamak [Parra]. The simplified flux model has also been used to generate radial ion temperature profiles with significant ITBs. Both the presence and size of these ITBs are strongly affected by the heat and momentum deposition profiles used in the model. Plots of the Gyro-Bohm normalized heat flux versus R/LT (constructed from our model by calculating the heat flux, temperature, and temperature gradient at each radial location in steady-state) show a striking similarity to those given for ASDEX Upgrade discharges (we have compared our results with Fig. 3 of [Wolff, PPCF 45 1757 (2003)]). We would like to construct these heat flux versus R/LT plots at different time slices of DIII-D discharges with varying deposition profiles. | ||
| Resource Requirements: | NBI at full power is needed/ on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. If the first experiment (#120 van Zeeland) that will put 150R/L off axis has occurred, then this experiment could also benefit from using the off-axis beam to alter the rotation and heating profiles. | ||
| Diagnostic Requirements: | MSE, CER, TS, ECE, magnetics, | ||
| Analysis Requirements: | ONETWO, TRANSP, our simplified flux model employing results from gyrokinetic simulations to predict onset conditions for ITB | ||
| Other Requirements: | |||
| Title | 185: Demonstration of QDT=10 ITER scenario with controlled ELMs and radiative divertor operation | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, W. Solomon, T. Petrie, M. Fenstermacher, T. Evans | ITPA Joint Experiment : | No |
| Description: | Demonstrate ITER operational scenario for QDT=10 with Pnet/Pth ~1.3, Prad/Ptot ~80% and ELM control (f_ELM increased by a factor of 5-10 versus uncontrolled ELM frequency) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Starting from an ITER demonstration discharge increase ELM frequency by pellet pacing (or oscillating RMP fields as comparison/back-up) by a factor of 5. Puff impurities (Neon and/or Nitrogen) from the divertor to increase plasma radiation. If needed increased input power to maintain appropriate power flow across the edge above L-H transition. Scan fuelling with gas and HFS pellet injector to achieve optimum density/confinement/radiation level. Repeat scan at a maximum controlled ELM frequency achievable. | ||
| Background: | ITER QDT=10 operation requires the control of both steady and transient loads while maintaining a sufficient level of energy confinement and plasma density. These requirements have not been yet demonstrated simultaneously in any existing devices. The purpose of this proposal is to make use of the new DIII-D capabilities for ELM control plus knowledge of confinement changes with edge power flow to approach or demonstrate the ITER control of steady and transient loads while optimising plasma confinement and maintaining edge power flow at the values expected for ITER | ||
| Resource Requirements: | 2 run days. Pellet pacing for ELM control and fuelling and oscillating RMPs | ||
| Diagnostic Requirements: | Pellet diagnostics, Pedestal diagnostics, Divertor powerr flus diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 186: Assessment of the role of electron & ion edge power flow channels in determining lambda-p in H-modes | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Thermal Transport in the Boundry | Presentation time: | Requested |
| Co-Author(s): | C. Lasnier, T. Leonard | ITPA Joint Experiment : | No |
| Description: | Determine the role of the ion and electron channels in setting the inter-ELM power heat flux width in H-modes | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start from an H-mode at the highest possible current in a configuration optimised for low natural density H-mode to achieve maximum Ti/Te at the edge at two input powers Pnet/Pth ~1 (3-4 MW) and Pnet/Pth ~ 4 (10-12 MW) NBI heated. Perform a gas fuelling scan to Type III ELMy H-mode at both powers to decrease the Ti/Te ratio. As a comparison two scans should be carried out : a) a QH mode plasma at similar current/field and levels of input power to study for highest Ti/Te and b) and H-mode scans at the low power level with ECRH for lowest Ti/Ti ratio. Measurements of the pedestal and edge profiles for electrons and ions and of the power flux at the divertor (and as far as possible electron and ion parameters) will be obtained in all these conditions. | ||
| Background: | Scaling of the power flux at the divertor target show a strong inverse Ip scaling in present experiments which extrapolates to few mm in ITER. It is important to understand to which degree this scaling is determined by transport in the ion channel or the electron channel to understand its extrapolation to IER conditions. | ||
| Resource Requirements: | 1 to 2 run days. High ECRH power for part of the experiments | ||
| Diagnostic Requirements: | Edge, Pedestal and Power flux diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 187: H-mode access and H-mode performance of H-modes with significant off-axis heating | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | General IP | Presentation time: | Requested |
| Co-Author(s): | M. Henderson, P, Gohil | ITPA Joint Experiment : | No |
| Description: | Determine the power requirements for H-mode access and maintaining a Type I ELMy H-mode (and evaluate its performance) for conditions with significant off-axis heating. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Power scans to determine the H-mode transition will we carried out in discharges with NBI only (co-injection) and with similar discharges with 50% NBI (co-injection) and ECRH at a level of 4-5 MW. The ECRH power deposition location will be scanned from central to as much as possible off-axis in 4 consecutive discharges. Following this, at input power level of Pnet/Pth ~1.5 will be chosen and stationary H-modes in Type I ELMy H-mode attempted at the same level of additional heating 50-NBI + 50%-ECRH for the 4 off-axis location studied. If the discharges do not stay in the Type I ELMy H-mode for some of the locations, the ECRH power will be increased until the Type I ELMy H-mode is achieved. For comparison 1 discharge with central NBI at the same power level and 4 discharges with NBI on-axis plus off axis at similar deposition locations as the ECRH ones will be carried out. The balance of co and counter NBI will be adjusted so as to maintain the same level of input torque as the 50%-NBI + 50 % ECRH shots | ||
| Background: | ITER heating systems comprise NBI, ICRH and ECRH. RF systems are optimised for central plasma heating at full toroidal field although they can also heat the plasma at other values of the toroidal field. ECRH, in particular, can provide central heating at full and half field. Deviation from these conditions means that ECRH heating is deposited off-axis. Scenarios of this type (with intermaediate values of the toroidal field) are likely to be carried out when developing H-mode plasmas in ITER from low current to high current both during the non-active and active operational phases. It is important to understand to which level off-axis heating influences the power requirements for H-mode access and for maintaining a Type I ELMy H-mode. | ||
| Resource Requirements: | 1 run day. ECRH heating | ||
| Diagnostic Requirements: | Pedestal, Core and ELM power flux measurements | ||
| Analysis Requirements: | Possibility for off-axis ECRH and NBI heating and associated heat deposition profiles | ||
| Other Requirements: | -- | ||
| Title | 188: Understanding the L-H Power threshold dependence on the X-Point height (Dup #35) | ||
| Name: | Gohil gohil@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Perform experiments to determine the key physics behind the dependence of the L-H power threshold on the X-point height above the divertor | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make all possible diagnostic measurements available that can reveal the key changes in the edge quantities as the X-point height is varied. In particular, examine the changes in quantities such as recycling, turbulence, SOL flows, profiles changes just inside the separatrix, etc. as the X-pint is changes and determine how these quantities relate to the changes in the L-H power threshold. | ||
| Background: | The L-H power threshold has been determined to have a strong dependence on the X-point height above the divertor surface for H, D and He plasmas. This indicates that there is common physics behind this effect, which can result in factors of 2 differences between the experimental L-H power threshold and the predictions from the present L-H power threshold scalings. Due to concurrent changes in several quantities as the X-point is varied, it is presently not possible to definitively explain the physics behind this effect. These experiments aim to separate out certain quantities such as recycling, reconnection lengths at the divertor, etc. in order to determine the important parameters that influence the power threshold. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | All available turbulence and divertor diagnostics, including all profile and SOL diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | Duplicate of #35. | ||
| Title | 189: High density H-mode characteristics & confinement versus heating method (ECRH-NBI) & input torque | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | P. Gohil, M. Henderson | ITPA Joint Experiment : | No |
| Description: | Determination of H-mode performance at high density versus dominant heating scheme and input torque | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start from an ITER demonstration discharge with co-NBI at 4-5 MW and increase plasma density by HFS pellet fuelling to | ||
| Background: | ITER H-mode operation is based on electron heated, high density H-modes with low input torque. While these three factors are known to be detrimental to the achievable plasma confinement no systematic characterisation of plasma confinement in these conditions has been carried out so far. | ||
| Resource Requirements: | 1 to 2 run days. ECRH heating. Pellet HFS fuelling. | ||
| Diagnostic Requirements: | Pedestal, core and ELM power flux diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 190: Assessment of the role of edge plasma density versus collisionality on ELM suppression with RMP | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | M. Fenstermacher, T. Evans, O, Schmitz | ITPA Joint Experiment : | No |
| Description: | Evaluation of the role of plasma collisionality versus edge particle sources on ELM suppression | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiment would consist of detailed fuelling scans with a plasma b_N ~2 at q95 = 3.5 (low and high Ip) and q95 ~ 7 to assess how much the density can be raised without getting ELMs back at constant I-coil current. Currents and q95 are chosen so as to expand a large collisionality range and two resonant windows for RMP supression. For every fuelling point for ELMs re-appear, perform an I-coil scan to identify at which level of current (if at all possible) ELM supression is recovered. As an added result measurements of the divertor profiles of power and particle flux and their splitting in a wide conditions including low and high divertor density, which would be helpful to understand what to expect in ITER, will be obtained | ||
| Background: | ELM suppression in DIII-D has been found to occur in stationary conditions at low densities. Increasing the plasma density by gas puffing leads to the re-appearance of ELMs and this is believed to be associated with the increase of the plasma collisionality and its effects on edge stability. For extrapolation to ITER it is very important to determine whether the lack of ELM suppression is linked to edge particle sources (density) or edge stability (collisionality) as in one case it would pose questions to the viability of the technique for ITER scenarios | ||
| Resource Requirements: | 1 to 2 run days. I-coils for RMP ELM suppression. | ||
| Diagnostic Requirements: | Edge and Pedestal diagnostics and divertor power flux measurements | ||
| Analysis Requirements: | Previous experiments with gas puffing at q95 ~3.5 and previous experiments at q95 ~7 and q95 ~ 3.5 and low Ip | ||
| Other Requirements: | -- | ||
| Title | 191: ELM suppression with n=2 perturbations | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | M. Fenstermacher, T. Evans, O. Schmitz | ITPA Joint Experiment : | No |
| Description: | Study ELM suppression with a dominant n=2 perturbation and compare the findings with the standard n=3 RMP suppression | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start from plasma conditions in which some windows of ELM suppression have been observed already with n=2 (128963) qith q95 ~3.4 and reproduce stationary ELM suppression by fine tuning of q95. Study ranges of ELM suppression with n=2 in terms of plasma beta and collisionality (i.e. gas puffing), rotation (co-counter NBI contribution) and width of the resonant window versus coil current. Repeat few discharges with n=3 for reference for parameters in which conditions are most similar and very different between n-2 and n=3 suppression based on previous evidence. | ||
| Background: | Understanding the requirements for ELM suppression in ITER remains an outstanding issue. ELM coil design is based on ELM suppression empirical criterion derived from experiments with n=3 perturbations; which does not describe satisfactorily findings in other experiments. It is important to compare ELM suppression in DIII-D with other coil spectrum to improve the physics basis for the evaluation of ELM suppression in ITER. | ||
| Resource Requirements: | 1 to 2 run days. RMP coils configured for n=2 and n=3 | ||
| Diagnostic Requirements: | Edge and Pedestal diagnostics, divertor power flux, X-point soft-X ray camera, etc. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 192: TBM Mock-Up Effects on Confinement at High Beta | ||
| Name: | Snipes Joseph.Snipes@iter.org | Affiliation: | ITER Organization |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | M Schaffer, P Gohil, et al | ITPA Joint Experiment : | No |
| Description: | Operate at high betaN > 2 conditions with and without the TBM mock-up fields at maximum current. Determine the optimum error field correction for n=1 error fields in the presence of the TBM fields and compare discharges with and without the optimum n=1 error field correction to determine the changes in energy confinement, plasma rotation, and density pump-out with and without the TBM fields. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Choose discharges at high betaN > 2 in the ITER Similar Shape like the highest betaN discharges from the previous TBM mock-up experiments. Apply previously determined optimum n=1 error field correction in the presence of the TBM fields. Quantify the effects on confinement. Scan betaN to near the highest values reasonably achievable that are not excessively unstable to MHD activity. | ||
| Background: | Previous TBM mock-up experiments in November 2009 showed that the effects of the TBM fields on particle, momentum, and energy confinement appeared to be largest at high betaN. It is important for ITER to understand if these effects can be mitigated with n=1 error field correction in the presence of the TBM fields as well as to determine how the effects may scale with increasing betaN. | ||
| Resource Requirements: | NBI at high power to reach high betaN. Error field correction with the I-coils and C-coils. TBM mock-up near the plasma at full TBM current. | ||
| Diagnostic Requirements: | Diagnostics for plasma rotation, kinetic measurements for energy confinement, magnetic measurements for beta and fluctuation measurements for tearing modes. | ||
| Analysis Requirements: | Energy confinement and plasma rotation analysis. | ||
| Other Requirements: | |||
| Title | 193: Develop low rotation QH-mode with NRMF as a low torque/rotation plasma for FNSF and ITER | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | A. Garofalo, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to demonstrate that low rotation QH-mode plasmas utilizing nonresonant magnetic fields (NRMF) can be created using start-up scenarios which employ low or slightly co-torque injection. Exploit the observation from the 2010 campaign that the NRMF makes the plasma much more robust against locked modes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize the previously developed QH-mode plasma with net zero NBI torque as the starting point. Alter the start-up phase, working towards an early phase with zero torque input by using a combination of balanced beam injection plus ECH for the initial heating. Utilize the balanced beam start-up and ramp up scenario developed in M. Austin's experiments on core transport barriers. Vary the start time of the NRMF to determine whether it can be used in the start-up phase to inhibit locked modes without affecting the power needed to produce the L to H transition. Once a robust plasma is obtained, do a power scan during the flattop phase to insure that good QH-mode operation is obtained over the power range where zero net input torque is available. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. The toroidal rotation in these plasma is much lower than in previous QH-modes with unbalanced NBI and is similar to what one might expect in future devices. Key questions for future devices are
1) Can high beta (betan_N >= 2) QH-mode plasmas with NRMF operate robustly at low rotation? 2) Can we create start-up scenarios with low or slighly co-torque which connect to QH-mode operation with NRMF without exciting locked modes? Work reported by Garofalo at IAEA 2010 demonstrated that the presence of NRMF made the plasma much more robust against locked modes. | ||
| Resource Requirements: | I-coil configured for n=3 NRMF. Reverse Ip. | ||
| Diagnostic Requirements: | All profile and fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 194: Off-axis NBCD in Presence of Stable n/m=1/1 Mode | ||
| Name: | POLEVOI none | Affiliation: | ITER Organization |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Requested |
| Co-Author(s): | J.M. PARK | ITPA Joint Experiment : | Yes |
| Description: | Evaluation of effect of stable n/m=1/1 mode and saw-tooth mixing on the NBCD for ITER-like off-axis current drive | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create magnetic plasma configuration which provides the shape of the NB driven current close to that predicted in ITER. Gradually increase NBI power until it will completely stabilize the saw-tooth mixing or make the ST period long enough to have q=1 surface near the off-axis NBCD maximum. Compare the results of current profile measurements with code prediction. Analyze the effect of the ST mixing on the NBCD and current density profile. | ||
| Background: | The NBI configuration of ITER with one NB aimed at the innermost direction and the second is aimed at the outermost direction is optimal to avoid TAEs.
In this configuration the NB driven current is mostly off axis with current maximum within Ā½ of plasma minor radius. In ITER simulations the fast ion pressure stabilizes the ST so, that the ST starts when q=1 is close to the NBCD maximum and ST area affects the whole area of the maximal NBCD. In the ASDEX-U off-axis NBCD experiments the n/m=1/1 mode activity becomes stable w/o ST mixing when the NBI power exceeds a certain level. For such power the current density suddenly becomes noticeably less than modeling predictions with neoclassical current resistivity and classical slowing down of the NBI fast ions. Presence of stable island can affect fast ion current, electron screening current and current resistivity. The detailed analysis is required to assess possible impact of the stable n/m=1 mode on the NBCD in ITER. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | MSE, neutron | ||
| Analysis Requirements: | Analyses of NBCD, current diffusion, equilibrium | ||
| Other Requirements: | |||
| Title | 195: Density Control and Active Impurity Removal from H-mode Plasmas | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | N. Brooks | ITPA Joint Experiment : | No |
| Description: | This experiment examines the feasibility of using changes in the magnetic balance to control core density and remove impurities from the core of DN and near-DN plasmas. Changing the magnetic balance from dRsep = 0 to dRsep = + 0.2-0.5 cm (with the ion gradB drift downward) can reduce pedestal (and line-averaged) density by up to one-half. Previous studies of impurity injection have shown that argon concentration was about a factor of three higher in double-null H-mode plasmas when compared with the dRsep = +0.5 cm cases with the ion gradB drift direction toward the lower divertor. The issue we want to examine here is that do we get preferential loss of core impurities with respect to deuterium when dRsep is changed slightly from magnetic balalnce and then returned to magnetic balance. With simultaneous pumping on both outer divertor legs of a magnetically balanced high-triangularity DN now possible, DIII-D IS UNIQUELY CONFIGURED TO MAKE A DEFINITIVE STATEMENT. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment is straightforward. Start with a DN shape and inject 60 torr l/s of deuterium gas from gasA, starting at t = 2.0 s. The direction of the toroidal field is forward, i.e., the ion gradB drift is toward the lower divertor. Argon impuries are injected at 0.3 torr l/s into the private flux region of both divertors. Wait for steady conditions; this should take the discharge out to about t = 4.0 s. Between t = 4.0 s and t = 4.4 s, change dRsep from 0 to +0.5 cm. Hold dRsep = +0.5 cm from t = 4.5 s to 4.7 s. Then return dRsep to 0, starting at t = 4.7 s and finishing up at 5.2 s. Compare argon impurity density before dRsep is changed with the impurity density after dRsep is restored to dRsep = 0. How long does it take the argon density to return to its original value? | ||
| Background: | The results of previous experiments have suggested the possibility of actively regulating plasma density by altering the magnetic balance of the plasma configuration. We also obtained a very limited set of data that suggested that impurities already in the core plasma can be exhausted by using this same regulating method. Demonstrating that we can (actively) control density and preferentially exhaust impurities from the core plasma of near-DN plasmas, provides a powerful tool that can significantly improve the prospects of futuristic tokamaks, which may have a serious problem with impurity accumulation in the core, including helium. | ||
| Resource Requirements: | Machine time: 0.25 (forward Bt), only the upper outer divertor and lower outer cryo-pumps are at liquid helium temperature, minimum 6 co-beam sources. | ||
| Diagnostic Requirements: | Asdex gauges inside both outer baffles, Asdex gauge in the upper PFR, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER. | ||
| Analysis Requirements: | UEDGE, ONETWO, MIST | ||
| Other Requirements: | -- | ||
| Title | 196: The Compatibility of Radiative Divertor with AT Plasma Operation | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | SSI Group | ITPA Joint Experiment : | No |
| Description: | This study will combine ALL the essential elements for making the first real test of a radiating divertor concept in an AT/hybrid DN (or near-DN) plasma, using realistic high triangular shape and particle exhaust configuration anticipated for high performance tokamaks. PRESENTLY, ONLY DIII-D HAS THE CAPABILITY TO MAKE THIS TEST. Argon is injected into the private flux region near the upper outer divertor separatrix target. Enhanced deuterium plasma flow toward the divertor in the low field SOL is enhanced by a combination of deuterium gas injected upstream of both outer divertor targets and active cryo-pumping from both outer divertor locations; advanced tokamaks will likely not use an inner pump for a variety of reasons that will not be discussed here. Previous experiments have shown that setting the ion gradB drift direction toward the lower divertor and taking dRsep = +0.5 cm will yield the best chances of optimizing the benefits of (near) double-null shape with maintaining a high performance relatively clean of impurity accumulation. An additional benefit of having the ion gradB drift direction out of the divertor is that we are running at a significantly lower core density than if the ion gradB drift direction reversed, so we expect to include RF in our advanced scenarios. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The plasmas are near-DN AT, and both outer divertor cryo-pumps are at liquid helium temperature. The gradB-ion drift direction is downward. dRsep = +0.5 cm. RF heating is used. This experiment is probably best done in as follows:
* First establish the sensitivity of AT plasmas to deuterium gas injection. Scan the deuterium gas puff rate, i.e., establish operational limit to how much D2 gas injection the AT plasma can accommodate before plasma degradation results. * Scan of the argon injection rate at a reasonable D2 injection rate, using a deuterium gas puff rate that maintains good AT properties as established above. Important measurables from this experiment are, first and foremost, the changes in energy confinement and current profile. Other important measurables include changes in the (poloidal) radiated power distribution and heat flux values, changes in the density and electron temperature at the divertor targets, and the accumulation of argon in the core and divertor plasmas. | ||
| Background: | High performance AT in the DN and near-DN configurations are attractive for future power reactor operation due to their high toroidal beta and energy confinement properties. However, for futuristic AT-machines (like ARIES-AT), there can be severe divertor power loading problems. A possible way of reducing excessive power loading at the divertor target(s) is to radiate significant power outside the main plasma, mainly in the divertor (hence, radiative divertor). But the resulting divertor cooling may also lead to a cooling of the upstream (core) plasma, which, in turn, may result in a marked degradation in AT-edge properties (e.g., bootstrap current). The expected increase in the argon presence in the pedestal can also be expected to affect the AT-pedestal adversely.
Previous work with radiating divertor H-mode DN plasmas has shown that the balanced DN results in overly rapid accumulation of the seeded impurity (argon) in the core plasma. Two important reasons for this are (1) the relatively easy penetration of an impurity specie from the high field side into the core plasma of the DN and (2) the particle drifts in the scrape-off layer plasma in one of the divertors that always assist in the escape of injected impurities from the divertor region to the vulnerable high field side SOL. On the other hand, the radiating divertor was shown to be effective in magnetically unbalanced DNs (dRsep=+0.5 cm with gradB drift down) for reducing divertor heat flux while still maintaining good H-mode properties. This configuration has also been show to produce the lowest density we can achieve in DIII-D for parameters generally used in AT experiments, and this should facilitate the use of RF heating and current density control. | ||
| Resource Requirements: | Initially, this experiment should be run in piggy-back. If the results are favorable, then a dedicated followup experiment can be planned. Both outer baffle cryo-pumps cold should be cold and there should be minimum of six co-beam sources. | ||
| Diagnostic Requirements: | sdex gauges inside both outer baffles, Asdex gauge in the upper PFR, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER. | ||
| Analysis Requirements: | UEDGE, ONETWO, MIST | ||
| Other Requirements: | -- | ||
| Title | 197: Fast Ion Profile Induced AlfvƩn Eigenmode Phase Variation | ||
| Name: | Tobias tobias@lanl.gov | Affiliation: | Los Alamos National Laboratory |
| Research Area: | Energetic Particle | Presentation time: | Requested |
| Co-Author(s): | D.A. Spong, R. Nazikian, M. Van Zeeland, W.W. Heidbrink, N.C. Luhmann, Jr. | ITPA Joint Experiment : | No |
| Description: | Directly measure modifications to RSAE and TAE structures due to the non-perturbative influence of fast ion profiles, particularly radial phase variations, i.e. eigenmode twisting. Shots with on-axis beam injection and off-axis beam injection will be compared. Varying beam power and off-axis beam injection angle will enable a scan of local fast ion pressure gradient with fast ion kinetic pressure peaking at different radial positions. Analysis will focus on the expected variation of 2D eigenmode phase structures for varying fast ion profiles and the expected repeatability of amplitude eigenfunctions for similar thermal ion profiles. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Establish L-mode plasma with early beam heating similar to 142111. 2) Introduce delayed off-axis beam injection to modify fast ion profiles while retaining thermal ion and current profiles as similar to 142111 as possible. 3) Vary off-axis beam power to obtain a scan of local fast ion pressure gradient in a region of low q shear. 4) If possible, vary off-axis beam injection angle in order to scan the location of maximum beam deposition with respect to q_min. 5) In each case, measure simultaneously the 2D temperature perturbation at inboard and outboard radii with ECEI for comparison with non-perturbative modeling. Fast ion loss detector (FILD) and fast ion d-alpha (FIDA) systems will identify modifications of the fast ion profile and coherent losses of fast ions due to AE activity. | ||
| Background: | AlfvƩn eigenmodes imaged during the 2010 campaign by the new ECEI diagnostic exhibited structural deformations which break with the symmetry constraints of ideal-MHD modeling. Analysis of the phase structure of these modes has led to a readily identifiable quantity which reveals the pervasive influence of non-perturbative effects [1,2]. This quantity provides a means for validating non-perturbative eigenmode solvers under conditions of known fast ion profiles, while a quantifiable relationship may yield diagnostic capability. The experiment proposed is intended to resolve remaining questions as to why 2D phase variations are manifest differently in different modes and at different times, as well as isolating this phenomenology from other non-perturbative effects. Further, modifications to coupled fast ion losses for well characterized changes in eigenmode structure are expected [3].
[1] B. Tobias, et. al., PRL (submitted) (2010) [2] M.A. Van Zeeland, et. al., PRL 97,135001 (2006) [3] M.A. Van Zeeland, et. al., Phys. Plasmas, (invited APS-DPP, 2010) | ||
| Resource Requirements: | All available beams. 150 off-axis beam with multiple tilt angles preferable. | ||
| Diagnostic Requirements: | ECEI will be primary imaging diagnostic. ECE radiometry and MSE also critical. All FILD and FIDA systems. | ||
| Analysis Requirements: | NOVA and TAEFL (ORNL) will be the primary AE modeling codes. | ||
| Other Requirements: | -- | ||
| Title | 198: Characterization of Sawtooth Reconnection | ||
| Name: | Tobias tobias@lanl.gov | Affiliation: | Los Alamos National Laboratory |
| Research Area: | Stability | Presentation time: | Requested |
| Co-Author(s): | R. Nazikian, H.K. Park, A. Turnbull, G.S. Yun, N.C. Luhmann, Jr. | ITPA Joint Experiment : | No |
| Description: | Combine images from 2D ECEI with available density fluctuation diagnostics in order to resolve electron plasma behavior and magnetic flux evolution during the fast reconnection associated with the sawtooth crash. Sawtoothing plasmas for which long-lived saturated precursors are present will be observed with temperature imaging diagnostics (ECEI) along with density diagnostics in new configurations motivated by questions raised during 2010 experiments. Sawtoothing plasmas in which the precursor mode is suppressed (i.e. ļæ½??beanļæ½?ļæ½ discharges) will be obtained in a similar manner for comparison. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Establish elongated, ļæ½??ovalļæ½?ļæ½ shaped L-mode plasmas with central safety factor below unity as in shot 118164. 2) Optimize plasma shape and position for available density diagnostics (BES, CO2 interferometers, SXR, core reflectometer). 3) Repeat for high-triangularity ļæ½??beanļæ½?ļæ½ case similar to shot 118162. | ||
| Background: | Previous studies of the sawtooth crash for ļæ½??ovalļæ½?ļæ½ and ļæ½??beanļæ½?ļæ½ shaped plasmas [1] concluded on the basis of stability analysis at times several ms before the crash and on the magnetic signature of the crash itself that crashes in ļæ½??ovalļæ½?ļæ½ discharges are of a quasi-interchange type, while ļæ½??beanļæ½?ļæ½ discharges exhibit crashes initiated by resistive internal kink instability. Imaging the sawtooth crash in ļæ½??ovalļæ½?ļæ½ discharges with ECEI during the 2010 campaign revealed a precursor consistent with quasi-interchange instability, but did not associate this precursor with the fast collapse at the end of the sawtooth cycle [2]. Rather, the crash is tentatively interpreted as a secondary instability, for which the underlying plasma displacement can only be determined with additional data pertaining to the density perturbation. Characterizing sawtooth crashes in ļæ½??ovalļæ½?ļæ½ and ļæ½??beanļæ½?ļæ½ discharges with 2D temperature fluctuation imaging and carefully arranged density diagnostics is aimed at resolving apparent discrepancies in these descriptions and providing a more complete description of this spontaneous plasma reorganization which is of general interest to the plasma physics community.
[1] E.A. Lazarus, et. al., PoP 14, 055701 (2007). [2] B. Tobias, et. al., PoP (invited APS-DPP 2010). | ||
| Resource Requirements: | Beams, no RF heating | ||
| Diagnostic Requirements: | ECEI, ECE radiometry, MSE, BES, core reflectometry, CO2 interferometers, SXR | ||
| Analysis Requirements: | MHD stability analysis to follow | ||
| Other Requirements: | -- | ||
| Title | 199: Investigate NRMF driven torque in low collisionality, low rotation QH-mode plasmas | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | K. Burrell, J.K. Park, W. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to systematically investigate the dependence of the NRMF driven torque on the magnetic field configuration and the plasma rotation, and compare with theoretical calculations. This is an essential step in order to progress toward better understanding and modeling of the torque driven by a NRMF.
This experiment will involve scans of the injected NBI torque in counter-rotating QH-mode plasmas with varying configurations of applied NRMFs. The scans will be guided by existing IPEC-NTV calculations of the NRMF torque from various possible coil configurations, based on a previous DIII-D discharge. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use discharge 138593 as a template: NBI counter torque ramps from ~6 Nm to ~0.5 Nm, with 6 kA of n=3 odd parity I-coil field.
Repeat NBI torque ramp with different NRMF configurations. Plan to repeat each NRMF configuration with density feedback (or pellet injection), in order to match the density obtained without NRMF. - reference discharge without NRMF - 6 kA of n=3 odd parity I-coil field, both polarities - 6 kA of n=3 odd parity I-coil (polarity of larger torque) + 5 kA n=3 C-coil (5 kA instead of 3 kA should be possible next year) - 6 kA of n=3 odd parity I-coil (polarity of larger torque) - 5 kA n=3 C-coil - 6 kA of n=3 even parity I-coil, both polarities - 6 kA of n=3 even parity I-coil (polarity of larger torque) + 5 kA n=3 C-coil - 6 kA of n=3 even parity I-coil (polarity of larger torque) - 5 kA n=3 C-coil Switch to using I-coil for n=1 error field correction, then repeat NBI torque ramp with: - reference discharge without NRMF - 7 kA of n=3 C-coil, both polarities | ||
| Background: | Comparison of DIII-D discharges 138593 (6 kA odd-parity n=3 I-coil) and 137234 (no I-coil reference) yields a measurement of the NRMF driven torque as a function of the plasma rotation, from rapid counter rotation to near-zero rotation.
IPEC-NTV calculations for DIII-D discharge 138593 calculate an NRMF torque evolution that compares with the measurment with varying degree of success. The theoretical calculation matches the measurement well at high rotation and very low rotation, poorly in the intermediate rotation range. The IPEC-NTV calculations also predict that the odd-parity n=3 I-coil gives the largest counter torque on the plasma. However, the relative magnitude of the torque from different NRMF configurations also depends on the plasma rotation. Benchmarking NTV calculations using the dependence of the NRMF torque on the magnetic field geometry and on the plasma rotation is an essential step before we can use with confidence NTV theory to predict the effect of NRMFs on future devices that may rely on the NRMF torque to access QH-mode, such as ITER and FNSF. | ||
| Resource Requirements: | Same as DIII-D discharge 138593.
New leads are been prepared to allow three C-supplies on the C-coil. | ||
| Diagnostic Requirements: | Same as DIII-D discharge 138593. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 200: Full Non-Inductive Current Drive at Low Plasma Current | ||
| Name: | Cunningham geoffrey.cunningham@ccfe.ac.uk | Affiliation: | CCFE |
| Research Area: | Heating & Current Drive | Presentation time: | Requested |
| Co-Author(s): | S. Yoon, Y-K Oh (KSTAR), B. Xiao (EAST), G. Jackson, A. Hyatt, J.
Leuer, (GA) |
ITPA Joint Experiment : | No |
| Description: | Goal of experiment is to generate a fully non-inductive CD in true steady state at end of the discharge at low plasma current and within the constraints of the DIII-D systems. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will re-establish one of our low current, noninductive current drive discharges (600kA, 119787). We will then adjust the power mix of ECCD, ECH and NB power to lower the overall plasma current to obtain steady state current levels at the end of the discharge. The assumption is that we can extend the discharge time as the current and power levels are reduced with the objective of achieving truly steady state profiles at the end of the discharge. Different levels of toroidal field will be investigated with lower levels allowing for longer overall pulse lengths. We will try to extend the pulse as long as possible at these lower current (and power levels) to seek truly steady state operation. Once we obtain a range of plasma currents driven fully non-inductively, we will investigate the ability to increase and decrease current by purely noninductively means. We will explore variations in density and plasma shape, particularly as the latter relates to divertor coil currents, which can contribute current increase/decrease through inductive coupling. These experiments will be carried out using a standard plasma startup with the E-coil frozen during the NICD phase. | ||
| Background: | Over many years DIII-D has worked extensively to establish fully non-inductive, steady state discharges in a variety of plasma configurations. However, many of these ā??stationaryā?? discharges have slowly evolving current profiles at the end of the discharge, which is limited primarily by machine constraints [ex. Politzer, NF 2005]. Some of these constraints could be eliminated if we operate at much lower current than is typically run in DIII-D (i.e. < 600kA). In addition, the recent introduction of more NICD capability and options, like added ECCD, stearable EC mirrors, Off-Axis NB, and Fast Wave, we now have much more flexibility in generating truly steady state plasmas. Recent experiments in KSTAR/EAST have achieved H-mode quality plasmas at current levels of order 300-800 kA, low toroidal fields (1.3-2T) and low injected power levels (1-1.3MW) [News-EAST, Xiao, Nov 2010; 1st H-mode KSTAR, Kwon, Nov 2010]. If DIII-D were to explore these low current levels with thrust at obtaining truly steady state, NI profiles we could potentially operated for extended times (>10s) which would allow for obtaining completely stationary discharges. This would also connect us with current experiments being performed in KSTAR and EAST and provide for an excellent collaboration and guidance as these machines evolve to tackle NICD issues. Additionally, this experiment would allow us to better connect with the solenoidless development work in DIII-D. | ||
| Resource Requirements: | Tokamak, H-mode plasma, ECH (5-6 Gyrotrons), Co-NB, (possibly FW) | ||
| Diagnostic Requirements: | Fast magnetics, MSE, Thompson Scattering, Spred, Visible camera view bumper limiter, Bolometers, IR camera, CO2 Interferometer, ECE, SXR | ||
| Analysis Requirements: | EFIT, Transp. | ||
| Other Requirements: | Quantification of the present machine flat-top time limits at reduced parameter set. | ||
| Title | 201: Does ITER need 1, 2 or 3 EF correction sets - understanding the limits of the ideal response model | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Requested |
| Co-Author(s): | A Boozer, JK Park, RJ La Haye | ITPA Joint Experiment : | Yes |
| Description: | Ideal plasma response models have not so far successfully explained behavior of error field correction. While the models predict that near perfect correction should be possible, experiments on DIII-D and JET show typically 40-50% correction at best, when measured in density ramp-down experiment. This means the predicted field in the plasma response model is an out by an order of magnitude for corrected error field cases. This is a critical question for ITER, as these codes are being used to predict (possibly extremely over-optimistically) correction requirements and address key questions such as how many coil sets are needed for correction. But we have no validation on the issues of how much correction is needed, or whether this can be achieved with one correction coil set. This is also an issue of fundamental understanding - if the ideal response models as applied to data are missing more than half of the intrinsic error effect, we need to learn how to better apply or extend the models to improve our understanding of the plasma response to 3-D fields. Possible explanations include combined effects from less dominant modes in the plasma, additional braking mechanisms (different qs, NTV), or other components of error field (eg n=2,3). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The approach is to be executed in two stages: (i) firstly test the basic principal of the ideal plasma response and potential limitations by using C coils as a simulated source of error and I coils as the correction field. This will enable testing of optimized correction at high amplitude, and in Ohmic or H mode plasmas. Can we correct 90 % of the C coil error or only 40%? Applied fields might be pulsed or rotated to do magnetic or other probing of plasma response as an additional diagnostic (ii) Explore whether improved intrinsic error correction is possible in DIII-D with more than one coil set by using C coils on top of I coils to optimize correction. This would use a C coil phase scan to measure any residual error field after optimal I coil correction, and then density ramps downs (ohmic) or torque ramp-downs (H mode) to measure improvement in correction. | ||
| Background: | The discrepancies of ideal with DIII-D error correction limitations have been acknowledged in publications [Park 2007] but dismissed as control system inadequacies at low density. however the effects are systematic [Scoville], and also manifest on JET [Buttery 2000]. So rather than an abrupt control system failure, they are a clear, well measured progressive effect. This is a fundamental issue for ITER that goes to the heart of how good error correction can be, and how many coil arrays are needed. To date, the only data we have on this issue invalidates the models being used. DIII-D is uniquely able to access these questions with its I and C coils. The issue also goes to the heart of understanding the plasma response to 3-D fields in other applications such as RWM or ELM control. | ||
| Resource Requirements: | 1 day for each stage. I and C coils. Modest other machine requirements | ||
| Diagnostic Requirements: | Usual MHD and locked mode sensors | ||
| Analysis Requirements: | Results should be self evident, but associated theoretical analysis to account for the ideal response limitations and new mechanisms should be pursued | ||
| Other Requirements: | |||
| Title | 202: Understanding the limits of the ideal response model (Dup 201) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | A Boozer, JK Park, RJ La Haye | ITPA Joint Experiment : | No |
| Description: | Ideal plasma response models have not so far successfully explained behavior of error field correction. While the models predict that near perfect correction should be possible, experiments on DIII-D and JET show typically 40-50% correction at best, when measured in density ramp-down experiment. This means the predicted field in the plasma response model is an out by an order of magnitude for corrected error field cases. This is a critical question for ITER, as these codes are being used to predict (possibly extremely over-optimistically) correction requirements and address key questions such as how many coil sets are needed for correction. But we have no validation on the issues of how much correction is needed, or whether this can be achieved with one correction coil set. This is also an issue of fundamental understanding - if the ideal response models as applied to data are missing more than half of the intrinsic error effect, we need to learn how to better apply or extend the models to improve our understanding of the plasma response to 3-D fields. Possible explanations include combined effects from less dominant modes in the plasma, additional braking mechanisms (different qs, NTV), or other components of error field (eg n=2,3). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The approach is to be executed in two stages: (i) firstly test the basic principal of the ideal plasma response and potential limitations by using C coils as a simulated source of error and I coils as the correction field. This will enable testing of optimized correction at high amplitude, and in Ohmic or H mode plasmas. Can we correct 90 % of the C coil error or only 40%? Applied fields might be pulsed or rotated to do magnetic or other probing of plasma response as an additional diagnostic (ii) Explore whether improved intrinsic error correction is possible in DIII-D with more than one coil set by using C coils on top of I coils to optimize correction. This would use a C coil phase scan to measure any residual error field after optimal I coil correction, and then density ramps downs (ohmic) or torque ramp-downs (H mode) to measure improvement in correction. | ||
| Background: | The discrepancies of ideal with DIII-D error correction limitations have been acknowledged in publications [Park 2007] but dismissed as control system inadequacies at low density. however the effects are systematic [Scoville], and also manifest on JET [Buttery 2000]. So rather than an abrupt control system failure, they are a clear, well measured progressive effect. This is a fundamental issue for ITER that goes to the heart of how good error correction can be, and how many coil arrays are needed. To date, the only data we have on this issue invalidates the models being used. DIII-D is uniquely able to access these questions with its I and C coils. The issue also goes to the heart of understanding the plasma response to 3-D fields in other applications such as RWM or ELM control. | ||
| Resource Requirements: | 1 day for each stage. I and C coils. Modest other machine requirements | ||
| Diagnostic Requirements: | Results should be self evident, but associated theoretical analysis to account for the ideal response limitations and new mechanisms should be pursued | ||
| Analysis Requirements: | Ideal response modeling should be undertaken to explore additional sources of error field | ||
| Other Requirements: | -- | ||
| Title | 203: Demonstration of disruption tolerant PFM surface | ||
| Name: | Wong wongc@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov (UCSD), E. Hollmann (UCSD), D. Humphreys, N. Brooks, J. Watkins SNL), C. Lasnier (LLNL), B. Chen, D. Wall, A. Hassanein (Purdue University), W. Wampler (SNL), A. McLean (ORNL) | ITPA Joint Experiment : | Yes |
| Description: | Plasma facing material (PFM) is a critical element of the high performance DT ITER and tokamak reactor design. It is the interface between the plasma and the high performance first wall and divertor components. Presently, C at the divertor is proposed to handle the early operation of ITER to withstand transient events like disruption, solid W is projected as the preferred PFM due to its low physical sputtering, high thermal performance at elevated temperatures, and high neutron fluence tolerant properties. Unfortunately, the commonly proposed material W could suffer radiation damage from implantation of alpha charged particles and experience blistering at the first wall and the formation of submicron fine structure at the divertor. Furthermore, it will melt under Type-I edge localized modes (ELMs) and disruption events. A potential remedy was demonstrated using linear machine material exposure which showed that, with a background of low-Z material, the W-surface damage could be prevented or significantly reduced. On transient tolerant, even for ITER during the all W-surface phase, as a conservative engineering design, the first wall and divertor PFM must withstand a few unanticipated disruptions even when the disruption and ELM mitigation techniques are fully engaged. Using a low-Z sacrificial material, like Si, covering about 50% of the W-surface and to a depth of ~1mm, will allow W to withstand type-I ELMs and disruptions without serious damage while retaining the capability of transmitting high heat flux for power conversion. An equivalent Si thickness of 10 Ī¼m is sufficient to form a vapor shielding layer during a disruption that protects the W-substrate from serious damage. Accordingly, transient tolerant PFM surface test buttons have been fabricated and initial results have been obtained with exposure in the DIII-D divertor, but without a disruption loading. This proposal is to demonstrate the vapor shielding effect during a disruption loading on DiMES and on the improved Si-W button samples with the use of the DiMES mechanism in DIII-D. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The proposed experiment would be the exposure of high power disruption on the Si-W buttons and the emission of Si, W and carbon will be monitored. After the exposure the sample will be examined to determine the amount of Si vaporized and the possible melting of the W material. A second disruption exposure would be used to determine the W buttons damage without the loading of Si. Results from these two exposures will be compared to determine the impacts from with and without the vapor shielding of Si. | ||
| Background: | For the development of advanced solid PFM, it is difficult to foresee any low activation metallic alloy that can outperform W-alloy. However, in addition to potential surface damage from helium ions, W or any metallic surface is sure to melt to some extent under the thermal dump of transient events such as a disruption. Therefore, in order to develop an acceptable robust PFM, an innovative approach for maintaining adequate material for vaporization to handle a limited number of transient events is needed. A proposal for holding an adequate low-Z material on the W-surface is to fill the indentations of a W surface with Si. The indentations on the W-surface are designed with a diameter of 1 mm and a depth of 1 mm, on a W surface thickness of 2 mm. With a sufficiently large surface area (~50%) filled with Si, this design would allow for the possibility of protecting the W-substrate from the thermal dump of a disruption or high power ELMs via the vapor shielding effect. The W-surface was selected to have a thickness of ~ 2mm to retain the capability of high heat transmission for power conversion. Button samples of this Si-W concept were fabrication and exposed in DIII-D. For the discharge we found that the thermal dump did not get onto the DiMES surface, and it failed to demonstrate the beneficial effect of a vapor shield. We propose to perform similar experiment in DIII-D in 2011 with better positioning of the thermal dump from a disruption and with the used of improved Si-W button samples. Instead of indentations, we will use slots, and with careful orientation with the magnetic field, we should be able to minimize the exposure of W-edges to the shallow angle impinging magnet field. However, even with the already exposed results we can report that as expected that nearly all the surface Si has been removed from the W-button, yet there is still a good amount of Si remaining in the W-indentations, which would be ready to provide a vapor shielding effect when a thermal dump reaches the Si-filled W-surface.
Based on higher magnification of the exposed Si-filled W-buttons from Energy Dispersive X-Ray (EDX) diagnostics, we can see the melted Si surface material and the relative clear surface of the W and graphite buttons, indicating the lack of Si transport during the discharges. During the discharge, line radiation from 176 ā?? 512 nm was monitored using an optical chord viewing the DiMES sample face directly using an Ocean Optics USB2000 spectrometer (Tint = 100 ms, 0.164 nm/pixel dispersion, 1.07 nm optical resolution). Within the sensitivity limit of the instrument, we found that essentially all the radiation lines were from carbon; no verifiable Si or W lines were found during the discharge. Therefore, it is essential to repeat the disruption exposure experiment in DIII-D, and with the use of better Si-W button samples. | ||
| Resource Requirements: | We will need to work with the disruption group and figure out the best way to put a high power disruption loading onto the DiMES sample, first exposure loading with Si-W buttons and second exposure with similar W-buttons but without the loading of Si. Set up shots will be used before the sample button sample is inserted to the divertor location for disruption exposure. Sample surface measurements will be performed by the material examination groups at GA and SNL-A. | ||
| Diagnostic Requirements: | Spectroscopic monitoring of Si, W and C lines before and during disruption.
Langmuir probes and IR camera should be used to characterize the discharges before and after the disruption discharges. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 204: Early Divertor Formation and H-Mode During Ramp-Up | ||
| Name: | Leuer leuer@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Not requested |
| Co-Author(s): | G. Cunningham [MAST], G. Jackson, A. Hyatt, D. Humphreys, N. Eidietis, P. Politzer, M. Walker | ITPA Joint Experiment : | No |
| Description: | Goal of experiment is develop a startup scenario, which obtains early divertor (<100ms, <300kA) and early H-mode during plasma ramp-up. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will begin by moving the wave enables (ability to generate coil current) on selected coils which will allow divertor operation from the nominal >=100ms to of order 50ms or less. Early plasma centroid detection, vertical stability and plasma control are all expected to be challenges during this early period of divertor generation. We will explore two approaches: 1) old-fashion Isoflux Control using a limited number of selected probes and 2) rtEFIT control. ECH is expected to be utilized early for preionization and burn-thru. Lower loop voltage will be utilized to minimize VV eddy-currents. Once a reasonable early diverted plasma is obtained, early beams will be utilized to push the plasma into H-mode. | ||
| Background: | Numerous plasma startup development campaigns have been performed over the years on DIII-D including the seminal plasma initiation work of [Lloyd, NF 1991] and more recent ITER studies [Jackson, IAEA 2010]. In all these startups, the plasma is limited on the inside or outside surfaces during the first approximately 200ms. Many of the coils are not enabled (ability to carry current) until 100ms. rtEFIT is not enabled until well into the discharge at plasma current levels typically in excess of 400kA. We expect to be able to actually dirvert much earlier and perhaps just after plasma initiation or less than 100ms. Early diverted plasma would remove it from contact with the wall and reduce impurity input to the plasma and wall heating. This would greatly reduce wall protection during early startup in ITER. In addition, early beams would possibly allow a diverted plasma to reach H-mode early in the discharge with comsenurate changes expected in flat-top profile parameters. A substantial reduction of resistive flux would also be expected over a standard Ohmic startup and this would have a substantial impact on ITER and next generation devices. These results would establish the lowest current at which divertor plasma operation could be achieved and lowest current at which NB can create H-mode plasmas. Much of this work would also be advantageous to other machines like KSTAR and EAST, allowing these machines to extend there present startup to early H-mode and reduction in flux demand. | ||
| Resource Requirements: | ECH, Standard Startup, ITER large Bore startup, NB-Co (2-sources). : Request several 2hr Thursday Nights for control development + 2 x 1/2 day experiment. (Note: successful implementation would allow remaining shot and flat top for other experiments (i.e. ITER discharges). | ||
| Diagnostic Requirements: | Fast magnetics, ECE, TS, MSE, filterscopes, Spread, Visible camera view bumper limiter, bolometers, IR camera, CO2 Interferometer, SXR | ||
| Analysis Requirements: | EFIT, JFIT, vertical stability analysis | ||
| Other Requirements: | Initial testing needed to test and validate expected new early plasma detection and control algorithms | ||
| Title | 205: Document response of low-k turbulence to Te increase using FW instead of ECH | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Verify that turbulence responds to increases in Te/Ti that are produced with FW+NBI the same way as when increases in Te/Ti are produced with ECH+NBI. To confirm that the response in CECE is NOT due to non-thermal EC emission generated by ECH and to ensure that comparisons with TGLF and GYRO are valid, a new experiment where Te/Ti is increased using FW should be performed under the same conditions where ECH was used in the past. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make simple, sawtooth free L-mode plasmas similar to past discharges used for TMV work. Reproduce the plasmas from a Te-scan pair 138040/138038. These are chosen because the shape used kept plasma in L-mode without resorting to inner wall limited discharges. 138038 was heated with 2.5MW co-NBi and 2.5 MW ECH deposited near rho = 0.2. Replace 2.5 MW of ECH used in 138038 with 1.0-2.5 MW of FW. Measure response in BES and CECE density and electron temperature fluctuations, respectively. Reproduce 138040 that had 2.5 MW of co-NBI only. Match the density. | ||
| Background: | Past measurements of the response of turbulence to Te/Ti in DIII-D L-mode and Ohmic plasmas has shown dramatic increases in Te-tilde/Te (low-k) but usually much smaller responses in n-tilde/n (low-k).
While this has been attributed to increases in the TEM drive that results from the increase in Te/Ti and the decrease in collisionality (an interpretation that is consistent with TGLF and GYRO predictions), there is an outstanding possibility that the CECE responds very strongly to the ECH rather than an underlying change in local turbulence drive terms. This could occur if the ECH drives even a slight perturbation to the EEDF, which cannot relax to a Maxwellian on the fast timescales relevant to the turbulence dynamics. Then the CECE (a radiometer based measurement) is responding to enhanced EC emission rather than enhanced turbulence. While there is no evidence from experiments to suggest that this is occurring (White POP 2010), it would be useful to use piggy-back/shared time during some of the FW experiments this year to dial up 138040/138038 and to check it out once and for all. | ||
| Resource Requirements: | FW (1.0- 2.5 MW) coupled to L-mode plasmas, equlibrium from 138038/138040 around t = 1600 ms). 150L (on axis for BES), 330L/30L for CER and MSE | ||
| Diagnostic Requirements: | MSE, CER, BES, CECE, standard profile diagnostics | ||
| Analysis Requirements: | ONETWO, TRANSP, CQL3D | ||
| Other Requirements: | -- | ||
| Title | 206: Broadband magnetic feedback-assisted steady state high beta operation | ||
| Name: | In inyongkyoon@unist.ac.kr | Affiliation: | Ulsan National Institute of Science and Technolog |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Requested |
| Co-Author(s): | M. Okabayashi, E. Strait, and RWM Physics group | ITPA Joint Experiment : | No |
| Description: | The proposal is to fully characterize and then quantify the contributions of the broadband magnetic feedback on steady state high beta operation, whose high performance was sustained much longer than otherwise. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Based on recent SSI discharges (e.g. 142347 with betaN > 3), the feedback response time, which determines the bandwidth of the magnetic feedback, will be systematically scanned, while all the other conditions, including ECCD deposition profiles, remain the same. Also, the magnetic feedback gain, which is desirable to be high enough in terms of error field correction (EFC) unless there is a hardware limit, will be optimized. All the key plasma performance parameters, such as betaN, density ne, and energy confinement time tauE , will be monitored, as well. (Estimated to be a half-day experiment). | ||
| Background: | Broadband magnetic feedback, whose feedback response time is shorter than the resistive wall characteristic time tauW, appeared to have contributed to the long sustainment of steady state high beta plasmas [1]. Specifically, several plasma performance parameters (e.g. betaN, ne, and tauE) remained high much longer than in the discharges without broadband magnetic feedback. A preliminary analysis result [2] suggests that broadband magnetic feedback enhanced the damping rates of n=1 resonant components associated with various bursty MHD events (e.g. fishbones), which might have helped high beta plasmas to be sustained longer. That is likely because a quick removal of the n=1 resonant component with the help of broadband magnetic feedback would prevent either resonant field amplification (RFA) or coupling with RWM from taking place in such high performance plasmas. Thus, more quantitative investigation, including the bandwidth dependence of magnetic feedback, may not only help us clarify whether such broadband feedback is essential in steady state high beta operation, but also allow us to quantify the specific requirements necessary for long-duration, high performance plasmas.
References [1] F. Turco et al, APS-DPP (2010) [2] Y. In et al., 23rd IAEA FEC (2010), to be submitted to NF (2010) | ||
| Resource Requirements: | All the available NBI sources, 4 gyrotrons for ECCD, and ICRF (if available for high-beta operation) | ||
| Diagnostic Requirements: | Magnetics | ||
| Analysis Requirements: | |||
| Other Requirements: | I-coils with Audio amplifiers, while C-coils with SPAs | ||
| Title | 207: Testing and correction of n=2 and n=3 error fields on DIII-D | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Requested |
| Co-Author(s): | Rob La Haye | ITPA Joint Experiment : | No |
| Description: | DIII-D error field correction only achieves a 40% reduction in density limit. We need to check for additional sources of error and try to correct them, in order to see (i) if this is an additional significant source of error that needs to be considered for ITER or DIII-D; and (ii) how this impacts plasma response models, which presently fail to account for the limitations in error correction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use I coils to apply n=2 and n=3 field scans (quite limited for n=2) to measure intrinsic error field threshold (Ohmic) and/or braking (in low torque NBI heated plasmas). It may be sensible to apply a constant low level of n=1 field to enhance the sensitivity of the plasma to the n=2 and n=3 fields - when the higher ne fields cause enough braking, this will enable the n=1 fields to penetrate, thereby providing a threshold measurement in the higher n field value. | ||
| Background: | Understanding error correction and plasma response to low n fields is a critical issue for ITER error correction, and important aspect of RWM and RMP-ELM control. The is a longstanding problem on a critical question for ITER - ho to achieve sufficient correction to avoid disruptions - made more urgent by recent studies showing increased plasma response to 3D fields in H modes | ||
| Resource Requirements: | I coils, and probably C coils, with enough PS | ||
| Diagnostic Requirements: | Usual MHD and locked mode | ||
| Analysis Requirements: | Results should be obvious, but further code studies to interpret any significant response obtained would be logical. | ||
| Other Requirements: | THIS ENTRY IS NOT TO LAY CLAIM TO THIS TOPIC, but to ensure relevant ideas go forward for consideration and joint planning by the group. I remain happy to help as team member or experiment leader as needed. | ||
| Title | 208: Testing and correction of n=2 and n=3 error fields on DIII-D (Dup 207) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | Rob La Haye | ITPA Joint Experiment : | No |
| Description: | DIII-D error field correction only achieves a 40% reduction in density limit. We need to check for additional sources of error and try to correct them, in order to see (i) if this is an additional significant source of error that needs to be considered for ITER or DIII-D; and (ii) how this impacts plasma response models, which presently fail to account for the limitations in error correction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use I coils to apply n=2 and n=3 field scans (quite limited for n=2) to measure intrinsic error field threshold (Ohmic) and/or braking (in low torque NBI heated plasmas). It may be sensible to apply a constant low level of n=1 field to enhance the sensitivity of the plasma to the n=2 and n=3 fields - when the higher ne fields cause enough braking, this will enable the n=1 fields to penetrate, thereby providing a threshold measurement in the higher n field value. | ||
| Background: | Understanding error correction and plasma response to low n fields is a critical issue for ITER error correction, and important aspect of RWM and RMP-ELM control. The is a longstanding problem on a critical question for ITER - ho to achieve sufficient correction to avoid disruptions - made more urgent by recent studies showing increased plasma response to 3D fields in H modes | ||
| Resource Requirements: | I coils, and probably C coils, with enough PS | ||
| Diagnostic Requirements: | Usual MHD and locked mode | ||
| Analysis Requirements: | Results should be obvious, but further code studies to interpret any significant response obtained would be logical. | ||
| Other Requirements: | THIS ENTRY IS NOT TO LAY CLAIM TO THIS TOPIC, but to ensure relevant ideas go forward for consideration and joint planning by the group. I remain happy to help as team member or experiment leader as needed. | ||
| Title | 209: 'Fast-track' error field correction (EFC) near no-wall stability limit | ||
| Name: | In inyongkyoon@unist.ac.kr | Affiliation: | Ulsan National Institute of Science and Technolog |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | M. Okabayashi, E. Strait, and RWM Physics group | ITPA Joint Experiment : | No |
| Description: | The proposal is to assess the applicability of a newly developed 'fast-track' error field correction (EFC) methodology, which requires far fewer iterations than conventional EFC method. If successful, this allows us to obtain the 'ideal' pre-programmed EFC waveform without any further iteration in all the experiments where the quality of EFC is critical (e.g. near ideal MHD no-wall stability limit). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In a high beta discharge where strong resonant field amplification (RFA) and/or locked mode is prevalent, the 'fast-track' EFC method [1] will be searched first (using high gain and a relaxation factor), and then conventional EFC procedures will be followed. Once a converged pre-programmed EFC waveform based on the conventional EFC is found, the comparison of both methods will show how accurate the pre-programmed EFC waveform is, as well as how many discharges can be saved. | ||
| Background: | A recent study [1] showed that the error field correction (EFC) strategy should be developed in consideration of the RWM stability conditions, in that the dynamic (feedback-controlled) EFC gain dependence would be remarkably diverse near the no-wall stability limit (stable, marginal and unstable RWM regimes). Specifically, the DEFC would be always either under-correcting in stable RWM regime or over-correcting in unstable RWM regimes, unless the RWM stability condition resides near the marginal limit where no feedback gain dependency is expected.
However, the same study [1] also showed that the most desirable pre-programmed EFC waveform can be easily obtained using high feedback gain in combination of a relaxation factor. According to this modeling, such scheme does not require any redundant discharge only for the purpose of accurate EFC. Considering that many tokamak experiments prefer the use of pre-programmed EFC, this 'fast track' EFC methodology is expected to provide the 'best' quality of EFC without any further iteration. Reference [1] Y. In et al., 23rd IAEA FEC (2010), to be submitted to NF (2010) | ||
| Resource Requirements: | More than 4 NBI sources to maintain high-beta operation above ideal MHD no-wall stability limit | ||
| Diagnostic Requirements: | Magnetics, ECE (including ECEI), SXR | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | I-coils/C-coils to be connected with all the available power supplies (SPA and AAs). | ||
| Title | 210: Advanced Inductive operation at low rotation | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Requested |
| Co-Author(s): | P.A. Politzer, T.C. Luce | ITPA Joint Experiment : | No |
| Description: | The goal of the experiment is to develop a low rotation Advanced Inductive target and characterize its performance relative to otherwise comparable all-co NBI discharges. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop an AI startup with low torque, ideally with ECH+RF only, but with additional balanced NBI as needed. Perform rotation scan in the counter-Ip direction to map out rotation dependence of confinement (in particular, is the performance symmetric around zero). An interesting question will be whether the current drive from the counter beam will provide the appropriate q-profile without need for the NTM typically associated with AI/hybrids. If performance is better with net counter torque than near-zero, substitute counter beam torque with n=3 NRMF to drive the rotation toward the (counter) offset rotation. In fact, there is a high chance that this technique may be necessary even for the near-zero torque cases, since the NRMF torque acting on counter rotation improves resilience to locked modes (Garofalo IAEA etc). | ||
| Background: | Previous efforts at investigating low rotation AI plasmas have focused on starting with the standard AI recipe with all co-NBI and then reducing the torque with the addition of counter torque. This approach has shown a significant reduction in confinement as the rotation is reduced. It is not clear how the performance will be when starting from a low rotation target. | ||
| Resource Requirements: | All co/ctr beams. n=3 I-coil. | ||
| Diagnostic Requirements: | Full profile diagnostics, complete pedestal set. FIDA and main ion CER (since low rotation, high beta is the conditions where difference between impurity and main ion rotation may be most significant) | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 211: Steady state scenario at reduced rotation | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of the experiment is to investigate elevated qmin, steady state scenario plasmas at reduced rotation using NRMF. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use typical high betaN steady state scenario discharge and compare performance and susceptibility to tearing modes associated with reducing rotation by: a) applying n=3 NRMF; and b) adding counter NBI. | ||
| Background: | Previous efforts at reducing the rotation in high betaN discharges by adding counter NBI generally experienced problems with tearing modes. It is not clear whether this is an effect of reduced rotation, modification to the current drive, increased power, or a combination thereof. Using n=3 NRMF to brake the plasma may be useful in isolating the rotation effect. | ||
| Resource Requirements: | All co/ctr beams. n=3 I-coil. | ||
| Diagnostic Requirements: | Full profile diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 212: Measure Error Field Threshold and Beta Limit for the ITER Baseline | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | R J La Haye | ITPA Joint Experiment : | Yes |
| Description: | Experiments have identified a greatly increased sensitivity of low beta H modes to error fields, and a low beta-N limit at low NB torque. In particularly these dictate exacting requirements for error field correction (significantly more severe than the previously considered worst situation of low density Ohmic plasmas), and a betan limit on DIII-D that is only slightly higher than ITER's baseline reference. However, these studies were performed with elevated q95~4.4. Therefore it is vital to assess these effects at q95~3.1, where the location of the q=2 surface may make the plasma even more sensitive. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dial up ITER-like baseline scenarios at zero NB torque and the ITER baseline betaN. Ramp error fields to measure threshold. Separately ramp betaN with optimal error correction to test betaN limit. | ||
| Background: | This is a critical issue to determining the scale of the challenge for ITER error correction and number of coil sets required. | ||
| Resource Requirements: | balanced beams, I coils | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 213: Test pedestal current models in ELMy H-mode discharges | ||
| Name: | xu xxu@llnl.gov | Affiliation: | LLNL |
| Research Area: | Pedestal Structure | Presentation time: | Requested |
| Co-Author(s): | Dan Thomas, Chris Holcomb, Rich Groebner | ITPA Joint Experiment : | No |
| Description: | Measure pedestal toroidal/parallel current profiles of H-modes using Li-BEAM and MSE and test the inductive current model vs neoclassical bootstrap current model. This is in the Pedestal Structure area because pedestal current drives edge kink/peeling mode, which leads to type-I ELMs. The proposal experiments will test a fundamental question for pedestal stability: current is calculated correctly using only bootstrap current model. Does the pedestal current drop at an ELM crash and then rebuild as predicted only by the bootstrap current or Is there a significant inductive parallel electric field that also helps rebuild the current. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop discharges with large ELMs and long ELM periods. The target discharges need to have pedestal current as high as possible so that we get the best possible edge current density measurements. The ITER baseline demo discharges are probably suitable. Measure the pedestal current density evolution during recovery from the ELMs. Use conditional averaging, if necessary, to improve the signal to noise of the measurements. | ||
| Background: | MAST MSE measurements show that toroidal current j_phi is not affected by the occurrence of a type-I ELM. BOUT++ simulations show that it is due to the compensating inductive current from a zonal field generated during an ELM event and relax slowly in-between ELMs in low collisionality regime. However, bootstrap current follows the pedestal pressure profile, collapsing and recovering. | ||
| Resource Requirements: | Standard beam-heated discharges in H-mode regime. | ||
| Diagnostic Requirements: | Li-BEAM and MSE operative. | ||
| Analysis Requirements: | Analyze LIBEAM and MSE data to obtain temporal evolution of pedestal current density. Obtain time history of kinetic EFITS to obtain best calculations of pedestal bootstrap current evolution during ELM recovery. Compare the calculated bootstrap current evolution to the measured current density evolution. Also consider using dpsi/dt of the kinetic equilibria to obtain a measurement of E-parallel by the Forest method. Determine if E_parallel varies significantly during ELM cycle. | ||
| Other Requirements: | |||
| Title | 214: Measure Error Field Threshold and Beta Limit for the ITER Baseline (Dup 212) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Requested |
| Co-Author(s): | Rob La Haye | ITPA Joint Experiment : | Yes |
| Description: | Experiments have identified a greatly increased sensitivity of low beta H modes to error fields, and a low beta-N limit at low NB torque. In particularly these dictate exacting requirements for error field correction (significantly more severe than the previously considered worst situation of low density Ohmic plasmas), and a betan limit on DIII-D that is only slightly higher than ITER's baseline reference. However, these studies were performed with elevated q95~4.4. Therefore it is vital to assess these effects at q95~3.1, where the location of the q=2 surface may make the plasma even more sensitive. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dial up ITER-like baseline scenarios at zero NB torque and the ITER baseline betaN. Ramp error fields to measure threshold. Separately ramp betaN with optimal error correction to test betaN limit. | ||
| Background: | This is a critical issue to determining the scale of the challenge for ITER error correction and number of coil sets required. | ||
| Resource Requirements: | balanced beams, I coils | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 215: Measure Error Field Threshold and Beta Limit for the ITER Baseline (Dup 212) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | Rob La Haye | ITPA Joint Experiment : | No |
| Description: | Experiments have identified a greatly increased sensitivity of low beta H modes to error fields, and a low beta-N limit at low NB torque. In particularly these dictate exacting requirements for error field correction (significantly more severe than the previously considered worst situation of low density Ohmic plasmas), and a betan limit on DIII-D that is only slightly higher than ITER's baseline reference. However, these studies were performed with elevated q95~4.4. Therefore it is vital to assess these effects at q95~3.1, where the location of the q=2 surface may make the plasma even more sensitive. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dial up ITER-like baseline scenarios at zero NB torque and the ITER baseline betaN. Ramp error fields to measure threshold. Separately ramp betaN with optimal error correction to test betaN limit. | ||
| Background: | This is a critical issue to determining the scale of the challenge for ITER error correction and number of coil sets required. | ||
| Resource Requirements: | balanced beams, I coils | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 216: Differentiate formation of edge heat and particle transport barriers using I-mode | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | L. Schmitz, T. Rhodes, P. Gohil, J. Rice | ITPA Joint Experiment : | Yes |
| Description: | I-mode plasmas provide an unprecedented opportunity to separately study the formation of an edge energy transport barrier (at a transition to I-mode) and particle transport barrier (at an L-H tranition). This should allow us to determine which fluctuations are critically associated with which transport channel. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Separate edge energy and particle barrier formation by operating plasmas with an I-mode phase, followed by H-mode. Obtain I-mode plasmas by operating just below L-H threshold in plasmas with "wrong" grad-B drift direction (grad-B away from X-point). | ||
| Background: | The formation of the edge H-mode barrier on DIII-D sees the simultaneous formation of particle and energy transport barriers, However, I-mode plasmas on C-Mod (improved L-mode plasmas), with a subsequent H-mode phase exhibit a double transport barrier formation - energy transport is reduced in the I-mode phase, while particle transport is unaffected until the conventional L-H transition occurs. This gives an unprecedented opportunity to separately study the formation of the energy and particle transport barriers using the local fluctuation diagnostics on DIII-D - diagnostics which C-MOD lacks. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | All edge turbulence measurements | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 217: Differentiate formation of edge heat and particle transport barriers using I-mode (dup 216) | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | L. Schmitz, T. Rhodes, P. Gohil, J. Rice | ITPA Joint Experiment : | Yes |
| Description: | I-mode plasmas provide an unprecedented opportunity to separately study the formation of an edge energy transport barrier (at a transition to I-mode) and particle transport barrier (at an L-H tranition). This should allow us to determine which fluctuations are critically associated with which transport channel. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Separate edge energy and particle barrier formation by operating plasmas with an I-mode phase, followed by H-mode. Obtain I-mode plasmas by operating just below L-H threshold in plasmas with "wrong" grad-B drift direction (grad-B away from X-point). | ||
| Background: | The formation of the edge H-mode barrier on DIII-D sees the simultaneous formation of particle and energy transport barriers, However, I-mode plasmas on C-Mod (improved L-mode plasmas), with a subsequent H-mode phase exhibit a double transport barrier formation - energy transport is reduced in the I-mode phase, while particle transport is unaffected until the conventional L-H transition occurs. This gives an unprecedented opportunity to separately study the formation of the energy and particle transport barriers using the local fluctuation diagnostics on DIII-D - diagnostics which C-MOD lacks. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | All edge turbulence measurements | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 218: Modulated Ion Thermal and Momentum Transport Studies Using Off-Axis NBI | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | The goal of the experiment is to take advantage of the off-axis NBI capabilty to do modulated heat and momentum transport studies. TRANSP calculations show that the off-axis beams deposit both heat and momentum to the ions over a very broad region of the plasma (past mid-radius), compared with the typical on-axis injection, which is highly peaked. The total power and momentum flow through the edge is approximately unchanged between on and off-axis NBI. This may also contribute profile stiffness experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate between on-axis and off-axis beams at constant beam power and/or betaN, and measure heat and momentum transport coefficients as a function of betaN, to investigate beta dependence of local transport coefficients versus global confinement scaling. | ||
| Background: | Modulated momentum transport experiments have been conducted at constant power/betaN but the ability to modulate the local ion heating at fixed total heating has not been possible until now. | ||
| Resource Requirements: | At least 4 co source (150s essential), ctr sources desirable. | ||
| Diagnostic Requirements: | Full profile diagnostics, especially CER, plus FIDA and main ion CER. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 219: Rapid shutdown with large shell pellet | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | P. Parks | ITPA Joint Experiment : | No |
| Description: | Assess ability of shell pellet concept to achieve rapid shutdown with collisional RE suppression. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Shut down stable, high-energy ITER-like shape (LSN) target plasmas with injection of a single large (D = 1 cm), boron powder-filled shell pellet. Increase target plasma thermal energy as necessary to ensure pellet shell breakup in plasma before thermal quench. | ||
| Background: | Rapid shutdown with collisional RE suppression is highly desired to avoid RE formation during disruptions in ITER. To date, only 20% of the required large density ncrit has been achieved (using multi-valve He MGI or using D2 SPI). Even if 100% ncrit could be achieved in ITER with MGI, the resulting gas reprocessing time of several days is deemed unacceptable. Shell pellets (thin sphere filled with powder) offer a possible alternative because the injected material (Be in the case of ITER) will coat the walls and not clog the pumping system. Proof-of-principle small (OD = 2 mm) shell pellet experiments were performed on DIII-D in 2008 and first large (OD = 1 cm) shell pellet experiments were performed in 2009. The large shell pellets did not burn through because the ablation rate of the 0.4 mm polystyrene shell was about 4x slower than expected. Now, we have made shell pellets with 0.1 mm polystyrene shells, which we expect to burn through. | ||
| Resource Requirements: | 1 run day. 4 gyrotrons, 6 beams. Large shell pellet launcher. | ||
| Diagnostic Requirements: | Fast camera, SPRED, SXR, interferometers. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | none | ||
| Title | 220: Modulated Ion Thermal and Momentum Transport Studies Using Off-Axis NBI (Dup 218) | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of the experiment is to take advantage of the off-axis NBI capabilty to do modulated heat and momentum transport studies. TRANSP calculations show that the off-axis beams deposit both heat and momentum to the ions over a very broad region of the plasma (past mid-radius), compared with the typical on-axis injection, which is highly peaked. The total power and momentum flow through the edge is approximately unchanged between on and off-axis NBI. This may also contribute profile stiffness experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate between on-axis and off-axis beams at constant beam power and/or betaN, and measure heat and momentum transport coefficients as a function of betaN, to investigate beta dependence of local transport coefficients versus global confinement scaling. | ||
| Background: | Modulated momentum transport experiments have been conducted at constant power/betaN but the ability to modulate the local ion heating at fixed total heating has not been possible until now. | ||
| Resource Requirements: | At least 4 co source (150s essential), ctr sources desirable. | ||
| Diagnostic Requirements: | Full profile diagnostics, especially CER, plus FIDA and main ion CER. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 221: Effect of different Z injection on RE plateau | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | P. Parks, T. Jernigan, N. Commaux | ITPA Joint Experiment : | No |
| Description: | Attempt to reduce RE current using large shell pellet injection and MGI gas injection of different species | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create large >100 kA RE beam by using IWL low density target plasma shut down with small Ar pellet. Then, fire large boron-filled pellet into RE beam and look for drop in RE current. Also, try Ne and Ar MGI. | ||
| Background: | Reliable, large > 100 kA RE beams are now created routinely in DIII-D using IWL low density targets shut down with small Ar pellets. These RE beams are remarkably resilient to external impurity injection ā?? small shell pellets, He MGI, and D2 SPI have all been fired at the RE plateau without any apparent effect. Only Ne MGI (one shot) appears to have had a measurable effect on the RE current. To study the effect of different species on the RE beam braking, we propose to fire boron (via large shell pellet), and Ne and Ar (via Medusa valve), and Ne (via shattered pellet, if available) into the RE plateau. Main diagnosics are the plasma current, visible spectroscopy, HXR emission, and SXR emission. | ||
| Resource Requirements: | 1/2 run day. Small argon pellet injector. Large shell pellet launcher. Medusa MGI system. SPI system with Ne, if available. | ||
| Diagnostic Requirements: | Fast camera, SPRED, SXR, MDS, CER, interferometers. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | none | ||
| Title | 222: Characterization of disruption heat loads for cross-machine comparison | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Requested |
| Co-Author(s): | R. Pitts, C. Lasnier, A. McLean | ITPA Joint Experiment : | Yes |
| Description: | Measure heat deposition time and footprint for hot VDE, beta limit, and MGI mitigated disruptions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Set up fast diagnostics, especially IR cameras, for accurate heat load measurements. Create intentional downward hot VDE (2 shots) and then intentional high-power (beta limit) disruption. These serve as worst-case scenarios for DIII-D heat loads. Then compare these with heat loads during mitigated (neon MGI) fast shutdowns. If time is available, also get good data for intermediate heat load cases (current limit disruptions, density limit disruptions, etc). Good spectroscopy data is desirable to attempt comparison of measured heat loads and first wall material erosion. | ||
| Background: | Understanding of disruptions heat loads is still largely empirical and cross-machine comparisons are few, except in the case of TQ and CQ time, where extensive comparisons have been done. It is desired to create a cross-machine database of main chamber heat loads and footprint from different types of disruptions and fast shutdowns to help make first wall material decisions in future devices. This has been made into a high-priority ITPA task for 2011 (DSOL-24). | ||
| Resource Requirements: | 1/2 run day. 6 beams, 4 gyrotrons. | ||
| Diagnostic Requirements: | IR fast cameras (aimed at lower divertor and at main chamber, if possible), fast visible cameras (aimed at main chamber to the extent possible), SPRED, SXR, interferometers, fast filterscopes, CER spectrometers. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | none | ||
| Title | 223: Low-Z collisional suppression of startup REs | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | T. Jernigan, N. Commaux | ITPA Joint Experiment : | No |
| Description: | Obtain data on effectiveness of low-Z impurities for collisional suppression of REs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create RE seeded discharges by starting with low density, IWL, and full gyrotron heating. Shut down early (while startup REs still exist in the discharge) with varying levels (1-6 valves) of low-Z (He or H2) MGI. | ||
| Background: | Presently, no tokamak sees disruption REs after shutdown with massive low-Z MGI. Nevertheless, present wisdom is that ITER will still have large REs even with low-Z MGI shutdown because of its ever-present RE seed and large CQ avalanche gain and because present experiments can only reach 20% ncrit (the theoretically required density for collisional RE suppression). To test if 100% ncrit is really necessary for collisional RE suppression, we propose to do very early (t ~ 500 ms) shutdown of gyrotron-heated targets, while startup REs still persist in the target plasma. By varying the amount of low-Z impurities injected, we hope to see the extent to which massive low-Z impurity injection damps the CQ RE avalanche. | ||
| Resource Requirements: | 1 run day. 4 gyrotrons, medusa 6-valve MGI system | ||
| Diagnostic Requirements: | Fast camera, SPRED, SXR, interferometers, BGO scintillators, fplastic. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | none | ||
| Title | 224: Use of magnetic feedback to avoid locking of a rotating mode | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | A. Garofalo, Y. In, H. Reimerdes, T. Strait | ITPA Joint Experiment : | No |
| Description: | - NTMs onset and mode locking is anticipated to occur at various circumstances from low plasma density to beta-collapse in the ITER operational scenarios. The avoidance of tearing mode locking is a critical issue in every step of discharge developments.
- Here, it is proposed to demonstrate a unique usage of internal coils for avoiding mode locking, if successful, this would present an important driving force for ITER to put in such coils. - As we reported in YER in 2010 and 2008, application of feedback can robustly synchronized tearing mode with a rotating external field in the limited piggyback experiments. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | - Apply the feedback with relatively slow bandwidth ( tau~ 10ms) and to decelerate the NTM rotation from a few kHz down to the order of tens Hz and sustain the mode rotation.
- In these previous piggyback experiments, we decelerated to 15 Hz ( tau= 40 ms) and 40 Hz (tau= 10 ms) and avoided the mode locking when the mode was rotating in the Ip direction. Goal - To answer several critical issues, - Capability to controlling the rotation frequency - Applicability range of betan - Minimum mode amplitude for sustaining the NTM - Process of mode synchronization and desynchronization (if occurs) - To assess the applicability to the coil and plasma surface separation like in ITER | ||
| Background: | |||
| Resource Requirements: | All the available NBI sources, | ||
| Diagnostic Requirements: | standard profile diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 225: Investigate formation of edge co-rotation layer at L-H transition | ||
| Name: | Muller mullersh@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | J.A. Boedo, K.H. Burrell, J.S. deGrassie, R.A. Moyer, D.L. Rudakov, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | Investigate the timescale and conditions under which the edge co-rotation layer forms after the L-H transition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce identical L-H transitions on a shot-to-shot basis and vary the probe plunge around the critical interval of -25 ms to +50 ms of the expected L-H transition, in which the layer is known to form. This sampling will be partially random due to the jitter in the L-H transition of about +/-25 ms in these condition, but a converged characterization of the profiles as a function of (r,t) in this critical phase can be obtained in about 10 good shots. Using a beam blip directly after the probe plunge on every shot, the purely intrinsic evolution of the core rotation profile in this critical period will also be obtained. | ||
| Background: | A strong edge co-rotation layer has been found with the midplane probe, which, in low-power LSN H-modes, forms within 50 ms after the L-H transition independently of the injected torque. The existence of the layer shows a clear correlation with core intrinsic rotation development. The experimental strategy of performing probe-plunge and beam-blip combos on a shot-to-shot basis at different time delays after the L-H transition turned out to be a major success, since it yielded unperturbed and directly interpretable information on the action of the intrinsic torques. The success of the first experiment suggests to exploit the developed experimental strategy further for a better characterization of the processes leading to edge-rotation-layer formation and their dependences on the plasma conditions. | ||
| Resource Requirements: | Midplane probe with Mach head, 2-3 gyrotrons, 1 co beam, 1 ctr beam, CER beams, all cryopumps | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 226: Confinement Enhancement via Resonant Radial Field Amplification of the Geodesic Acoustic Mode | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | S. Cowley [1], A. Garofalo [2], K. Hallatschek [3], C. Holland [4], G. Jackson [2], Z. Yan [5]
[1] Imperial College, London; CCFE, Abingdon, UK [2] General Atomics [3] Max Planck-IPP, Garching, Germany [4] UCSD [5] U. Wisconsin |
ITPA Joint Experiment : | No |
| Description: | Amplify the naturally-occuring Geodesic Acoustic Mode, to control and suppress turbulence and associated transport near the plasma separatrix while maintaining a non-ELMing L-mode-like plasma edge. The goal is to achieve enhanced global energy confinement via the resulting turbulence suppression. The experiment would exploit the high-frequency radial B-field capability of the DIII-D I-Coils, and measure the resulting turbulence and GAM response to this resonant radial field perturbation at the GAM frequency. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish plasma conditions were the GAM has been clearly observed and has a relatively large "natural" amplitude: Upper-Single-Null L-mode plasmas at moderate power (2 sources, co-injected). q95-scaling experiments [McKee, PPCF-2006] demonstrate that the GAM oscillation amplitude is higher in higher q95-discharges. Thus, establish a high q95 condition, e.g., Ip=0.8 MA, B_t=2.0 MA, q95~6.5 (119531). The I-Coil will be configured in an n=0, m=1 configuration (upper and lower coils 180Āŗ out of phase) and run near 15 kHz (SPAs operate at up to 100 kHz, so this is technically feasible).
Establish basic plasma conditions and benchmark GAM parameters with the 2D 8x8 BES array and toroidally-displaced DBS systems. Turn on I-Coil in above configuration at ~15 kHz, near the known GAM frequency range. Scan frequency in the expected GAM range (14-18 kHz). The radial field produced by the I-coil at these frequencies is relatively low: it is predicted to be of the order Br < 1 Gauss at this high frequency, based on measurements by G. Jackson. The relatively low field results from image currents in the wall at this frequency. It will need to be experimentally assessed whether this field is adequate to interact with and perturb or resonantly amplify the GAM. Turbulence dynamics and the GAM amplitude will be experimentally measured. Success would be evident via GAM amplification, turbulence suppression, and enhanced confinement. Whether or not the GAM amplification is successful, this resulting fluctuation data set will provide a very quality set of wide-field fluctuation measurements with the 8x8 BES array that can be used to study the nonlinear properties and dynamics of turbulence. In particular, a newly developed basis-operator bispectral analysis method [Baver, Terry, PoP-2009] will be applied to this data as well as to GYRO simulations of these plasmas. This may help elucidate why the simulations systematically have difficulty quantitatively reproducing turbulence and transport measurements in these plasma conditions and locations, an issue that will need to be resolved to fully validate transport simulations. | ||
| Background: | The Geodesic Acoustic Mode (GAM), an electrostatic, coherent, radially-sheared zonal-flow oscillation, has been observed in DIII-D in the outer radial region of L-mode discharges. High-frequency poloidal velocity analysis of BES turbulence measurements have provided a detailed characterization of the GAM structure, which is also observed with the Doppler Backscattering diagnostic. The electrostatic potential and corresponding radial electric field is radially localized with well-defined k_radial, but is poloidally and azimuthally symmetric (m=0, n=0). Theoretically, it is predicted to have an m=1, n=0 pressure sideband as a result of the non-uniform ExB flow on a flux surface, which has been observed in some experiments (AUG, HL-2A). The pressure oscillation, peaking at the "top" and "bottom" of the plasma, relaxes via a radial drift current which gives rise to the very coherent GAM oscillation under the right plasma conditions.
Typically, the GAM is observed near 15 kHz, consistent with its predicted frequency of omega=c_s/R, and peaks spatially near r/a = 0.85-0.98. The GAM can shear turbulence, and thus reduce the saturated level of turbulence and resulting transport. It interacts nonlinearly with the turbulence, driving a forward transfer of internal energy to higher frequency/wavenumber [C. Holland, PoP (2007)]. Shearing rate estimates from the poloidal flow shear of the GAM, obtained from the time-varying radial gradient of poloidal velocity, suggests that its shearing rate is comparable to the turbulence decorrelation rate and thus should play a role in turbulence saturation and decorrelation. If it were possible to amplify the GAM, it might be feasible to control and reduce turbulence and resulting transport, thus improving energy confinement. The high frequency I-Coils and audio amplifiers implemented on DIII-D provide a possible mechanism to amplify the GAM. The concept would be to generate an n=0, m=1 (odd-parity) radial magnetic field perturbation at the GAM frequency with the I-Coils. It has been proposed (S. Cowley, Imperial College) that this field may interact with and amplify the GAM by creating a small pressure perturbation through equilibrium shape modulations, thus enhancing the pressure sideband by resonantly "squeezing" the flux surface at the GAM frequency. It is also possible that the radial field will interact with or amplify the radial drift current that creates the periodic pressure relaxation. | ||
| Resource Requirements: | I-coils set up in n=0, m=1 (odd-parity) configuration, connected to SPAs operating at high frequency (14-20 kHz). 2 NBI, USN plasma. | ||
| Diagnostic Requirements: | BES (8x8 array configuration), DBS-5, DBS-8, CECE, Reciprocating probe with Reynolds Stress head | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 227: Test physics of rotational screening | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | W. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to test a fundamental assumption at the basis of many works on modeling of RMP screening and induced transport [eg recent papers by Nardon, Waelbroeck, Beucolet], namely that the rotational screening of RMPs results from the motion of the electron fluid across the field lines at the resonant surfaces [Yu Q. and Gunter S. 2009 Nucl. Fusion 49 062001]. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose to investigate the plasma flow requirements for a nonaxisymmetric magnetic field to penetrate and open a magnetic island, by thouroghly documenting the plasma profiles in plasmas like DIII-D discharge #138611.
In this and in many similar discharges, we observed a locked plasma state that was sustained for several energy confinement times in conditions of high beta and high confinement. These conditions offer the unique opportunity to carry out high quality profile measurements in presence of a locked island. - Radial jog techniques can be applied to improve the radial resolution of the measured profiles. - Additional RMPs could be applied and rotated to change the toroidal phase of the island. - The new capability to directly measure the main ion rotation would add tremendous new insight into what are the flow requirements for RMP shielding. | ||
| Background: | Recent theoretical work describes the rotational screening of RMPs as resulting from the motion of the electron fluid across the field lines at the resonant surfaces [Yu Q. and Gu Ģ?nter S. 2009 Nucl. Fusion 49 062001]. The velocity of the electron fluid across the field lines can be written as Vā?„e = VE + Vā??e, where VE = Er/B is the E Ć? B drift velocity and Vā??e = peā?²/eneB is the electron diamagnetic drift velocity.
This fundamental hypothesis is at the basis of many works on modeling of RMP screening and induced transport [eg recent papers by Nardon, Waelbroeck, Beucolet]. Recent DIII-D discharges (138611 and similar) showed a locked plasma state that was sustained for several energy confinement times in conditions of high beta and high confinement. These QH-mode low rotation discharges offer a unique opportunity to carry out high quality profile measurements in presence of a locked island. | ||
| Resource Requirements: | Same as discharge 138611. | ||
| Diagnostic Requirements: | Main ion rotation measurements. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 228: Stability boundary of Energetic Particle (EP)-driven RFA/RWM at low rotation with q_min ~2: | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Requested |
| Co-Author(s): | Go Matsunaga, John deGrassie, Bill Heidbrink, Yongkyoon In, Yueqiang Liu, H, Reimerdes, Ted Strait | ITPA Joint Experiment : | Yes |
| Description: | - We have carried out the initial study of the off-axis-fishbone-driven RWM in low rotation plasmas in FY2010 experiments
- This exploration have accelerated our understanding of the EP contribution to RWM stabilization through direct coupling as well as the non-ambipolar rotation drop caused by EP losses. - In this proposal, we would like to develop quantitative analysis and provide the prediction to the alpha particle influence on the EP-driven RFA/RWM in ITER and tokamak-oriented reactors - This operational regime is useful since the EP effects is well separated from the thermal contribution with higher precession drift frequency and less sensitive to collision term approximation. - EP diagnostics should clarify the EP velocity /space distribution sensitivity: - The rotation drop due to EP losses during RWM formation will be documented for the analysis of the impact to RWM stability |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Approach / Target:
- q_min~2 with off axis NBI, with ECCD NTM suppression. - including co/ctr NBI to characterize the precession drift frequency sensitivity Hypothesis: - EP velocity / spatial distribution plays the key role with paradigm given in APS-invited talk Physics outcome - OFB is external kink or internal kink? - Impact of simultaneous EP loss and transport-related issue :rotation drop? - Mode distortion and EP loss relation Code for EP M3D-K (Gu-Yong Fu), Spiral (G. Kramer) and MARS-K | ||
| Background: | |||
| Resource Requirements: | All the available NBI sources, 4 gyrotrons for ECCD | ||
| Diagnostic Requirements: | standard profile diagnostics, EP diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 229: First-principles Model-based Current Profile Control during the Ramp-up Phase in DIII-D | ||
| Name: | Schuster schuster@lehigh.edu | Affiliation: | Lehigh U |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Requested |
| Co-Author(s): | Yongsheng Ou - Shenzhen Institute of Advanced Technology, China
Chao Xu ā?? Zhejiang University, China John Ferron, Tim Luce, Mike Walker, Dave Humphreys ā?? General Atomics Didier Moreau, Didier Mazon ā?? CEA, France |
ITPA Joint Experiment : | No |
| Description: | Establishing a suitable current profile has been demonstrated to be a key condition for the achievement of advanced scenarios with improved confinement and possible steady-state operation. The approach at DIII-D focuses on creating the desired current profile during the plasma current ramp-up and early flattop phases with the aim of maintaining this target profile during the subsequent phases of the discharge. The controller used for the q evolution during the ramp-up phase is presently a simple PI (proportional-integral) algorithm with empirically determined gains. The q profile is obtained in real time from a complete equilibrium reconstruction using data from the Motional Stark Effect (MSE) diagnostic. The controller requests a power level to the actuator (electron cyclotron heating (ECH) or neutral beam heating (NBH)) which is equal to preprogrammed feed-forward value plus the error in q times a PI gain. Present limitations of this controller (oscillations and instability), the high dimensionality of the problem, and the strong coupling between the different variables describing the dynamics of the current profile of the plasma motivates the design of a model-based, multi-variable controller that takes into account the dynamics of the q response to the different actuators. The Advanced Scenario thrust is interested in developing a model based controller to be used in forming desirable current profiles during the plasma current ramp-up. Some characteristics of the problem that make it difficult are the limited actuator power, the need to avoid unstable MHD regimes, and the significant nonlinearity of the problem.
The objective of this experiment is to implement first-principles model-based controllers developed for the regulation of the q profile evolution during the early phase of the discharge, including ramp-up and beginning of the flattop, with the ultimate goal of achieving a desired target profile at some time during the first part of the flattop phase. This experiment intends to continue the tests performed during the last experimental campaign. The 2009 experiments allowed for the identification of some spatial coordinate conflicts between diagnostics and PCS algorithms. In addition, since the 2009 experiments the magnetic diffusion equations has been implemented in simserver simulations. This provides a more effective tool to debug the implementation in the PCS of the control algorithms before experimental testing. The experiment will evaluate the performance of optimal closed-loop controllers designed based on first-principles models. It is expected that closed-loop controllers will add robustness to previously tested open-loop controllers. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Open-loop optimal control laws will be expressed as time trajectories for the actuators: total plasma current, average plasma density, and non-inductive current drive (NBI, ECH) power. The closed-loop controller will regulate in real-time these three actuators based on real-time measurements of the poloidal flux or q profiles. We will assess the ability of the combined open-loop and closed-loop controllers to drive the current profile from an initial condition different from (but close to) the nominal one to a specific target profile. Different initial and target profiles will be considered mainly in L-mode. The first-principles model-based current profile control experiment will require half a day to possibly one day. However, it is absolutely necessary to dedicate at least two short (2 hours) preliminary sessions to debugging and testing of the control algorithms implemented in the PCS. | ||
| Background: | The control group at Lehigh University (LU) headed by Prof. Eugenio Schuster has been working on this problem for more than four years. A preliminary first-principle control-oriented model of current profile evolution in response to auxiliary heating and current drive systems (NBI, EC) and electric field due to induction was developed for the plasma current ramp-up and early-flattop phases [1]. Optimal open-loop control schemes were developed based on the simplified control-oriented model [2, 3]. These algorithms predict the open-loop (or feedforward) actuator waveforms that are necessary to drive the plasma from a specific poloidal flux initial profile to a predefined final profile during the current ramp-up. Data obtained from the 2008 1/2ā??day experiment showed: 1- qualitative agreement with the q profile evolution predicted by the simplified model, 2- actuators constraints were correctly taken into account during the control synthesis, 3- success in achieving monotonic target profiles with positive and near-zero shear near the axis. Nevertheless, reversed shear target profiles could not be achieved. This motivated the use of CORSICA (full predictive model instead of simplified control-oriented model) for the development of optimal open-loop controllers and further refinement of the simplified control-oriented model. A reduced-order first-principles model is obtained from the original simplified control-oriented infinite-dimensional model and combined with Optimal Control and Robust Control theory to synthesize closed-loop controllers [4, 5]. Initial testing of these algorithms was carried out in 2009, identifying some implementation issues. Based on initial results obtained in simulation studies, it is anticipated that the scheme can play an important role in experiments at the DIII-D tokamak. The development of model-based current profile controllers aims at saving long trial-and-error periods of time currently spent by fusion experimentalists trying to manually adjust the time evolutions of the actuators to achieve the desired current profile at some pre-specified time during the early flattop phase.
[1] Y. Ou et al., ā??Towards Model-based Current Profile Control at DIII-D,ā?? Fusion Engineering and Design 82 (2007) 1153ā??1160. [2] Y. Ou et al., ā??Extremum-Seeking Open-Loop Optimal Control of Plasma Current Profile at the DIII-D Tokamak,ā?? Plasma Physics and Controlled Fusion, 50 (2008) 115001. [3] C. Xu et al., ā??Ramp-Up Phase Current Profile Control of Tokamak Plasmas via Nonlinear Programming,ā?? IEEE Trans. on Plasma Science, vol.38, no.2, p.163, 2010. [4] Y. Ou et al., ā??Optimal Tracking Control of Current Profile in Tokamaks,ā?? IEEE Transactions on Control Systems Technology, in press. [5] Y. Ou et al., ā??Robust Control Design for the Poloidal Magnetic Flux Profile Evolution in the Presence of Model Uncertainties,ā?? IEEE Trans. on Plasma Science, vol.38, no.3, p.375, 2010. | ||
| Resource Requirements: | Machine time: At least two 2-hour evening sessions + 1/2 day experiment. | ||
| Diagnostic Requirements: | Core and tangential Thomson, CER, CO2, magnetics, MSE, ECH diagnostics, a reasonable set of fast ion instability diagnostics (UF interferometers, FIR scattering, ECE at 500 kHz, fast magnetics with fast delay set in the current ramp), FIDA. For closed-loop experiment real-time magnetic measurements and equilibrium reconstruction including the q-profile (EFIT and RTEFIT with MSE) are essential, and real-time measurements of the ion and electron temperature profiles, as well as line-averaged plasma density or density profile are required. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 230: Fuelling efficiency and retention for pellets with shallow penetration | ||
| Name: | POLEVOI none | Affiliation: | ITER Organization |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | Larry BAYLOR | ITPA Joint Experiment : | No |
| Description: | Systematic analysis of fuelling efficiency and retention for HFS and LHS pellets with shallow penetration in L,H mode and Ohmic plasmas | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This task does not require special experiments. It is necessary in all phases of all scenarios inject a few small pellets with shallow penetration (r/a > 0.8) to create a database for scaling for fuelling efficiency dN/Npel, and retention time as a function of local (n,T,q, ā?¦) and global (B,Ip,..) parameters. | ||
| Background: | Gas puffing cannot provide sufficient fuelling in ITER. Pellet injection is considered as the a main tool for the core fuelling. The Pellet Injection System (PIS) can deliver particles to the outer ļ??ļ? ļ?¾ 20% of the minor radius. Thus, PIS just supports the pedestal height at sufficient level. Unfortunately, theory based models are not applicable in this area even for description of the heat transport. Thus, for ITER predictions we need to extrapolate the experimental data. Fortunately, sources of particles injected by pellets are much more certain than particle sources from the gas puffing since we know and can easily measure dN and retention time for such particles. Therefore, it is possible to do regression analysis to find the retention time and fuelling efficiency as a function of local parameters (n, T, ļ??, pellet and plasma atomic mass, etc) and global (B,Ip) parameters in H-,L- and Ohmic modes in a range of parameters expected in ITER. That is necessary for assessment of fuelling needs and accessible ITER densitiess for H, He, DD and DT phases of ITER operation for ITER operation planning.
Example: simple scaling with current, H-factor and isotopic effect predicts that support of the density required for full power NBI in L-mode is 3 times more demanding for PIS than for the basic DT operation. Example: Possible residual fuelling from LFS pellet for pace making affects possibility of the independent density control. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 231: Fuelling efficiency and retention for pellets with shallow penetration | ||
| Name: | POLEVOI none | Affiliation: | ITER Organization |
| Research Area: | General IP | Presentation time: | Requested |
| Co-Author(s): | Larry BAYLOR | ITPA Joint Experiment : | No |
| Description: | Systematic analysis of fuelling efficiency and retention for HFS and LHS pellets with shallow penetration in L,H mode and Ohmic plasmas | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This task does not require special experiments. It is necessary in all phases of all scenarios inject a few small pellets with shallow penetration (r/a > 0.8) to create a database for scaling for fuelling efficiency dN/Npel, and retention time as a function of local (n,T,q, ā?¦) and global (B,Ip,..) parameters. | ||
| Background: | Gas puffing cannot provide sufficient fuelling in ITER. Pellet injection is considered as the a main tool for the core fuelling. The Pellet Injection System (PIS) can deliver particles to the outer ļ??ļ? ļ?¾ 20% of the minor radius. Thus, PIS just supports the pedestal height at sufficient level. Unfortunately, theory based models are not applicable in this area even for description of the heat transport. Thus, for ITER predictions we need to extrapolate the experimental data. Fortunately, sources of particles injected by pellets are much more certain than particle sources from the gas puffing since we know and can easily measure dN and retention time for such particles. Therefore, it is possible to do regression analysis to find the retention time and fuelling efficiency as a function of local parameters (n, T, ļ??, pellet and plasma atomic mass, etc) and global (B,Ip) parameters in H-,L- and Ohmic modes in a range of parameters expected in ITER. That is necessary for assessment of fuelling needs and accessible ITER densitiess for H, He, DD and DT phases of ITER operation for ITER operation planning.
Example: simple scaling with current, H-factor and isotopic effect predicts that support of the density required for full power NBI in L-mode is 3 times more demanding for PIS than for the basic DT operation. Example: Possible residual fuelling from LFS pellet for pace making affects possibility of the independent density control. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 232: Low torque/rotation ITER baseline scenario | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | Develop a low torque/rotation version of the ITER baseline scenario using ECH and/or balanced NBI. This is a critical "missing element" in the ITER demonstration discharges operated to date in terms of matching expected ITER operating conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Already have baseline scenario plasmas with about 50% RF (ECH) power (~2 MW). With ~4 MW of ECH power available in 2011, should be able to generate baseline plasmas which are ECH dominated, with reduced torque input. Remaining NBI input (if any) can be using balanced beams. May require development of locked mode error field compensation | ||
| Background: | Very limited data with reduced rotation in ITER baseline scenario plasmas. | ||
| Resource Requirements: | Full gyrotron set. Counter beams. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 233: Low torque/rotation ITER baseline scenario development (Dup 232) | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | Develop a low torque/rotation version of the ITER baseline scenario using ECH and/or balanced NBI. This is a critical "missing element" in the ITER demonstration discharges operated to date in terms of matching expected ITER operating conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Already have baseline scenario plasmas with about 50% RF (ECH) power (~2 MW). With ~4 MW of ECH power available in 2011, should be able to generate baseline plasmas which are ECH dominated, with reduced torque input. Remaining NBI input (if any) can be using balanced beams. May require development of locked mode error field compensation | ||
| Background: | Very limited data with reduced rotation in ITER baseline scenario plasmas. | ||
| Resource Requirements: | Full gyrotron set. Counter beams. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 234: Improvement of Dynamic Error Field Correction(DEFC ) | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | Yongkyoon In, Piero Martin, Paolo Piovesan, L. Piron, E. Strait, and RWM Physics group | ITPA Joint Experiment : | No |
| Description: | For "3D effects" exploration and SSI target development, the improvement of DEFC should be very helpful to provide better error field correction.
Goal: (1st step), To Minimize the amplitude of n=1 RFA, simultaneously minimizing undesirable n=2,3 components (2nd step), To include the minimization of undesirable poloidal m-components. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1st step:
- Minimization of toroidal phase mismatching due to the rotation and finite betan between uncorrected error field and the applied correction - At present, in the process of dynamic error field correction, assuming the mode rigidity, we simply try to reduce the RFA without paying attention to the phase shift between uncorrected error field and the applied correction - As its consequence, we may be increasing n = 2, 3 components as well as undesirable poloidal m-components. - Prepare the PCS logic with the toroidal phase shift explicit including preprogrammed wave form (new F-matrix), rather than specifying three coil currents 2nd step: - Poloidal m components to be more precise by upper/lower independent operation with similar type of new F-matrix. Depending upon preliminary survey, we may need more audio amplifier purchase | ||
| Background: | Target
The suppression of RFA includes the fast transiently-excited RFA bubbles driven by various MHDs, like peeling-mode mode well below no-wall limit. This can be achieved with applying simultaneously the AC compensation developed by RFX group. - transiently-excited RFA/RMP - peeling-mode-driven RFA condition | ||
| Resource Requirements: | All the available NBI sources, 4 gyrotrons for ECCD
PCS logic programming. In 2nd step : after preliminary survey, we may need more audio amplifier purchase | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 235: Application of improved DEFC feedback in high betan SSI plasmas | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | Yongkyoon In, Piero Martin, Paolo Piovesan, L. Piron, E. Strait, and RWM Physics group | ITPA Joint Experiment : | No |
| Description: | For successful SSI target development, the improvement of DEFC should be very helpful to provide better error field correction.
Goal: (1st step), To Minimize the amplitude of n=1 RFA simultaneously minimizing undesirable n=2,3 components (2nd step), To include the minimization of undesirable poloidal m-components. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1st step:
- Minimization of toroidal phase mismatching due to the rotation and finite betan between uncorrected error field and the applied correction - At present, in the process of dynamic error field correction, assuming the mode rigidity, we simply try to reduce the RFA without paying attention to the phase shift between uncorrected error field and the applied correction - As its consequence, we may be increasing n = 2, 3 components as well as undesirable poloidal m-components. - Prepare the PCS logic with the toroidal phase shift explicit including preprogrammed wave form (new F-matrix), rather than specifying three coil currents 2nd step: - Poloidal m components to be more precise by upper/lower independent operation with similar type of new F-matrix. Depending upon preliminary survey, we may need more audio amplifier purchase The improvement is to be developed in collaboration with "error field" category Target As we noticed in SSI discharges in FY2010, there several types of MHD-driven-RFA were excited with fast repetition rates. These RFAs were considered as possible causes leading to lower transport properties. In principle, fast DEFC can reduce the RFA amplitude. However, the reduction of the mode amplitude with less-optimized coil current easily add undesirable error field m/n components, which could create the NTM seeds or accelerate the formation of high m/n peeling-mode-driven RMPs. The suppression of RFA includes the fast transiently-excited RFA bubbles driven by various MHDs, like peeling-mode mode well below no-wall limit. This can be achieved with applying simultaneously the AC compensation developed by RFX group. | ||
| Background: | |||
| Resource Requirements: | All the available NBI sources, 4 gyrotrons for ECCD,
PCS logic programming. In 2nd step : after preliminary study, we may need more audio amplifier purchase | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 236: Data-driven Model-based Current, Rotation and Kinetic Profile Control during the Flattop Phase | ||
| Name: | Schuster schuster@lehigh.edu | Affiliation: | Lehigh U |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Requested |
| Co-Author(s): | Yongsheng Ou ā?? Shenzhen Institute of Advanced Technology, China
Chao Xu ā?? Zhejiang University, China John Ferron, Tim Luce, Mike Walker, Dave Humphreys ā?? General Atomics Didier Moreau, Didier Mazon ā?? CEA, France |
ITPA Joint Experiment : | No |
| Description: | Establishing a suitable current profile has been demonstrated to be a key condition for the achievement of advanced scenarios with improved confinement and possible steady-state operation. The current approach at DIII-D focuses on creating the desired current profile during the plasma current ramp-up and early flattop phases with the aim of maintaining this target profile during the subsequent phases of the discharge. A closed-loop controller is necessary to regulate the current and kinetic profiles around the target values during the flattop.
The objective of this experiment is to implement model-based controllers developed for the regulation of the current profile, rotation profile and temperature/pressure profile evolutions during the flattop phase of the discharge. The model-based control algorithms are synthesized based on data-driven models identified during the last campaign. The actuators are (i) co-current NBI power, (ii) counter-current NBI power, (iii) balanced NBI power, (iv) total ECCD power from all gyrotrons in a fixed off-axis current drive configuration, and (v) loop voltage. Off-axis NBI could possibly be included as an additional actuator depending on the execution of a parallel experimental proposal (Experimental identification of the plasma response to off-axis NBI ā?? D. Moreau et al.). The multivariable, model-based controllers developed within this project differ from non-model-based, empirically-tuned, PID (proportional-integral-derivative) controllers, in two distinctive aspects: 1- knowledge of the system (identified model) is incorporated during the synthesis of the controller, ii- the relationships among all input and output variables are taken into account during the synthesis of the controller. These two distinctive aspects are indeed the reasons for which improved performance is expected from advanced multivariable model-based controllers. Indeed, the strong coupling between the different physical variables involved in the plasma transport phenomenon and the high complexity of its dynamics make unavoidable the use of information of the to-be-controlled system, i.e., dynamic models, during the synthesis of plasma profile controllers. It is important to emphasize at this point that the PCS (plasma control system) at DIII-D does have infrastructure for implementing such advanced controllers. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dynamic response data was used to identify state-space dynamic models for the evolution of q, rotation and temperature/pressure profiles using subspace identification techniques. The identified modes were used for the synthesis of reduced-order controllers that exploit the time-scale separation between kinetic and magnetic variables and optimally regulate the profiles around the nominal values. The data-driven model-based current profile control experiment will require half a day to possibly one day. However, it is absolutely necessary to dedicate at least two short (2 hours) preliminary sessions to debugging and testing of the control algorithms implemented in the PCS. | ||
| Background: | A group of researchers at JET, including Didier Moreau and Didier Mazon, have been working for more than five years now on the development of model-based controllers for the regulation of an equilibrium profile during the flattop phase of the discharge. Different current and temperature gradient target profiles have been reached and sustained for several seconds at JET during the flattop current phase. The control schemes rely on the experimental identification of linearized static and dynamic response models, using lower hybrid current drive (LHCD), ion cyclotron resonance heating (ICRH) and neutral beam injection (NBI) as actuators. The controller designed based on a static response model, which finally reduces to a proportional integral regulator incorporating information of the static response of the system, has been shown effective when rapid plasma events are absent. If the controller is expected to respond to rapid transients, such as MHD phenomena, which may displace the system on a short timescale during the slow evolution of the current density profile towards its desired shape, information of the dynamic response of the system must be incorporated into the controller synthesis. Exploiting the different time scales of kinetic and magnetic variables, a dynamic model has been recently identified and used for the synthesis of a two-timescale controller at JET [1] and DIII-D [2].
The control group at Lehigh University (LU) headed by Prof. Eugenio Schuster has started working during the 2008 experimental campaign on the identification of a dynamic response model for the q profile evolution during the flattop phase [3, 4]. Further experiments were carried out during the 2009 experimental campaign [5]. A reduced-order state-space model obtained from data using subspace identification techniques can be combined with Optimal and Robust Control theory to synthesize closed-loop controllers that optimally regulate current, rotation and kinetic profiles. It is anticipated that the scheme can play an important role in experiments at the DIII-D tokamak. [1] D. Moreau et al., ā??A two-time-scale dynamic-model approach for magnetic and kinetic profile control in advanced tokamak scenarios on JET,ā?? Nucl. Fus. 48 (2008) 106001. [2] D. Moreau et al., "Plasma Models for Real-Time Control of Advanced Tokamak Scenarios", submitted to Nuclear Fusion (see 2010 IAEA FEC, paper EXW/P2-07). [3] C. Xu et al., ā??Current Profile Evolution Modeling via Subspace Identification Algorithms,ā?? DPP Annual Meeting of the American Physical Society (APS), 2008. [4] C. Xu et al., ā??Transport Parameter Estimations of Plasma Transport Dynamics using the Extended Kalman Filter,ā?? IEEE Trans. on Plasma Science, vol.38, no.3, p.359, 2010. [5] W. Wehner et al., ā??Feedback Tracking Control of Safety Factor and Rotation Profile Evolutions in the DIII-D Tokamak via System Identification,ā?? DPP Annual Meeting of the American Physical Society (APS), 2010. | ||
| Resource Requirements: | Machine time: At least two 2-hour evening sessions + 1/2 day experiment. | ||
| Diagnostic Requirements: | Core and tangential Thomson, CER, CO2, magnetics, MSE, ECH diagnostics, a reasonable set of fast ion instability diagnostics (UF interferometers, FIR scattering, ECE at 500 kHz, fast magnetics with fast delay set in the current ramp), FIDA. For closed-loop experiment real-time magnetic measurements and equilibrium reconstruction including the q-profile (EFIT and RTEFIT with MSE) are essential, and real-time measurements of the ion and electron temperature profiles, as well as line-averaged plasma density or density profile are required. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 237: Parametric studies of pedestal-localized high frequency coherent modes | ||
| Name: | Yan yanz@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Pedestal Structure | Presentation time: | Requested |
| Co-Author(s): | Rich Groebner, George McKee, Phil Snyder, Tom Osborne, Keith Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to study the parametric dependencies of the High Frequency Coherent modes (HFC) observed with BES during a high density QH mode plasma in 2010 campaign. These HFC modes exhibit several features predicted for Kinetic Ballooning Modes. In particular, perform a toroidal field scan to investigate the diamagnetic drift velocity dependence of these modes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment method is to reproduce the high density QH mode plasmas from the 2010 campaign (e.g., 137253) in which HFC modes were observed. Then vary toroidal field/plasma current(?) to see if the frequency spacing between two successive modes of HFC can be varied, as well as to vary the diamagnetic drift contribution to investigate if the mode intrinsic frequency of HFC still shows ~0.2-0.3 times diamagnetic drift frequency, which is one of the features predicted for KBM. This will confirm (or not) the link between the experimentally observed HFC and theoretically predicted KBM. | ||
| Background: | It has been predicted that KBMs limit the H-mode pedestal pressure gradient, e.g., the EPED1 model assumes KBMs limit pedestal structure, but experimental evidence is lacking. Experimental results from 2010 campaign on DIIID in a high density QH mode plasma have shown KBM like features in a set of high frequency coherent modes (HFC) of the long wavelength density fluctuations from BES measurement [1]. These HFC exhibit KBM like features. The intrinsic mode frequency is ~0.2-0.3 times the diamagnetic drift frequency, medium-n structure (inferred from poloidal wavenumber). It appears that these higher-density QH-mode plasmas establish a unique experimental case where these modes exist in a quasi-stationary saturated state, with long ELM-free phases, possibly limited by the high ExB shear.
[1] Z. Yan, et al., APS2010 invited talk, 2010, to be submitted to Phys. Plasmas (2010). | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | BES, CER, TS, DBS, CECE, PCI | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 238: Dependence of edge co-rotation layer on plasma current and X-point location | ||
| Name: | Muller mullersh@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | J.A. Boedo, K.H. Burrell, J.S. deGrassie, R.A. Moyer, D.L. Rudakov, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | In the 2010 experiment, the only parameter variation that led to a change in the edge co-rotation layer was the reduction of Ip from 1.3 to 0.8 MA. However, the shot in question was also subject to a change in location of the X-point, such that it was not possible to distinguish between the two effects. This experiment aims to reproduce this variation and disentangle the two possible effects. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce the experimental conditions of proposal 225 with different plasma currents, while making sure that the shape does not vary (which the 2010 experiment failed at). Then vary the position of the X-point at nominal plasma current. | ||
| Background: | (see proposal 225) | ||
| Resource Requirements: | (see proposal 225) | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 239: Is reduction in confinement with TBM purely a rotation effect? | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Test whether the reduction in energy confinement seen with the TBM is purely a result of the reduced rotation associated with the NRMF braking torque. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce TBM amplitude scan to recover documented degradation of momentum and energy confinement. For each condition, repeat with TBM turned off, and apply counter-NBI to match rotation and compare energy confinement with/without TBM for matched rotation. | ||
| Background: | TBM experiments last year showed a significant reduction in rotation when the TBM was turned on, as well as a degradation of confinement. It is unknown whether the reduced ExB shear associated with reduced rotation can account for all the change. | ||
| Resource Requirements: | TBM, counter-NBI | ||
| Diagnostic Requirements: | Full profile diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 240: Understand 2/1 tearing limits in low torque high beta scenarios like advanced inductive or hybrid | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER like baselines at low torque have been found to be highly susceptible to error fields, and even with optimal error correction, they encounter tearing modes at low betan, ~2.2. This raises questions for higher beta scenarios like advanced inductive or hybrid, not least because plasma response to error fields is well established to rise with betan - leading to increased braking and more likely triggering of tearing modes. Also the higher beta will increase bootstrap currents which potentially makes it easier for small islands to bifurcate to large amplitude. Current profile is likely a key parameter governing the whole process, and an important factor in establishing scenario viability. Therefore it becomes particularly important to evaluate the prevalence and sensitivities of 2/1 mode threshold in advanced inductive, to assess: (i) whether the changes in current profile for the more advanced regimes improves stability (and how to improve further); (ii) to establish viability and limits for the regime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish low torque variant of advanced inductive plasma. Test prevalence of modes by varying current profile formation recipe (eg early hearing timing) and betan, between standard advanced inductive values and relaxed (later heating start, lower betan) ITER baseline like regimes. Test 3-D field role with n=1 I coil ramps in some cases. Key goals are to determine how stability and 3-D field sensitivity vary with current profile and beta at low rotation. | ||
| Background: | |||
| Resource Requirements: | Varies from quick checks (few shots) on the back of low rotation regime development, to dedicated scans to achieve complete goals. | ||
| Diagnostic Requirements: | usual MHD | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 241: Optically thick pedestals for documenting Te-tilde and nT cross-phase angle in QH-mode and I-mode | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | D. Whyte, A. Hubbard, J. Rice, K. Burrell, T. Rhodes, M. E. Austin | ITPA Joint Experiment : | No |
| Description: | Explore possibility of steady operation of QH-mode and I-mode plasmas at DIII-D with optically thick pedestals. The identification and understanding of the EHO (QH-mode) and the Weakly Coherent Mode (WCM) (associated with I-mode) can be greatly augmented with measurements of Te-tilde (CECE) and nT cross-phase angle (CECE-reflectometer) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In QH-mode plasmas, steadily increase ne and Te in the pedestal to ensure that optical depth tau2X > 3-4 in conditions where the EHO and WCM are observed on ECE radiometer channels.
In I-mode on DIII-D (see proposal by D. Whyte), do the same: steadily increase ne and Te in the pedestal to ensure that optical depth tau2X > 3-4 in conditions where the EHO and WCM are observed on ECE radiometer channels. | ||
| Background: | The EHO in QH-mode on DIII-D is often seen in edge radiometer channels and is very clearly seen in CECE channels. However, the region where the EHO exists radially is also the region over which the optical depth of the 2nd harmonic ece falls off very rapidly, making it difficult to interpret measured ECE perturbations from the EHO as Te-tilde/Te. (White, PhD thesis, 2008). The same difficulties are envisioned for any measurements of the WCM in I-mode that may be attempted using CECE and coupled reflectometer/CECE to measure the nT phase angle. | ||
| Resource Requirements: | QH-mode and I-mode plasmas | ||
| Diagnostic Requirements: | CECE, coupled CECE-Reflectometer | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 242: Diagnostic checkout of for main-ion CER diagnostic. | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | New diagnostic capabilities are now available on DIII-D for measurement of the thermal deuterium temperature and rotation. In order for the measurements and interpretation of the measurements to be properly validated in a wide range of plasma conditions | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiments we will perform will investigate the ion temperature dependence of the atomic physics corrections for charge exchange measurements. The experiments will vary the ion temperature over a wide range of values, because the atomic physics corrections scale most directly with Ti. The other contribution to the interpretation of the spectral emission is the neutral halo and fast ion emission (FIDA). Therefore, the beam power, torque, electron density and temperature (for fast ion slowing down) will all be systematically varied. The response to each variable will be compared to simulations and co+counter view analysis. | ||
| Background: | The ion temperature, toroidal rotation and C6+ impurity concentration is measured by charge-exchange recombination (CER) spectroscopy on DIII-D. While A newly installed diagnostic on DIII-D aims to measure the bulk ion (deuterium) toroidal rotation, temperature and density. In order for this diagnostic to provide reliable data to the DIII-D experimental program, a proper diagnostic checkout must be performed. | ||
| Resource Requirements: | 1 Day Experiment. Modulation of 30LT and 210RT neutral beams. | ||
| Diagnostic Requirements: | CER (CVI). Profile reflectometer for accurate electron density. ECE for accurate electron temperature. | ||
| Analysis Requirements: | Analysis of charge-exchange data will be done by CER group (Grierson). | ||
| Other Requirements: | |||
| Title | 243: Dependence of pedestal turbulence, width and profiles on Ī²p | ||
| Name: | Yan yanz@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | J. Callen, Rich Groebner, George McKee, Phil Snyder, Tom Osborne | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to vary Ī²p and investigate whether the pedestal width and turbulence characteristics scale in a predicted manner with Ī²p; in particular, the correlation length scaling with pedestal width will be explored. This will provide an experimental test about whether the observed long wavelength density fluctuations are correlated to pedestal width. The turbulence characteristics between low betap and high betap conditions will also be compared. It will provide additional test whether these turbulences are KBM related, and will also examine the connection between the observed long-wavelength fluctuations and pedestal transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main ideal of the experiment is to scan betap while keeping the other non-dimensional parameters (collisionality, rho* and q) as constant as possible at the pedestal top. It will be LSN, low triangularity H-mode plasma (similar to shot 136051). Diagnostics like BES (8Ć?8 2D array), DBS and CECE will be used to measure the turbulence spatial structure and other characteristics from the outer core, across the pedestal, and into the SOL. The outcome of the experiment will provide for investigation of the turbulence characteristics scaling with pedestal width for comparison with theoretical models. This data will also be used to examine the connection between turbulence and pedestal transport. | ||
| Background: | The EPED1 model, based on peeling-ballooning and kinetic ballooning mode (KBM) instability theories, has successfully predicted the pedestal height and width in many experiments [1]. KBM theory predicts that the pedestal width scales with square root of Ī²p. A previous JET/DIIID Ļ?* experiment found that the pedestal width has no or weak dependence of Ļ?*[2]. Consistently the long wavelength density fluctuations radial correlation length is also found to depend weakly on Ļ?*[3]. By varying the pedestal width through varying betap we can exam if the turbulence characteristics vary with pedestal width. In addition, the transport resulting from these fluctuations will be explored so see if they plausibly explain the transport across the pedestal, or whether other mechanism need to be invoked (e.g., paleoclassical transport in the outer half of the pedestal).
[1] P.B.Snyder, et al., Phys. Plasmas, 16, 056118, 2009 [2] M.N.A Beurskens, et al., PPCF, 51,124051, 2009 [3] Z. Yan, et al., APS2010 invited talk, 2010 | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | BES, CER, TS, DBS, CECE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 244: Thermal Deuterium Particle Transport Studies Using Modulated Gas Injection | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | Particle transport studies in DIII-D have previously focused on electron density and impurity density. New measurement capabilities enabling radial profiles of thermal deuterium ion temperature, bulk deuterium toroidal rotation and thermal deuterium charge-exchange photoemission provide the necessary measurements for determining thermal ion particle transport in DIII-D. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform modulated edge fueling under steady conditions in a variety of discharges. Determine the phase of propagation of fuel ions by comparing adjacent diagnostic channels. Similar analysis will be done for electron data. | ||
| Background: | The ion particle transport channel has not be investigated before for deuterium plasmas in DIII-D. | ||
| Resource Requirements: | Gas injection. Modulation of 30LT and 210RT neutral beams. | ||
| Diagnostic Requirements: | CER (CVI), profile reflectometer. | ||
| Analysis Requirements: | Analysis of charge-exchange data will be done by CER group. | ||
| Other Requirements: | |||
| Title | 245: Understand tearing limit in low torque high beta scenarios like advanced inductive/hybrid (Dup 240 | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER like baselines at low torque have been found to be highly susceptible to error fields, and even with optimal error correction, they encounter tearing modes at low betan, ~2.2. This raises questions for higher beta scenarios like advanced inductive or hybrid, not least because plasma response to error fields is well established to rise with betan - leading to increased braking and more likely triggering of tearing modes. Also the higher beta will increase bootstrap currents which potentially makes it easier for small islands to bifurcate to large amplitude. Current profile is likely a key parameter governing the whole process, and an important factor in establishing scenario viability. Therefore it becomes particularly important to evaluate the prevalence and sensitivities of 2/1 mode threshold in advanced inductive, to assess: (i) whether the changes in current profile for the more advanced regimes improves stability (and how to improve further); (ii) to establish viability and limits for the regime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish low torque variant of advanced inductive plasma. Test prevalence of modes by varying current profile formation recipe (eg early hearing timing) and betan, between standard advanced inductive values and relaxed (later heating start, lower betan) ITER baseline like regimes. Test 3-D field role with n=1 I coil ramps in some cases. Key goals are to determine how stability and 3-D field sensitivity vary with current profile and beta at low rotation. | ||
| Background: | |||
| Resource Requirements: | Varies from quick checks (few shots) on the back of low rotation regime development, to dedicated scans to achieve complete goals. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 246: Understand tearing limit in low torque high beta scenarios like advanced inductive/hybrid (Dup 240 | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | ITER like baselines at low torque have been found to be highly susceptible to error fields, and even with optimal error correction, they encounter tearing modes at low betan, ~2.2. This raises questions for higher beta scenarios like advanced inductive or hybrid, not least because plasma response to error fields is well established to rise with betan - leading to increased braking and more likely triggering of tearing modes. Also the higher beta will increase bootstrap currents which potentially makes it easier for small islands to bifurcate to large amplitude. Current profile is likely a key parameter governing the whole process, and an important factor in establishing scenario viability. Therefore it becomes particularly important to evaluate the prevalence and sensitivities of 2/1 mode threshold in advanced inductive, to assess: (i) whether the changes in current profile for the more advanced regimes improves stability (and how to improve further); (ii) to establish viability and limits for the regime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish low torque variant of advanced inductive plasma. Test prevalence of modes by varying current profile formation recipe (eg early hearing timing) and betan, between standard advanced inductive values and relaxed (later heating start, lower betan) ITER baseline like regimes. Test 3-D field role with n=1 I coil ramps in some cases. Key goals are to determine how stability and 3-D field sensitivity vary with current profile and beta at low rotation. | ||
| Background: | -- | ||
| Resource Requirements: | Varies from quick checks (few shots) on the back of low rotation regime development, to dedicated scans to achieve complete goals. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 247: Control of the safety factor in steady-state scenarios using off-axis neutral beam injection | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Requested |
| Co-Author(s): | C. Holcomb, J. Ferron, M. Walker, D. Humphreys, D. Moreau | ITPA Joint Experiment : | No |
| Description: | In this experiment, the safety factor will be controlled during the ramp-up and initial flattop phase of the plasma current using an optimized PID controller, real-time equilibrium reconstructions based on external magnetic diagnostics and internal pitch angle measurements from MSE, and on/off-axis and co/counter neutral beam injection (NBI). The ultimate goal of the experiment will be to demonstrate control of the quantity q(0)-q_min. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The new on/off-axis NBI capability, the upgraded core and edge MSE systems, and off-line simulations will be used to improve the performance of the PID-based safety factor controller in the PCS. The experimental plan is outlined by the following. (i) Reproduce and setup a previous AT discharge used for q-profile control (such as 126814), or an improved target identified by the initial off-axis NBI experiments. Work to establish a reproducible density waveform. (ii) Setup and tune PID controller starting with optimized gains identified by off-line simulations. Demonstrate control of q_min. If successful, attempt control of q(0)- q_min. Work to eliminate the need for a feedforward waveform. (iii) Maximize the pulse length by minimizing the number of early NBI pulses for MSE. Investigate using counter NBI during startup since the 210 beam is generally not used in the high-beta phase of the discharge. (iv) Vary mix of on/off-axis NBI power. Work to establish how much additional (if any) off-axis NBI is necessary to control q(0)- q_min over an appreciable range. (v) Vary the radial location of q_min using mix of on/off-axis NBI and ECCD. | ||
| Background: | Robust control methods are needed in future tokamak reactors to achieve profiles that will enable high fusion gain and noninductive sustainment of the plasma current [M. Walker et al, Fus. Eng. and Design 82 (2007), 1051]. In present day devices, control methods are necessary to access high performance operating regimes near instability boundaries in the presence of variations in the current profile following discharge breakdown, in the plasma density, and in the plasma impurity profiles [T.C. Luce, Nucl. Fusion 45 (2005) S86]. Previous DIII-D experiments have focused on the development of feedback algorithms (based on a PID controller) to control the safety factor during the ramp-up of the plasma current in advanced tokamak scenarios [J.R. Ferron et al, Nucl. Fusion 46 (2006) L13-L17]. Inputs to the PID controller include the safety factor profile calculated by a real-time equilibrium reconstruction, constrained by external magnetic field and flux measurements and by internal poloidal field measurements from the motional Stark effect diagnostic. Experiments have shown that NBI heating is the most efficient actuator [J.R. Ferron, IAEA 2008]. The NBI heating acts by modifying the conductivity profile, which control the relaxation rate of the inductive current. Comparisons of the measured current evolution with transport simulations over multiple years (since 2005) have shown that the radial localization of the NBCD has a significant effect on the agreement with theory. The role of Alfven eigenmode instabilities may be important for understanding the observed variations in the current evolution. Results from these initial experiments suggest a path forward toward full optimization of the PID controller, which would be an important tool for improving steady-state non-inductive scenarios. An optimal PID controller would also provide an important benchmark for vetting more sophisticated model-based controllers [Ou, PhD. Thesis 2010; D. Moreau et al, Nucl. Fusion 48 (2008), 106001].
During the 2011 campaign, two new tools may enable improved control using existing PID controllers. As a result of upgrades to the core and edge MSE diagnostic, improved measurements of the magnetic field line pitch angle will be available to constrain the real-time reconstructions. This is significant since there is an increased demand on the quality of the MSE data during the ramp-up of the plasma current when the q profile is flat and q_min is relatively high. The new off-axis neutral beam capability in may provide the means to further tailor the conductivity profile for control of the quantity q(0)-q_min in high q_min scenarios. Successful q profile control with the off-axis beam would demonstrate an advanced understanding of the physics associated with off-axis injection. | ||
| Resource Requirements: | 8 NB sources including off-axis beam. All available gyrotrons. I-coil error field correction. RT EFITs | ||
| Diagnostic Requirements: | Equilibrium diagnostics including magnetic sensors, flux loops, core and edge MSE. Fast ion instability diagnostics (including UF interferometers, FIR scattering, ECE and ECEI, fast magnetics, BES, FIDA, FILD, BILD). CER. Thomson. CO2. ECE. Profile reflectometer. | ||
| Analysis Requirements: | PID controller optimization via off-line simulations with CORSICA and Simulink. Comparisons with ONETWO/TRANSP. Analysis of initial off-axis neutral beam experiments. | ||
| Other Requirements: | -- | ||
| Title | 248: Compare Toroidal Rotation of Deuterium Ions and Carbon Impuritiy Ions with Neoclassical Theory | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | Determination of thermal deuterium toroidal rotation is important for assessing the stability of plasmas in the presence of RWMs and NTMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish baseline discharge with flat pressure profiles. Then move on to increased pressure profile steepness with core ion heating provided by NBI. This scan should be performed in a wide range of densities. | ||
| Background: | The ion temperature, toroidal rotation and C6+ impurity concentration is measured by charge-exchange recombination (CER) spectroscopy on DIII-D. While A newly installed diagnostic on DIII-D aims to measure the bulk ion (deuterium) toroidal rotation, temperature and density. Previous measurements of bulk ion rotation [De Grassie, J. S. et. al. Phys. Plasmas 2007] in helium plasmas displayed differences between the bulk ion rotation and neoclassical predictions based on impurity measurements. We will explore the rotation differences in deuterium plasmas. The control knob on differential velocity is the thermal ion pressure profile (mostly grad-Ti), and this will be varied through NBI heating. | ||
| Resource Requirements: | 1 Day Experiment. Modulation of 30LT and 210RT neutral beams. | ||
| Diagnostic Requirements: | CER (CVI), profile reflectometer, ECE. | ||
| Analysis Requirements: | Analysis of charge-exchange data will be done by CER group. | ||
| Other Requirements: | |||
| Title | 249: Measurement of pedestal deuterium toroidal rotation, temperature and density profile. | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | We aim to measure the edge pedestal profile of bulk deuterium toroidal rotation, temperature and density to characterize the structure of the pedestal. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use interleaved edge tangential CER measurements of both impurity carbon and deuterium charge-exchange spectra to produce profiles of density, temperature and toroidal rotation. | ||
| Background: | We want to understand the characteristics of the bulk ion species in the pedestal. | ||
| Resource Requirements: | Modulation of 330LT beams source. | ||
| Diagnostic Requirements: | CER (CVI). | ||
| Analysis Requirements: | Analysis of charge-exchange data will be done by CER group. | ||
| Other Requirements: | |||
| Title | 250: Investigate stabilization of turbulence in low rotation QH-mode plasmas | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | K. Burrell, G. McKee, L. Schmitz, W. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to test basic mechanisms of turbulence suppression in low rotation QH-mode discharges with applied NRMFs.
These discharges showed evidence of turbulence suppression and confinement improvement when the plasma rotation was lowered toward zero. This only happened at sufficiently high betan value. These low rotation QH-mode discharges offer a unique opportunity to isolate beta effects on turbulence, because of the possibility to run at zero rotation, therefore ruling out effects of ExB shear stabilization. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | - Reproduce discharges 138604 (lower beta) and 138605 (higher beta) with full fluctuation documentation. BES would benefit from carrying out these experiments in normal Ip, whcich could be possible with some development.
Also, BES would benefit from a more regular modulation of the 150 NBI sources. - Test if a stronger confinement improvement is obtained at higher beta. | ||
| Background: | Reference plasmas are DIII-D discharges 138604 (lower beta) and 138605 (higher beta).
When the NBI counter torque is reduced toward zero, the rotation is reduced, but confinement goes up in the higher beta case. No confinement improvement is observed at lower beta. Measurements of density fluctuations from DBS and BES show strong turbulence reduction at minor radius between 0.5 and 0.9. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 251: Collisionality scaling of the coupled turbulence/zonal flow during LH transition | ||
| Name: | Yan yanz@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Jose Boedo, George McKee, Dimitry Rudakov, Rich Moyer, L. Schmitz | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the collisionality scaling of the coupled turbulence/zonal flow system before, during and after the LH transition through varying electron density at different momentum input. Try to understand the role of Reynolds stress and zonal flow playing in triggering LH transition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main idea is to vary electron density to vary collisionality at different momentum input (co, counter and balanced) to investigate zonal flow effects on LH transition. The beam power will be kept low to favor the Langmuir probe measurement of Reynolds stress at plasma edge. BES, DBS and CECE will be used to measure plasma turbulence. | ||
| Background: | The existence of the geodestic acoustic mode (GAM) and the zero-mean-frequency (ZMF) zonal flow predicted to be generated by the plasma turbulence has been clearly identified experimentally in tokamak and stellarator plasmas [1,2] and may relate to the mechanism for L- to H- mode transition [3]. It is predicted that the zonal flow can be damped linearly through collisions. In the 2010 campaign the ion gyro-radius scaling of the coupled turbulence/zonal flow has been investigated through varying toroidal field. However the collisionality dependence has not yet been investigated yet thought to be very important. It is predicted that increasing ion-ion collisionality will damp zonal flows, potentially leading to higher ambient turbulence and transport levels [4]. ZF damping via collisionality may play a role in the strong density dependence of the L-H transition power threshold. Understanding the underlying physics and scaling of the threshold dependence is a key issue for ITER.
[1] G.R.Mckee, et al., Phys. Plasmas, 10, 1712, (2003) [2] A.Fujisawa, et al., Phys. Rev. Lett., 93, 165002 (2004) [3] K. H. Burrell, Phys. Plasmas 4, 1499 (1997) [4] Z. Lin, Phys. Rev. Lett. 83, 3645 (1999). | ||
| Resource Requirements: | 4 neutral beams | ||
| Diagnostic Requirements: | BES, CER, TS, DBS, CECE, PCI, Mid-plane probe with Reynolds stress head | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 252: Comparison of Impurity Carbon and Deuterium Toroidal Rotation for RWM Rotation Dependence | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | Plasma rotation is generally a stabilizing for RWMs and NTMs. The plasma rotation is measured with charge exchange spectroscopy of light impurities such as fully-stripped carbon in DIII-D, and this impurity rotation profile is readily available for assessment of the plasma toroidal rotation velocity. However, neoclassical theory predicts that the toroidal rotation of the main ions (deuterons) and impurity carbon can be significantly different when there are strong pressure gradients. New diagnostic capabilities on DIII-D will enable the measurement of the bulk thermal deuterium toroidal rotation, rather than relying on neoclassical calculations, to determine the rotation profile of the main ions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Discharge beam timing discussed with RWM group to enable diagnostic capabilities. | ||
| Background: | |||
| Resource Requirements: | 30LT and 210RT neutral beam modulation. | ||
| Diagnostic Requirements: | CER | ||
| Analysis Requirements: | CER Group (Grierson) | ||
| Other Requirements: | Discussion with CER group to determine appropriate diagnostic beam modulation. | ||
| Title | 253: Determination of bulk deuterium toroidal rotation at low rotation. | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | Measure the bulk ion rotation in addition to standard impurity CER rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In coordination with CER group (Grierson), develop beam modulation patters which maximize the data quality of the main-ion CER measurements. | ||
| Background: | The plasma rotation is measured with charge exchange spectroscopy of light impurities such as fully-stripped carbon in DIII-D, and this impurity rotation profile is readily available for assessment of the plasma toroidal rotation velocity. However, neoclassical theory predicts that the toroidal rotation of the main ions (deuterons) and impurity carbon can be significantly different when there are strong pressure gradients. New diagnostic capabilities on DIII-D will enable the measurement of the bulk thermal deuterium toroidal rotation, rather than relying on neoclassical calculations, to determine the rotation profile of the main ions. | ||
| Resource Requirements: | 30LT and 210RT neutral beam modulation. | ||
| Diagnostic Requirements: | CER | ||
| Analysis Requirements: | CER Group (Grierson). TRANSP. | ||
| Other Requirements: | Discussion with CER group to determine appropriate diagnostic beam modulation. | ||
| Title | 254: Determination of bulk deuterium toroidal rotation for ITER baseline discharges. | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Not requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | Measure baseline scenario bulk-ion rotation, and compare to standard impurity measurements. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In coordination with CER group (Grierson), develop beam modulation patters which maximize the data quality of the main-ion CER measurements. | ||
| Background: | The plasma rotation is measured with charge exchange spectroscopy of light impurities such as fully-stripped carbon in DIII-D, and this impurity rotation profile is readily available for assessment of the plasma toroidal rotation velocity. However, neoclassical theory predicts that the toroidal rotation of the main ions (deuterons) and impurity carbon can be significantly different when there are strong pressure gradients. New diagnostic capabilities on DIII-D will enable the measurement of the bulk thermal deuterium toroidal rotation, rather than relying on neoclassical calculations, to determine the rotation profile of the main ions. | ||
| Resource Requirements: | 30LT and 210RT neutral beam modulation. | ||
| Diagnostic Requirements: | CER | ||
| Analysis Requirements: | CER Group (Grierson). TRANSP. | ||
| Other Requirements: | Discussion with CER group to determine appropriate diagnostic beam modulation. | ||
| Title | 255: Measurement of deuterium beam neutral halo | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | K. H. Burrell, W. M. Solomon | ITPA Joint Experiment : | No |
| Description: | The halo surrounding neutral beam injection is a second source (besides the beam itself) of neutral particles in the core of hot fusion plasmas. These halo neutrals a charge donors for charge-exchange phenomena. Diagnostics such as FIDA and main-ion CER are sensitive to the density and spatial extent of the neutral halo. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Tune core vertical CER to D-alpha and modulate 330RT beam. The parameters which affect the halo emission are electron density, electron temperature, ion temperature, neutral beam injector voltage, neutral beam injector power, and plasma rotation to a lesser extent. | ||
| Background: | Need to benchmark FIDA simulation code for interpretation of spectral amplitude for all FIDA diagnostics and main-ion CER. Preliminary experiments in fall ā??10 demonstrated feasibility, but with little plasma parameter variation. | ||
| Resource Requirements: | 330RT beam modulation. | ||
| Diagnostic Requirements: | Vertical CER to D-alpha, but TANG CER to CVI for Vtor, Ti and carbon density. Need good electron density and temperature as well for simulation. | ||
| Analysis Requirements: | CER group, equilibrium and complete profile analysis. | ||
| Other Requirements: | |||
| Title | 256: Change in stiffness with density peaking | ||
| Name: | Staebler none | Affiliation: | GA |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | Terry Rhodes | ITPA Joint Experiment : | No |
| Description: | Using the TGLF transport model it has been predicted that there is a significant change in the stiffness (ratio of incremental diffusivity to power balance diffusivity) with density gradient. For flat density profiles ther is a critical temperature gradient and the stiffness is high. If the power is low so that the profile is near to marginal the turbulence should have a distivtive bursty character seen in GYRO simulations. For somwhat peaked density profiles there is no critical temperature gradient and for low power the stiffness is near one. The turbulence should always be in a steady broadband condition without large bursts. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The detailed experimental plan will be determined by the Profile Stiffness and Critical Gradient group and may be very different from this first idea since ther are many diagnostic constraints. The goal is to compare the stiffness and turbulence characteristics of flat and peaked density profile discharges with no net external torque (since rotation also impact stiffness). An inner wall limited L-mode could profide the peaked density case and a similary shaped discharge which is not limited would give an H-mode with a flat density profile. The H-mode will need to be pumped to achieve a similar density at r/a=0.5 as the L-mode. The L-mode may need to be low density in order to have a sufficiently peaked density. Low density also gives the best fluctuation diagnostic access. ECH pulses would be used to evaluate the incremental electron thermal diffusivity at the mid radius. | ||
| Background: | Target discharges for the L-mode limited and H-mode shapes need to be found. This has been done before on DIII-D. | ||
| Resource Requirements: | balanced beams, ECH, | ||
| Diagnostic Requirements: | full compliment of fluctuation and profile diagnostics. | ||
| Analysis Requirements: | TGLF modeling of target discharges to determine expected results and fine tune the plasma conditions. | ||
| Other Requirements: | |||
| Title | 257: A Powered VFI for robust MIMO control | ||
| Name: | Hyatt hyatt@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Requested |
| Co-Author(s): | Michael Walker, David Humphreys | ITPA Joint Experiment : | No |
| Description: | We propose to finish the experimental tests begun in 2009 in which plasmas are generated and controlled with F-coils subject to the "VFI bus constraint" but with the voltage on the VFI bus regulated by a dedicated supply. This yearā??s experiments will complete the tests with the data required to predict if a powered VFI removes enough complexity to create MIMO controllers robust over a wide range of parameters in DIII-D while using the present choppers and retaining the VFI constraint. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will use the powered VFI patch and LSN plasma successfully developed in 2009. The LSN control algorithm used will be modified to allow simultaneous Fcoil current control and fast vertical stabilization control. Results from this experiment will indicate if the powered VFI approach is sufficient to allow a robust MIMO controller to be utilized with the present Fcoil power supplies and the VFI constraint. The results will inform a later decision to purchase and install a large (~10 kA/600V) voltage regulated supply that is engineered and dedicated to this purpose.
Need to do: - Add combined stabilization/F-coil current control logic to the special VFI-control version of shape control algorithm (which uses the standard PID feedback for shape control). - Run 2-3 shots where we switch from standard PID control, but with VFI voltage controlled, to regulation of F-coil currents (no shape control feedback) with the VFI voltage controlled. - Run 6-8 shots where the controlled F-coil currents are used as actuators by the standard PID feedback shape control, with the VFI voltage being controlled simultaneously. | ||
| Background: | Analysis of MIMO experiment data in the DND configuration completed in 2008 demonstrates that it is essentially impossible to obtain routine robust model-based plasma shape control because of the combination of VFI bus constraint and severe chopper nonlinearities. In addition, operational experience with the present empirically tuned PID controllers has found the VFI connection to be problematic for shape control for a number of plasma equilibria. In large measure these difficulties arise because the VFI constraint as it is presently implemented using simple coils to handle the net Fcoil current forces some of the Fcoil supplies into saturation, and this in turn places some choppers deeply into nonlinear territory. If, however, the VFI bus voltage is regulated to a fixed small value, such as zero, each Fcoil is effectively decoupled from all others and all choppers can be maintained in relatively linear operation.
Experiments with a powered VFI on a LSN configuration in 2009 demonstrated improved control with the standard PID shape control algorithms. Experiments to evaluate the potential for breaking the gridlock on MIMO control caused by the VFI/nonlinear choppers combination were inconclusive due to the inherent inability of the algorithm to allow plasma stabilization with the standard PCS vertical control algorithm when all F-coil currents were being simultaneously controlled. However, MIMO controllers which simultaneously stabilize the plasma and control currents in the 2, 6, and 7 coils have already been demonstrated in DND configurations, so we expect that modifying the LSN PID controller used in the 2009 experiments to simultaneously stabilize and control all F-coilsā?? currents is straightforward. | ||
| Resource Requirements: | Need a few beams (1-3), 30L for MSE, standard error field correction. Need 2 2-hour sessions to verify new control logic is working plus 1/2 day to conduct tests of coil current control that will generate sufficient data to predict MIMO robustness using a powered VFI. | ||
| Diagnostic Requirements: | Standard magnetics, coil currents, chopper and power supply voltages, CO2 interferometers for density control. | ||
| Analysis Requirements: | Auto EFITs | ||
| Other Requirements: | Mike Walker needs approximately 1-2 weeks of coding time to modify the existing PCS LSN P-VFI algorithm to support the required testing plan. | ||
| Title | 258: Validation of theoretical Geodesic acoustic mode predictions and turbulence models at large r/a | ||
| Name: | Hillesheim jon.hillesheim@ukaea.uk | Affiliation: | CCFE |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | W.A. Peebles, T.A. Carter, L. Schmitz, T.L. Rhodes, G. Wang | ITPA Joint Experiment : | No |
| Description: | The goal of the experiment will be to acquire a larger set of fluctuation measurements at large r/a (~0.7-0.9) than has previously been done in order to investigate both the Geodesic acoustic mode (GAM) and observed discrepancies between gyrokinetic simulation results and measurements in previous experiments. In particular, a high priority will be to acquire data on the GAM, zonal flows, and measurements of the crossphase between electron temperature and density fluctuations in this radial region, in addition to measurements of fluctuation levels. Whereas previous ne/Te measurements have focused on turbulence, this experiment will attempt also acquire ne/Te crossphase measurements of a coherent mode, the GAM, if present. By looking at more facets of the turbulence experimentally one purpose of the experiment is to better understand why the observed differences with gyrokinetic simulations occur. Another goal of the experiment will be to test theoretical predictions about the GAM, in particular its radial structure and the dependence of its radial wavenumber on the temperature ratio, as well as to perform quantitative comparisons of measurements of the interaction between the GAM and turbulence with simulations. With both DBS and BES data it should be possible to simultaneously investigate GAM interactions with both low-k and intermediate-k density fluctuations. Gathering both DBS and BES data also provides a cross-check on GAM measurements, while GAM data can still be acquired on shots where beams are used to get CER/MSE data. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use a beam-heated L-mode plasma with early heating to delay the onset of sawteeth for the acquisition of the turbulence data. The experiment needs to be in L-mode since the GAM is not observed in H-mode plasmas. The primary goal will be to acquire spatially and wavenumber resolved measurements of the GAM and turbulence, which will require X-mode DBS access to at least r/a ~0.6-0.7. This must be balanced with keeping the optical depth sufficiently high to acquire CECE and ne/Te crossphase data at large r/a. This can be attempted by running low density, high heating power discharges to raise the electron temperature as high as possible towards the edge. Once the target plasma is created, fluctuation diagnostics will scan radial position and wavenumber within r/a ~0.7-0.9 (and DBS/BES to the LCFS). During the latter portion of the shots a Te/Ti ratio scan will be performed by varying levels of ECH to change the radial wavenumber of the GAM. | ||
| Background: | Previous transport validation model experiments have systematically found significant discrepancies between the predictions of nonlinear gyrokinetic simulations and experimental measurements at large r/a, beyond what could be resolved by profile sensitivities. Previous measurements of the crossphase between electron temperature and density fluctuations have mostly been in the core (r/a ~0.5-0.8), and have been of turbulence, not coherent modes such as the GAM. Previous comparisons of GAMs to gyrokinetic simulation have been qualitative; no quantitative comparisons have been performed. There exists disagreement between measurements of the radial dependence of the GAM frequency and simulations/theory--generally, the former observes a GAM frequency that is constant in radius while the latter predicts a frequency continuum depending on the temperature with radius. A more complete set of measurements may help to better understand this discrepancy. | ||
| Resource Requirements: | 1 day. All beams, gyrotrons. | ||
| Diagnostic Requirements: | All localized fluctuation measurements: DBS, reflectometry, CECE, BES, etc.
Profile measurements: Thomson, ECE, profile reflectometry, CER, MSE | ||
| Analysis Requirements: | Comparison to gyrokinetic simulations. | ||
| Other Requirements: | -- | ||
| Title | 259: Pedestal and core transport and turbulence response in ELM-suppressed plasmas via RMP modulation | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans, R. Moyer, L. Schmitz, O. Schmitz, L. Zeng | ITPA Joint Experiment : | No |
| Description: | Examine the temporal response of low-k turbulence, particle transport, and density profile modification in the pedestal and core region to ascertain whether the observed fast turbulence enhancement with RMP may be causing the increased particle transport. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Establish a "standard" RMP ELM-suppressed discharge condition (e.g., 142250). Then, modulate the I-coils as rapidly as is technically possible with the power supplies, for as long a duration as is feasible (~2 seconds). The modulated data allows for phase-locked averaging of the fluctuation data to obtain fast time-response with the low-amplitude H-mode fluctuations. Shot 142250 & 142251 successfully established this modulation, but where unable to repeat it. Obtain fluctuation measurements with the 8x8 BES array in the plasma edge/pedestal region to complement previous core measurements and complete the radial scan in the important edge region. Obtain high-time-resolution Thomson Scattering measurements at the modulation pulses, along with Profile Reflectometry measurements.
This data will contribute to Milestone 178 and the goals of elucidating the transport mechanisms behind RMP ELM suppression (Task Force), of great interest to ITER. This proposal also stands as a general placeholder for other necessary scans and parametric variations that may be required to complete a study of turbulence and transport, particle pump-out (TBD): parity, collisionality, rho*, RMP amplitude, q95, etc. A data mining survey is in progress and will be used to clarify what additional experiments are required. | ||
| Background: | Understanding the physical mechanisms behind RMP-induced ELM-suppression is critical to exploitation and extrapolation of this technique to ITER and other experiments. The general mechanism is thought to result from enhanced particle transport (associated with particle pump-out), and reduction of the pedestal gradients. The prevailing hypothesis is that the resulting pedestal pressure and current density gradients are reduced below the peeling-ballooning limit, inhibiting ELMs. Several mechanisms for the transport enhancement have been proposed, including production of a stochastic edge, island overlap, tearing modes at pedestal top.
An experimental observation is that turbulence increases significantly when RMP field is applied and ELMs are suppressed, suggesting that this may be an important mechanism driving the resulting particle pump out and gradient reduction. These measurements have been obtained with BES in RMP ELM-suppressed discharges [McKee, APS-2009; TTF-2010; Yan-IAEA-2010]. It has been further postulated that zonal flow damping by the RMP may lead to the enhanced turbulence and particle transport [P. Diamond, UCSD/NFRI]. A lingering question has been whether the turbulence enhancement is leading and driving the particle transport, or is a response to the increasing density gradients (or both). Fast profile reflectometer measurements show rapid profiles changes in response to RMP. An experiment (3/2010) applied modulated RMP via operating the I-coils at 5 Hz for 2 seconds. This allowed for an examination of the turbulence and transport response, which showed large rapid changes (< 3 ms) in turbulence in the range of 0.8 < r/a < 0.85 as the I-coils were modulated, while ELMs remain suppressed. A critical question is what happens to turbulence in the pedestal region, but these measurements were not obtained for technical reasons (conditions evolved and proper plasma conditions were not restored during a 3D experiment, 3/2010). | ||
| Resource Requirements: | I-coils (up to 7 kA), modulated at up to several Hz. | ||
| Diagnostic Requirements: | BES, DBS, Profile reflectometer, high-speed Thomson Scattering | ||
| Analysis Requirements: | Phase-locked characterization of turbulence response. | ||
| Other Requirements: | |||
| Title | 260: Role of the plasma response in ELM suppression by non-axisymmetric magnetic fields | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | H. Reimerdes, O. Schmitz, J.M. Hanson, T.E. Evans, E. Lazarus, M.E. Fenstermacher, C. Lasnier, M. Schaffer, I. Joseph | ITPA Joint Experiment : | No |
| Description: | Document the dependence of the n=3 plasma response on the plasma pressure and q95 in ISS and balanced double null discharges using modulated n=3 magnetic perturbations. The goal of this experiment is to obtain data to validate plasma response and 3D equilibrium models, and obtain plasma response measurements in scenarios where ELM suppression has been studied. The data are important for establishing what role the plasma response plays in ELM suppression, and for studying ELM suppression in discharges with stellarator symmetry. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Obtain plasma response measurements in ISS discharges (like 126006 or 138344) at betan/li~ 1.5 and a range of q95 including 2.8. Use n=3 even parity I-coil field with a modulated n=3 amplitude (standing wave) to probe the plasma. Look for a change in the phase of the plasma response, which is predicted by MARS-F calculations. Increase betan/li to 2.0 and repeat. The phase shift (if observed) should occur at larger values of q95. Repeat measurements in the balanced double null (BDN) shape (e.g. 142585) and attempt ELM suppression. Look for correlations between ELM suppression and the measured and predicted structure of the plasma response. Document divertor structures. | ||
| Background: | MARS-F calculations of the linear ideal MHD plasma response to n=3 even parity I-coil fields in ISS discharges reveal a dependence not only on the plasma pressure, but also on the q profile [Lanctot et al, APS 2010]. It is well known that the pressure dependence of the plasma response results from a change in the stability of damped stable kink modes [Boozer, Phys. Rev. Lett. 2001]. This dependence has been previously documented, and is in good agreement with magnetic plasma response measurements [Lanctot et al, Phys. Plasmas 17 (2010)]]. What is less well known is that the q profile affects how the external field couples to the kink mode, with the coupling between the even parity I-coil and the kink mode reaching a maximum at lower values of q95. The MARS-F results show that at values of q95~3.6 and betan~2.0, the resonant field and the coupling between the even parity n=3 I-coil field and the n=3 kink mode is such that the plasma response field at the midplane attains a minimum. This occurs in the resonant q95 window for ELM suppression. A relatively large change in the phase of the plasma response occurs across this minimum. This correlation suggests a possible connection between the structure of the plasma response and ELM suppression. To investigate this correlation further, similar simulations of the plasma response are planned in other scenarios where ELM suppression has been attempted but not yet achieved such as the balanced double null shape, and in scenarios where ELM suppression has been achieved with I-coil configurations other than even parity. These plasma response calculations will be used to hypothesize what plasma conditions and I-coil configurations may be important for obtaining ELM suppression. | ||
| Resource Requirements: | Special F-coil patch panel (20DNRDPM12) for BDN. n=3 even parity I-coil for experiments in ISS discharge. Experiment should follow boronization to minimize likelyhood of locked modes. | ||
| Diagnostic Requirements: | Equilibrium diagnostics including kinetic profiles. Visible divertor cameras, divertor IR cameras, divertor (ISP) Langmuir probes, fast reciprocating probe, ECE, ECE-I, fast profile reflectometer, BES, UCLA reflectomers | ||
| Analysis Requirements: | EFIT, MARS-F, VMEC, SIESTA, PIES, HINT2 | ||
| Other Requirements: | |||
| Title | 261: Disruption avoidance with ECH | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Most disruption processes involve q=2 locked modes at some stage, even if they originate from other causes. FTU had some success in deploying q=2 heating to arrest the progression towards disruption of a particular radiative induced tearing mode. We should explore whether this technique has general applicability to delay or prevent disruptions in high performance ITER relevant conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Deploy q=2 tracking systems and real time mirror control for ECH. Induce a potentially disruptive even - for example by exceeding the density or injecting an Argon pellet. Fire gyrotrons at q=2 to see if event can be slowed or stopped. Explore variations with ECCD:ECH and nature/severity of disruption event induced, and q95. I coils may be deployed to hold island in a given phase in some cases. | ||
| Background: | Disruptions remain the critical issue for the tokamak generally, and for ITER. Development of an avoidance strategy is critical - it is not enough to simply rely on mitigation of potentially quite frequent events. This technique has high potential and builds on previous progress on DIII-D, its unique capabilities, and results from elsewhere around the world. | ||
| Resource Requirements: | 1-2 days on the machine, Good ECH, mode tracking in PCS, steerable mirrors in real time, Anders Welander, I coils | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 262: Disruption avoidance with ECH (Dup. 261) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Most disruption processes involve q=2 locked modes at some stage, even if they originate from other causes. FTU had some success in deploying q=2 heating to arrest the progression towards disruption of a particular radiative induced tearing mode. We should explore whether this technique has general applicability to delay or prevent disruptions in high performance ITER relevant conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Deploy q=2 tracking systems and real time mirror control for ECH. Induce a potentially disruptive even - for example by exceeding the density or injecting an Argon pellet. Fire gyrotrons at q=2 to see if event can be slowed or stopped. Explore variations with ECCD:ECH and nature/severity of disruption event induced, and q95. I coils may be deployed to hold island in a given phase in some cases. | ||
| Background: | Disruptions remain the critical issue for the tokamak generally, and for ITER. Development of an avoidance strategy is critical - it is not enough to simply rely on mitigation of potentially quite frequent events. This technique has high potential and builds on previous progress on DIII-D, its unique capabilities, and results from elsewhere around the world. | ||
| Resource Requirements: | 1-2 days on the machine, Good ECH, mode tracking in PCS, steerable mirrors in real time, Anders Welander, I coils | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 263: ELM control by edge parallel current drive | ||
| Name: | Svidzinski svidzinski@far-tech.com | Affiliation: | FAR-TECH, Inc. |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | Yongkyoon In, Jin-Soo Kim | ITPA Joint Experiment : | No |
| Description: | The proposal is to examine the feasibility of plasma edge stabilization and ELMs control by driving a strong parallel current at the plasma edge in Tokamak. Possible way to examine this is to shortly increase toroidal loop voltage during H mode such that a strong edge localized parallel current is driven transiently. The strong parallel current creates negative magnetic shear (qā??<0) at the edge which might stabilize peeling-ballooning modes and increase the pedestal height. In a reactor scenario noninductive edge parallel current drive techniques might be used. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | During pulsed parallel current drive regime in the Madison Symmetric Torus RFP a strong parallel current is driven at the plasma edge region. Plasma boundary is stable in spite of a relatively strong edge pressure gradient, high plasma beta in RFP and a strong unfavorable curvature of equilibrium magnetic field at the edge in RFP. Ten-fold confinement improvement is observed. A strong negative magnetic shear is created at the edge by the driven parallel current.
Numerous experiments were performed on JET to study formation of internal transport barrier (ITB) in discharges with reversed magnetic shear. A strong of-axis current was driven by Lower Hybrid Current Drive to form a reversed q-profile with a large negative shear (qā??<0) in the plasma core. Analysis of these discharges [1] showed that ITBs were formed at location of a strong negative magnetic shear (location of the driven current), meaning that plasma instabilities were suppressed there. In our proposal a strong negative shear is created at the plasma edge by the driven edge parallel current such that plasma stabilization is expected there as well. By analogy with JET results we propose to move the ITB to the plasma edge and place it on top of the edge transport barrier (pedestal region) in the H-mode discharge to further enhance and stabilize the latter. Theoretical analysis of the plasma edge stabilization by the driven edge parallel current in a simplified tokamak-like cylindrical equilibrium shows that the plasma boundary is stabilized by a sufficiently strong edge current [2]. The stabilization is due to a strong negative magnetic shear created by the driven current at the edge and due to the proximity of the plasma boundary to a conducting wall. Stability of plasma edge in Tokamak in a fully toroidal geometry in the limit of large toroidal mode numbers was analyzed in [3]. Equation (11) in this reference shows that the edge modes are stable when qā??<0 at the edge. We propose to experimentally examine the effect of negative magnetic shear (qā??<0 is localized at the plasma edge) on the plasma edge stability on DIII-D. [1] Yu. F. Baranov, et. al., Plasma Physics and Control. Fusion, Vol. 46 p. 1181 (2004). [2] V. Svidzinski et. al., Nuclear Fusion Vol. 50 p. 045009 (2010). [3] J. W. Connor, et. al., Physics of Plasmas, Vol. 5 p. 2687 (1998). | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 264: Disruption avoidance with ECH (dup. 261) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Most disruption processes involve q=2 locked modes at some stage, even if they originate from other causes. FTU had some success in deploying q=2 heating to arrest the progression towards disruption of a particular radiative induced tearing mode. We should explore whether this technique has general applicability to delay or prevent disruptions in high performance ITER relevant conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Deploy q=2 tracking systems and real time mirror control for ECH. Induce a potentially disruptive even - for example by exceeding the density or injecting an Argon pellet. Fire gyrotrons at q=2 to see if event can be slowed or stopped. Explore variations with ECCD:ECH and nature/severity of disruption event induced, and q95. I coils may be deployed to hold island in a given phase in some cases. | ||
| Background: | Disruptions remain the critical issue for the tokamak generally, and for ITER. Development of an avoidance strategy is critical - it is not enough to simply rely on mitigation of potentially quite frequent events. This technique has high potential and builds on previous progress on DIII-D, its unique capabilities, and results from elsewhere around the world. | ||
| Resource Requirements: | 1-2 days on the machine, Good ECH, mode tracking in PCS, steerable mirrors in real time, Anders Welander, I coils | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 265: Revolutionizing EFIT analysis by adding a constraint on ff' | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | B.A. Grierson, N.A. Pablant, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to do a detailed checkout of |B| measurements using the newly developed main ion CER system. The initial checkout will be done in piggyback experiments. However, to fully verify the measurement prior to using it for routine EFIT analysis, we will need dedicated experimental time. This dedicated checkout will exercise the system under a broad range of conditions, including extreme conditions where we expect the maximum difference between the vacuum field and the field with plasma. These include high beta_p discharges and discharges with rapid toroidal field ramps. The |B| results from the main ion system will be compared with kinetic EFIT analysis including MSE measurements | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The shot sequence will include variations in toroidal field, toroidal field ramp rate, plasma current, plasma density and neutral beam voltage. These will primarily be two point scans; for example, the toroidal field variation will be from 1.4 to 2.16 T. Plasma current and beam power will be used to change beta_p from below unity to as high as possible to investigate the plasma diamagnetism and paramagnetism. Both upward and downward ramps in the toroidal field will be used to induce poloial currents, thus changing the toroidal field from the vacuum value. | ||
| Background: | EFIT MHD equilibrium analysis is fundamental to all the stability and transport analysis done on D III-D. In order to better constrain the analysis, internal magnetic field line pitch measurements from MSE are coupled with the external magnetic measurements in magnetics-only EFITs. However, even this does not fully constrain the fitting problem, since the parameterization used for the p' and ff' terms can allow these two terms to trade off against each other in the fitting. Further constraints can be added by doing a full kinetic EFIT although such analysis can be quite time consuming, since kinetic efits require analysis of all the plasma profiles plus calculation of the fast ion pressure. The kinetic EFIT process constrains the EFIT much better because the p' term is now specified from the kinetic measurements. By adding measurements of |B| at eight spatial locations from the main ion CER system, we can achieve an equivalent constraint, since the |B| profile is directly related to ff'. Work by N.A. Pablant for his recently completed thesis has demonstrated that the main ion CER system can measure |B| with a random error of about 2 mT and a time resolution of at least 10 ms. The difference between the vacuum field and the field with plasma is clearly resolved. Use of |B| constraints in magnetic equilibrium analysis has been discussed by Foley et al [Nuclear Fusion 48, 085004 (2008)]. If we can fully develop this capability for EFIT analysis including the |B| measurement, we will bewhich is potentially as accurate as a kinetic EFIT but which would require significantly less analysis time. In addition, because the analysis using |B| determines the plasma pressure without need to calculate the fast ion fraction, we will be able to provide information on fast ion loss. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Main ion CER system and MSE system. Full set of profile diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 266: Dynamic error correction at low torque | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The enhanced response of the plasma to error fields at low rotation may provide a basis for magnetic feedback correction & measurement of intrinsic error. We should test this | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish modest rotation low beta H mode. Ramp down torque to negative values (to ramp rotation to near zero) while deploying n=1 RWM feedback. If suitable response detected, optimize by varying the pre-programmed correction and raising gain. If not explore variations in beta and density to generate improved response, while avoiding rotating mode limits. Could also look to diagnose impact by plasma rotation effect. | ||
| Background: | Real time error field measurement and control needs to be developed for future devices, particularly at low torque, which this targets. | ||
| Resource Requirements: | May be quick test as part of a wider campaign - given more time if proves promising. RWM feedback system needed. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 267: Dynamic error correction at low torque (Dup. 266) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The enhanced response of the plasma to error fields at low rotation may provide a basis for magnetic feedback correction & measurement of intrinsic error. We should test this | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish modest rotation low beta H mode. Ramp down torque to negative values (to ramp rotation to near zero) while deploying n=1 RWM feedback. If suitable response detected, optimize by varying the pre-programmed correction and raising gain. If not explore variations in beta and density to generate improved response, while avoiding rotating mode limits. Could also look to diagnose impact by plasma rotation effect. | ||
| Background: | Real time error field measurement and control needs to be developed for future devices, particularly at low torque, which this targets. | ||
| Resource Requirements: | May be quick test as part of a wider campaign - given more time if proves promising. RWM feedback system needed. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 268: Identifying the origin of critical gradients | ||
| Name: | Hillesheim jon.hillesheim@ukaea.uk | Affiliation: | CCFE |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | W.A. Peebles, T.A. Carter, L. Schmitz, T.L. Rhodes, G. Wang | ITPA Joint Experiment : | No |
| Description: | It is thought that critical gradients in fusion experiments arise due to the microinstabilities ($k_{perp} rho_i sim 1$) that produce transport. The goal of the experiment will be to try identify the underlying instability by fluctuation measurements in two target plasmas, aiming to create one case that is ITG dominated and, if possible, one case that is TEM dominated. Presumably, if there is a different dominate instability, there will be a different critical gradient and stiffness level. The identification of the underlying instabilities will be attempted through a combination of measurements that are expected to produce different results if the underlying instability changes: the crossphase between temperature and density fluctuations, the ratios of relative temperature to density fluctuations, the slope of the dependence of density fluctuation amplitude on wavenumber, and by the direction of propagation of the turbulence. As a secondary goal, during the latter portion of each shot, profile stiffness can be quantified through modulation of ECH and gas puffs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use an ECH heated Ohmic plasma that has been shown to have significant TEM at some radii from previous experiments. For the ITG case, use the same plasma, but with balanced beams instead of ECH. This should minimize the number of changes made while still being enough to change the underlying instability. Pre-experiment TGLF/linear gyrokinetic calculations should be done to establish that a significant change in the underlying instability is expected--while it would be ideal to have one case with only ITG and one with only TEM, this is a difficult experimental proposition; to see a change in the fluctuation measurements a change in which is dominate should be sufficient. The higher fluctuation levels in L-mode, in comparison to H-mode, allow for a wider range of fluctuation measurements--some of the following measurements are not feasible in H-mode due to low fluctuation levels. For each case, the same set of fluctuation measurements will be acquired, including the crossphase between temperature and density fluctuations with coupled reflectometry/CECE, temperature fluctuations with CECE, density fluctuations with BES (use beam blips for TEM case), and scan the wavenumber to which DBS is sensitive.
The direction of propagation of the turbulence will be attempted to be measured locally for the first time in the core of a tokamak by using beam blips to send the rotation back and forth across zero. The center of the Doppler shifted peak detected with DBS depends on the lab frame velocity of the turbulence (ExB velocity plus the phase velocity of the turbulence), but the width of the peak is determined primarily by Doppler broadening and occurs when the background rotation is zero--it has been previously observed that the minimum peak width occurs at a different time than when the spectrum is centered at zero frequency, but not in a controlled experiment. Provided the data possesses sufficiently high fidelity, the phase velocity of the dominate microinstability may be able to be resolved. The first portion of each shot will be used to gather the complete set of fluctuation measurements, scanning DBS wavenumber shot-to-shot. The second will be used to apply a sequence of beam blips to force the rotation back and forth across zero. The third will be used to apply profile modulations to quantify stiffness. | ||
| Background: | Models of microinstabilities have often been invoked to explain critical gradient behavior and transport in tokamaks, but neither the ITG or TEM instabilities have been unambiguously identified through fluctuation measurements. | ||
| Resource Requirements: | 2 half days or 1 day. All beams, gryotrons. The proposed measurements may be able to be folded into other profile stiffness/critical gradient experiments. | ||
| Diagnostic Requirements: | All fluctuation and profile diagnostics. | ||
| Analysis Requirements: | Comparisons to model predictions from GYRO/TGLF. | ||
| Other Requirements: | |||
| Title | 269: Extract Transport Coefficients from well diagnosed DIII-D discharges | ||
| Name: | Elder david.elder42@gmail.com | Affiliation: | U of Toronto |
| Research Area: | Thermal Transport in the Boundry | Presentation time: | Requested |
| Co-Author(s): | To be determined
tentatively: P.C. Stangeby (U. Toronto), A.W. Leonard (GA), C.J. Lasnier (LLNL), M.A. Makowski (LLNL), J.A. Boedo (UCSD), B.D. Bray (GA), N.H. Brooks (GA), J.G. Watkins (SNL) |
ITPA Joint Experiment : | No |
| Description: | It is important to understand the physics underlying the particle and heat transport in the scrape off layer (SOL). The distribution of density and temperature in the scrape off layer is directly related to the particle and heat transport both along the field lines and perpendicular to the field lines. If a sufficiently accurate description of the SOL plasma can be developed and combined with knowledge of the sources it is possible to extract, as a function of radial position averaged over a specific flux surface, the effective cross field transport coefficients (Dperp, Xperp) contributing to the plasma solution.
This approach has already been applied using the OEDGE code with promising results [1] to a selection of discharges from the power width Joule milestone experiments in the 2010 campaign. However, in order to have sufficient confidence in the results for publication, additional data with improved diagnostics are required to further constrain the plasma solutions. During the recent torus opening, improvements have been made to both the Thomson scattering systems and the Langmuir probes. In addition, the previous experiments did not regularly include significant magnetic sweeping nor were they well positioned to optimize the divertor Thomson scattering data obtained. This proposal is to repeat selected discharges from the power width Joule Milestone experiments with sufficient improvement in diagnostic measurements and optimization of the plasma position and sweeping so that the confidence in the transport coefficients extracted from the plasma reconstruction process can be improved and these results published. [1] "Application of OEDGE to Transport Coefficient Extraction in DIII-D Joule Milestone Discharges", J.D. Elder et al., APS poster 2010 |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The proposed experiment would repeat the most important discharges from the SOL power width Joule Milestone experiments from previous campaigns exploiting upgrades in the diagnostic systems during the recent vent as well as magnetic sweeping to give substantial improvements in the reliability of the diagnostic data.
Each different plasma condition examined would require one dedicated discharge (preferably two identical) with sweeping of the outer strike point to optimize the data obtained from the divertor Thomson scattering system and Langmuir probes. The objective is to obtain as complete a profile as possible of the target and divertor plasma conditions from the Langmuir probes and divertor Thomson. Upstream conditions will be obtained from Thomson scattering measurements and reciprocating probes. These two sets of boundary conditions will be connected using SOL models in OEDGE and source terms calculated by EIRENE. Combining all of these data will yield flux surface averaged estimates of the transport coefficients in the outer SOL of these discharges. | ||
| Background: | Transport coefficients can be extracted from a simulated background plasma solution if the dominant sources and sinks are known.
The OEDGE code has already been applied to the extraction of transport coefficients for several different conditions from the Ip scan of the power-width Joule milestone discharges. The Xperp values were extracted over the power width profiles for five discharges in the previous campaign. The value of Xperp found was very similar for each plasma condition and varied within the range of 0.12+/-0.05 m2/s when averaged over the power width. It might seem surprising that similar values of Xperp are found until the situation is compared to the results from a very simple model (eq 1). [equation 1] Xp = Gamma_sheath * c_s * (lambda_q)^2 / (6 * Lconn) Gamma_sheath=sheath heat transmission coefficient C_s=sound speed at target Lconn=connection length=distance outer target to outer midplane Lambda_q=midplane power width Taking the experimentally measured quantities used in this equation, calculating a rough Xperp estimate and comparing these to the Xperp values obtained by extraction using OEDGE averaged over the power width for each condition gives the following comparison. Ip=1.48MA, Xp_estimated=0.10 , Xp_extracted=0.08 Ip=1.3MA, Xp_estimated=0.14, Xp_extracted=0.10 Ip=1.04MA, Xp_estimated=0.21, Xp_extracted=0.15 Ip=0.8MA, Xp_estimated=0.16, Xp_extracted=0.12 Ip=0.5MA, Xp_estimated=0.18, Xp_extracted=0.16 Two features are clear from this list. The rough estimate from the simple model agrees quite well with the values extracted from the code simulation. In addition, the estimated values have the same trends in terms of magnitude. It is not clear whether Xp has a dependency on Ip or whether Xperp should be considered constant over this range of plasma conditions. It would appear from this analysis that changes in the fundamental heat transport are not responsible for the changing heat deposition profile at the target as measured by infra red thermography. However, one factor that is changing in these cases is the connection length (defined here as the distance from the outer target to the outer midplane roughly averaged over the power width). Better input data to the modeling are required before any firm conclusions can be reached regarding the transport in these discharges. The results presented here are indicative of weak or no dependence of Xperp on the plasma current, leading to the conclusion that differences in target heat deposition may be due to factors other than changes in transport. It would be worthwhile to apply the OEDGE transport coefficient extraction process to better diagnosed plasmas similar to the ones initially examined. In addition, it would be useful to choose one or two other cases from the same Joule Milestone series where changes in Xperp might be expected (e.g. different B-field) | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | This experiment requires good data from the Thomson system, both upstream and divertor, Langmuir probes, and reciprocating probes. It also requires infra-red target measurements of the heat deposition profile. In addition, filterscope measurements of Dļ?” at the outer target, TTV measurements of hydrogenic emissions in the outer divertor, and other spectroscopic measurements (such as CIII) which can be used to constrain the plasma solution would be very valuable. | ||
| Analysis Requirements: | 1) OEDGE code analysis for plasma reconstruction and transport coefficient extraction.
2) Calculation of power deposition widths. | ||
| Other Requirements: | -- | ||
| Title | 270: Is high-k turbulence a significant contributor to the transport dynamics in the H-mode pedestal? | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Pedestal Structure | Presentation time: | Requested |
| Co-Author(s): | T.L. Rhodes, R. Groebner | ITPA Joint Experiment : | No |
| Description: | The goal is of this work is to determine if (1) high-k ETG scale fluctuations exist in the H-mode pedestal and (2) if they significantly contribute to pedestal transport and control the pedestal electron temperature gradient or pedestal structure. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In this experiment, a comprehensive study of high-k pedestal turbulence is proposed. The pedestal electron temperature gradient, and the ratio of electron temperature and density gradient scale length in the pedestal will be controlled by changing the line-averaged density and by adding off axis ECH power on select shots (0.7 < r/a < 0.9). The linear ETG mode critical gradient is expected to be close to R/L_ne for sufficiently short L_ne. Plasmas with reduced pedestal density gradient (as achieved in C-mod and in DIII-D with unfavorable grad_B drift direction) can be used to discriminate whether ETG-scale fluctuations depend on L_ne/L_Te. Pedestal gradients also depend on plasma shape (shape will be chosen in consultation with pedestal group based on expected response). In the experiment we will use a combination of low through high-k turbulence diagnostics to determine existence and parametric dependence of ETG scale fluctuations. The high-k scattering diagnostic can measure density fluctuations in the wavenumber range where the ETG growth rate typically peaks (k rho_s ~ 10-15 or k rho_e ~0.2-0.3) . Doppler Backscattering complements this capability by covering the intermediate TEM/ETG range (k rho_s <3) where ETG transport would be expected to peak. The expected ETG phase velocity (in the electron diamagnetic direction) is sufficiently large in the pedestal to be measurable by DBS. Independent high resolution CER measurements of the pedestal radial electric field would be required to extract the ETG phase velocity from the DBS data, potentially allowing direct ETG identification. | ||
| Background: | ETG scale fluctuations (Jenko, PoP 2001) are conjectured to be a significant contributor to the electron thermal flux in the pedestal region of H-modes. Linear stability calculations show increased ETG growth rates at the pedestal top and in the outer core plasma. Due to the increased growth rate and the strong ExB shear affecting or quenching low-k, ITG/TEM-scale turbulence, ETG to become the dominant instability to regulate electron transport and control the pedestal electron temperature gradient. However, no experimental evidence exists so far to support this conjecture. | ||
| Resource Requirements: | Beams, ECH for off-axis heating | ||
| Diagnostic Requirements: | All fluctuation diagnostics, Core/edge CER, main ion CER if available | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | Vertically Shifted plasmas may be considered to optimize diagnostic resolution and high-k measurements. | ||
| Title | 271: Active Error Correction by Locked Mode Feedback | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Locked modes tend to orientate themselves into the 'notch' created by the error field, though are also subject to viscous talks from the background plasma. We should explore whether this provides an effective way of optimizing the error field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create a benign locked mode in a high q95~5 plasma. Explore variations in I coil currents to try to shrink the mode and enable it to spin up. This might be done in feed-forward initially, or progress to feed-back. Low beta/zero torque or Ohmic regimes may be best to see minimize torques on the island. Modest beam torque or variations in density might be deployed to control the phase where the island unlocks, varying the sensitivity of the island to underlying error field. Negative NB torque might be explored to cancel the offset rotation, so the island most closely aligns to the intrinsic error phase. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | THIS ENTRY IS NOT TO LAY CLAIM TO THIS TOPIC, but to ensure relevant ideas go forward for consideration and joint planning by the group. I remain happy to help as team member or experiment leader as needed. | ||
| Title | 272: Collisionality dependence of turbulence and transport and tests of code predictions | ||
| Name: | Rhodes rhodes@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Staebler, Kinsey | ITPA Joint Experiment : | No |
| Description: | The predictions of turbulence and transport simulations will be tested and verified by utilizing a scan in collisionality keeping the most relevant dimensionless quantities fixed and also by vary-ing several of them in a controlled manner. If Thomson is sufficient will consider partial or full B field scan to keep most other non-dimensional parameters fixed or nearly fixed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A scan in L and H-mode plasma density will be used to vary the collisionality over a large range (~4 or more) while keeping most other relevant dimensionless quantities fixed. This approach is a fairly good match to the restrictions placed upon the magnetic field by the ECH and the various millimeter wave diagnostics (ECE, CECE, DBS, reflectometry). The target discharge will be a sawtooth free L-mode. Varying density will also vary plasma beta, however that has not been seen to be a strong controller of turbulence and transport. After varying density, the plasma tem-perature would be varied allowing both rho_* (=ion gyroradius calculated using the electron temperature / radius) and collisionality to vary simultaneously. This experiment will make full use of the multi-scale (ITG to ETG scales) and multi-field (Ʊ, Ttilde, flow) turbulence measure-ments that are now available on DIII-D (i.e. BES, CECE, DBS, FIR, high-k backscattering, PCI, reflectometry, correlation reflectometry/DBS). | ||
| Background: | Simulations predict strong variation in transport at low collisionality with fluctuation driven ther-mal fluxes increasing rapidly with decreasing nu*. Particle flux predicted to change sign at low collisionality. Test this prediction as well as collision model used in simulations. | ||
| Resource Requirements: | Machine Time: TBD
Number of Neutral Beam Sources: required ECH: required | ||
| Diagnostic Requirements: | All fluctuation and profile diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 273: Multi-scale, multi-field turbulence in reversed shear and ITB plasmas | ||
| Name: | Rhodes rhodes@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Doyle | ITPA Joint Experiment : | No |
| Description: | Measurements utilizing DIII-Dā??s unique array of multi-scale and multi-field fluctuation diagnostics will be carried out in the core of reversed shear and ITB plasmas. New and unique information will be obtained on the behavior of the fields in normal/reversed magnetic shear and ITB conditions. Importantly, information on potential sources of electron thermal transport in these conditions will be obtained. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Early beam injection will be used to control the magnetic shear profile. The formation of ITB plasmas will be of particular interest. | ||
| Background: | Reversed shear and internal transport barrier (ITB) plasmas often exhibit core regions of flat temperature and density profiles where the turbulent drive is expected to be low and yet the transport is high. We will probe the turbulence and transport response in these regions utilizing DIII-Dā??s unique measurement capability for profiles of low, intermediate, and high k density fluctuations, low k temperature, and fast turbulence flow velocity measurments. A particular focus will be on radial distribution and response of intermediate and high-k turbulence and low k temperature turbulence and their possible contribution to electron thermal transport. The effect of magnetic shear will also be investigated since in these plasmas q naturally varies in time. Since transport remains anomalous in the core of these plasmas these measurements will provide a broad wavenumber range with which to challenge non-linear turbulence simulations. All available fluctuation diagnostics together with CER and MSE will be utilized. | ||
| Resource Requirements: | Machine Time: 1
Number of Neutral Beam Sources: all required ECH: required | ||
| Diagnostic Requirements: | All fluctuation and profile diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 274: Optimizing Error Correction at High Beta | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | There is evidence that the plasma 'sees' the DIII-D intrinsic error somewhat differently at high beta, requiring higher error correction. This suggests that the mode through which the RMP couples to the plasma is changing, or that the coupling across the palsma itself may be varying. It is important to understand these issues of palsma response in roder to understand the action and correction/application of 3-D fields more generally. it is also highly desirable to minimize error fields in high beta plasma. Thus a method based on Reimerdes' torque scans is proposed to measure intrinsic error at high beta. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In a low torque high beta plasma, ramp up the I coils with different phasing to determine equivalent phase and amplitude of optimal correction. Test this correction with a torque ramp down shot. Repeat with neighboring I coil phasing (start with I-240, then do 180 and 300) to see optimal phasing is the same as low beta plasmas | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | THIS ENTRY IS NOT TO LAY CLAIM TO THIS TOPIC, but to ensure relevant ideas go forward for consideration and joint planning by the group. I remain happy to help as team member or experiment leader as needed. | ||
| Title | 275: Stability boundaries for off-axis fishbone | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | General SSI | Presentation time: | Requested |
| Co-Author(s): | Okabayashi, Matsunaga, Takechi | ITPA Joint Experiment : | No |
| Description: | The primary goal of this experiment is to determine ways to stabilize the off-axis fishbones observed in high betan plasmas with 2.0> qmin > 1.5. Variations in beam parameters are the primary tool. The data will also improve identification of the instability, which may assist in its avoidance. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Reestablish the instability (shot 141076).
2) Try for an unstable condition with slightly lower density & beam power. 3) Vary the angle of injection as much as possible at approximately constant betan. | ||
| Background: | Off-axis fishbones can trigger resistive wall modes in both DIII-D and JT-60U. We speculate that these modes are an energetic-particle branch of the external kink. If so, they should depend sensitively on the fast-ion distribution function. Off-axis injection with normal BT alters the trapped-particle population appreciably. | ||
| Resource Requirements: | All 8 sources; 150 beams tipped down.
Minimum 4 gyrotrons for ECCD. | ||
| Diagnostic Requirements: | EP diagnostics
ECE Active MHD spectroscopy | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 276: Starvation H-modes | ||
| Name: | Muller mullersh@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Not requested |
| Co-Author(s): | J.A. Boedo, K.H. Burrell, J.S. deGrassie, R.A. Moyer, D.L. Rudakov, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | It is well known that the plasma density after the L-H transition in the ELM-free phase continuously increases, which eventually results in ELMs when either the stored number of particles or the stored energy become too large. In the low-power ELM-free H-modes used in the 2010 Reynolds stress experiment (see also proposal 225), ELM-free phases of about 1 s were obtained, during which the plasma density rose by a factor of 3 without any gas puffing. Estimations suggest that the plasma is "sucking in" essentially all available neutrals from the SOL, ionizing and confining them, and that towards the end of the density rise, the neutrals in the SOL are already strongly depleted. It appears possible that the plasma can be kept in an ELM-free steady state when one minimizes the available number of particles in the vacuum vessel and simultaneously minimizes the heating power, such that sources at the divertor due to sputtering can be minimized.
The experiment aims at developing a "starvation H-mode" scenario, in which the plasma density can no longer rise since all neutrals in the SOL are depleted and part of the plasma. Due to zero cross-field transport in the ELM-free H-mode, sputtering from the main chamber walls can essentially be excluded as an additional particle source. Trying to minimize the energy input to 2 gyrotrons (no beams obviously), the sputtering in the divertor can probably also be prevented (with some luck and cryopumps). The simplified conditions of such a steady-state "starvation" ELM-free H-mode are extremely attractive for fundamental physics studies (transport studies, intrinsic rotation, neutral dynamics and sources, etc.), modeling and validation. Regardless of whether a formula for a steady-state ELM-free H-mode can be found, the attempt will almost certainly lead to ground-breaking insights on the neutral sources and dynamics, and the plasma ionization sources. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start off with very good vacuum conditions (probably skip the daily reference shot). Put in the minimum amount of gas for startup. Keep the plasma away from the wall in L-mode to reduce sputtering, or attempt to go to H-mode as early as possible during startup. Use only ECH, since beams would provide a core particle source, even if small. After the plasma goes to H-mode, attempt to maximize the length of the ELM-free period. See if conditions can be found such that the density rise bottoms out, and if the properties of the ELM-free H-mode can be maintained in steady-state.
Use the 2010 Reynolds stress experiment as a reference condition. | ||
| Background: | |||
| Resource Requirements: | Cryopumps, ECH | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 277: CQ deconfinement of REs using pulsed impurity injection | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | A. Loarte, E. Hollmann, P. Parks, S. Putvinski | ITPA Joint Experiment : | No |
| Description: | Attempt to deconfine seed REs with rapid impurity injection into CQ plasma edge | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create disrupting target plasma with large (~ 10 kA) RE seed by firing Ar killer pellet into low density IWL plasma. Then, during thermal CQ (before RE current becomes dominant), inject strong pulse of neon gas into plasma edge (with rupture disk or SPI) to try to deconfine RE seed population. Look for seed population loss to wall using HXR scintillators. | ||
| Background: | It has been proposed to deconfine RE seeds in ITER by using pulsed rupture disk bursts of gas fired into the CQ plasma, thus shrinking the current channel and destabilizing tearing modes. This scheme has not been tested experimentally, however. We propose to test this using the DIII-D two-shot rupture disk system with neon. If the rupture disk is not available, we would use neon in the SPI system. | ||
| Resource Requirements: | 1 run day. Small argon pellet injector. Rupture disk system. SPI system with neon. | ||
| Diagnostic Requirements: | BGO scintillator array, fplastic, fast camera, SPRED, SXR, interferometers. | ||
| Analysis Requirements: | None | ||
| Other Requirements: | None | ||
| Title | 278: Further exploration of NTV rotation dependence | ||
| Name: | Cole andrew@woodruffscientific.com | Affiliation: | Woodruff Scientific Incorporated |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Requested |
| Co-Author(s): | Callen, Hegna, Garofalo, Solomon | ITPA Joint Experiment : | No |
| Description: | Our approach can categorized in two thrusts: 1) revisit the low toroidal rotation peak discharges from 2009, such as 138574, and attempt to force pure superbanana-plateau dominated peak NTV and compare with peak measurements both from 2009, and perhaps new shots with higher ion collisionality... making the peak purely in the 1/nu regime. The goal would be a further confirmation of our understanding of NTV at low toroidal rotation. The challenge is to create a dB profile peaked towards the core versus the edge. 2) The second thrust would involve a broad sweep of the NTV torque vs rotation from nearly balanced to very co-rotating plasmas. There are two reasons for this goal. Firstly, it would be nice to have the capability to measure the torque profile versus rotation in a single shot, and secondly, at larger co rotation, it should be possible to measure the effects of electron NTV, a novelty in tokamaks and a direct connection to stellarator transport. This may involve heating electrons near the plasma edge to make electron transport worse. Away from the ion dominated peak, the electrons in the 1/nu regime can compete with the remaining ion TTMP torque. The ratio of electron NTV to ion-TTMP scales as
(Te/Ti)^2.5 x sqrt(m_e/m_i) omega_(bounce,ion)/ [q * nu_(eff,ion)]. Preliminary analysis indicates this should easily be larger than unity, ie, that electron 1/nu torques should dominate over ion TTMP away from the low rotation ion peak. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The first thrust of this proposal would follow closely with the approach used in 2009, i.e. nearly balanced NBI, with beam feedback. The challenge would be to peak the dB profile towards the core and then again towards the edge. The former might involve stable n=1 plasmas...? But this might be a stretch without locked modes at nearly balanced rotation. Ideas in this regard are welcome. The second thrust would involve either trying to develop the "quick scan method" of NTV torque measurement, and it that fails, going to Co-rotating plasmas, say from 10 - 30 krad/sec. and look for electron 1/nu NTV. | ||
| Background: | Space holder here, from 2009 RoF proposal
This is a continuation and extension of experiments started in FY08 (MPs 2008-02-01 and 2008-02-02) operating in reversed-Ip. We will use the C-coil for optimal correction of the n=1 error field, determined via DEFC. We will use odd-parity on the I-coil for an almost purely non-resonant n=3 field to apply braking. The braking is applied after all profiles have reached nearly stationary conditions. We will apply the braking field with a fast step and maintain it for at least a couple of momentum confinement times, before ramping down the field slowly, through a succession of torque equilibrium states, in order to carry out a full scan of the braking amplitude, and test torque dependence on the plasma rotation. Feedback control of the NBI power is required in order to maintain both beta and the torque constant during the application of the braking. We will carry out shot-to-shot scans of the NBI torque such that the initial toroidal plasma rotation prior to activation of the I-coil varies from 'slightly' co-rating to nearly balanced, and repeat the braking measurement mentioned above for each case. | ||
| Resource Requirements: | Same as shot 138574 to start with | ||
| Diagnostic Requirements: | All standard magnetics and internal profile diagnostics, including reflectometer for density profile measurements. FIR and BES fluctuation
measurements should also be acquired. | ||
| Analysis Requirements: | Analysis of the magnetics for extraction of the n=3 plasma response. Kinetic equilibrium reconstruction for accurate n=3 stability modeling. Time-dependent profile fitting for TRANSP modeling of the discharge evolution. | ||
| Other Requirements: | -- | ||
| Title | 279: ITER Baseline scenario operation with dominant electron heating | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | ITER will operate with dominant electron heating and Te~Ti. Previous DIII-D baseline discharges have operated with Te~Ti, but not with dominant electron heating. This is important for confinement and transport, as changing to dominant electron heating may change the dominant turbulence modes in the plasma | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Increase ECH heating fraction from about 50% (~2 MW) in previous low collisionality baseline plasmas to dominant, using full ECH power of ~4 MW. | ||
| Background: | Low collisionality baseline scenario plasmas operated in 2009 with 50% RF heating showed local changes in calculated turbulence and transport. Need to repeat with higher (dominant) electron heating fraction, and more central deposition. | ||
| Resource Requirements: | Full gyrotron compliment. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 280: Understanding plasma resonances to 3-D fields with *ELMing* plasmas | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | A critical element in understanding RMP effects on ELMs is the q95 resonance. This may imply particular effects associated with resonant surfaces, such as zero shear conditions, NTV effects, etc. Distinguishing and resolving physics of these effects would be assisted by better measurement of the degree of resonance (ie proximity to ELM suppression) as a function of q95. Some effects will link to rational q, or integer q surfaces, or have no particular q surface resonance of interest. Proximity to an ELM suppression effect may be gauged by ELM frequency, so with sub critical levels of I coil field for ELM suppression, sweeps in q95 could determine whether there are 1 or multiple resonances, and whether the resonances are evenly spaced (eg every 1/3 in q95). this will give a structure to the response vs 95, which may help inform us about which models best match the ELM suppression effect. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Execute q95 ramps with sub critical I coil fields to measure ELm frequency changes. One might repeat this with different plasma configurations (eg weak/strong shape/DND) or parameters (collisionality) in order to test for near resonances or variation in the resonances further - for example getting data to constrain models in regions where they do not lead to complete ELM suppression. Further, it would make sense to execute this for n=2 and n=3 (and possibly n=1) fields to see if spacing of any resonances changes (eg gaps of 0.5 in q95 with n=2, while 0.33 in n=3) | ||
| Background: | Key issues are expanding the available data basis to constrain models with novel new kinds of mechanisms. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 281: Improvement of Dynamic Error Field Correction(DEFC ) | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | Yongkyoon In, Piero Martin, Paolo Piovesan, L. Piron, E. Strait, and RWM Physics group | ITPA Joint Experiment : | No |
| Description: | For "3D effects" exploration and SSI target development, the improvement of DEFC should be very helpful to provide better error field correction.
Goal: (1st step), To Minimize the amplitude of n=1 RFA, simultaneously minimizing undesirable n=2,3 components (2nd step), To include the minimization of undesirable poloidal m-components. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1st step:
- Minimization of toroidal phase mismatching due to the rotation and finite betan between uncorrected error field and the applied correction - At present, in the process of dynamic error field correction, assuming the mode rigidity, we simply try to reduce the RFA without paying attention to the phase shift between uncorrected error field and the applied correction - As its consequence, we may be increasing n = 2, 3 components as well as undesirable poloidal m-components. - Prepare the PCS logic with the toroidal phase shift explicit including preprogrammed wave form (new F-matrix), rather than specifying three coil currents 2nd step: - Poloidal m components to be more precise by upper/lower independent operation with similar type of new F-matrix. Depending upon preliminary survey, we may need more audio amplifier purchase Target The suppression of RFA includes the fast transiently-excited RFA bubbles driven by various MHDs, like peeling-mode mode well below no-wall limit. This can be achieved with applying simultaneously the AC compensation developed by RFX group. - transiently-excited RFA/RMP in SSI - peeling-mode-driven RFA ( type 1 ELM?) condition in SSI | ||
| Background: | |||
| Resource Requirements: | In 2nd step : after preliminary survey, we may need more audio amplifier purchase | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 282: Experimental Tests of Thermodiffusive Particle Transport | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | G. McKee, C. Angioni, C. Holland, T. Rhodes, L. Schmitz, I.U. Uzun-Kaymak, G. Wang, A. White, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Vary collisionality in L-mode and H-mode plasmas and apply ECH (or FW) electron heating to test whether a thermodiffusive mechanism is operative and driving enhanced particle transport, which might explain ECH-driven particle pump-out. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a high collisionality (maximum density possible) L-mode discharge. Characterize fluctuations across mid-radii. Apply ECH to increase Te and Te-gradient and study turbulence and transport response. Repeat at low collisionality. Also repeat in a relatively high-q95 H_mode discharge, possibly hybrid (more specifics will follow).
BES measurements will be employed to look not just at turbulence characteristics, but to investigate turbulent transport by applying 2D velocimetry techniques to examine eddy trajectory (as pertaining to turbulent particle flux). Codes predictions that ion temperature fluctuations will respond differently, which will provide an excellent opportunity to test the newly redesigned UF-CHERS system (2011). | ||
| Background: | Experiments on ASDEX-U (Angioni, Nuclear Fusion-2004) suggest that a temperature gradient-driven thermodiffusive mechanism may be driving enhanced outward particle transport in low-collisionality, Trapped-Electron-Mode dominated plasmas, while at higher-collisionality, ITG dominated plasmas, this particle-drive mechanism is suppressed. Furthermore, there are simulations showing that the particle transport may be strongly wavenumber dependent [Angioni-ITPA-2008].
This topic is important for transport model validation, mode identification, and exploits new diagnostic capability available (8x8 BES) or under development (ion temperature fluctuations). A collisionality scan will also provide an excellent data set for testing zonal flow damping via ion-ion collisions [Z. Lin, PRL-1999]. | ||
| Resource Requirements: | ECH, FW, NBI | ||
| Diagnostic Requirements: | BES (8x8), UF-CHERS, CECE, DBS, FIR, PCI. | ||
| Analysis Requirements: | Velocimetry analysis of 2D BES fluctuation data. TGLF, GYRO. | ||
| Other Requirements: | |||
| Title | 283: Measurement of RE final loss power balance | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Requested |
| Co-Author(s): | A. Loarte, E. Hollmann, N. Eidietis | ITPA Joint Experiment : | No |
| Description: | Measure power energy deposited by RE beam striking the wall | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create large (200 kA+) RE beam using small Ar pellet injection into low-density IWL target plasma. Then, use control system to move RE beam down into lower divertor shelf at various speeds. By comparing divertor shelf leg current monitors, IR measurements and HXR signals, reconstruct power balance (magnetic vs kinetic energy) and timescales of final loss phase of RE beam into lower divertor shelf. | ||
| Background: | RE beams are believed to contain dominantly (75% + ) magnetic energy and less kinetic energy. However, recent experiments on JET indicate that during a RE beam/wall strike the magnetic energy is converted into toroidal electric fields and then into kinetic energy, so nearly 100% of the total RE beam energy goes into local tile heating (as opposed to ohmic heating of the conducting structure via wall currents). This result has not been checked on other machines yet. | ||
| Resource Requirements: | 1 run day. Small argon pellet injector. | ||
| Diagnostic Requirements: | BGO scintillator array, fplastic, fast camera, SPRED, SXR, interferometers, lower tile shelf leg current monitors, IR camera, fast filterscopes, CER spectrometers | ||
| Analysis Requirements: | halo current analysis | ||
| Other Requirements: | none | ||
| Title | 284: A DiMES test aimed at establishing if high-Z material net erosion << gross erosion | ||
| Name: | Stangeby peter.stangeby@utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | ITER First Wall Issues | Presentation time: | Requested |
| Co-Author(s): | TBD | ITPA Joint Experiment : | No |
| Description: | The net erosion rate at the divertor strike points of future high power devices such as ITER will not be acceptable unless net erosion is reduced substantially relative to gross erosion. Fortunately, just such a reduction is theoretically expected to occur due to two effects: 1) < A definitive test requires that both net and gross erosion rates be measured directly in the same experiment. Gross erosion rates can be measured from the intensity of sputtered neutral WI (400.8, 429.4, 498.2, and 505.3 nm) and MoI lines (550.6, 553.3, 557.0 nm). This requires reliable knowledge of inverse photon efficiencies, S/XB(Te), in order to convert spectroscopic intensities into atom influx densities. Very recently, S/XB values have been directly measured on PISCES for WI [Nishijima, Phys Plasmas 16 (2009) 122503] and MoI [Nishijima, J Phys Mol: At Mol Phys, in press], increasing the reliability of neutral influx measurements for these high-Z elements. The new PISCES SX/B values for WI differ by an order of magnitude from previous measurements, made 1997-2002, which is attributed in part to a ne-dependence of S/XB that the PISCES experiments uncovered. Net erosion rates are measured by using thin deposited metallic films, of order 10-100 nm, whose thickness is measured ex situ before and after exposure. The sample is prepared with a depth marker and thickness is measured using ion beam analysis [Wampler, J Nucl Mater 233-237 (1995) 791; Krieger, J Nucl Mater 241-243 (1997) 684]. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Use low density SAPP (Simple as Possible Plasma) attached divertor conditions [J Nucl Mater 313-316 (2003) 883] which are the best characterized divertor conditions achieved to date in DIII-D. L-mode, thus the complication of ELMs avoided. Particularly extensive Divertor Thomson and probe data. The sputtering of high-Z PFCs is primarily by low-Z plasma impurities and by self-sputtering; for low density SAPP we have especially good information on the dominant low-Z impurity, carbon, due to comprehensive CI, CII, CIII measurements. While for this experiment it is not essential to be able to calculate the gross sputtering rate, since it will be measured spectroscopically, it is important to get as complete as possible a handle on the entire situation since this could provide a clue to solving the puzzle. The plasma conditions at the OSP, Te ~ 20 eV, ne ~ 2e19 are similar to those employed successfully in earlier DiMES net erosion studies of W and Mo [Wampler, J Nucl Mater 233 (1995) 791; Brooks and Whyte, Nucl Fusion 39 (1999) 525; note that gross erosion was not measured in these earlier experiments; the DTS data were also too scattered to be usable]. Dennis Whyte notes that hotter, thus more eroding, conditions could be achieved by using RMP H-mode (no ELMs) discharges.
2. Use repeat shots of a single plasma condition. Since the C-mod measurements were campaign-integrated, it is hard to rule out that the net erosion could have occurred in off-normal conditions. 3. Use Jeff Brooks' advanced computational modeling code set for the interpretation of the net vs gross erosion - the same code set used for the C-mod Mo analysis. The< 4. Bill Wampler proposes to deposit 10 - 20 nm of W on a Mo substrate sample to install on DiMES. Based on previous DiMES experiments it should be possible to measure the net erosion due to ~ 10 shots (1/2-day experiment) similar to the low density SAPP ones. The W sample should be large, limited by the 5 cm diameter of DiMES, in order to create a net erosion situation, which requires a sample large compared with the mean free path for ionization of the sputtered W-atoms, which is of order 1 mm for low density SAPP conditions. Dmitry Rudakov advises that a 2.5 cm disk would be large enough for most of the eroded W to be re-deposited locally and that whatever is not deposited back on the disk will deposit on the graphite cap around it (in earlier DiMES experiments Wampler measured e-folding length of ~3 mm). Rudakov, also Karl Krieger, note that W coverage on the cap can be measured and compared with the decrease of the W film thickness which will provide information on how much tungsten migrates on a longer scale. It is also desirable that most of the eroded W redeposits on the cap, since tiles around DiMES will not be contaminated. The disk can be well aligned with the cap with almost no gap, so that leading edges should not be a problem. | ||
| Background: | In progressing from present fusion devices to ITER (reactors), the annual energy load, P-heat X T-annual will increase by ~ 3 (~5) orders of magnitude. If the crude assumption is made that total gross erosion scales with the energy load, then the rate of total gross erosion in ITER (reactors) could be ~3 (~ 5) orders of magnitude higher than in present devices. Erosion poses a number of serious problems including PFC wear, plasma contamination, dust production, mirror coating, disruptions due to exfoliated/spalled material, etc. Fortunately it is net, not gross, erosion that matters and the plasma conditions foreseen at the divertor targets of devices like ITER and FDF are such that it is theoretically expected that there will be strong suppression of net erosion relative to gross erosion due to strong redepostion of sputtered particles.
Although the theoretical ideas underlying these processes are basic and while there is also evidence for their validity, e.g., [Krieger, J Nucl Mater 266-269 (1999) 207], there is also recent indication that something important may be missing in our understanding of the physics that controls prompt local deposition: in C-Mod the measured net erosion rate of Mo at the outer strike point (campaign-integrated) is found to be about an order of magnitude larger than expected from simple basic considerations and also contrary to detailed WBC code modeling: < The C-mod measurements were campaign-integrated, making them difficult to interpret. As noted by Brooks, short exposure conditions are required using well characterized, repeat discharges to provide the interpretable data needed for a definitive test of the relation between net and gross erosion. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 285: Improving error field correction at high beta | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The proposed research seeks to address whether error field correction (EFC) in DIII-D can be improved for high beta and low rotation discharges. A large body of existing data from both stable and unstable discharges, with various applied EFC methods will be examined in order to establish what vacuum error field harmonics are most important, and to determine how to best apply currents in correction coils to address deleterious effects associated with error fields. The knowledge gained from this analysis will be used to prescribe and test improvements in EFC for high beta discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The SURFMN code will be used to quantify the dominant corrected and uncorrected error field harmonics in the existing data set. Analysis with a perturbed ideal MHD code, such as IPEC or MARS-F will then be performed to estimate the plasma response to the vacuum spectrum. The correction coil currents needed to mitigate the effects of the most important error field harmonics can then be determined, and compared with the applied correction. The simultaneous operation of both the I and C coil sets to optimize the correction spectrum will be considered. | ||
| Background: | The dominant vacuum error field sources in DIII-D are small tilts and shifts with respect to axisymmetry of the poloidal and toroidal field coils (F-coils, B-coils), and central solenoid (E-coil) [J. L Luxon, et al., Nucl Fusion 43 1813 (2003)]. These misalignments are represened in the SURFMN code. Variations on two basic approaches to error field correction are presently in use on DIII-D. In the first, which is usually called "standard" or "pre-programmed" EFC, correction currents are determined based on known equilibrium coil currents and the known important resonant vacuum error field components [J. T. Scoville and R. J. La Haye, Nucl Fusion 43, 250 (2003)]. The resonant components of interest were determined from low-density locked mode thresholds. In the second approach, referred to as "dynamic" correction (DEFC), correction coil currents are determined by slow time constant feedback on the plasma response measured with external magnetic sensors [A. M. Garofalo, et al., Nucl Fusion 42 1335 (2002)]. At high beta, the required correction current amplitude for maximizing plasma rotation, or the amplitude found by DEFC is roughly a factor of 1.5 higher than that determined by consideration of low-density locked mode thresholds. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | SURFMN analysis, perturbed ideal MHD simulations (eg IPEC or MARS-F). | ||
| Other Requirements: | Also to be submitted to RWM Physics, General SSI, Error field and TBM mockup effects | ||
| Title | 286: Improving error field correction at high beta (Dup 285) | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | General SSI | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The proposed research seeks to address whether error field correction (EFC) in DIII-D can be improved for high beta and low rotation discharges. A large body of existing data from both stable and unstable discharges, with various applied EFC methods will be examined in order to establish what vacuum error field harmonics are most important, and to determine how to best apply currents in correction coils to address deleterious effects associated with error fields. The knowledge gained from this analysis will be used to prescribe and test improvements in EFC for high beta discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The SURFMN code will be used to quantify the dominant corrected and uncorrected error field harmonics in the existing data set. Analysis with a perturbed ideal MHD code, such as IPEC or MARS-F will then be performed to estimate the plasma response to the vacuum spectrum. The correction coil currents needed to mitigate the effects of the most important error field harmonics can then be determined, and compared with the applied correction. The simultaneous operation of both the I and C coil sets to optimize the correction spectrum will be considered. | ||
| Background: | The dominant vacuum error field sources in DIII-D are small tilts and shifts with respect to axisymmetry of the poloidal and toroidal field coils (F-coils, B-coils), and central solenoid (E-coil) [J. L Luxon, et al., Nucl Fusion 43 1813 (2003)]. These misalignments are represened in the SURFMN code. Variations on two basic approaches to error field correction are presently in use on DIII-D. In the first, which is usually called "standard" or "pre-programmed" EFC, correction currents are determined based on known equilibrium coil currents and the known important resonant vacuum error field components [J. T. Scoville and R. J. La Haye, Nucl Fusion 43, 250 (2003)]. The resonant components of interest were determined from low-density locked mode thresholds. In the second approach, referred to as "dynamic" correction (DEFC), correction coil currents are determined by slow time constant feedback on the plasma response measured with external magnetic sensors [A. M. Garofalo, et al., Nucl Fusion 42 1335 (2002)]. At high beta, the required correction current amplitude for maximizing plasma rotation, or the amplitude found by DEFC is roughly a factor of 1.5 higher than that determined by consideration of low-density locked mode thresholds. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | SURFMN analysis, perturbed ideal MHD simulations (eg IPEC or MARS-F). | ||
| Other Requirements: | Also to be submitted to RWM Physics, General SSI, Error field and TBM mockup effects | ||
| Title | 287: Improving error field correction at high beta (Dup 285) | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The proposed research seeks to address whether error field correction (EFC) in DIII-D can be improved for high beta and low rotation discharges. A large body of existing data from both stable and unstable discharges, with various applied EFC methods will be examined in order to establish what vacuum error field harmonics are most important, and to determine how to best apply currents in correction coils to address deleterious effects associated with error fields. The knowledge gained from this analysis will be used to prescribe and test improvements in EFC for high beta discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The SURFMN code will be used to quantify the dominant corrected and uncorrected error field harmonics in the existing data set. Analysis with a perturbed ideal MHD code, such as IPEC or MARS-F will then be performed to estimate the plasma response to the vacuum spectrum. The correction coil currents needed to mitigate the effects of the most important error field harmonics can then be determined, and compared with the applied correction. The simultaneous operation of both the I and C coil sets to optimize the correction spectrum will be considered. | ||
| Background: | The dominant vacuum error field sources in DIII-D are small tilts and shifts with respect to axisymmetry of the poloidal and toroidal field coils (F-coils, B-coils), and central solenoid (E-coil) [J. L Luxon, et al., Nucl Fusion 43 1813 (2003)]. These misalignments are represened in the SURFMN code. Variations on two basic approaches to error field correction are presently in use on DIII-D. In the first, which is usually called "standard" or "pre-programmed" EFC, correction currents are determined based on known equilibrium coil currents and the known important resonant vacuum error field components [J. T. Scoville and R. J. La Haye, Nucl Fusion 43, 250 (2003)]. The resonant components of interest were determined from low-density locked mode thresholds. In the second approach, referred to as "dynamic" correction (DEFC), correction coil currents are determined by slow time constant feedback on the plasma response measured with external magnetic sensors [A. M. Garofalo, et al., Nucl Fusion 42 1335 (2002)]. At high beta, the required correction current amplitude for maximizing plasma rotation, or the amplitude found by DEFC is roughly a factor of 1.5 higher than that determined by consideration of low-density locked mode thresholds. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | SURFMN analysis, perturbed ideal MHD simulations (eg IPEC or MARS-F). | ||
| Other Requirements: | Also to be submitted to RWM Physics, General SSI, Error field and TBM mockup effects | ||
| Title | 288: Scaling of Intermittency/turbulence | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Thermal Transport in the Boundry | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov, D. Elder, C. Lasnier, J. Watkins, R. Moyer, boundary group | ITPA Joint Experiment : | No |
| Description: | Scaling of turbulence with a few plasma parameters (pressure, Ip) will be carried out to establish a basic scaling of intermittency that can be compared to turbulence codes and heat footprint in the divertor. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Change Ip and pressure and measure decay length of Ne, Te in the divertor and midplane and Lambda_q in the divertor. | ||
| Background: | Previous work established dependence of turbulent flux with density and Ip, Since then theoretical work has been performed that can explain the dependence based on fundamental principles, so need to check those. | ||
| Resource Requirements: | ECH, NBI | ||
| Diagnostic Requirements: | ALl edge diagnostics, turbulence diagnostics | ||
| Analysis Requirements: | UEDGE/SOLPS analysis and turbulence analysis (Lodestar/BOUT) | ||
| Other Requirements: | |||
| Title | 289: Measurement of campaign-integrated carbon erosion and co-deposition in DIII-D | ||
| Name: | Stangeby peter.stangeby@utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Requested |
| Co-Author(s): | TBD | ITPA Joint Experiment : | No |
| Description: | The principle draw-back to the use of graphite PFCs in ITER is the, apparently, high rate of tritium retention due to co-deposition. The most exhaustive assessment carried out to date of D-retention in an all-carbon tokamak is the DITS (Deuterium Inventory in Tore Supra) project, see Background. DITS has identified a major discrepancy: < The last time systematic post mortem measurements were made in DIII-D was 2 decades ago: <>, CPC Wong, et al, J Nucl Mater 196-198 (1992) 871. Since that time, the alignment of tiles in DIII-D has been improved markedly; pre-2006 the floor tiles had normal spacing of 2-4 mm, now reduced to 0.6 mm, with tile-to-tile steps less than 0.1 mm. It is possible that a significant source of carbon creating C co-deposits comes from plasma machining of exposed tile edges, a process that will eventually saturate but which may be a significant source of C in most operating tokamaks today. The qualitative impression reported from recent visual inspections inside DIII-D is that it is hard to find much deposition, and what is found is typically only a few microns thick. By contrast the carbon co-deposits in JET and TS range up to ~ 0.5 mm. The different situation in the present DIII-D may be due to its exceptional tile alignment. Jim Davis has made the following rough estimate of the rate of C deposition in DIII-D: < |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The proposed experiment would repeat the 1989 Wong experiment to measure the rate of carbon net erosion, carbon deposition and D co-deposition in an all-carbon PFC tokamak with excellent tile alignment, DIII-D. Freshly-cleaned tiles would be installed at a representative set of poloidal locations, including inside the cryopump duct entrances. A few toroidal locations would also be assessed to check for toroidal symmetry. The tiles would be exposed for the 2011 campaign and removed at the end of the campaign for ex situ analysis. | ||
| Background: | Graphite is almost an ideal PFC material for a device like ITER where neutron damage is not an issue. Its only significant draw-back is the, apparently, high rate of tritium retention due to co-deposition. Co-deposition is a non-saturating process, in contrast with all other know processes of hydrogen uptake by solid materials.
On most tokamaks particle balance experiments have been performed showing that a large fraction of the D input to the torus for a discharge, 50% say, remains in the vessel at the end of the shot. This has been reported for both carbon PFC tokamaks and also for the Mo PFC C-mod tokamak. It is not clear what fraction of this D retention is due to saturable processes, often termed dynamic retention, and what fraction is due to co-deposition. It is only by carrying out post mortem analysis of the co-deposits on tiles removed from the tokamak that the amount of D-retention by co-deposition can be identified with any certainty. The most systematic study to date of D retention in a tokamak is the Tore Supra DITS project (Deuterium Inventory in Tore Supra) where two weeks of repeat, dedicated discharges were used: <>, B Pegourie, et al, J Nucl Mater 390-391 (2009) 550. Since the TS discharges are very long, ~ 400 s, it had been thought possible to use particle balance analysis to identify the separate role played by dynamic retention and by co-deposition, see Fig. 1 in T Loarer, et al, Nucl Fusion 47 (2007) 1112. However, the amount of D-retention assigned to carbon co-deposition by this approach is significantly discrepant with what is then measured post mortem in the tiles: the post mortem analysis finds only 10 to 40% of the D in the carbon codeposits that particle balance indicated should be there: < The suggestion in (2) is that there are dynamic retention wall reservoirs with very long time constants, longer than the TS discharge times, and therefore what appears from particle balance to be co-deposition retention is actually retention by (ultimately) saturable retention processes. This would seem to confirm that the only reliable way to measure the rate of D(T) retention by co-deposition is by post mortem analysis of the co-deposits. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 290: Demonstrate ELM Suppression in ITER Baseline Scenario Discharges | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | T. Evans, R. Moyer | ITPA Joint Experiment : | No |
| Description: | While RMP ELM suppression has been demonstrated in ITER-like conditions on DIII-D, it has not been applied to the ITER-demonstration baseline scenario discharges, which provide a much closer match to expected ITER conditions than previously. These plasmas provide a unique test bed, where RMP ELM suppression can be attempted at the ITER q_95, at the ITER operating beta_n, at high and low collisionality, with very large or "normal" sized ELMs, and operating with P~P_LH, which is a sensitive operating point. Success in ELm mitigation in these plasma conditions would increase confidence in success on ITER itself. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Vary operating q_95 of the ITER baseline scenario discharges and map out operating window in which ELM suppression is obtained. | ||
| Background: | ELM suppression on DIII-D has been demonstrated at q_95 down to 3.2, while the ITER baseline scenario plasmas operate at q_95 ~ 3.1. Thus, we should be able to apply the RMP system to the ITER baseline discharges. | ||
| Resource Requirements: | LSN plasmas, ITER shape, RMP coils | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 291: Demonstrate ELM Suppression in ITER Baseline Scenario (Dup 290) | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans, R. Moyer | ITPA Joint Experiment : | No |
| Description: | While RMP ELM suppression has been demonstrated in ITER-like conditions on DIII-D, it has not been applied to the ITER-demonstration baseline scenario discharges, which provide a much closer match to expected ITER conditions than previously. These plasmas provide a unique test bed, where RMP ELM suppression can be attempted at the ITER q_95, at the ITER operating beta_n, at high and low collisionality, with very large or "normal" sized ELMs, and operating with P~P_LH, which is a sensitive operating point. Success in ELm mitigation in these plasma conditions would increase confidence in success on ITER itself. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Vary operating q_95 of the ITER baseline scenario discharges and map out operating window in which ELM suppression is obtained. | ||
| Background: | ELM suppression on DIII-D has been demonstrated at q_95 down to 3.2, while the ITER baseline scenario plasmas operate at q_95 ~ 3.1. Thus, we should be able to apply the RMP system to the ITER baseline discharges. | ||
| Resource Requirements: | LSN plasmas, ITER shape, RMP coils | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 292: RWM stability control using NBI feedback | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Requested |
| Co-Author(s): | M. J. Lanctot, H. Reimerdes | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to extend preliminary results in controlling the plasma response to an applied magnetic perturbation using feedback modulation of the neutral beams injected power. In a previous experiment in 2010, the control of the plasma response was demonstrated over a range of normalized beta, between 1.1 and 1.9, below the no-wall limit. Once it is demonstrated and optimized above the no-wall limit, this control technique could become a useful tool for avoiding disruptions in high performance discharges and for validating kinetic resistive wall mode (RWM) stability models. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Slowly rotating n=1 I-coil waveforms, in the range of 10-20 Hz, will be used to obtain plasma response measurements. Neutral beam power feedback using the toroidal amplitude of the plasma response as a feedback value will be used. A linear ramp in the plasma response target will be used to map out the range of achievable plasma response and beta values with the available beam power. Step function plasma response target waveforms will then be used to aid in optimizing gain settings and validating models for the feedback dynamics. | ||
| Background: | The response of a stable plasma to applied, low-n magnetic perturbations has been shown to be a sensitive indicator of RWM stability. This measurement technique is feasible above the ideal MHD no-wall limit as long as the plasma remains stable. Equilibrium control algorithms that use the plasma response as an input parameter may be able to facilitate high performance operation on the cusp of marginal stability. Control of the plasma response using NBI power feedback was demonstrated on DIII-D in 2010, but normalized beta values above the no-wall limit were not reached. | ||
| Resource Requirements: | At least 5 co-Ip beam sources | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | Also submitted to RWM physics including rotation dependence, Stability. | ||
| Title | 293: RWM stability control using NBI feedback (Dup 292) | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | M. J. Lanctot, H. Reimerdes | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to extend preliminary results in controlling the plasma response to an applied magnetic perturbation using feedback modulation of the neutral beams injected power. In a previous experiment in 2010, the control of the plasma response was demonstrated over a range of normalized beta, between 1.1 and 1.9, below the no-wall limit. Once it is demonstrated and optimized above the no-wall limit, this control technique could become a useful tool for avoiding disruptions in high performance discharges and for validating kinetic resistive wall mode (RWM) stability models. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Slowly rotating n=1 I-coil waveforms, in the range of 10-20 Hz, will be used to obtain plasma response measurements. Neutral beam power feedback using the toroidal amplitude of the plasma response as a feedback value will be used. A linear ramp in the plasma response target will be used to map out the range of achievable plasma response and beta values with the available beam power. Step function plasma response target waveforms will then be used to aid in optimizing gain settings and validating models for the feedback dynamics. | ||
| Background: | The response of a stable plasma to applied, low-n magnetic perturbations has been shown to be a sensitive indicator of RWM stability. This measurement technique is feasible above the ideal MHD no-wall limit as long as the plasma remains stable. Equilibrium control algorithms that use the plasma response as an input parameter may be able to facilitate high performance operation on the cusp of marginal stability. Control of the plasma response using NBI power feedback was demonstrated on DIII-D in 2010, but normalized beta values above the no-wall limit were not reached. | ||
| Resource Requirements: | At least 5 co-Ip beam sources | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | Also submitted to RWM physics including rotation dependence, Stability. | ||
| Title | 294: Can the Non-Resonant RMP/Puff and Pump be Effective in Divertor Heat Flux Reduction? | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | M. Fenstermacher, S. Mordijck, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | A direct comparison of resonant (q95=3.5) - and non-resonant (q95=4.5) cases will establish whether or not resonance is absolutely necessary for successful ELM mitigation during puff and pump operation. The even parity I-coil configuration is used. Comparisons between the resonant and the non-resonant cases include (1) a scan in I-coil current, (2) a density scan with fixed I-coil current, and (3) an I-coil scan at fixed density. This experiment is focused on determining whether or not it makes a difference if the plasma is in resonance or not with regard to transient and between-ELM heat flux at the divertor targets after ELMs return during resonant RMP/radiating divertor operation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | High performance discharges, e.g., hybrid, are the preferred starting-point plasmas for this comparison, although standard high confinement ELMing H-modes are also acceptable. The base-case plasmas are long-pulse with Ip-flattop time ~5 s (minimum) fixed 1.2 MA. The direction of the ion grad-B drift is toward the lower SN X-point. Deuterium injection is done from the top of the vessel and particle pumping is done at the lower divertor outer target. The approach is straightforward, as described below:
The first step is to verify ELM suppression window for the resonance case (q95=3.5) with Bt=-1.8 T, and no ELM suppression at the non-resonance case (q95=4.5) with Bt=-1.94 T. Then: I-coil scan: Assess transient and steady state heat flux at the divertor targets at low-, medium-, and high values of I-coil current for q95 = 3.5 (resonant) and 4.5 (non-resonant). In principle, the I-coil scan at fixed q95 can be done within a single shot. Sample plots: H98, nped, Qp and Qp,ave versus I-coil current. I-coil scan at fixed density: Assess transient and steady state heat flux at the divertor targets at low-, medium-, and high values of I-coil current for RMP cases at fixed nped for q95 = 3.5 and 4.5. Target nped should reflect pre-RMP value and density-feedback mode should be used. In principle, the I-coil scan at fixed q95 can be done within a single shot. Sample plots: H98, Qp and Qp,ave versus I-coil current. Density scan with fixed I-coil: Assess transient heat flux Qp and steady state heat flux Qp,ave at the divertor targets, over the widest range in density (and collisionality) for RMP with q95 = 3.5 and 4.5. Use most promising I-coil current value, as established in (1) and (2) above. In principle, density scans for a give q95 can be done over 2-3 shots. Sample plots: H98, Qp and Qp,ave versus nped. | ||
| Background: | Recent studies at DIII-D have provided a first step toward evaluating the response of ELMing H-mode plasmas to applied RMP over wide ranges in density, pedestal collisionality, and maximum pressure gradient in the pedestal. We found that the RMP/radiating divertor plasmas examined had a relatively narrow operating window in density and q95 that provided complete ELM suppression. On the other hand, significant ELM mitigation using RMP under radiating divertor conditions was shown to provide a much broader window to work from and provide significant reduction in both the transient (ELM-generated) component and the steady-state component of the heat flux at the divertor targets. These results strongly suggest that a radiating divertor in the presence of RMP may provide a effective way of running a radiating divertor at lower core density than traditional radiating divertor approaches, even if the q95 chosen is not in resonance. | ||
| Resource Requirements: | Seven beam sources. This is a 0.5 day experiment. | ||
| Diagnostic Requirements: | IR camera, CER, Thomson scattering, SPRED, bolometers, CO2 interferometers. | ||
| Analysis Requirements: | ONETWO, ELITE, UEDGE | ||
| Other Requirements: | -- | ||
| Title | 295: Doublet Discharges with non-inductive EC startup | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | J.S. deGrassie, A. Turnbull, H. Reimerdes | ITPA Joint Experiment : | No |
| Description: | Obtain Doublet Discharges in DIII-D and compare stability, beta, and transport to "conventional" discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Extend Doublet duration from 140857. First apply Feedforward control and scan Vloop and F6/F7 current ratio. Develop feedback algorithms to heat both Doublet lobes and maintain Iup/Idown ~ 1. Characterize the stability and transport. Increase kappa > 1. | ||
| Background: | The present DIII-D tokamak began it's life 32 years ago with the mission to study Doublet plasmas. That work was limited by a primitive feedback system and lack of power supplies such that the "droplet" instability could not be controlled. Improvements in plasma control, a recently discovered startup scheme, and improved diagnostics suggest that doublets can be revisited in the present device. | ||
| Resource Requirements: | Standard resources, off-axis NBI (lower doublet lobe) and ECH (upper doublet lobe) | ||
| Diagnostic Requirements: | Any off-axis diagnostics are especially needed, e.g. ECE imaging, fast camera viewing CIII. | ||
| Analysis Requirements: | EFIT capable of fitting Doublets, 2D Abel inversions of fast camera data, JFIT generated g files. | ||
| Other Requirements: | Preparatory work to specify requirements for F-coil currents (feedforward) and PCS feedback algorithms to maintain doublets | ||
| Title | 296: Differences in Impurity Accumulation Between Resonant and Non-resonant RMP Plasmas | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | M. Fenstermacher, S. Mordijck, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | A direct comparison of resonant (q95=3.5) - and non-resonant (q95=4.5) cases for plasmas with applied RMP will establish whether or not their respective impurity ion transport is measurably different. The even parity I-coil configuration is used. The main comparison test involves density scans at different deuterium injection rates at a fixed impurity (argon) injection rate. This experiment will determine whether the buildup of injected impurity in the main plasma has any dependence on whether the plasma is set up in an RMP resonant or an RMP non-resonant state. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | High performance discharges, e.g., hybrid, are the preferred starting-point plasmas for this comparison, although standard high confinement ELMing H-modes are also acceptable. The base-case plasmas are long-pulse with Ip-flattop time ~5 s (minimum) fixed 1.2 MA. The direction of the ion grad-B drift is toward the lower SN X-point. Deuterium injection is done from the top of the vessel and argon impurities are injected into the lower divertor private flux region. Particle pumping is done at the outer divertor target. The approach is straightforward. There are three values of I-coil current (0, 3, and 6 kA) and three gas puff rates (0, 40, and 80 torr l/s) to be considered for each of the two q95 cases (3.5 (resonant) and 4.5 (non-resonant)). | ||
| Background: | In recent experiments, we compared impurity buildup in the main plasma for a control case without RMP with two cases at high and intermediate I-coil currents. ELMs were completely suppressed at both I-coil currents. During the H-mode phase of the discharges, argon was injected at a near-trace level but no deuterium was puffed. For these shots, the buildup in argon density was 20%-25% higher for plasmas with the I coil activated than without. There was little difference in argon accumulation between the intermediate and high I-coil cases. We also showed that argon accumulation inside the main plasma as a function of the deuterium injection rate for RMP-resonant and the corresponding non-RMP ELMĀ¬ing H-mode plasmas. For both RMP and non-RMP cases, the concentration of argon in the main plasma trended downward with increasing the deuterium injection rate. The return of Type-1 ELMing activity at higher pedestal density (and pedestal collisionality) may be responsiĀ¬ble for the simiĀ¬larity in argon impurity accuĀ¬mulation in the main plasma between RMP-resonant and non-RMP disĀ¬charges. Our analysis suggests that these RMP resonant results for impurity transport at higher density would carry over to RMP non-resonant cases. This is important to establish since ELM mitigation and heat flux reduction with RMP under non-resonant conditions provide considerable flexibility in selecting plasma parameters. | ||
| Resource Requirements: | Seven beam sources. This is a 1.0 day experiment. | ||
| Diagnostic Requirements: | IR camera, CER, Thomson scattering, SPRED, bolometers, CO2 interferometers, and SPRED. | ||
| Analysis Requirements: | MIST, ONETWO | ||
| Other Requirements: | -- | ||
| Title | 297: Excitation and control of the Edge Harmonic Oscillation using ICRF beat waves | ||
| Name: | Carter tcarter@physics.ucla.edu | Affiliation: | UC, Los Angeles |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | W.A. Peebles, T. Rhodes, L. Schmitz (UCLA); P.B. Snyder, V. Chan (GA); R. Goldston (PPPL) | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to explore the potential use of ICRF
beat waves to nonlinearly excite and control low-frequency modes in the pedestal, in particular the Edge Harmonic Oscillation (EHO). The ability to excite the EHO outside of the parameter regime where it is usually active may enable the extension of the QH mode regime to a wider range of operational scenarios. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | A single ICRF antenna would be used to simultaneously broadcast at two
closely spaced frequencies (equivalent to feeding it with a modulated RF signal). The antenna located at 285-300 degrees would be used, and two frequencies, e.g. 60MHz and 60MHz + $f_{rm IF}$, where $f_{rm IF}$ is the frequency of the EHO, would be feed into the amplifier chain driving the antenna. It may also be possible to use amplitude modulation at low frequency. An important part of the proposed experiment would be establishing that a nonlinear response of non-neglible amplitude can be driven in DIII-D using two ICRF waves with closely-spaced frequencies. The frequency of separation would be scanned, looking for response in turbulence diagnostics (BES, DBS, Reflectometry, FIR scattering, etc) at the beat frequency. The size of the modulation in the ICRF could be varied as well as its frequency. Experiments in QH mode plasmas with an naturally occuring EHO would be performed to look for resonant drive of the EHO. Matching the wavenumber characteristics of the beat-driven response to the EHO would be optimal (e.g. having $n=3$ content in the ICRF and the driven beat response). If resonant drive/modification of a naturally occuring EHO is observed, the next step would be to see if the resonant drive can be used to persistently drive the EHO through a plasma parameter change that would typically cause a QH to H-mode transition. These three goals ((1) establish nonlinear response to the beat wave, (2) drive an existing EHO, (3) make the EHO persist through a QH- to H-mode transition) could be accomplished using discharges with an L-mode phase followed by a transition to QH-mode and then to H-mode (e.g. discharges like those used by Schmitz, et al. in Phys. Rev. Lett. 100, 035002 (2008), or as shown in Fig. 13 of Burrel, et al. Phys. Plasmas 12, 056121 (2005)). It is likely that the nonlinear esponse to the modulation/beating would be spatially localized, and care would be taken to develop an experimental plan to allow variation in the localization of the effect. For example, the response might be enhanced near locations where the ICRF is being absorbed. On the other hand, strong absorption could generate fast ions that could complicate the experiment, and weak damping would enhance the effective cavity Q for the fast waves. In planning the experiment, having the ability to vary the degree and location of absorption is therefore important. If active excitation of the EHO can be acheived using ICRF beat waves, it is possible that particle transport in the pedestal might be controllable, potentially reducing or eliminating ELMs in H-mode discharges. | ||
| Background: | ICRF beat waves have been successfully used to nonlinearly excite
Alfven Eigenmodes in JET [A. Fasoli, et al., Nucl. Fusion 36, 258 (1996)] and ASDEX-U [K. Sassenberg, et al., Nucl. Fusion 50, 052003 (2010)]. We propose to establish the capability to drive ICRF beat waves in DIII-D. We additionally propose a novel application for ICRF beat waves: for interacting with and controlling low frequency modes, in particular the Edge Harmonic Oscillation. It might be possible to modify and perhaps control this mode and the associated transport using this technique. This proposal is motivated in part by recent experiments in LAPD. It was found that co-propagating kinetic shear Alfv'{e}n waves can drive a surprisingly large nonlinear response at their beat frequency; this response in turn scatters the pump waves to generate an extended series of sidebands [T.A. Carter, et al., PRL 100, 155001 (2006)]. Recently we have explored interaction of this beat response with gradient driven instabilities [D.W. Auerbach, et al., PRL 105, 135005 (2010)]. Field-aligned density depletions are created in LAPD by selectively blocking plasma production using an obstruction; unstable drift waves grow around these depletions. An experiment was conducted in which two co-propagating KAWs are broadcast along this striation with varying frequency separation. Resonant drive of the instability is observed when the frequency and parallel wavenumber of the beat wave matches the unstable mode. The instability is controlled (suppressed) when nearby damped modes are excited by the beat-wave drive. Similar effects have been observed by using an external array of electrodes to directly excite drift-waves in a laboratory plasma [Klinger, et al., PRL 86, 5711 (2001)]. We are interested in exploring the ability to nonlinearly excite and control a range of modes, including the EHO, TAEs, drift-type modes, etc. | ||
| Resource Requirements: | We will need ICRF, potentially at full power, using the 285/300
antenna, with the capability to feed two closely spaced frequencies to the antenna. | ||
| Diagnostic Requirements: | It is desirable to have the full suite of fluctuation diagnostics to
diagnose the plasma response at the beat frequency: reflectometers, DBS, BES, scattering, PCI, etc. Diagnosing the RF fields themselves (reflectometry? PCI?) would also be desirable but not required. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 298: ELM Suppression in H-Mode Plasmas with RMP in Odd Parity | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | M. Fenstermacher, S. Mordijck, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | Plasma conditions delimiting ELM suppression in an odd parity I-coil configuration, such as collisionality-, density-, and pressure gradient in the pedestal, are compared with those in an even parity I-coil configuration. Data and analysis for the even parity cases are already completed. All that is now necessary is to run a relatively small number of comparative shots in odd parity. The main scan involves variations in pedestal density obtained at different deuterium injection rates. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For the odd parity configuration, set I-coil current to 5.8 kA, in order to match up with the even parity cases already analyzed (i.e., shots 138541-51, (2010)). The target H-mode plasmas will have the same working parameters as those same shots, e.g., Bt = -1.78 T and Pinj = 6 MW. An Ip scan identical to the one from shot 138538 is done to establish the ELM suppression window, which can then be compared with the q95-ELM suppression window from the 138538 even parity case. Based on some earlier data, we expect this ELM suppression window to be q95 = 3.4 to 4.0. Next, set q95 to fall in the middle of this range in ELM suppression and change the density and collisionality via gas puffing. The methodology of the RMP- and gas puffing timing during the shot is identical to types used in shots 138541-51. Gas puff rates of 0, 40-, and 80 torr l/s are used to vary density and collisionality, and comparisons made with the corresponding even parity cases. | ||
| Background: | By far, the lions share of the previous studies of RMP ELM suppression and mitigation have used even parity. While doing so has been successful in many respects, the cost of applying RMP in even parity has typically been in lower energy confinement and significant density pump-out of the main plasma. Over the years, a very modest amount of work has been directed toward the odd parity configurations, where ELM suppression was demonstrated for conditions with q95~3.5 and pedestal collisionality ~ 1. What was particularly interesting with the odd parity configuration was that density and energy confinement was affected only minimally when RMP was activated. ELM suppression for collisionality ~1 is much more in-line with DIII-D AT and hybrid pedestal conditions than collisionality <0.25, which is needed for ELM suppression with even parity. In principal, unless low density, low collisionality is essential, odd parity I-coil configuration may offer significant advantages over an even parity I-coil configuration, particularly in a radiating divertor environment. The reason for these differing behaviors are far from understood. We think that a carefully crafted systematic comparison between even and odd parity cases would shed light on this issue. No systematic comparisons between even and odd parity-based ELM suppression are presently available. | ||
| Resource Requirements: | Eight beam sources. Odd parity I-coil configuration. This is a five shot experiment, meaning this is <0.5-day experiment. | ||
| Diagnostic Requirements: | IR camera, CER, Thomson scattering, SPRED, bolometers, CO2 interferometers. | ||
| Analysis Requirements: | ONETWO, UEDGE | ||
| Other Requirements: | -- | ||
| Title | 299: Investigation of Sheath Power Transmission at the DIII-D Divertor | ||
| Name: | Buchenauer none | Affiliation: | Sandia National Laboratories |
| Research Area: | General PBI | Presentation time: | Not requested |
| Co-Author(s): | Jon Watkins, Dmitry Rudakov, Charlie Lasnier, Josh Whaley | ITPA Joint Experiment : | No |
| Description: | The sheath is the ultimate regulator for the interaction of plasmas and materials, and of particular importance is the rate at which power flows through the sheath to the surface, since this sets the thermal loading on PFCs. However, experimental observations often find that the standard sheath theory to predict heat loading is inadequate, often by factors of 2-10, which is completely unacceptable for predictive capabilities of heat loading. Our initial measurements aimed to improve our understanding of this phenomenon using DiMES on DIII-D were incomplete due to instrumental problems. Using the redesigned cabling system, it should be possible to continue this investigation by improving on the electron temperature, plasma density, and heat flux measurements. Initial trials of prototype Langmuir probes which could test theories of probe response showed that the probes could survive well beyond the expected heat loads. More recent evidence from embedded thermocouples suggest that a dedicated experiment to compare infrared measurements with modeling of near surface thermocouples are warranted. DiMES serves as a unique platform to perform these critical experiments and play a role in resolving the sheath power transmission conundrum. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In order to investigate the flux of ions reaching the probes, this study will compare the long-used domed design with a similar domed design elevated above the magnetic sheath of the divertor tiles and a flat surface probe whose normal is aligned along the local magnetic field (surface normal probe). As these probes have surfaces normal to the divertor heat flux, they will require that the outer strike point be swept across the DIMES location to retain thermal integrity. Possible studies have included
1. determination of equilibration time during strike point sweep across DIMES probe 2. power scaling 3. variation with magnetic field angle 4. density and/or divertor neutral pressure scaling 5. toroidal field scaling For this proposal, we expect that we can reasonable cover studies 1 and 2 using the second half of a shot. If the run goes well, we expect that study 4 could also be accommodated by the main experiment. The measurements would be made between 3.5 to 5.0 seconds of flattop current, with operation well within core plasma stability boundaries to minimize damage to the probe tips and L-mode plasmas to avoid transient effects at the divertor (ELMs). Power levels would be limited to 3-4 sources based on thermal calculations and previous measurements of divertor heat and plasma parameters. We will require that the magnetic configuration be optimized to provide a radial x-point sweep for which the outer strike point moves inward from R=151.5 cm to R=143.0 cm (approximately) with minimal change in the x-point height (nominally 12-15 cm above the divertor floor). A reference shot (80136) was used to provide a similar sweep for a DIMES exposure on October 21, 1993, however some development will likely be required for the present divertor geometry. The rate of the sweep would vary during study 1 and be fixed during study 2. Procedure 1. Start with ohmic plasma with 250 msec inward sweep. Setup line scan for 165 degree IR camera to scan across the domed DIMES probe. Concurrently, determine peak parallel heat flux using the other lower divertor camera to setup sweep with beams (3 shots). 2. Decrease sweep duration from 250 msec to 50 msec (3 shots) and look for variation in ion flux and electron temperature at peak of profile from the DIMES probes. 3. At an appropriate sweep duration determined in step 2, obtain heat flux and probe data for the DIMES probes using 165 degree camera in line scan (3 shots needed to switch line scan from regular dome to raised dome). 4. At the shortest appropriate sweep duration as determined in step 2, increase beam power to 1, 2, and 3 beams to increase parallel heat flux up to 50 MW/m^2, keeping a close watch on divertor spectroscopy at the higher power levels (3 shots). 5. During second scan of main experiment, repeat measurements from step 3 and 4 with the higher target density plasma (higher divertor neutral pressure). | ||
| Background: | Since its installation in 1992, the Divertor Materials Evaluation System (DiMES) on the DIII-D tokamak has provided a unique platform for the study of plasma surface interactions. Early experiments performed many first-of-kind observations at a divertor surface: quantification of the net erosion rate of carbon, demonstration of reduced erosion during plasma detachment, elucidation of the role of chemical sputtering, quantification of deuterium retention in carbon and metallic coatings, and identification of a critical issue of MHD interaction between liquid lithium and a divertor plasma. These passive measurements have provided data for the benchmarking of PSI codes and helped to improve the operation of DIII-D.
Less well known perhaps is that DiMES can also be a platform for the development of plasma diagnostics. Early design issues have now been improved to provide 12 electrical feedthroughs (+ one pair for a thermocouple) for active measurements (microsecond time response). Sandia California designed the first active DiMES head and has tested Langmuir probes and H-microsensors using the platform. With the improved cabling, we propose to utilize the DiMES platform to address the sheath power trasmission conundrum. Experiments from DIII-D and other tokamaks have demonstrated that the power transmission through the sheath, as determined by divertor Langmuir probes and infrared camera images, is still a mystery. Profiles of the sheath power transmission factor (ratio of heat flux to product of ion saturation current and electron temperature) show this ratio varying across the outer divertor strike point, to values much less than the nominally expected value of ~7 near the peak heat and ion flux (more recent data shows a similar trend). Since this determination relies on the interpretation of probe characteristics and the thermal response of the tiles to the heat flux, the use of DiMES to better understand the plasma measurements and heat load would be beneficial. Our first attempt to make measurements above the magnetic sheath of the floor and with a normal-incident probe tip were plagued by instrumental problems, although the probe design performed well at the high parallel heat loads (7 divertor strike point sweeps over the probe with 25 MW/m^2). This DiMES head was refurbished in early 2010 and reinserted into the DIII-D divertor for discharges in March and April. These ohmically heated discharges had a modest amount of ECH power, but resulted in x-point sweeps that did not provide for strike points to reach the radius of DiMES (OSP was near 130 cm while the center of DiMES is at 148.5 cm). Despite having conditions that provided useful data on the low values of sheath power transmission, they did allow for resolution of the cabling problems that have plagued the earlier measurements, and routine probe measurements have been demonstrated. | ||
| Resource Requirements: | The hardware for the DiMES probe is available, along with instrumentation provided by the divertor Langmuir probe array. The experiment would require an IRTV at the 165 degree toroidal location (we are currently exploring alternatives to the TEXTOR camera), in additional to the present divertor camera view. Run time of approximately 1/2 day would also be required, including NBI availability (no cryo-pumping needed or desired). | ||
| Diagnostic Requirements: | Required Diagnostics
A desirable element of the experiment would be to use longer focal length optics for the IR camera at 165 degrees (we also would require operation in line scan mode to improve time resolution during the x-point sweeps). This would provide better resolution of the heat flux to the domed probes Divertor Langmuir probes Probes mounted on the DiMES system, using instrumentation from the divertor probe racks 165 degree lower divertor camera with long focal length optics and line scan mode Other lower divertor camera in customary video mode Divertor spectroscopy (Boron / Nitrogen) Magnetics for EFIT determination of field angles Zeff C02 interferometer Thomson scattering (burst mode for main experiment, but regular mode for piggyback) Fast filterscope channels viewing the lower divertor Other useful diagnostics Tile current array Fast tile thermocouple array Bolometers Edge CER for ion temperatuers | ||
| Analysis Requirements: | Analysis of the probe signals and IR data would be critical. Magnetics (EFIT) evaluation of the strike point locations and geometry changes would also be needed. More detailed evaluation of other desirable diagnostics would be welcome to begin to understand the relationship between the sheath power transmission factor and collisional effects in the divertor plasma. | ||
| Other Requirements: | |||
| Title | 300: Dependence of particle transport on collisionality | ||
| Name: | Zeng zeng@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | E. Doyle, et al | ITPA Joint Experiment : | No |
| Description: | Investigate the dependence of particle pinch and diffusion coefficient on collisionality, in order to test low-k turbulence theory predictions and compare with the parameter dependence results in momentum transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Scanning neu* from 0.1 to 0.8 in L-mode plasma, and use D2 gas puff modulation technique to calculate particle pinch velocity and diffusion coefficient. In H-mode plasma, also scan collisionality and use D2 gas puff modulations. | ||
| Background: | Have limited data in L-mode plasma. Need to expend the range of neu* in L-mode. Also try to get neu* scan data in H-mode. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | profile reflectometer, turbulence diagnostic | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 301: rho* scaling of intrinsic torque | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | T. Tala | ITPA Joint Experiment : | Yes |
| Description: | Test the rho* scaling of the intrinsic torque by comparing dimensionlessly similar discharges between DIII-D and JET. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce matched pairs of DIII-D/JET discharges (fixed q, betaN, nu*) with different rho* (perhaps using recent hybrid shots by Politzer to look at rho* scaling of hybrid confinement). Make NBI torque perturbations to infer intrinsic torque as described eg in Solomon et al PoP 2010. Since edge intrinsic torque is believed to depend on pedestal gradients and edge temperature, perform combinations of betaN and Ip scans as indirect ways of varying these. | ||
| Background: | Measurements on DIII-D have shown a clear dependence of the edge intrinsic torque on both the pedestal pressure and edge temperature, and an expression predicting the intrinsic drive has been derived. However, it still remains unclear how this torque scales to ITER with rho*. Theoretical arguments may suggest as unfavorable as rho*^3, while other simulations seems to indicate only a weak rho* dependence. | ||
| Resource Requirements: | Co and counter NBI | ||
| Diagnostic Requirements: | Full profile diagnostics, especially CER, plus FIDA and main ion CER. | ||
| Analysis Requirements: | TRANSP + post-processing | ||
| Other Requirements: | -- | ||
| Title | 302: ELM Suppression in High Performance AT Plasmas with RMP in Odd Parity | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | M. Fenstermacher, S. Mordijck, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | This experiment examines whether or not RMP ELM suppression in the odd parity coil configuration can be successfully applied to high performance AT-class plasmas. Odd parity has been little used in the last six years or so, mainly because complete ELM suppression at the very low collisionalities representative of the ITER pedestal (i.e., nu-star = 0.1). The odd parity configuration may be more appropriate to AT plasmas because both density and energy confinement are affected only minimally when RMP was activated, unlike the application of even parity RMP. Direct comparison of the density/collisionality range will be made with corresponding previously analyzed H-mode plasmas that had even parity. With a modest investment in experimental time needed to establish the ELM-suppressed operating regime, this experiment can be piggybacked onto appropriate AT SSI experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish the range in q95 for ELM suppression in an odd parity configuration. For the odd parity configuration, set I-coil current to 5.8 kA, Bt = -1.78 T, and Pinj = 6 MW. An Ip scan similar to the one from shot 138538 is done to establish the ELM suppression window. Based on some earlier data, we expect this ELM suppression window to be q95 = 3.4 to 4.0. Next, set q95 that falls nearest to q95=4. If this selection of q95 for ELM suppression is of interest to AT studies, then the effects of RMP on high performance plasmas using the odd parity configuration can be studied in piggy-back mode. | ||
| Background: | Virtually all the previous studies of RMP ELM suppression and mitigation have used even parity. While doing so has been successful in many respects, the cost of applying RMP in even parity has typically been in lower energy confinement and significant density pump-out of the main plasma. Over the past several years, only a very modest amount of work has been directed toward the odd parity configurations, where ELM suppression had been demonstrated (for some cases) for conditions with q95~3.4 to 4.0 and pedestal collisionality ~ 0.6. What was particularly interesting in using the odd parity configuration was that density and energy confinement was affected only minimally when RMP was activated, which is unlike our even parity results. ELM suppression for collisionality ~0.6 is much more in-line with DIII-D AT and hybrid pedestal conditions than collisionality <0.25, which is needed for ELM suppression with even parity. In principal, unless low density, low collisionality is essential, odd parity I-coil configuration may offer significant advantages over an even parity I-coil configuration, particularly in a radiating divertor environment. | ||
| Resource Requirements: | Five beam sources. Odd parity I-coil configuration. This experiment can be piggybacked onto appropriate SSI plasma experiments. | ||
| Diagnostic Requirements: | IR camera, CER, Thomson scattering, SPRED, bolometers, CO2 interferometers. | ||
| Analysis Requirements: | UEDGE, ONETWO | ||
| Other Requirements: | -- | ||
| Title | 303: Steady state L-mode discharges with rational q_min > 1 using off-axis NBI | ||
| Name: | Austin austin@fusion.gat.com | Affiliation: | U of Texas, Austin |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The purpose of this experiment is to study a low-order rational q_min transport barrier in steady-state L-mode conditions. The crux of the experiment would be using the new off-axis NBI capability to drive current to stabilize a hollow current profile and create an enduring reverse shear q profile with q_min = 2, for example. If a recipe could be found for creating a steady-state reverse shear profile, the experiment would proceed to study the transport barrier, measuring turbulent fluctuation levels and looking for zonal flows, convective cells, or whatever it is that makes transport barriers at rational q surfaces. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The starting point for these discharges would be L-mode discharges with early NBI with on-axis beams to slow the current penetration and produce slowly evolving reverse shear current profiles. At a certain point in time the on-axis beams would be replaced by off-axis beams that will maintain the hollow current profile. The off-axis NB current drive could possibly be assisted with ECCD; a significant problem could be maintaining L-mode with NB plus EC power. | ||
| Background: | Electron heat transport barriers have been observed transiently at low-order in discharges with slowly evolving reverse shear q profiles Studying these barriers has been difficult because they are short-lived. Finding a way to freeze the hollow current profile to establish a long-lived barrier would permit the measurement of quantities that may show the nature of the barrier. In particular, certain fluctuation measurements require integration times of order 100 ms. | ||
| Resource Requirements: | Four neutral beams: 30L, 330L, off axis: 150L, 150R
Possibly ECH, min 4 sources. | ||
| Diagnostic Requirements: | Fluctuation diagnostics, MSE. | ||
| Analysis Requirements: | Normal profile analysis programs. | ||
| Other Requirements: | |||
| Title | 304: Density peaking and turbulent particle transport | ||
| Name: | Zeng zeng@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | E. Doyle, T. Rhodes, et al | ITPA Joint Experiment : | Yes |
| Description: | Investigate parameter dependence of density peaking, e.g., neu*, beta, Te gradient, q profile etc., in order to compare to turbulent transport model predictions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Scan neu*, beta, with and without ECH deposition at the center, Ip ramp up and down as well. Obtain del_ne/ne in the core in each condition. Use gas puff modulation technique to calculate particle pinch velocity and diffusion coefficient also. | ||
| Background: | In JET and ASDEX-U, density peaking has been observed to increase with decreasing collisionality. In DIII-D, it hasn't been observed (Park, IAEA 2010). But it is necessary to investigate this phenomenon in details with advanced diagnostic tools. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | Profile reflectometer, turbulence diagnostic measurements | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 305: Investigation of the transport barrier in bat-eared Te profiles | ||
| Name: | Austin austin@fusion.gat.com | Affiliation: | U of Texas, Austin |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | T.L. Rhodes, L. Schmitz | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to understand the nature of the transport barrier in the bat-eared Te profile discharges. Analysis of the heat transport barrier in these peculiar off-axis EC-heated discharges shows that there is factor of 10 reduction in chi_e in a narrow region around the q=1 surface. More data is needed to confirm that the barrier is at q=1, to measure the q-profile (see if it is reversed), and to look for zonal flows or other transport suppression mechanisms. Also, more shots would allow the exploration of the question: Why is the transition to H-mode necessary? The experiment could possibly be enhanced by attempting the bat-eared discharges with the new off-axis NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce the discharges from 2009 except raise the field to 2 Tesla to allow broader coverage of the Te profile by ECE (a subsequent toroidal rotation experiment serendipitously revealed that bat-ears can be made in a 2 T discharge). Use NB blips to get MSE data. | ||
| Background: | The bat-eared Te profile discharges represent a unique case where a transport barrier is created with the application of power outside the barrier, hence clearly marking the barrier location by the peculiar profile shape that ensues. Since the barrier location is known and relatively long lived, diagnostic measurements to explore the barrier are straightforward. | ||
| Resource Requirements: | NB sources: 2-4, possibly off-axis beam. ECH sources: min. of 4 | ||
| Diagnostic Requirements: | Fluctuation diagnostics, ECEI | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 306: Development of H, He versions of Baseline Scenario for ITER Low Activation Phase | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Placeholder - need better definition of ITER wants/needs | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 307: Checkout of real-time steerable mirror NTM control | ||
| Name: | La Haye rbrtlahaye@gmail.com | Affiliation: | Retired from GA |
| Research Area: | NTM Stabilization | Presentation time: | Requested |
| Co-Author(s): | J. Lohr, B. Penaflor, R. Prater, A. Welander | ITPA Joint Experiment : | No |
| Description: | The new capability for real-time steering of ECCD launch mirrors with control of "poloidal" angle by the PCS makes routine use possible; no need to move plasma (Rsurf or Zsurf) or change BT for alignment. A preliminary slow sweep in April was successful. PCS control of pre-programmed sweeps, search and suppress, and active tracking all need to be developed and checked out using the mirrors. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a steady m/n=3/2 NTM as in 142650 for example.
For all gyrotrons, one at a time, do preprogrammed slow sweeps across q=3/2 and back to check biggest dip and mirror locations. Repeat with 100 Hz modulation of gyrotron power for ECE measure of where absorption occurs. Repeat (no mod) with pre-programmed single "instantaneous" steps of each mirror (0.5 degrees) close to biggest dip to check PCS/Ethernet latency and mirror response. Multiple gyrotrons powered simultaneously to study the power and alignment requirements for complete stabilization, i.e. map out rate of decrease of n=2 MIRNOV Btheta with misalignment and/or Peccd; condition for complete stabilization. | ||
| Background: | ITER will use real-time steerable mirrors, not plasma shifts or changes in BT for alignment as we have had to do up to now. Our successful control techniques of "search and suppress" and "active tracking" can be modified for mirror steering. | ||
| Resource Requirements: | All gyrotrons (6 in 2011) one at a time to start. Fast mirror upgrade by PPPL in Jan desirable but not necessary. Reduction of PCS latency needed as planned. Safety interlocks on hardware and PCS to avoid mirror failures. | ||
| Diagnostic Requirements: | ECE in particular. All kinetics (MSE, CER, Thomson) for TORAY-GA. | ||
| Analysis Requirements: | Auto analysis codes for surveying OK. | ||
| Other Requirements: | -- | ||
| Title | 308: Study of 3D effect on RWM stability | ||
| Name: | Takechi none | Affiliation: | Japan Atomic Energy Agency |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | M. Okabayashi(PPPL), G. Matsunaga(JAEA), A. Isayama(JAEA),
Y. In (FAR-TECH) |
ITPA Joint Experiment : | No |
| Description: | RWM control (coils) will change the 3-D plasma configuration, because of finite size of RWM control coils. To clarify the 3-D effect on RWM stability, investigation of RWM control with reduced set of coils is proposed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Control of n=1 RWM with reduced set of coils, e.g. two toroidal coils sets in the opposite position (total coil number is four). Plasma feels n=2 sideband and change the 3-D configuration. Study the RWM stability with MHD spectroscopy on marginal RWM stable plasma or growth rate on RWM unstable plasma. | ||
| Background: | We performed RWM control experiment with reduced coil sets on RFX-mod. It is found that RWM can be stabilized with very few coils. Large sideband effect was observed. For example, Marginal unstable mode is destabilized by sideband effect and terminated the plasma. We will have RWM control experiments of TOKAMAK discharges on RFX-mod on early 2011. We can compare the RFX-mod and DIII-D experimental results. We will also study the 3D configuration with VMEC code and check the MHD stability with NIFS code. | ||
| Resource Requirements: | 1 day experiment with
NBs to exceed no-wall beta limit, ECCD for NTM suppression | ||
| Diagnostic Requirements: | Standard for RWM | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 309: ITER startup experiments--completing the matrix | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Not requested |
| Co-Author(s): | A.W. Hyatt, D.A. Humphreys, T.C. Luce | ITPA Joint Experiment : | Yes |
| Description: | Some remaining ITER startup experiments need to be completed:
1. Startup in helium with oblique EC launch 2. EC assist, applying Vloop first, then Bpol for field null and vertical field (ITPA IOS 2.3) 3. Compare HFS startup: Ohmic vs. EC assist |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Set up is similar to previous ISAR work. This experiment requires t ā?¤ 200ms. After that, piggyback experiment is acceptable | ||
| Background: | ITER startup experiments have been largely completed, and presented as an invited talk at APS 2009 and the recent FEC 2010. However there are some loose ends that need additional experimental time (see description above). Item 1 was direct request from ITER personnel, Item is an ITPA joint experiment (IOS 2.3), and item 3 was work that wasn't completed in 2010. | ||
| Resource Requirements: | ECH | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 310: Multiple low-n RWM identification and feedback control | ||
| Name: | In inyongkyoon@unist.ac.kr | Affiliation: | Ulsan National Institute of Science and Technolog |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | J.S. Kim, M. Okabayashi, E. Strait, V. Svidzinski | ITPA Joint Experiment : | No |
| Description: | The proposal is first to identify the presence of multiple low-n (up to n=3) RWMs in steady state high beta operation, and then to provide the corresponding multiple low-n mode feedback control. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The first step will be to identify the presence of the multiple low-n RWMs in steady state high beta plasmas beyond transiently observed multiple low-n modes in high beta plasmas. The target discharges are likely to be ELM-induced RWMs (possibly, off-axis fishbone-mode (OFM)-driven RWMs, as well) in high beta plasma, which are often accompanied by significant multiple low-n modes (at least for both n=1 and n=3 modes).
Once these modes are identified, the feedback control will be provided to suppress each mode simultaneously using individually controlled I-coils. If the non-RWM noise affects the feedback performance, a new Kalman filter compatible for n = 1 and 3 RWMs will be applied. | ||
| Background: | A recent paper [1] identified the multiple low-n RWMs above each low-n no-wall limit (betaN ~ 3.76) in DIII-D discharge, though these multiple modes were transiently observed. Specifically, while the n=1 mode was suppressed likely due to both rotational stabilization and n=1 RWM feedback, the n=3 mode appeared dominant, leading to beta collapse. Also the presence of the n=2 mode was also identified by post-processing B-dot probes. However, lacking in the multiple-mode identification at the time of the experiments, the n=1 mode feedback was greatly influenced by the aliasing component from n=3. Since DIII-D has the capability to provide the multiple low-n mode identification and feedback control (at least for n=1 and n=3), it is suitable to demonstrate the multiple low-n RWM feedback control scheme beyond the ideal MHD multiple low-n no-wall limits in steady state high beta plasmas in DIII-D. The fully independent I-coil control needs to be confirmed, prior to the experiments. When the non-RWM noise becomes problematic, a multiple-mode compatible Kalman filter which FAR-TECH has recently developed will be applied for the first time.
Reference [1] Y. In et al, Phys. Plasmas 15, 102506 (2008) | ||
| Resource Requirements: | All the available NBI sources with ECCD for preemptive NTM suppression | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | n=2 component extraction from B-dot signals | ||
| Other Requirements: | I-coils need to be independently controllable, applicable to n=1 and 3 (at least) | ||
| Title | 311: Validation of the kinetic RWM stability model using off-axis NBI in Positive B_T Discharges | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Requested |
| Co-Author(s): | M.J. Lanctot, H. Reimerdes, J.W. Berkery, S.A. Sabbagh, I.T. Chapman, M.A. Van Zeeland, W.W. Heidbrink, M. Okabayashi, E.J. Strait, Y. In, R La Haye | ITPA Joint Experiment : | Yes |
| Description: | Test the drift kinetic RWM stability model by producing an unstable pressure-driven RWM in DIII-D. The goal of the experiment will be to probe the experimental conditions where the RWM is predicted by the MISK code to be most unstable. The dependence of RWM stability on the energetic particle distribution will be explored by varying the NBI configuration and probing the plasma stability using slowly-rotating magnetic fields. The effect of n=1 and n=3 fields on the mode stability will be investigated in the parameter space where the RWM is found to be least stable. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Feedback control of the NBI will be used to attain the normalized plasma beta, toroidal rotation, and the trapped energetic ion fraction of the total pressure (beta_fast) predicted by the MISK code to result in an unstable RWM. We will reduce beta_fast using the off-axis neutral beam to heat discharges where the toroidal field is in the same direction as the plasma current (positive B_T), and by maximizing the plasma current and thermal particle density. Throughout the experiment, the growth rate of the RWM will be measured using modulated n=1 and n=3 magnetic fields. When an unstable RWM is found, the effect of n=3 magnetic braking on the stability of the mode will be documented using mostly non-resonant magnetic fields (i.e. odd parity). The ramp rate of the plasma current will be varied to change the effect of fast ions on the mode. | ||
| Background: | Previous DIII-D and NSTX experiments have shown that the observed RWM stability is consistent with the kinetic stability model in the MISK code, which identifies the trapped thermal and energetic ions as being responsible for stabilizing the RWM above the no-wall limit [Berkery Phys. Plasmas 17 (2010), 082504; Reimerdes IAEA 2010]. One main difference between the two experiments is that NSTX can access the unstable RWM regime at "intermediate" values of rotation while, in DIII-D, the RWM is stable over the entire range of plasma rotation profiles. However, analysis of DIII-D experiments suggests that the RWM is only marginally stable in weakly-shaped LSN H-mode discharges when beta normalized is 2.3, the plasma rotation is 40 km/s (or 0.9 percent of the inverse Alfven time) at CER chord T6, and beta_fast is approximately 20 precent. The RWM is predicted to be unstable when the stabilizing effect of fast ions is excluded. In 2011, we will have three knobs to reduce the trapped energetic ions. The off-axis neutral beam is expected to lead to a decrease in beta_fast when the toroidal field is in the positive toroidal direction (i.e. co-Ip). Also, the ratio of the energies stored in hot ions and in the thermal plasma is inversely proportional to the plasma current and the plasma density. This identifies the optimal conditions for this experiment: a LSN H-mode discharge with zero upper triangularity (to minimize the stabilizing effect of poloidal shaping), positive B_T, Ip consistent with B_T for q95<4.0, pellet fueling and gas puffing to maximize the density. | ||
| Resource Requirements: | Positive B_T (co-Ip). 8 NB Sources are required. Pellet fueling. I-coil error field correction and testing waveforms. Experiment should follow boronization. | ||
| Diagnostic Requirements: | Equilibrium diagnostics including kinetic profiles (Thomson, CER, MSE) and fast ion diagnostics. | ||
| Analysis Requirements: | TRANSP/ONETWO calculations of fast ion pressure, Kinetic EFIT, MISK, MARS-K, and HAGIS. Analysis will be ongoing as more information on the performance of the off-axis neutral beam becomes available. | ||
| Other Requirements: | -- | ||
| Title | 312: Plasmas with high rotational Mach number | ||
| Name: | Zerbini marco.zerbini@enea.it | Affiliation: | ENEA C.R. Frascati |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Study of plasmas with high rotational Mach number in the discharge core. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Increase torque to increase Vrot, starting from shots 140221/2 and similar. This will also increase Ti hence Cs, so plasma regime will have to be carefully adjusted to obtain M>1. | ||
| Background: | In DIII-D database there are examples where a situation with M~0.5 or more is reached for a limited time, at least in the center of the discharge. The idea is to obtain a few shots with very high Mach number (M >2) established for an experimentally consistent interval of time, to study clear supersonic conditions. This can give information about equilibrium in presence of non-linear shock effects, capable to locally increase the plasma parameters. The shock waves could be intrinsic, i.e. self generated in the supersonic fluid, or induced by an external cause (pellet). | ||
| Resource Requirements: | Standard Tokamak with NBI and possibly pellet | ||
| Diagnostic Requirements: | Standard diagnostics, fast ECE, neutrons, X-ray | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | The shots will not have to be dedicated, since the only requirement is high core Mach number, other programs with similar plasma conditions can use them. | ||
| Title | 313: Reduce separatrix density via sweeping divertor strike point outward to reduce ELM frequency | ||
| Name: | Callen jdcallen@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | R. Groebner, T. Petrie | ITPA Joint Experiment : | No |
| Description: | As described in the Background section below, a new paleoclassical-based model (UW-CPTC 10-6, August 30, 2010) predicts that a key parameter governing the pedestal structure and its evolution is the electron density on the separatrix. Specifically, the model predicts that, if the separatrix density is reduced the density at the pedestal top will be reduced and the evolution toward an ELM will take longer -- or perhaps the evolution into an ELM can be prevented? Also, T_e at the top of the pedestal and toroidal plasma rotation Omega_t throughout the pedestal are predicted to remain constant; but the T_e pedestal width is predicted to decrease.
The density on the separatrix is usually reduced in H-mode plasmas by increasing the divertor pumping during a sequence of shots. The proposal here is to transiently move the divertor strike point outward (from an initial poor pumping position to where the divertor pumping is greater) after an L-H transition but before the first ELM and/or during long repetition time ELMs. The physics objective would be to explore the effects of the transiently increased pumping on the separatrix density, pedestal profiles of n_e, T_e and Omega_t, and ELM period. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | An H-mode plasma with a long temporal evolution between the L-H transition and the first ELM and/or with low frequency ELMs needs to be set up with initially poor divertor pumping (via a divertor strike point set at about 1.33 m). Then, just after the L-H transition (and NBI power reduction to facilitate a long ELM-free period) the divertor strike point should be moved outward (on ~ 50 ms time scale?) to 1.37 m (for best pumping) to transiently increase the neutral pumping. The main effects to be explored during this transient phase would be the evolution of the electron density, temperature and toroidal plasma rotation at the separatrix, within the pedestal and at the top of the pedestal, and concomitant changes in the time to the first ELM and/or the time between ELMs. | ||
| Background: | Recently, predictions have been developed for the structure of plasma parameter profiles of H-mode pedestals in transport quasi-equilibrium in tokamak plasmas -- in "A Model Of Pedestal Structure," report UW-CPTC 10-6, which is available via http://www.cptc.wisc.edu. The predictions are based on assuming paleoclassical radial plasma transport processes dominate throughout the pedestal. Model predictions have been given for the profiles and magnitudes of the electron density and temperature, and plasma toroidal rotation in the pedestal. All the predictions have been shown to agree quantitatively (within a factor of about two) with properties of the recently studied low density 98889 DIII-D pedestal [J.D. Callen et al., Nucl. Fusion 50, 064004 (2010)]. Also, recent SOLPS modeling by John Canik has shown that (as reported in his invited talk JI2.1 talk at the recent Chicago DPP-APS meeting) these model predictions for transport quasi-equilibrium electron heat transport and density profile shape "are consistent with" NSTX pedestal data both with and without Lithium wall coatings. Further, John has estimated the Z_eff variation in the 98889 pedestal and also obtained "pretty good" agreement between the new pedestal structure model predictions and the SOLPS modeling results for that pedestal's chi_e(rho) and n_e(rho).
In the UW-CPTC 10-6 report 4 fundamental tests, 4 secondary tests and 4 improvement scenarios are identified. Improvement scenarios #1 and #4 suggest that the density at the pedestal top and slowing the evolution to an ELM can be controlled by reducing the pedestal electron density on the separatrix. One of the key effects of RMPs in preventing ELMs in H-mode plasmas is to reduce the separatrix density through "density pump-out," which is not well understood. This ROF proposal seeks to try a different approach to ELM control based on an idea motivated by this new pedestal structure model for forestalling the onset of ELMs or perhaps even eliminating them -- by transiently increasing the pumping in an H-mode plasma and thereby reducing the separatrix density. The n_e, T_e and Omega_t pedestal profile responses to a transiently reduced separatrix density could also help validate or invalidate the paleoclassical-based pedestal structure model. | ||
| Resource Requirements: | A "poorly pumped" low power H-mode DIII-D plasma with a long time between the L-H transition and the first ELM and/or with long repetition time ELMs needs to be established -- i.e., discharges similar to those discussed in Groebner et al., Nuclear Fusion 49, 045013 (2009). Then, the divertor strike point needs to be moved out radially to a "maximal" pumping position. Because the particle confinement time improves significantly during the ELM-free period following the L-H transition, it is not completely clear whether or not particles can be removed fast enough to make the requisite difference before the first ELM arrives. However, a test of the particle removal rate can be done before this experiment is attempted and so this question can be resolved before actually spending dedicated experimental time, e.g., during vessel cleanup in April, 2011 or during the "start up" day on the Friday prior to multi-week operation. It should be pointed out that additional pumping, if needed --> Diag | ||
| Diagnostic Requirements: | can be provided by simultaneously using the cryo-pumps from the divertor opposite the main cryo-pump; this can be done by using a plasma shape with finite dRsep configured to take advantage of all three available pumps. After proof-of-principle is established, this experiment would likely require at least half a day for setup and exploration of transient increases in the divertor strike point.
The new edge Thomson system is needed to better measure evolution of electron density at the "separatrix" (identify via where T_e ~ 90 eV?) and n_e, T_e throughout the pedestal. In addition, CER measurements of the Carbon toroidal flow velocity (and deuterium as well?) in the pedestal are needed. Also, some filterscope and divertor probe data would be useful to facilitate SOLPS and/or UEDGE modeling of the effects of the transiently increased divertor pumping. | ||
| Analysis Requirements: | Some divertor pumping modeling with SOLPS and/or UEDGE would be useful to confirm the transient effects of increased divertor pumping. | ||
| Other Requirements: | -- | ||
| Title | 314: LSN without Return Current control shape distortion | ||
| Name: | Hyatt hyatt@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Requested |
| Co-Author(s): | John Ferron | ITPA Joint Experiment : | No |
| Description: | Develop a VFI based (possibly a powered VFI) LSN control algorithm that does NOT require active control of the Return Current magnitude using carefully chosen VFI configurations and return current coil. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | LSNs at DIII-D always use a return current control that distorts the plasma shape. Almost always the distortion is placed on the lower outer plasma boundary. The distortion can be minimal to very pronounced depending on the details of the LSN shape and plasma parameters explored by the experiment. There have been a few attempts to try to minimize the distortion in specific experimental shape, such as the ITER SC2 baseline, but these efforts concentrated on modifying the VFI configuration to minimize the distortion requirements, but still operating with return current control. It is very possible to operate WITHOUT return current control as long as the return current isn't too large and/or cause distortion by itself. It should be possible to devise suitable VFI configurations that minimize these issues. It may also be possible to devise simple modifications to the existing LSN control algorithm that will lessen the effects of direct return current distortion. The end result will be a modified LSN control scheme that will allow the experimenter to specify and realize a desired LSN shape rather that accept today's return current control induced shape distortion. | ||
| Resource Requirements: | 4-5 neutral beams+ all standard plasma ops resources required. | ||
| Diagnostic Requirements: | Standard plasma operations diagnostics and MSE | ||
| Analysis Requirements: | EFITs only | ||
| Other Requirements: | |||
| Title | 315: Particle transport analysis via impurity and D2 puffing modulations | ||
| Name: | Zeng zeng@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Requested |
| Co-Author(s): | E. Doyle, et al | ITPA Joint Experiment : | No |
| Description: | The goal is that gas puff modulation technique should be used for a routine particle transport analysis. Develop transient particle transport analysis via impurity modulated puffing technique, as a routine, part of all detailed transport analysis. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Ar or Ne impurity modulated at low frequency puffs into plasma, and use SPRED and CER to measure impurity density profile evolution. Then particle pinch velocity and diffusion coefficient can be calculated. Repeat the same plasma discharge but with D2 modulated gas puffs instead of impurity, use profile reflectometer to measure ne profile evolution. D and V also can be calculated from the evolution. The plasma conditions can be Ohmic, L-mode and H-mode discharges. | ||
| Background: | D2 main gas modulated puffing technique has been successfully applied for particle transport analysis in DIII-D. But it is lack to have impurity modulated puffing as a routine transient transport analysis. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | CER, SPRED, Profile reflectometer, and Turbulence diagnostic measurements | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 316: ELM Suppression with RMP in Odd Parity | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | M. Fenstermacher, S. Mordijck, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | Collisionality-, density-, and pressure gradient in the pedestal that delimit ELM suppression in an odd parity I-coil configuration are compared with those in an even parity I-coil configuration. Data and analysis for the even parity cases are already completed. All that is needed is to obtain a small additional number of shots at comparable parameters but in odd parity. The main scan involves variations in pedestal density obtained at different deuterium injection rates. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For the odd parity configuration, set I-coil current to 5.8 kA, in order to match up with the even parity cases already analyzed (i.e., shots 138541-51, (2010)). The target H-mode plasmas will have the same working parameters as those same shots, e.g., Bt = -1.78 T and Pinj = 6 MW. An Ip scan identical to the one from shot 138538 is done to establish the ELM suppression window, which can then be compared with the q95-ELM suppression window from the 138538 even parity case. Next, choose q95 from the center of the ELM suppression window (which should be between 3.4 and 4.0, based on earlier studies.), change the density and collisionality via gas puffing. The methodology of the RMP- and gas puffing timing during the shot is identical to types used in shots 138541-51. Gas puff rates of 0, 40-, and 80 torr l/s are used to vary density, collisionality, pressure gradient in the pedestal, and comparisons are then made with the corresponding even parity cases. | ||
| Background: | The lions share of the previous studies of RMP ELM suppression and mitigation have used even parity. While doing so has been successful in many respects, the cost of applying RMP in even parity has typically been in lower energy confinement and significant density pump-out of the main plasma. Over the years, a very modest amount of work has been directed toward the odd parity configurations, where ELM suppression was demonstrated for conditions with q95~3.4-4.0 and pedestal collisionality ~ 0.6. What was particularly interesting with the odd parity configuration was that density and energy confinement was affected only minimally when RMP was activated. ELM suppression for collisionality ~0.6 is much more in-line with DIII-D AT and hybrid pedestal conditions than collisionality <0.25, which is needed for ELM suppression with even parity. In principal, unless low density, low collisionality is essential, odd parity I-coil configuration may offer significant advantages over an even parity I-coil configuration, particularly in a radiating divertor environment. The reason for these differing behaviors are far from understood. We think that a carefully crafted systematic comparison between even and odd parity cases would shed light on this issue. No systematic comparisons between even and odd parity-based ELM suppression are presently available. | ||
| Resource Requirements: | Seven beam sources. Odd parity I-coil configuration. This is approximately a five shot experiment, meaning this is <0.5-day experiment. | ||
| Diagnostic Requirements: | IR camera, CER, Thomson scattering, SPRED, bolometers, CO2 interferometers. | ||
| Analysis Requirements: | ELITE, ONETWO, UEDGE | ||
| Other Requirements: | -- | ||
| Title | 317: Comparison of resistive wall mode feedback techniques using internal and external coils | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Requested |
| Co-Author(s): | J. Bialek, O. Katsuro-Hopkins, G. Navratil | ITPA Joint Experiment : | No |
| Description: | The proposed experiment seeks to compare the resistive wall mode (RWM) feedback performance of internal versus external coil sets. A comparative study of this nature may help inform decisions on whether to keep the in-vessel, ELM control coils in the ITER baseline design, as these coils could also be used for RWM feedback. A VALEN model for DIII-D will be used to design and debug control algorithms for both I-coil and C-coil feedback. Such simulations will be useful in determining the most effective feedback algorithm for each coil set. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | If a pressure driven RWM target discharge is not available, a low beta, current driven target based on shot 133021 will be used. This target makes use of a plasma current ramp to obtain a broad current density profile. The edge safety factor q95 decreases as the discharge evolves, with the result that the plasma becomes increasingly more kink unstable as q95 progresses through low-order rational numbers. This property of the discharge presents a built in criterion for evaluating the efficacy of feedback schemes ā?? the most effective feedback method will be the one that enables stable operation at the smallest value of q95.
After obtaining a suitable no-feedback reference discharge, RWM feedback with the C-coil followed by the I-coil will be tested in separate discharges. Standard, preprogrammed error field correction will be applied with the coil set that is not presently being used for RWM feedback. The best performing feedback algorithms and parameter settings from VALEN simulations will serve as a starting point. | ||
| Background: | VALEN is finite element, electromagnetic code that is capable of simulating RWM feedback with realistic coil geometries in the presence three-dimensional, non-axisymmetric conducting structures [J. Bialek, et al., Phys Plasmas 8, 2170 (2001)]. The present VALEN DIII-D model has roughly 1300 elements and includes the vacuum vessel wall with port holes and flanges, poloidal and radial magnetic field sensors, and three-dimensional representations of the I and C-coils. In addition to DIII-D, VALEN has been used to design and simulate RWM feedback schemes for the NSTX, HBT-EP, KSTAR, and ITER experiments.
For KSTAR and ITER, a linear RWM observer and feedback controller designed using model reduction and optimal control theory is expected to bring about significant feedback power reduction compared with proportional gain only feedback [O. Katsuro-Hopkins, et al., Nucl Fusion 47 1157 (2007); O. Katsuro-Hopkins, et al., Nucl Fusion 50 025019 (2010)]. This design technique has been used to create a controller for DIII-D that can be tested in the proposed experiment [J. M. Hanson, et al., proc 37th EPS Conference on Plasma Physics, P4.127 (2010)]. Feedback control of pressure-driven RWM with the DIII-D I and C-coils has been compared previously, with the finding that I-coil feedback enabled operation at significantly higher values of normalized beta divided by the plasma internal inductance and open-loop RWM growth rate compared with C-coil feedback [E. J. Strait, et al., Phys Plasmas 11, 2505 (2004)]. However, a new experiment will be able to take advantage of a recently implemented capability to drive the C-coils with the low-latency, high-bandwidth, voltage control ā??audioā?? amplifiers that are used for I-coil RWM feedback. The current control, switching power amplifiers previously used for feedback with the C-coils suffered from significant latencies, on the order of 100 microseconds. Additionally, VALEN simulations predict that the C-coils can be used achieve comparable performance to the I-coils, stabilizing modes with growth rates near the ideal-wall limit, when a controller designed with the model reduction and optimal control theory technique referenced above is used. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 318: Roles of thermal and energetic particles in passive RWM stabilization | ||
| Name: | In inyongkyoon@unist.ac.kr | Affiliation: | Ulsan National Institute of Science and Technolog |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | H. Reimerdes, M. Okabayashi, Y.Q. Liu, G. Matsunaga and RWM Physics group | ITPA Joint Experiment : | Yes |
| Description: | The proposal is to clarify the individual roles of thermal and energetic particles in passive RWM stabilization. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Based on a recent discharge (e.g. 141069) where off-axis fishbone-modes (OFM) frequently occurred, the probing frequency of the MHD spectroscopy will be linearly scanned to observe the variations of the plasma response. Considering that the previous study already observed the relevancy of the thermal particles in a slow probing frequency near 20 Hz in [1], the emphasis will be given to the possible plasma response associated with the energetic particles in higher frequency (e.g. up to 400 Hz). The probing field will be also scanned in both co- and counter-Ip directions to clarify any sign dependence of particle motions in theoretical calculations. | ||
| Background: | In recent experiments, strong plasma response was measured in an intermediate frequency between ion precession drift (omega_D) and bounce (omega_b) frequencies of thermal particles based on MHD spectroscopy, supporting a theoretical prediction that the least stable weakly damped RWM would be peaked somewhere between omega_D and omega_b, based on a perturbative approach in MISK code [1].
Meanwhile, in another study associated with the off-axis fishbone mode (OFM) induced RWMs, the precession-drifting energetic particles are predicted to be primarily responsible for passive RWM stabilization based on a self-consistent approach in MARS-K, while thermal particles alone were not sufficient to explain the experimental results [2]. Since both studies were more or less based on similar target discharges, it remains to be resolved which particles (thermal or energetic) are the main contributors to passive RWM stabilization. Considering that the MHD spectroscopy reflects plasma response directly, a full scan of MHD probing frequency in the relevant frequency range of both thermal and energetic particles in terms of precession and bouncing motions will clarify which particles (thermal or energetic) are playing more responsible for passive RWM stabilization. In parallel, this will help us to resolve the discrepancies between perturbative approach and self-consistent theoretical calculations. References [1] H. Reimerdes et al, to be submitted to PRL (2010) [2] M. Okabayashi et al., APS-DPP invited talk (2010), to be submitted to PoP (2010) | ||
| Resource Requirements: | more than 4 NBI sources with ECCD for pre-emptive NTM suppression | ||
| Diagnostic Requirements: | Energetic particle (EP) diagnostics, magnetics, internal diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | I-coils with Audio amplifiers and C-coil with SPAs | ||
| Title | 319: Impact of the current evolution on the ITER demo discharges - 2 | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Requested |
| Co-Author(s): | T. Luce | ITPA Joint Experiment : | No |
| Description: | We plan to investigate the role of the current profile, and of the li measurement of the current density peaking, in the tearing stability of DIII-D ITER demonstration discharges. From recent studies, it seems that the central factor leading to the appearance of the n=1 instabilities that limit these discharges is the evolution of the J profile, and not the betaN level as previously believed. The analysis of the database shows that all the discharges end (with a tearing mode or at the end of the NBI power) when the li trace reaches a precise range of values, all included between 0.85-0.95. For this reason it is important to explore different starting li levels, at different betaN values, to understand whether any ITER discharge of this type will inevitably be terminated due to the natural evolution of the current density profile. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce an ITER demo discharge, with q95~3, and evolve the current profile ina way that the starting li value at the beginning of the high-betaN phase is kept fixed at the highest possible value. Run the discharge at progressively higher betaN flattop levels. Run the discharge at 2 betaN levels, increasing betaN after 2 s, to check if the tearing mode appears later in the discharge (due to the current evolution) or early in the discharge (due to the high-beta effects) | ||
| Background: | In DIII-D discharges the target scenario in the ITER shape at q9=3 and betaN=1.8 is often limited by the appearance of a tearing mode with toroidal mode number n=1, which
significantly reduces the confinement and often ends the discharge with a disruption. The standard approach to define the operational limits due to these instabilities characterizes this stability limit as a beta limit. In the general approach, it is inferred that the plasmas become unstable above a marginal beta limit where the NTMs are metastable and can be triggered by a seed island, provided usually by sawteeth, ELMs or fast particles instabilities like fishbones. The classical stability term is not taken into account for the stability limit in this approach. In the presence of ELMing and sawtoothing activity during the discharge, it is implied that the instability will be triggered if the discharge is run above a certain beta value, and the classical delta-prime term is considered constant and negative. This approach is also implied in discussion of DIII-D results. However, it has been shown that in the experimental data of the ITER-like DIII-D discharges, the betaN level is not a discriminator for the stability of the discharge, and that the destabilizing bootstrap term is not varying as the plasma evolves towards the triggering of the instability. Moreover, the li traces for all the discharges is still decreasing when the modes start. For these reasons we believe that the change in the classical stability term related to the evolution of the current profile plays the main role in the destabilization of the modes. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 320: Lengthen ELM period via ECH in SOL, which reduces separatrix density? | ||
| Name: | Callen jdcallen@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | R. Groebner, J. Lohr | ITPA Joint Experiment : | No |
| Description: | As described in the Background section below, a new paleoclassical-based model (UW-CPTC 10-6, August 30, 2010) predicts that a key parameter governing the pedestal structure and its evolution is the electron density on the separatrix. Specifically, the model predicts that if the separatrix density is reduced the density at the pedestal top will be decreased and the evolution toward an ELM will take longer -- or perhaps the evolution into an ELM can be prevented? Also, T_e at the top of the pedestal and toroidal plasma rotation Omega_t throughout the pedestal are predicted to remain constant; but the T_e pedestal width is predicted to decrease.
The density on the separatrix is usually reduced in H-mode plasmas by increasing the divertor pumping. Electron cyclotron heating (ECH) in the Scrape-Off-Layer (SOL) was shown to extend the ELM period in H-mode plasmas long ago -- in J. Lohr et al., "The Effect of Edge Resonant Electron Cyclotron Heating On Edge Localized Modes In A Tokamak," GA-A20182, May 1991. This proposal seeks to reproduce and extend that early DIII-D ELM experiment to explore the degree to which ECH in the SOL and near the separatrix region can be used to control ELMs. The key physics objective would be to see if the effect of the ECH on the ELM period was caused by a reduction of the separatrix density. Also, presuming this is the case, the induced changes in the pedestal profiles of n_e, T_e and plasma toroidal rotation will be compared against paleoclassical predictions. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment requires an H-mode plasma with a fairly long temporal evolution between the L-H transition and the first ELM which is then followed by fairly long (> 30 ms) period ELMs. Just after the L-H transition (and NBI power reduction to facilitate a long ELM-free period), one or a string of ~ 1 MW pulses (>~ 10 ms) of ECH power should be injected into the SOL. It would also be good to inject a string of ECH pulses into the SOL between long period (>~ 30 ms) ELMs. The main effects to be explored during the resultant transient phases would be the evolution of the electron density, temperature and toroidal plasma rotation at the separatrix, within the pedestal and at the top of the pedestal, and concomitant changes in the time to the first ELM and/or the average ELM period. It would be important to try various ECH power levels, pulse lengths and deposition positions ranging from say 4 cm outside the separatrix on the outer mid-plane in to the separatrix and perhaps up to the tanh symmetry point (typically ~ 0.6 cm inside the separatrix on the outboard mid-plane) of the H-mode pedestal. | ||
| Background: | Recently, predictions have been developed for the structure of plasma parameter profiles of transport quasi-equilibrium H-mode pedestals -- in "A Model Of Pedestal Structure," report UW-CPTC 10-6, which is available via http://www.cptc.wisc.edu. The predictions are based on assuming paleoclassical radial plasma transport processes dominate throughout the pedestal. Model predictions have been given for the profiles and magnitudes of the electron density and temperature, and plasma toroidal rotation Omega_t in the pedestal. All the predictions have been shown to agree quantitatively (within a factor of about two) with properties of the recently studied 98889 DIII-D pedestal [J.D. Callen et al., Nucl. Fusion 50, 064004 (2010)]. Also, recent SOLPS modeling by John Canik has shown that (as reported in his invited talk JI2.1 talk at the recent Chicago DPP-APS meeting) these model predictions for transport quasi-equilibrium electron heat transport and density profile shape "are consistent with" NSTX pedestal data both with and without Lithium wall coatings. Further, John has estimated the Z_eff variation in the 98889 pedestal and also obtained "pretty good" agreement between the new pedestal structure model predictions and the SOLPS modeling results for that pedestal's chi_e(rho) and n_e(rho).
In the UW-CPTC 10-6 report 4 fundamental tests, 4 secondary tests and 4 improvement scenarios are identified for the new pedestal structure model. Improvement scenarios #1 and #4 suggest that the density at the pedestal top and slowing of the evolution to an ELM can be controlled by reducing the pedestal electron density on the separatrix. One of the key effects of RMPs in preventing ELMs in H-mode plasmas is to reduce the separatrix density via "density pump-out," which is not well understood. The average electron pressure gradient within the SOL may be limited by interchange-instability-induced "blobs" etc. Thus, if ECH increases the T_e in the SOL, the density within and on the separatrix should maybe decrease correspondingly. This ROF proposal seeks to try a different approach to ELM control based on an idea motivated by this new pedestal structure model for forestalling the onset of ELMs or perhaps even eliminating them -- via reducing the electron density on the separatrix by introducing ECH in the SOL, which has been shown (in GA-A20182, May 1991) to increase the period of ELMs in DIII-D (by a factor ~ 3). The n_e, T_e and Omega_t pedestal profile responses to a transiently reduced separatrix density could also help validate or invalidate the paleoclassical-based pedestal structure model. | ||
| Resource Requirements: | An H-mode DIII-D target plasma with a long time between the L-H transition and the first ELM and/or with long period ELMs needs to be established -- i.e., discharges similar to those discussed in Groebner et al., Nuclear Fusion 49, 045013 (2009). Then, ECH should be applied in perhaps >~ 10 ms, ~ 1 MW pulses after the L-H transition before the first ELM and between long period ELMs. One might need to back off the NBI to a no ELM condition and then add equivalent ECH inside the separatrix to get the ELMs back. Also, different radial deposition positions ranging from 4 cm outside the separatrix to the mid-point (~ 0.7 cm inside the mid-plane separatrix position) should be explored. This experiment would likely require a full day for setup and exploration of various options for determining the effects of SOL-region ECH pulses on the electron density on the separatrix, pedestal profiles of n_e, T_e and Omega_t, and the ELM period. | ||
| Diagnostic Requirements: | The new edge Thomson system is needed to better measure evolution of electron temperature and density at and just outside the "separatrix" (identify via where T_e ~ 90 eV?) and n_e, T_e throughout the pedestal. In addition, CER measurements of the Carbon toroidal flow velocity (and deuterium as well?) in the pedestal are needed. | ||
| Analysis Requirements: | Some ECH multiple pass ray tracing would be needed to determine the toroidal field strength and gyrotron frequencies that are needed to deposit significant ECH in various regions of the SOL. | ||
| Other Requirements: | Since the ECH is to be deposited in a lower T_e (< 100 eV) plasma region, its single pass absorption is likely to be small. Thus, a significant fraction of the ECH power is likely to be rattling around the entire machine. Hence some diagnostics may have to be shuttered of to prevent them from being damaged. | ||
| Title | 321: Analysis of Pedestal Fueling Using OEDGE | ||
| Name: | Elder david.elder42@gmail.com | Affiliation: | U of Toronto |
| Research Area: | Pedestal Structure | Presentation time: | Requested |
| Co-Author(s): | To be determined | ITPA Joint Experiment : | No |
| Description: | Recycling of hydrogen in the pedestal region may be a significant contributor to pedestal formation and behavior. There is no way at the present time to directly measure the fueling of the pedestal. However, this problem can be approached by using codes to simulate the background plasma and the associated hydrogen recycling.
OEDGE is an appropriate tool for this task. By the process of empirical plasma reconstruction, OEDGE is used to reconstruct a plasma solution (ne, Te, Ti, vb and E) that is consistent with as many diagnostic measurements as possible. This solution is then representative of ā??averageā?? plasma conditions during a substantial time in the specific discharge. The EIRENE hydrogen Monte Carlo code is then used as part of OEDGE to simulate the neutral sources, transport and ionization of hydrogen recycling inside the vessel. This includes measured ionic fluxes to surfaces from Langmuir probe measurements as well as estimated sources from recombination resulting from the reconstructed plasma solution. In addition, it is possible to extrapolate plasma conditions at the edge of the modeling grid to wall surfaces in order to estimate contributions from hydrogen recycling at the main chamber walls. The code tracks many statistics related to hydrogen including neutral atom and molecule densities as well as ionization. The contributions to core and pedestal fueling due to neutral ionization from the various sources can then be reported. This includes a 2D distribution of over the SOL, pedestal and core regions of the modeling grid as well as identification of the dominant sources contributing to core and pedestal fueling as a function of the local plasma conditions. In order to perform such an analysis with a reasonable level of confidence, good diagnostic measurements are required to constrain the reconstruction process. This requires as much diagnostic data as possible for both the inner and outer divertor as well as inner and outer SOL plasmas. Key elements are good Langmuir probe data for both the inner and outer targets. Ideally, plasma configurations would be chosen which yield this kind of data. In addition, infra red power flux measurements, Thomson scattering measurements (both main chamber and divertor), reciprocating probes, filterscopes and Tangential TV camera spectroscopic measurements are crucial to try to constrain the global plasma solution as much as possible. After these constraints are in place it is possible to develop a plasma solution that is consistent with the measurements and then simulate the pedestal fueling. If discharges already exist that are sufficiently well diagnosed then this research proposal may not require additional machine time. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Analyse existing shots with sufficient diagnostic data first. In addition, try to ensure that discharges included for pedestal analysis in the 2011 campaign include gathering of sufficient diagnostic data for empirical plasma reconstruction. This would include using some magnetic sweeping during setup discharges to characterize divertor plasma conditions as well as collection of infra-red, Langmuir probe, and divertor Thomson diagnostic data. | ||
| Background: | Similar analyses as those proposed here have been done in the past by A.W. Leonard [1] and J.D. Elder [2]. In the latter case OEDGE was used to analyse core fueling from different sources in four different discharges. However, this analysis needs to be extended to a more complete set of related discharges with good diagnostics. In addition, the sensitivity and distribution of core leakage to variation in the background plasma solutions used needs to be assessed.
[1] ā??Pedestal fueling through interpretive analysis of measured main chamber and divertor target flux in DIII-Dā??, A.W. Leonard et. al., Journal of Nuclear Materials (June 2009), 390-391, pg. 470-473 [2] ā??OEDGE Modeling of Deuterium Recycling in DIII-Dā??, J.D. Elder et al., APS poster 2009. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 322: Why Is It So Difficult to Suppress ELMs With RMP When the ion gradB Drift Is Away From the X-point? | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | M. Fenstermacher, S. Mordijck, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | In this experiment, we investigate whether or not it is possible to suppress ELMs of a SN plasma with the ion gradB drift directed away from the X-point. The even parity I-coil configuration is used. This experiment varies pedestal collisionality and pedestal pressure gradient (via gas puffing) and plasma rotation (via beam balance). ELM suppressed plasmas using RMP with the ion gradB drift toward the X-point already is in-hand in large part to compare with, so that this experiment can be completed using relatively little experimental time. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | First establish the q95 ELM suppression window exists when the ion gradB drift direction is away from the X-point. Fix Bt = -1.78 T, Pinj = 6 MW, and I-coil = 3.2 kA sweep Ip, as done for comparison shot 138538. If not suppression, keep repeating this approach at higher I-coil until either suppression occurs or the plasma falls out of H-mode. If necessary, repeat at higher power level. If a q95 ELM suppression window is established where (presumably) q95 ~ 3.5 is well within this window, scan density (and collisionality) via gas puffing (~80 torr l/s). Then, raise plasma rotation by increasing Pinj and repeat gas puff. Repeat. | ||
| Background: | From a previous experiment we have shown the importance of the ion gradB drift on ELM suppression using the RMP method. For example, with I-coil ~ 3.2 kA and q95 = 3.5, ELM suppression was achieved with ease when the ion gradB drift was toward the X-point of a lower SN diverted plasma; when the ion gradB direction was reversed, there was no suppression, even though the pedestal collisionality for the two cases was virtually identical (i.e., ~0.25). (Since SN plasmas run at typically lower density when the ion gradB drift is directed away from the X-point than when the this drift is opposite, we had suspected that lower pedestal density and lower collisionality would result in relatively prompt ELM suppression, which was not the case at all.) We would like to understand why ELM suppression is more difficult when the ion gradB drift is away from the X-point, as this result is very likely to shed real light on the nature of how RMP ELM suppression. The focus in this experimental proposal is to look more systematically at the roles of plasma rotation and collisionality in ELM suppression with RMP. | ||
| Resource Requirements: | Machine time 0.5 day (in forward Bt), I-coil, dome- and upper baffle cryo-pumps cold, minimum 6 co-beams. | ||
| Diagnostic Requirements: | Asdex gauges (in particular, dome and upper baffle locations), core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER. | ||
| Analysis Requirements: | UEDGE, ONETWO, ELITE | ||
| Other Requirements: | -- | ||
| Title | 323: MHD stability characterization for Off-axis NBI | ||
| Name: | Murakami masanori_murakami@att.com | Affiliation: | Retired |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Requested |
| Co-Author(s): | JM Park, C. Petty, W. Heidbrink, M. Vanzieland, A. Polevoi, T. Suzuki, ..E. Lazarus, et al | ITPA Joint Experiment : | No |
| Description: | ā?¢ Characterize MHD stabilities for off-axis NBI for variations of +/- BT direction; NBI steering variation
ā?¢ This will help for steady state experiments and also ITER planning ā?¢ Beam-driven MHD stabilityies (Fish-bon, sawtooth; sawtooth; monster sawtooth) are related to magnetic pitch just like off-axis NBCD ā?¢ These have important implication to ITER planning |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Variation of BT direction
2) off-axis LT and RT beam 3) Steering angle scan | ||
| Background: | ā?¢ Efficiency of Off-axis NBCD is sensitive to the beam alignment to the magnetic field line , as demonstrated in the DIII-D 2007 NBCD experiment using small plasmas.
ā?¢ Fast-ion driven istability, such as fish-bone instability is driven by trapped particles , therefore instability would not occur for injection paralllel to the magnetic field.,. This situation is exactly the same as the off-axis NBCD. ā?¢ There was some suspicion that the recent steady state high-beta experiments that exclusively operated in the normal BT direction my have been suffered from fish-born instabilities.. ā?¢ In the DIII-D steering configuration, reverse BT has much better CD efficiency than normal BT.. So typical NBCD experiment could lead to MHD-quite operation. ā?¢ ITER has the wrong BT direction for the off-axis NB steering, so it is a problem if off-axis NBI configuration is prone to fishbone instability. ā?¢ Certain results of off-axis NBI experiments in other devices could be confused by this MHD instability effect ā?¢ It is important to characterize MHD stability operation space for off-axis NBI experiments in DIII-D utilizing the steering capability. | ||
| Resource Requirements: | 3 days(?) with shared other experiments for
ā?¢ +/- BT direction ā?¢ steering angle scan | ||
| Diagnostic Requirements: | MSE; magnetics;
FIDA / neutrons; neutral particle analyzer, Loss detecto; | ||
| Analysis Requirements: | NUBEAM; TRANSP; MHD codes | ||
| Other Requirements: | |||
| Title | 324: Divertor startup in DIII-D | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Not requested |
| Co-Author(s): | A.W. Hyatt, J. Leuer | ITPA Joint Experiment : | No |
| Description: | Obtain diverted discharges in DIII-D as early as possible. Quantify V-sec reduction and changes in the current profile, which may be a useful "knob" in other DIII-D experiments. Evaluate potential for ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a typical DIII-D discharge, move the divert time earlier in steps. Goal is t_divert < 100ms. This might also be a useful target for obtaining an early H-mode which could produce changes in the current profile evolution. Note: This proposal does not lend itself to "piggyfront" status, as failed shots are expected, i.e. it needs dedicated time. | ||
| Background: | Normal divert time in DIII-D is typically 250 to 400 ms, Ip > 0.4 MA. Flux savings may be obtained by diverting earlier in time. For ITER, the earliest divert time may also reduce main chamber heating. This experiment would seek to address and understand some of the challenges in diverting early. | ||
| Resource Requirements: | May want to change the breakdown voltage and initial Ip ramp. This could be scanned as part of the experiment. | ||
| Diagnostic Requirements: | Increase gain on magnetics arrays
mse is a crucial diagnostic for this experiment | ||
| Analysis Requirements: | May require new EFIT snap file at lower currents | ||
| Other Requirements: | -- | ||
| Title | 325: High Performance Plasma for FNSF | ||
| Name: | Navratil navratil@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The new initiative in the US FES Program to define a Fusion Nuclear Science Facility as the centerpiece of a Fusion Nuclear Science Program calls for a DT plasma source of 14 MeV neutrons that is quite different in character from the burning plasma state which ITER aims to achieve. Rather, to minimize cost and reduce risk, proposals for an AT and ST based FNSF would use copper coils would use a low Q driven system with Q in about 2. In such plasmas, alpha heating from fusion is not the dominant heating source, and such plasmas could be based on more direct extrapolations of high performance. hot ion plasmas previously studied on DIII-D. Studies should begin to establish scalable hot-ion driven target plasmas as a design basis for an AT or ST FNSF at low Q~2. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The capability of DIII-D to suppress ELMS, stabilize the RWM with active feedback, and ECRH to sustain elevated central Q AT plasmas would be applied to target plasmas similar to the high Q-DD performance plasmas from the 1995 campaign. | ||
| Background: | In 1995, Ed Lazarus and I used a controlled H-mode transition to produce a very substantial increase in fusion reactivity (Q-DT equivalent =0.32) , energy confinement time (tau-E ~ 400 msec), neoclassical ion confinement over most of the plasma cross-section [ PRL 77 (1996) 2714 and Nucl. Fusion 37 (1996) 7]. Based on what we have learned about beta limiting behavior in such plasmas in the past 15 years, it is clear that these plasmas were transient due to RWM onset stimulated by the first large ELM and the peaking of the current profile. We now have tools to address these limiting phenomena that were unavailable in 1995. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 326: 2D imaging of core and edge modes during the L-H transition | ||
| Name: | Yun gunsu@postech.ac.kr | Affiliation: | Pohang U of Science and Technology |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Hyeon K. Park, Benjamin Tobias, Tobin Munsat | ITPA Joint Experiment : | No |
| Description: | Simultaneous 2D images of core and edge modes evolving in the course of the L-H transition have been clearly captured by the KSTAR ECEI system during the 2010 KSTAR campaign. It is observed that the edge becomes less affected by the core MHD activities (sawtooth) and develops filamentary structures (m~25) as the plasma evolves into deeper H-mode. A strong shear in the poloidal flow has been observed during the appearance of filaments. In fully developed H-mode state, the ELM bursts are often preceded by the filaments, i.e., ELM-precursor. In addition, while the toroidal rotation of the core is driven by the NBI (co-current in KSTAR) as expected, the toroidal flow in the outer edge appears to be in the opposite direction. It is our interest to study the role of the layer where the rotation is sheared the most.
Similar study on DIII-D for 2D visualization of the L-H transition physics and ELMs can be done for wider range of plasma conditions. In particular, L-H transitions in counter-/co-current and on-/off-axis ECH cases will be interesting because ECH directly affects the core sawtooth activities, which in turn affects the edge pressure profiles during the L-H transition. Furthermore, co- and counter-NBI capability can be used to study the effect of rotation on the confinement and edge instability such as reduction of the sawtooth crash heat pulse at the edge observed in the KSTAR H-mode. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish double-null L-mode plasmas (kappa~1.8, delta~0.6, NBI~2MW co- & counter-, varying ECH injection angle: ref.shot.???).
(2) Obtain the 2D ECEI images across the poloidal cross-section from the q~1 surface through the scrape-off layer as follows. This will require 3 reproducible shots. (3) Turn on ECH and Repeat for varying injection angle. | ||
| Background: | The study of sawtooth-initiated H-mode transitions helped to understand that the improved conļ¬?nement is caused by the development of a transport barrier right at the edge but inside the separatrix [1,2]. Although the physics behind the H-mode transition has not been clariļ¬?ed yet, there is substantial experimental and theoretical evidence that turbulent ļ¬?ows, which enhance transport and limit the conļ¬?nement, are diminished by sheared poloidal ļ¬?ows residing at the plasma edge. Rather reciprocally, this argument led to a theory that ļ¬?uctuations themselves induce the ļ¬?ows via Reynolds stress, which then act back and annihilate the turbulence. Also the shear of toroidal rotation is considered important for edge instability and confinement of the inner region although the role of the toroidal shear is not clear yet.
In this context, it is important to understand how the sawtooth crash and the subsequent heat flow towards the edge affects the generation of shear flow. The ECEI system [3] has a unique capability to measure 2D Te fluctuations with simultaneous coverage of both core and edge regions in 1~2cm spatial resolution and ļ?sec time resolution, suitable for detailed measurement of 2D structure of heat and plasma flows. Although it may be difficult to resolve the Reynolds process and the reduction of turbulence directly, the ECEI system will provide substantial information for (or against) the conceived H-transition process. [1] F. Wagner et al., Phys. Rev. Lett. 53 (1984) [2] F. Wagner, Plasma Phys. Control. Fusion 49 (2007) [3] B. Tobias et al., Rev. Sci. Instr. 81 (2010) | ||
| Resource Requirements: | NBI (co- and counter-injection)
ECH (perpendicular, co- and counter-current injection) | ||
| Diagnostic Requirements: | ECEI, ECE radiometry, MSE, BES, CXR spectroscopy, CO2 interferometers, SXR | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 327: Enhancement of FW antenna loading with local gas puffing | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Heating & Current Drive | Presentation time: | Requested |
| Co-Author(s): | M.-L. Mayoral, V. Bobkov, M. Goniche, J.C. Hosea, S.J. Wukitch,
S. Moriyama, F.W. Baity, L. Colas, F. Durodie, A. Ekedahl, G.R. Hanson, P. Jacquet,P. Lamalle, I. Monakhov, M. Murakami, A. Nagy, M. Nightingale, J.-M. Noterdaeme, J. Ongena, M. Porkolab, P.M. Ryan, M. Vrancken, J.R. Wilson, ASDEX Upgrade Team |
ITPA Joint Experiment : | Yes |
| Description: | In FY09-10 experiments, DIII-D tested the effect of local D2 puffing on FW antenna loading with a single orifice adjacent to the 285/300 antenna. Whether or not a substantial part of the loading enhancement is a truly *local* effect is a topic of great interest to the ITER IOS Topic 5.2 group (listed as co-authors). AUG results in L-mode discharges show a substantial fraction of the loading enhancement they see is associated with the local puffing, as opposed to an increase in the global far SOL density which can be caused by puffing from any location in the machine. In DIII-D to date the results are not as clear. It seems that under some circumstances the effect can have a local part, while in other cases the majority of the enhancement, at least, is due to a global change in the ELM character or other global change in the far SOL.
In the present LTO, we are adding local puff locations to the other two FW antennas, at 0 deg and at 180 deg, and are adding a multi-orifice ('comb') gas puffing injector near 285/300 to compare with a single orifice. Here we propose to continue the experimental study of the effect of gas puffing on the far SOL density and on the FW antenna loading, in both L- and ELMing H-mode plasmas, with the goal being discovery of a regime in which the antenna loading between ELMs can be increased to a level that will permit the coupling of large FW power levels without undue deleterious effects on confinement. A big hole in our data from 2009 is the lack of clear measurements of the effect of puffing on the far SOL density with either the ORNL reflectometer, near 300 deg, or with the UCLA profile reflectometer. This is an important measurement for quantitative studies of the effect of puffing on the far SOL. Ideally, we'd also like to use the midplane reciprocating Langmuir probe as well. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The basic experiment is to create a stationary condition, either an L-mode or a regularly ELMing H-mode condition, and then compare the effects of puffing from individual midplane puff locations on antenna loading, both on the antennas nearest the active puffing location and on the other two. The basic scans for a given plasma condition are of puffing location, puffing level and the timing of the puffing relative to the start of the ICRF pulse. The relative timing appeared to significantly affect the power handling of the FW antenna in the earlier experiments. If time permits, we would like to try puffing neon or nitrogen from these orifices, to follow up on an intriguing result seen on Alcator C-Mod where injection of these 'exotic' gases seemed to have a very positive effect on the FW antenna performance (strongly reduced antenna arcing). | ||
| Background: | |||
| Resource Requirements: | All three FW antenna puff locations, calibrated; all three FW systems, NBI at moderate power levels. Special diagnostics as listed below. | ||
| Diagnostic Requirements: | Both ORNL and UCLA reflectometers, midplane reciprocating Langmuir probe. Other Langmuir probes desired (divertor, etc.). Visible images of as many of the FW antennas and their surrounding area as possible (including the UCSD fast camera looking at 0 deg). | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 328: FW-only L/H Transition Power Study (Dup 79) | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | A. Nagy, M. Porkolab, P.M. Ryan, J. Hosea, G. Wang | ITPA Joint Experiment : | No |
| Description: | ITER may not have enough auxiliary heating power to exceed the L/H transition power in the Day 1 configuration (hydrogen ops). For this reason, several machines have remeasured the L/H transition power in hydrogen with hydrogen or helium beams and compared those results with deuterium. Furthermore, recent work has shown that the L/H transition power has a dependence on plasma toroidal rotation speed, with lower rotation speeds being associated with lower L/H transition power levels. Even 20 percent-level effects may be important in this context. With this in mind, the fact that the Fast-Wave only H-mode observed in 1991 had a distinctly lower power threshold than with NBI heating in the same discharge may be of importance. The fact that H-mode transitions were observed with fast wave (FW) power as the only auxiliary heating source, under conditions of rather low single-pass absorption was an important piece of evidence that multiple-pass absorption of the FW power can be efficient. By expanding the range of FW frequencies, densities (and hence target electron temperatures), and using 3rd harmonic ECH, we can get a more quantitative measurement of the edge losses by determining the L-H transition threshold power under varying single-pass absorption conditions. This is important to ITER, both from the point of view of improving knowledge of access to H-mode in plasmas with only intrinsic rotation (no torque) and also to improve understanding of FW edge losses under varying edge conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment consists of scans of target density, rf power (at two different frequencies: 60 MHz and 90 MHz), toroidal field, and whether 3rd harmonic ECH is added (at the appropriate toroidal field), and comparison of co-, counter-current, and push-pull phasing. A beam is used for comparison, later in the shot. Minimal beam blips are used for CER, MSE diagnostics. At each condition, the power threshold for L-H transition is observed for FW, for the comparison beam, and for ECH (at the appropriate fields). | ||
| Background: | H-modes with fast wave heating by direct electron absorption as the only form of auxiliary heating were discovered at DIII-D in July 1991, and have not been studied since. In particular, the fast waves in that experiment were launched with the shortest available parallel wavelength ("Pi phasing") at 60 MHz at around 1 T, and we have never studied H-modes with current drive phasing, either co- or counter-current, or at higher frequency than 60 MHz. Furthermore, the great interest in the dependence of L/H transition power levels on rotation and/or applied torque in recent years has provided a new motivation for this experiment, as mentioned in the description section above. Finally, insofar as this study provides further data on FW edge losses, the emphasis on this area on NSTX in the past several years has increased the need to obtain data at higher toroidal fields than can be run on NSTX to see how these effects scale with BT, to provide data both for possible future STs and for ITER FW heating.
In a piggyback experiment in 2010, it was found that H-mode transitions could be rather easily obtained with a very low NBI power plus 2-3 MW of FW power in directional (counter-current) phasing. This reduces the uncertainty that FW-only H-modes can be obtained in the lower-k-parallel 90 deg phasing. | ||
| Resource Requirements: | Machine Time: 1 day Experiment
Number of gyrotrons: 4 Number of neutral beam sources: 4 Three FW systems, one at 60 MHz and the others at 90 MHz. | ||
| Diagnostic Requirements: | Edge reflectometry with the antennas adjacent to the 285-300 FW antenna, along with the UCLA profile reflectometers, would be a very helpful addition to the usual diagnostic set for this experiment. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 329: Stabilization of NTMs with ECCD | ||
| Name: | Isayama isayama.akihiko@qst.go.jp | Affiliation: | QST |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | R.J. La Haye, R. Buttery, M. Austin, G. Matsunaga, M. Takechi | ITPA Joint Experiment : | Yes |
| Description: | This proposal includes study of ECCD effect on NTM, which supplements previous experiments, taking into account the limited diagnostic capability in ITER. The experiments are related to the ITPA Joint Experiment MDC-8, "Current Drive Prevention/ Stabilization of NTMs". In this proposal, the following experiments are included: (a) Modulation effect: identify the minimum EC wave power for complete stabilization both for modulated and unmodulated ECCD cases, (b) Phase effect: Investigate the stabilization effect (including destabilization for X-point ECCD) for different phase difference between NTM rotation and modulated EC wave power, (c) Preemptive (or very early) stabilization: identify the minimum EC wave power for preemptive (or very early) stabilization and (d) ECCD width effect: investigate the stabilization effect for different ECCD deposition width by changing the toroidal injection angle. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Experimental condition to obtain an m/n=2/1 NTM is based on the previous experiments (e.g. Volpe et al., Phys. Plasmas 16, 102502 (2009)). After the discharge scenario for obtaining 2/1 NTM is established, ECCD parameters are changed with the same plasma parameters. Data for ECH (i.e. zero toroidal injection angle) and counter-ECCD are also taken for comparison. | ||
| Background: | Stabilization with modulated ECCD was successful, and some results have been reported from DIII-D, JT-60U and ASDEX-U. To supplement the results, in particular, to investigate the degradation of the stabilization effect due to deviation from the optimum condition, data on the above topics are taken. The result will have an impact on establishing the NTM stabilization scenario in ITER with limited diagnostic capability. | ||
| Resource Requirements: | neural beams, ~6 gyrotrons, real-time system for NTM stabilization | ||
| Diagnostic Requirements: | magnetic probe, ECE (also oblique ECE if available), MSE, CER | ||
| Analysis Requirements: | REVIEW, NEWSPEC, TORAY | ||
| Other Requirements: | -- | ||
| Title | 330: Role of edge electric fields in 3D RMP transport | ||
| Name: | Zarnstorff none | Affiliation: | PPPL |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | R. Nazikian | ITPA Joint Experiment : | No |
| Description: | Hypothesis: The RMP induced particle transport may be due to EXB transport, due to || electric fields on open flux surfaces and field lines. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Compare the effect of the RMP on pedestal density tranport, while varying the spatial correlation between magnetic field structure and electric-potential structure by
a) scanning the x-point height away from the divertor plates, to vary the || transport length on stochastic field lines, and thereby the || electric field distribution. b) comparing single- and balanced double-null discharges, both as a function of (balanced) x-point height. In the up-down symmetric case, the effects of the drifts may cancel. c) In balanced double-null configurations, comparing RMP excitations that are up-down symmetric vs asymmetric. | ||
| Background: | willing to talk if desired. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Analysis of pedestal density and pedestal profiles; equilibrium measurements and analysis with RMP; edge measurements and simulations with RMP structures if available. | ||
| Other Requirements: | |||
| Title | 331: Testing the zero shear rational q model of ELM suppression | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The rational q zero shear model of RMP ELM suppression predicts a number of trends which can be tested against experimental data. Some of these require numerical analysis of existing data to test, or more careful review of trends. Others would benefit from new data. I will try to summarize both aspects in the plan. The model is based on edge bootstrap current leading to a region of weak or zero shear in the pedestal. This region of zero shear may then be easily perturbed by RMP fields (or the associated shielding response) to lead to transport. tests therefore rest on three main planks: (i) parameter space dependence - where we get significant edge bootstrap and where this can lead to weak shear in the pedestal; (ii) visible structures associated with these perturbations; (iii) careful correlation of effects against detailed models. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | To test parameter dependencies of the effect, we should look for correlations with key parameters. Edge bootstrap should be high for high shaped, but less for low shape - so harder to get ELM suppression effect. Magnetic shear is harder to reverse in double null, so the suppression should go away as we approach DND condition as edge magnetic shear increases. Current ramps may perturb the edge current and should lead to model-able changes in the q95 value at which the zero shear coincides with rational q. Finally we should undertake careful diagnosis to look for associated structure (see separate proposal). New experiments in 2011 may be desirable to fill some gaps in existing data, and get better diagnosis - eg of edge BS through new TS. | ||
| Background: | ANALYSIS TASKS: We need to make accurate kinetic EFITs for ELM suppressed & non-suppressed cases for a range of conditions (SND/DND, low/high nu*. We should also run MARS to look at plasma response to fields as a zero shear rational q point is approached. Some vacuum modeling may be instructive to see how easily field lines are perturbed by I coil RMP fields. | ||
| Resource Requirements: | Analysis of past cases and modeling over coming months. Experiment time to add new data. | ||
| Diagnostic Requirements: | Needs careful though - how can we best see structures hypothesized - thinks like finger in the edge. | ||
| Analysis Requirements: | see above... | ||
| Other Requirements: | These types of study may naturally combine with tests of other ideas. | ||
| Title | 332: Investigate non-inductive EC startup and Increase Plasma Current | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | A.W. Hyatt, N. Eidietis, D.A. Humphreys, R. Pinsker | ITPA Joint Experiment : | No |
| Description: | To date, up to 33 kA has been obtained in DIII-D for non-inductive EC startup. This experiment would vary parameters to increase this current and further study the physical mechanisms. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Increase the non-inductive duration by delaying turn-on of the OH drive. Then further increase Ip
1. Evaluate Oblique launch vs. Radial. 2. Prefill scan at EC=2 MW until decrease in Ip is observed 3. P_EC scan to obtain higher Ip 4. Compare EC radial and oblique launch 5. B_VF scan at highest Ip 6. BT scan 7. At highest parameters, add NB power (slowly increase duration) and measure NB CD, to obtain highest current. Evaluate effect of off-axis NB, since measured Ip to date peaks off-axis. 8. Rerun shape #78769 (with present diagnostic set) with high trapped electron fraction for comparison. | ||
| Background: | Non-inductive EC driven startup may be important for future burning plasma devices including the commissioning phase of ITER. To demonstrate the efficacy of this technique, higher currents need to be demonstrated as well as a better understanding of the physical mechanisms. | ||
| Resource Requirements: | ECH (6-7 gyros), NB (including off axis), delay onset of VloopB for longer experimental time.
Special patch panel if work from C. Forest (#78769) is to be compared to more normal startup to evaluate effect of trapped electrons. | ||
| Diagnostic Requirements: | ECE imaging, CER "blips" at t = 0, higher gain on magnetics arrays, fast camera on CIII, | ||
| Analysis Requirements: | 2D Abel inversion of fast camera data, New EFIT files to accommodate non-standard plasma shape. | ||
| Other Requirements: | -- | ||
| Title | 333: High Beta, Steady State Hybrids | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Steady State) | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | This experiment will integrate a high beta hybrid plasma with the reactor relevance of Te~Ti and full noninductive current drive. In 2011, the addition of a sixth gyrotron and optimization of the six co-beams will allow us to eliminate the residual 9 mV loop voltage of our best previous case, and hopefully lower q_95 from 5.85 to 5.0 at the same B_T. Additionally, the higher heating power should allow us to increase beta closer to the ideal wall limit, which is around beta_N=4.
This experiment will demonstrate that H-mode (hybrid) discharges with q_min~1 are capable of high beta (beta_N~4) operation with >50% bootstrap current fraction. The remaining noninductive current will will be supplied by on-axis sources at high efficiency. The poloidal magnetic flux pumping that is self-generated in hybrid will suppress the sawteeth despite the strong on-axis current drive, which is important for avoiding the 2/1 mode. The higher efficiency for on-axis current drive will offset the modest bootstrap current fraction such that this scenario will satisfy the requirements for FDF as well as (or better than) the high q_min scenario with strong off-axis current drive. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Start by repeating shot 133881. (2) Inject all six gyrotrons with central current drive. For the six co-NBI sources, increase the injection voltages as much as possible while maintaining a plasma pulse length of at least 5 seconds. (3) Optimize the dynamic error correction (may use broadband feedback), adjust the plasma shape for optimal pumping. (4) Attempt to increase beta_N using the full heating power. If plasma current is overdriven (i.e. negative loop voltage), then increase plasma current to compensate. | ||
| Background: | The current proposal for FDF envisions a high q_min advanced tokamak scenario with 70% bootstrap current fraction. While this is compatible with the US view of DEMO, the physics of the high q_min AT scenario is still being developed. There is also an issue regarding the high off-axis current drive efficiency needed for FDF in this proposal.
Here I propose that the low q_min hybrid scenario is compatible with the requirements of FDF, and it has several advantages. First, the physics basis is well advanced. Long duration hybrid discharge with high beta and high confinement are routinely achieved. Second, because q_min=1 in the hybrid scenario, all of the external current drive can be deposited near the plasma center where the current drive efficiency is the highest (because of the lack of trapped particles and the high electron temperature). While the bootstrap current fraction will be lower in this low q_min hybrid scenario (50% rather than 70%), the increase in the current drive efficiency for central deposition more than makes up for this. Experiments on DIII-D have come very close to demonstrating this scenario using five co-beams and five gyrotrons. Hybrid plasmas with beta_N=3.4 were stably produced with a loop voltage of 9 mV. The loop voltage was a strongly decreasing function of heating power. While the ion and electron temperature were nearly the same outside of rho=0.2, the H-mode confinement factor remained high, H_98=1.4. This result is better than for the typical hybrid regime on DIII-D and is correlated with better than usual electron thermal transport in this LSN plasma shape. Therefore, this proposal will likely lead to the development of a high beta, high confinement, steady state scenario based on the hybrid regime. A half-day experiment in 2010 did not result in improved parameters despite the additional of a sixth co beam source because of 2/1 NTM issues. My hypothesis is that the 2.1 mode onset in hybrids, at least for cases well below the ideal wall limit, is related to having a too peaked pressure profile. This could explain several facets of the 2/1 mode onset, such as the dependence on the current evolution and the dependence on the confinement factor. We will need to pay close attention to the peakness of the pressure profile and find ways to decrease it if necessary, such as using the off-axis beam, changing the wall conditions or gas pre-fill levels. | ||
| Resource Requirements: | NBI: 6 co sources are needed. 210RT may be used to collect MSE data.
ECH: 6 gyrotrons with 4 MW of injected power. FW: It is desirable to couple 1 MW or more, but core absorption needs to be demonstrated. I-coils: Dynamic error field correction will be used (possibly broadband feedback). | ||
| Diagnostic Requirements: | MSE is critical. | ||
| Analysis Requirements: | TRANSP for current drive and transport, DCON for stability. | ||
| Other Requirements: | -- | ||
| Title | 334: Higher Beta ELM-Suppressed Hybrids | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Continue RMP ELM control experiments in the ITER shape by controlling the 3/2 NTM amplitude to allow higher normalized beta to be achieved without rotational slowing and mode locking. The 3/2 NTM amplitude can be decreased by either (1) optimizing the error field correction to obtain higher rotation rates, (2) using ECCD at the q=1.5 surface, or (3) trying higher q95 (>4) if a RMP resonant window exists. The goal is to obtain RMP ELM-suppressed hybrids with beta_N~3 (close to the ideal no-wall limit). If naturally dominant 4/3 NTM hybrids can be produced, then the ECCD suppression of the 3/2 NTM is not necessary. The higher q95 cases should work because the coupling between the 3/2 NTM and the wall is weakened as the resonant layer is moved closer to the plasma center. This ELM suppression experiment should first use only co-NBI, but once optimized results are obtained then lower rotation plasmas should be studied using balanced NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Repeat previous best ELM-suppressed hybrid case 129958. (2) Optimize error field correction to obtain highest rotation rates. (3) Use ECCD at q=1.5 surface to reduce and/or eliminate the 3/2 NTM. (4) Determine new beta limit during RMP ELM suppression. (5) Add counter NBI to slow rotation rate such that M<0.1. (6) Steer gyrotrons not needed for NTM control to core deposition to obtain Te=Ti. | ||
| Background: | Experiments in August 2007 used the I-coil to completely suppress ELMs in high beta hybrids for q95=3.7. If a dominant 4/3 NTM was present, then beta_N up to at least 2.5 could be achieved during the I-coil phase (actual beta limit not known). However, if a 3/2 NTM was present, then for beta_N>2.2 the plasma rotation slowed down rapidly during the I-coil phase and a locked mode terminated the hybrid discharge. While reducing or suppressing the 3/2 mode is one method to overcome this problem, improved error field correction may also help to reduce the amount of drag and help sustain the plasma rotation, especially in the pedestal region. | ||
| Resource Requirements: | NBI: All 8 sources are requested for long pulse operation.
EC: Minimum of 3 gyrotrons, prefer to have 6 gyrotrons. I-coil: Required with n=3 setup. Optimal error field correction is also needed. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 335: Hybrid Beta Limit at Low Rotation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Determine the beta limit (most likely due to the onset of a 2/1 mode) for low rotation hybrid plasmas. Compare the beta limit to the ideal no-wall limit, and to the limit in rapidly rotating plasmas. First study q95=4.2, and if time permits study q95=3 and q95=5 as well. Determine how sensitive this limit is to error field correction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish hybrid plasma with q95=4.2 using PCS feedback control of beta_N and rotation. (2) Use counter NBI to reduce rotation rate to minimal value. (3) Program a slow ramp up in beta_N to determine the stability limit (likely limit is onset of 2/1 mode). (4) Try different types of error field correction (i.e., optimal, dynamic, broadband, etc.) to see if this effects the beta limit. (5) Repeat beta_N ramp using only co-NBI. (6) If time permits, repeat at q95=3 and q95=5. | ||
| Background: | The bulk of hybrid study on DIII-D has been with co-NBI, producing rapid plasma rotation. For q95~4, the beta limit appears to be no less than the ideal no-wall stability limit (as long as the current profile is fully penetrated and the walls are properly "conditioned"). At q95~3, the beta limit has been reported to be 80% of the ideal no-wall limit. At q95~5 at low density, or at lower q95 at high density (and thus a broad pressure profile), hybrid plasmas have considerably exceeded the ideal no-wall limit.
However, it is expect that in future larger devices like ITER the normalized rotation rates will be much smaller. Experiments in standard H-mode plasmas indicate that the beta limit decreases at lower rotation. Therefore, we need to measure the beta limits for low rotation hybrid plasmas at different q95 as part of our validation of this scenario for ITER. Previous experiments on DIII-D showed that there was a limit in how low the rotation could be reduced even for beta_N<3 before the onset of a 2/1 mode (Politzer NF 2008). Further experiments showed that this low rotation limit was not due to the presence of the 3/2 NTM. A profile analysis of these low rotation cases shows that the toroidal rotation near the pedestal region is near zero or slightly negative, a consequence of the counter torque density near the edge for balanced-NBI. If this edge feature is the cause of the 2/1 mode limit for balanced-NBI, we should look for ways to drive the edge rotation away from zero, either by improved error field correction or by a different edge NBI torque (either more positive or more negative). | ||
| Resource Requirements: | NBI: All 8 sources are requested for long pulse operation. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 336: Diagnose to death ELM and no-ELM suppression & plasma topology | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | A key element of understanding the RMP-ELM suppression effect is being able to see the processes involved. This has two considerations at least. Firstly diagnosing the field topology on which the effect is based. And secondly diagnosing the particular processes that lead to transport. This proposal suggests some ideas for steps to address these two key aspects, which may relate to various models of the effect. The other theme of this proposal is that we should make very careful comparison of such measurements between ELM suppressed and non-ELM suppressed plasma to see if we can identify any key differences that may elucidate the physics mechanisms | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiment plan here would be to repeat a sequence of key discharges in ELM suppress or not conditions, with optimized diagnostics to see the various process / conditions outlined below. A key element may to see how things vary between ELM suppressed and not conditions - eg by varying q or I coil current.
DIAGNOSIS: A starting point is we need to ensure we get the best possible edge plasma reconstruction, so we can have an accurate knowledge of the symmetric part of the equilibrium structure before RMP is applied. Here, the new edge TS is important to provide better reconstructions, probably in a range of plasma of ELKM/non-ELM RMP suppression. In addition, diagnostics that may directly measure the field structure or plasma response should be considered - can we identify where the laminar (ideal response) nexted flux surface structure exists and where "ergodisation" or islands form? Use of RCP, interferometer or ECE may help. Can we see finger, filaments or island directly - will impurity injection help illuminate these? Use of interferometer or new TS may help identify density jumps or perturbations associated with the hypothesized NTV from singular currents idea. If we pulse RMP fields off can we structure start to 'spin up' past diagnostics? Looking for structure on magnetics may be good, but may only be possible if they are rotating. Finally we should deploy turbulence diagnostics - can we temperature fluctuations, and do these correlate with the ELM? CONCLUSION: Probably there is much else - we should try and collect all such techniques and have a concerted go at getting everything in one set of discharges with varying conditions | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | THIS ENTRY IS NOT TO LAY CLAIM TO THIS TOPIC, but to ensure relevant ideas go forward for consideration and joint planning by the group. I remain happy to help as team member or experiment leader as needed. | ||
| Title | 337: Current profile modification in Hybrid Discharges by ECH in current ramp | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General SSI | Presentation time: | Not requested |
| Co-Author(s): | P.A. Politzer | ITPA Joint Experiment : | No |
| Description: | Apply ECH during Ip ramp and evaluate access to hybrid discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Compare a normal AI discharge and then add EC
1. EC power scan 2. EC heating vs. ECCD configuration 3. q95 scan, compare a low q95 standard discharge to one with best results from #1 and #2 | ||
| Background: | An ansatz of AI work is that the current profile plays a major role in access to this regime. Changing the current profile with EC during the entire ramp and up to the start of the hybrid phase (both EC heating and ECCD) will allow us to evaluate this | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | mse is a crucial diagnostic | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 338: Dependence of Stiffness on Elongation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Scan the temperature gradient at fixed density and fixed pedestal height using a tilt scan of the 150 beam plus an ECH deposition scan to determine the stiffness of transport. Repeat this at various values of plasma elongation. Compare results with GYRO, TGLF, GLF23, and MM theory-based models. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish H-mode plasmas with density control using primarily the 150 beamline and 6 gyrotrons. (2) In steps, scan the deposition of the heating power by tilting the 150 beamline and ECH launcher angles. If the 150 beam can be only tilted overnight, then do an ECH deposition scan with one beam tilt angle in the afternoon, and then the following morning repeat the ECH deposition scan for the other beam tilt angle. (3) Try to maintain constant ExB shear, Ti/Te ratio and pedestal height. This may be obtained maturally, but if not use the counter beams to control the rotation, and the overall heating power can be adjusted to keep the pedestal height fixed. (4) Repeat scans at elongations between 1.5 and 2.0. | ||
| Background: | Confinement databases have not been able to clearly distingush between the GLF23 and Multimode theory-based models, probably because the most sensitive parameters are not clearly varied. Perhaps the largest difference between GLF23 and MM is the level of transport stiffness, especially in the outer regions of the plasma. In addition, these models have a very different elongation scaling of the stiffness value. Studying this experimentally should help us to validate (or disprove) these two theory-based models. | ||
| Resource Requirements: | NBI: 150 beamline critical; other beams desirable.
EC: 6 gyrotrons required. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 339: Measurement of Inductive Poloidal Current | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | P. Politzer | ITPA Joint Experiment : | No |
| Description: | Measure the poloidal current density profile induced by ramping the toroidal field coil. Compare with the poloidal current expected from the parallel Ohm's law to determine if the perpendicular conductivity is large enough to give a significant contribution. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Study H-mode plasmas with beta_pol near unity so that the "natural" poloidal current is negligible. Compare discharges with positive and negative ramps of the toroidal field to cases with no BT ramping. Keep the plasma current, density, and temperature constant during these ramps. Study two cases, a low electron temperature plasma with NBI heating only, and a high electron temperature plasma using ECH in addition to the diagnostic beams. | ||
| Background: | The magnitude of the perpendicular conductivity has not been measured to my knowledge in tokamaks. In this experiment, ramping the toroidal field will induce a poloidal electric field that can be exactly computed using Faraday's law. Multiplying this E_pol by the parallel conductivity gives the parallel contribution to Ohm's law, while multiplying E_pol by the perpendicular conductivity gives the perpendicular Ohmic current density. Using the MSE data (although not necessarily equilibrium reconstruction), both the poloidal current density and the parallel current density can be measured. By comparing plasmas with and without a BT ramp, it will be possible to determine if the measured change in the parallel current density is enough to explain the total measured poloidal current (i.e., the perpendicular conductivity is negligible). | ||
| Resource Requirements: | NBI: Co and counter MSE beams, and at least 2 other sources.
EC: Minimum 6 gyrotrons. | ||
| Diagnostic Requirements: | MSE is critical. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 340: Looking for plasma resonant response to 3D fields | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | The response of the plasma to 3-D fields remains puzzling, particularly near the edge with the interaction on ELM suppression. In particular the degree of penetration, island formation or ergodic response remains a a key aspect to understand. Thus two techniques are proposed to perturb the resonant response and look for changes either in plasma itself, or ELM suppression effect. These are: (i) Change plasma rotation - lower rotation should enable fields to penetrate more readily, thereby enhancing any potential ergodization effects; (ii) Change field from resonant to non-resonant - switching parity of n=3 fields changes the balance from resonant to non-resonant fields [Lanctot APS 2010] when plasma response is included. this may be a good way for again looking for penetrative process, or processes that depend on shielding. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | (i) Apply pulsed RMPs (for maximum diagnosis) in high and low rotation in plasmas which are either successfully ELM suppressed by RMP or not. Some work may be need to make a suitable lower rotation clearly ELMy scenario - ideally this should have some close to zero rotation (not zero torque). (ii) Apply odd and then even parity RMP in matched cases. Again use two types of case, one in which ELMs can be suppressed, another outside the window. Further insight may also be gained by mixing in C-coil fields to cancel out particular resonances. NB: In both cases we want to deploy maximum diagnostics (eg see prop 336, which may synergize) to see what is going on with plasma structure and transport. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | We should check that this basis does not already exist! But may be valid anyway, if we have better diagnosis possible. Need to do modeling of I coil + C coil response cases to check that we will apply resonant/non-resonant and for more complex mixes how we tailor fields further. | ||
| Other Requirements: | THIS ENTRY IS NOT TO LAY CLAIM TO THIS TOPIC, but to ensure relevant ideas go forward for consideration and joint planning by the group. I remain happy to help as team member or experiment leader as needed. | ||
| Title | 341: Extend zero NBI torque QH-mode operation toward ITER baseline objectives | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | K. Burrell | ITPA Joint Experiment : | No |
| Description: | The goals of this experiment are:
- Demonstrate zero-torque QH-mode using NRMFs at q95 ~ 3.0 in ISS discharges. - Develop zero-torque ELM-free path to zero-torque QH-mode state of DIII-D discharge 141439 (3 |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | - Starting from discharge 141439, find NBI torque requirements for QH-mode operation at q95~3. Use Bt and Ip ramps to get to target q95 with minimal perturbations of the early phase of the discharge.
- Reduce NBI torque in early phase, 1 | ||
| Background: | The ELM-stable regime of QH-mode is sustained in DIII-D discharge 141439 for ~1 s with zero NBI torque and at or above the ITER baseline target values of H89 and betan: H89ā?„2 and betanā?„2. Important steps to improve the parameter match of ITER baseline scenario are to:
- change the value of q95 from ~5 to ~3, - extend the zero NBI torque operation to the entire discharge duration, - reduce Zeff from ~4 to ~2. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 342: Modulation of Bootstrap Current | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | D. Thomas | ITPA Joint Experiment : | No |
| Description: | Directly measure the bootstrap current profile near the H-mode pedestal by modulating the pedestal gradient using an oscillating I-coil current and measuring the oscillating MSE/LIB response. Ideally this should be done in an ELM-suppressed discharge, but ELMs will be tolerated if they cannot be avoided. It would be useful to make a fiducial comparison by modulating the edge ECH power in place of modulating the I-coil current. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish RMP ELM-suppressed discharge with q_95=3.5 with good MSE and Lithium Beam diagnostic coverage at relatively high field. (2) Modulate the I-coil current at 5-20 Hz to vary the pedestal gradients. Make the modulation depth as large as possible without having ELMs return. (3) Make the I-coil modulation depth 100% even if ELMs return. (4) Repeat previous step with q_95 out of the ELM suppression window. Changing q_95 should vary the bootstrap current. (5) Aim ECH for power deposition near the top of the H-mode pedestal. Modulate all gyrotrons using several different frequencies (5-20 Hz). | ||
| Background: | The bootstrap current profile near the H-mode pedestal strongly effects the plasma stability. If the bootstrap current density can be modulated, then the flux surface average value of the oscillating component can be determined by Fourier analyzing the pitch angles measured by MSE/LIB via the poloidal flux diffusion equation. This can be exploited to determine if the modifications in the pedestal gradients caused by the RMP really result in a change in the edge pressure-driven currents. The best method of modulating the bootstrap current is therefore to modulate the I-coil current. To check the method, it would be good to obtain a fiducial by applying modulated ECH near the H-mode pedestal [core ECH is not as desirable owing to (a) pulse pile up and (b) electron-ion collisional exchange]. | ||
| Resource Requirements: | NBI: Co and counter MSE beams, and at least 2 more sources.
EC: 6 gyrotrons. | ||
| Diagnostic Requirements: | MSE and Li Beam are critical. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 343: Effectiveness of ECEI for RMP characterization | ||
| Name: | Tobias tobias@lanl.gov | Affiliation: | Los Alamos National Laboratory |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | N.C. Luhmann, Jr., C.W. Domier, R. Boivin, M.E. Austin, R. Nazikian | ITPA Joint Experiment : | No |
| Description: | Correlate ECEI images in the optically thin or grey edge regions of H-mode plasmas with modulated even and odd parity RMPs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish H-mode plasma with modulated, slowly rotating RMP fields. Correlate ECEI data to RMP modulation, obtaining 2D images of fluctuating T_rad in (1) high resolution near midplane and (2) wide zoom for data 25-30 cm above and below midplane. Obtain density profile and fluctuating density correlated to RMP to determine (1) T_rad/Te and (2) relative contributions of fluctuating Te and ne to T_rad. Additionally, simultaneous measurement of fluctuating Te in the optically thick core (also correlated to RMP) will provide a boundary condition for interpretation. | ||
| Background: | The utility of ECEI in optically thin and grey edge regions has not been established experimentally. A comparison of T_rad to known fluctuating magnetic field quantities and measured fluctuating ne will allow for an evaluation of proposed schemes for interpretation and provide further direction. This experiment is proposed for H-mode plasmas with limited overdense regions, allowing the contribution ECE emission reflected from the vessel walls to be ignored. This work is motivated by the prospect of applying ECEI to a wide variety of experiments including ELM control and suppression, non-inductive start-up, 3D field penetration, H-mode pedestal physics, L-H transition, etc. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | ECEI, edge density (Thomson, reflectometers) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 344: Can Nitrogen impurity Seeding Improve Radiating Divertor Performance? | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor Detachment and Plasma Flows | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | A direct comparison of an argon-seeded radiating divertor with a nitrogen-seeded radiating divertor is the focus of this experiment. Two major tokamak research facilities (Asdex-U and C-mod) have reported that nitrogen has several advantages over other seeding impurities, such as argon. We propose to test this hypothesis on DIII-D, as described below. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Our base case H-mode plasma is a lower single-null divertor with the ion gradB drift direction toward the X-point: Bt = -1.8 T, Ip=1.43 MA, q95 = 3.5, and Pinj=6-7 MW. The model plasma is shot 138548. These parameters are selected in order to facilitate direct comparisons with previous experiments, such the RMP argon-based radiating divertor. Deuterium gas is injected from he top of the vessel (UOB), while argon and nitrogen are injected into the private flux region of the lower divertor. First, nitrogen is injected into the private flux region at a non-perturbing level and three levels of a steady deuterium puff are used on successive shots, that is, 0, 40, and 80 torr l/s. At the highest deuterium injection rate, take two additional shots at perturbing levels of nitrogen. This process is repeated for the argon injection case, so that, in total, there are ten shots. Impurity accumulation in the core plasma, distribution of radiated power, energy confinement time, and heat flux reduction as a function of pedestal density and collisionality. | ||
| Background: | Both Asdex-U and C-Mod have reported significant improvement in energy confinement by seeding nitrogen in their H-mode plasmas. Low Zeff plasmas were typical of these plasmas. In addition, nitrogen also led to a sharp drop in divertor heat flux. DIII-D has focused on argon seeding, and while getting reasonably good results in terms of maintaining energy confinement time reasonably well and reducing divertor heat flux, the results from Asdex-U and C-Mod suggest that DIII-D could do much better. In fact, A. Kallenbach has suggested that DIII-D at least examine this possibility. This experiment does just this but with only a relatively small investment in experimental time. | ||
| Resource Requirements: | Machine time 0.5 day (in forward Bt), minimum 5 co-beams. | ||
| Diagnostic Requirements: | Asdex gauges (, core Thomson scattering, lower divertor fixed Langmuir probes, and CER. | ||
| Analysis Requirements: | UEDGE and ONETWO | ||
| Other Requirements: | -- | ||
| Title | 345: RMP ELM Suppression at the NTV Offset Rotation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Establish RMP ELM suppression in a plasma with mild counter rotation. Allow the rotation to "lock" to the offset rotation given by NTV, using additional NTV torque from the n=3 C-coil. Evaluate the confinement and stability properties of this discharge. Compare even and odd parity to vary the relative contributions of resonance and nonresonant effects. The NTV offset rotation frequency should be made as large as possible by operating at low Ip (i.e. low Bp) and low density (i.e. high Grad_Ti). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Use reverse Ip configuration so that most of the neutral beams are injecting in the counter direction. (2) Establish ELMy H-mode plasmas with Ip=1.0 MA and q_95=3.6. Lower Ip may be used if the beam ion confinement is good enough. (3) Start with even parity of I-coil. Apply RMP to suppress ELMs, with the n=3 C-coil added for additional (counter) NTV torque. Allow the density to pump out to a low level to obtain a high gradient in the ion temperature. (4) Determine the sensitivity of the toroidal rotation rate during RMP application with the amount of counter-torque injection. If the effect of nonresonant braking is large, then the toroidal rotation should be a stronger function of the NTV offset velocity than of the NBI torque. (5) Compare even and odd parity of I-coil, ideally in same discharge if SPAs are used. | ||
| Background: | For co-rotation discharges, applying the RMP to suppress ELMs results in a reduction of the toroidal rotation. This reduces the confinement time, and also can lead to locking of the plasma if the resonant braking effect becomes large. It is predicted that the nonresonance braking effects of an RMP coil on ITER may dominate over the co-torque injection from neutral beams, in which case the toroidal rotation on ITER should "lock" to the NTM offset value. This experiment proposes to study the consequences of this effect by starting with a counter rotation frequency close to the NTM offset value. | ||
| Resource Requirements: | Reverse plasma current configuration.
RMP I-coil configuration. Use SPAs so that even and odd parity can be compared in same discharge. C-coil in n=3 configuration. | ||
| Diagnostic Requirements: | |||
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| Other Requirements: | |||
| Title | 346: Effect of Islands on ECCD | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | R. Prater | ITPA Joint Experiment : | No |
| Description: | ECCD is an important tool to control MHD, such as tearing modes. While DIII-D has done detailed studies of ECCD, these have been for an axisymmetric plasma. The helical perturbations from tearing modes may significantly change the ECCD profile, which in term could affect its application to MHD control. This experiment will examine two facets of the effect of islands on ECCD. First, the flux-surface-average parallel current density will be compared for deposition at the island O-point or X-point. Second, the ECCD profile will be decomposed into separate toroidal and helical components. This second case requires a slowly rotating island, which can be achieved using entrainment with the I-coil. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Part I: Effect of islands on flux-surface-average parallel EC current density. (1) Target plasma is to be taken from successful modulated ECCD experiment to stabilize the 2/1 NTM. Probably a mixture of co/counter NBI will be used to slow the island rotation frequency to <5 kHz. (2) During the ECCD measurement phase, the 30LT and 210RT beams should be on continuously for MSE data acquisition, (3) With EC deposited at the island O-point, compare co/radial/counter ECCD injection. For the co-ECCD case, the power should be limited so that the island is NOT stabilized. (4) Repeat last step for ECCD deposition at the island X-point. (5) Repeat last step with continuous ECCD (i.e. not modulated).
Part II: Helical current from ECCD (1) The target plasma should be taken from a successful entrainment experiment where the I-coil is used to force a 2/1 tearing mode to rotate at a frequency <1 kHz. (2) Apply co/counter/radial ECCD at the q=2 location continuously (i.e. not modulated). (3) Compare co/radial/counter ECCD injection. (4) Compare modulated ECCD at island O-point or X-point for co/radial/counter injection. | ||
| Background: | Experiments on DIII-D over the last 10 years have made detailed comparisons between ECCD theory and experiments on the local level. However, these experiments specifically avoided MHD such as sawteeth and tearing modes. Thus, the ECCD studies were done in a axisymmetric plasma configuration. The highly localized region of ECCD led to the development of methods for direct analysis of the MSE signals without equilibrium reconstruction. This direct analysis method was able to determine the ECCD profile with spatial resolution limited only by the MSE diagnostic itself. Later, this methodology was extended to include the helical perturbations from tearing modes. The helically perturbed current for a m/n=2/1 "quasi-stationary" mode was successfully determined using MSE data and was reported at the 2006 EPS meeting. | ||
| Resource Requirements: | NBI: Both co and counter beams are required.
EC: 6 gyrotrons are required. I-coil: Entrainment of rotating 2/1 mode required for Part II of this experiment. | ||
| Diagnostic Requirements: | MSE is critical, with highest time resolution possible. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 347: Use top coil to develop multiharmonic error correction | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The best error correction on DIII-D achieves is a 38% reduction in the error field low density limit. It is possible that ITER may need much better correction - this will require more than one 'knob' (ie coil set) to achieve. DIII-D can research how to do this if it has a much larger source of error field, in addition to its I and C coils for correction. Use of the top coil as a source of simulated intrinsic error will achieve this, and provide a known source of error - key for model testing - and with very different structure to the correction fields (likely ITER situation, and also key for model testing). Therefore it is proposed the top coil be re-installed on DIII-D ASAP, and experiments commence with it to understand how to achieve the better error field correction that ITER needs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Installed Top coil. Energize. Conduct experiments with I coils and C coils separately to measure equivalent error field correction. Then combine I and C coils to see if correction can be improved. | ||
| Background: | (This experiment would also provide a large source of field to test plasma response and stability in high beta plasma - where the plasma response may inform us about mode non-rigidity issues). | ||
| Resource Requirements: | Top coil connected to PS (and cooling?) | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 348: New Optimal Plasma Shape for AT Scenario? | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Steady State) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | For the high q_min, steady-state AT scenario, switch the plasma shape from the standard unbalanced DND shape to the lower SND shape in shot 129323. The new plasma shape is proved to have high beta limits and low electron heat transport for the low q_min hybrid scenario. If these properties are present in the q_min>1.5 AT scenario, the result will be (1) higher electron temperature (and higher confinement), and (2) higher noninductive current fraction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main objective of this experiment is to repeat the high-beta, steady-state AT scenario but with the plasma shape given by shot 129323. The heating waveforms during the current ramp up phase will been to be optimized to raise q_min above 2 at the beginning of the flat top phase. If stronger cryopumping is desired to reduce the plasma density, than reverse BT direction may be required. | ||
| Background: | During an ECCD stabilization experiment in 2007, it was recognized that the discharges developed had some interesting properties (example: shot 129323). Although RWM feedback stabilization was not being used, the plasma beta exceeded the ideal no-wall limit with beta_N reaching 3.5 before the beam power topped out. Even more interesting was the fact that the core electron temperature was ~1 keV higher than normal for the hybrid scenario. This was a result of a much lower than typical electron heat transport. Usually for the hybrid scenario in the standard AT plasma shape, heat loss through the electron channel is dominant. This is attributed to ETG-scale turbulence. However, for the lower SND shaped used in this ECCD experiment, the electron heat loss was much lower than the ion heat loss. This plasma shape was used for high-beta, steady-state hybrid experiments in 2008. Here it was found that even with 3.0 MW of ECCD and Te=Ti except near the axis, the confinement time remained high with H_98=1.4. This is a much better transport result than for ECH hybrid experiments in the standard AT plasma shape where H_98 normally drops below 1.1. | ||
| Resource Requirements: | NBI: All co beams required.
EC: All 6 gyrotrons required. BT: Reverse BT direction may be desired for improved density control in lower SND shape. I-coil: Dynamic error field correction is desired. | ||
| Diagnostic Requirements: | |||
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| Other Requirements: | |||
| Title | 349: Dust transport in detached divertor plasmas | ||
| Name: | Smirnov none | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | S.I. Krasheninnikov, D.L. Rudakov | ITPA Joint Experiment : | Yes |
| Description: | The previous dust injection experiments performed on DIII-D with carbon dust introduced in the lower divertor demonstrated that ~2% of the injected dust material penetrated to the core plasma resulting in a factor ~2 increase of the core carbon concentration and the total radiated power. However, these experiments were performed in attached divertor plasma regimes with relatively high plasma temperatures ~30-35eV near the dust launch location. The mobilization and transport of dust in detached divertor regimes may significantly differ from that due to modification of divertor plasma flows and dust ablation patterns. Studies of dust transport in detached divertor can be particularly important, considering that divertor detachment will be required in ITER for the heat load management. Characterization of the dust transport in the detached divertor and penetration to the core plasma can be done with fast camera imaging of the dust injection region and simultaneous monitoring of the core impurity density and radiated power. Analysis of time evolution of core impurity density and comparison of the obtained results with the previous dust injection experiments in attached divertor regimes can provide additional data on dust dynamics and mobilization in divertor. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Carbon dust in amount of a few tens of mg will be introduced in the lower DIII-D divertor using the DiMES manipulator. Dust will be placed on a holder leveled with the divertor floor tiles; it will be mobilized by sweeping the outer strike point (OSP) over the holder. Dust mobilization and dynamics is monitored with the fast framing cameras simultaneously with the measurements of the core impurity density and total radiation power. | ||
| Background: | It is expected that large quantities of dust in future fusion devices, such as ITER, will be formed in divertor region, where particle and heat load on plasma wetted surfaces is the largest. The dust formed in divertor is then accelerated by various forces, most notably plasma drag force, and transported through the plasma, where it partially or completely ablates providing volumetric source of impurities. Due to small charge to mass ratio the dust can penetrate deeply toward the plasma core as compared to ion impurities. | ||
| Resource Requirements: | 2-3 high-density LSN H-mode discharges with detached OSP. LSN patch panel allowing sweeping OSP over DiMES is required. | ||
| Diagnostic Requirements: | Lower divertor tangential TVs, DiMES TV, UCSD fast camera, bolometers, filterscopes, SPED, CER. Fast camera(s) coupled to tangential and/or vertical view of DiMES highly desirable. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 350: ECE Imaging of ELM-NTM coupling | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The ECE imaging camera being developed by U.C. Davis and collaborators will be used to measure the changes in the electron temperature profile during ELM events in hybrid plasmas. The 2D images will help us understand the physics behind this coupling, and perhaps improve our understanding of magnetic flux pumping in hybrids that maintains the safety factor minimum slightly above unity. Both n=2 and n=3 tearing modes will be studied. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment can piggyback on another hybrid experiment as long as the toroidal magnetic field is high enough (BT~2 T). The target hybrid plasmas should have type-I ELMs with ~40 Hz frequency and q95>4 so that sawteeth are suppressed. We want to image both the usual hybrid case with a 3/2 NTM as well as hybrids with a dominant n=3 NTM (such as 4/3 or 5/3). We will likely not want to use ECCD to stabilize the 3/2 NTM because the required filtering needed to remove the 110 GHz radiation will compromise the ECE images (this needs further study, however). Conditional averaging over many ELMs will be used to improve the SNR of the ECE imaging diagnostic. | ||
| Background: | Using the ECE radiometer array, a modification in the electron temperature profile was observed previously during ELM events near the rational surfaces for 3/2 and 5/3 NTMs (but interestingly, not for 4/3 NTMs). This demonstrated a clear coupling between ELMs and NTMs, but the physical mechanism is not clearly understood. | ||
| Resource Requirements: | NBI: 6 co sources are requested for long pulse lengths. | ||
| Diagnostic Requirements: | ECE imaging and MSE are critical. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 351: Dust formation in MARFE plasmas | ||
| Name: | Smirnov none | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | S.I. Krasheninnikov, D.L. Rudakov | ITPA Joint Experiment : | No |
| Description: | The calculations done in support of recent experiment on JET indicate that MARFE emission in IR can not be explained by bremsstrahlung radiation along. It was suggested that the excess IR emission may be produced by thermal radiation from nano-scale dust grains. However, verification of this result is required. If confirmed, the formation of dust in MARFE plasmas may have significant implications on development of MARFE due to plasma recombination on dust surface. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use IR and near-IR filtered camera imaging to measure radiation from MARFE. Thomson scattering and FIR interferometer can be used to measure electron temperature and density in the MARFE. Additional deuterium gas puffing may be requireed to induce MARFE formation toward the end of a shot. | ||
| Background: | The nano-scale hydro-carbon dust has been collected from limiter in TS tokamak [1] and from below diveror region in JET [2]. It is thought that such dust can be formed in relatively cold and impurity reach detached divertor and MARFE plasmas. Formation of the nano-scale dust in ITER may have implications for MARFE instability process and tritium retention due to large dust surface area. It was also suggested theoretically [3] that such dust may affect collective behavior of plasma due to large dust number density, while the total dust quantity may remain low.
[1] P. Roubin et al., J. Nucl. Mater. 390-391 (2009) 49. [2] A. Murari et al., Plasma Phys. Control. Fusion 50 (2008) 124043. [3] S.I. Krasheninnikov et al., (to be published in J. Nucl. Mater., 2011) | ||
| Resource Requirements: | 3 high density shots (dedicated of shared, operation mode and configuration TBD) with additional gas puffing towards the end to induce MARFE | ||
| Diagnostic Requirements: | IR camera, lower divertor tangential TV with near IR filter, UCSD fast camera with near IR filter, Thomson scattering, FIR interferometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 352: Determine if NRMF from C-coil allows zero-NBI-torque QH-mode | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, J.M. Park, W.M. Solomon, M. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to demonstrate that low rotation QH-modes can be created using nonresonant magnetic fields (NRMF) from the C-coil only. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize the basic plasma which was exploited in ROF proposal 199 while energizing the C-coil at maximum allowable current in an n=3 configuration. Scan neutral beam power to scan beta, since theory and previous experiments indicate that the NRMF torque is higher at higher beta. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. The best results were obtained using both the C-coil and the I-coil. For work on future devices, it would be advantageous if the nonresonant fields could be produced using coils outside the vacuum vessel. Accordingly, it would be of interest to see if the low rotation QH-mode could be produced using the C-coil only. A decision on this experiment should be made after the completion of ROF proposal 199, which investigates the effect of various I and C-coil combinations and compares these with predictions from IPEC. If those results indicate a reasonable probability of success, then this experiment should be done. | ||
| Resource Requirements: | Reverse Ip. C-coil equipped with new leads to allow three C-supplies to simultaneously feed the C-coil | ||
| Diagnostic Requirements: | Standard profile and fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 353: Improving error field resilience with q=2 ECCD | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Error fields now appear to be a major concern for inducing disruptions in the ITER baseline H mode, as well as Ohmic plasmas. The process of error field penetration involves island formation. And particularly in H-mode, the island often forms rotating rather than being a direct penetration - ie the process is not simply complete loss of torque balance, but rather an island being destabilized. Therefore it is likely that pre-emptive ECCD may prevent the tearing process directly, by improving the underlying tearing stability or heating/driving currents in any small islands that form. This may be crucial for the ITER baseline. it may also be highly relevant for other scenarios where error fields can brake the plasma enabling tearing. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Apply pre-emptive ECH to low torque H mode plasmas with an error field ramp to see if the ECH raises the field threshold. Scans should be made in ECH deposition location relative to q=2, and in heating: current drive fraction. | ||
| Background: | |||
| Resource Requirements: | |||
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| Title | 354: Test of Turbulence Spreading Using Turbulence Propagation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The question of turbulence spreading, that is, whether turbulence is or is not a strictly local phenomenon, can be precisely tested by modulating the turbulence (and plasma profile) at a fixed location and then monitoring the propagation of the turbulence (and plasma profiles) away from this region. If the turbulence propagation speed is much faster than the temperature or density propagation speed, then this can be attributed to turbulence spreading. For this purpose it does not matter much how the turbulence is modulated; it can be a simple amplitude modulation or something more sophisticated such as a modulation of the radial correlation length. The most likely source of modulation is ECH, either as a monopolar change in the electron temperature profile or as a "swing" experiment where the ECH deposition is alternated between two (closely spaced) location. The turbulence diagnostic must be capable of covering a large radial range, so the 32 channel linear array of the BES diagnostic is ideally suited for this experiment. An 8 channel DBS diagnostic would also be useful to monitor the propagation of intermediate k turbulence. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To minimize MHD, this experiment will use an L-mode plasma with 1-2 sources of continuous NBI for diagnostic purposes (BES, CER, MSE) and 6 gyrotrons for turbulence modulation. If the beam power needs to be limited to 1 source, repeat shots can be taken to switch between beams. The ECH modulation rate should be relatively high (~100 Hz) to allow an accurate measurement of the propagation speed. Actually it is preferable to study several different modulation rates, so repeat shots will be taken to cover the range 25-200 Hz. | ||
| Background: | While the ECH "swing" experiment led by Jim DeBoo has similarities to this proposal, in that case the ECH modulation was too slow to obtain the phase delay information that is crucial to this proposal. Also the radial spread of the tubulence modulation was limited in DeBoo's case, perhaps a consequence of the "swing" arrangement. Therefore, a monopolar modulation of the ECH at relatively high frequency is preferred for this proposal. | ||
| Resource Requirements: | Beams: 30LT, 330LT, 150LT
ECH: Six gyrotrons | ||
| Diagnostic Requirements: | BES 32 channel linear array
DBS 8 channel array | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 355: NRMF-assisted QH-mode in wave-heated discharges | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | deGrassie, Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to obtain QH-mode in discharges that are mostly heated by ECH and FW, similar to ITER discharges. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make ELMing ECH H-mode. Maximize betaN by using maximum ECH and trying to couple FW power. Apply I-coil n=3 field during ~steady state phase. Vary I-coil current.
Work on keeping density constant using feedback controlled gas puffs. If this deteriorates confinement try injecting pellets. If confinement becomes too low to keep betan~1.8 with ECH only, add some balanced NBI power. Measure velocity profile with a) standard first blip and b) the continuous, sparse balanced blips developed by deGrassie in 2008. | ||
| Background: | Experiments in 2009 tested for the first time the effect of n=3 fields on ECH H-modes. Only odd parity n=3 I-coil fields were used. At largest field amplitude, very strong
density pump-out was observed, accompanied by strong reduction of beta and rotation. In one case, the rotation went down to zero with still a finite beta H-mode (137227). If we can maintain betaN in such a discharge, we may observe the rotation cross into counter direction, and enable QH-mode for the first time in wave-heated plasmas. | ||
| Resource Requirements: | ECH and FW, in addition to balanced NBI. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 356: Investigate elongation limit in low li discharges with/without applied 3D magnetic fields | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Requested |
| Co-Author(s): | L. Lao, R. Buttery, M. Chu, A. Collier, N. Ferraro, A. Reiman, A. Turnbull | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to produce high kappa ~ 2.7 discharge with low li to provide target for high beta + 3D perturbation study | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using the formation technique employed to produce low-li DIII-D discharge #122976 (early beam heating, Bt and Ip ramps), attempt to reproduce the profiles of 122976 in a discharge of smaller minor radius (a~50 cm).
At li~0.5, the elongation is predicted stable even beyond k~2.6. Look for n=0 stability limit. Apply maximum n=3 I-coil perturbation, look for an effect. Compare with MHD calculations. | ||
| Background: | A possible upgrade of DIII-D, currently under study, consists in the installation of nonaxisymmetric coils above and below the midplane, capable of applying a field large enough to improve the vertical stability of highly elongated plasmas [Rieman, PRL 99, 135007 (2007)].
High elongation is expected to be beneficial for both confinement quality and stability. FNSF and DEMO studies rely on high elongation to reach very high fusion performance. Stable highly elongated plasmas can also be produced by using very low li (~0.5). Recent calculations by Lang Lao show that a low li DIII-D discharge like 122976 would be stable to n=0 with kappa up to 2.66 with the DIII-D plasma-wall distance. This proposed experiment would provide us low li plasmas with the very high elongation that 3D fields could enable in moderate li as well. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 357: How is intrinsic rotation affected by ELMs | ||
| Name: | Nave none | Affiliation: | IPFN, IST |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | M.F.F.Nave, J. de Grassie, | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
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| Title | 358: Test up/down asymmetry effects on intrinsic rotation | ||
| Name: | Nave none | Affiliation: | IPFN, IST |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | M.F.F.Nave, J. de Grassie, | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
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| Title | 359: Stabilize tearing modes directly with ECCD in SS regime | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | As many steady state plasma attempts are terminated by the occurrence of tearing modes, particularly the 2/1 mode, we should explore whether these can be effectively stabilized or kept small by direct ECCD mode control in the island itself, with modulated tracking ECCD, switching to pre-emptive ECCD when the mode goes away. While it is of course desirable to use ECCD to try to tailor a more intrinsically tearing stable plasma, this technique has born little fruit in 2009-10, while there is a pressing need to provide target plasmas to explore the development of regimes with the off axis beams. Switching ECCD to this mode control role, may be a efficient practical step towards this end - we can go back to more generalized profile tailoring with ECCD in later years. Application of the technique may also tell us something useful about the requirements for mode control in AT regimes - for example if they are delta-prime triggered they may be hard to eliminate entirely, and a more global profile modification may be needed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In SSI plasmas, deploy real time q=2 tracking of ECH, while regimes developed with off axis beams. Could also be switched to q=5/2 or q=3 modes if these become the problem. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | THIS ENTRY IS NOT TO LAY CLAIM TO THIS TOPIC, but to ensure relevant ideas go forward for consideration and joint planning by the group. I remain happy to help as team member or experiment leader as needed. | ||
| Title | 360: Maintain low-rotation QH-mode with ECH only | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to demonstrate that low-rotation QH-mode can be sustained with ECH, which provides plasma heating with no input torque. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | As discussed in the background section, the plasma configuration for this work will be the one developed in ROF proposal 83. The low-rotation QH-mode will be established using balanced neutral beam injection run up until about 3000 ms in the shot. At this point, we will switch from NBI to ECH to see if the plasma can sustain low-rotation QH-mode. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. Although the net neutral beam torque was zero for these shots, there were small variations in the torque density profile as a function of radius owing to fast ion orbit effects. A demonstration that low rotation QH-mode could be sustained by ECH only would make clear that these residual radial variations were irrelevant. The best plasma for this demonstration will be the one developed in ROF proposal 83 which allows QH-mode operation with the most co-NBI torque. This plasma will have a shape and C and I-coil configuration optimized for minimum intrinsic torque and maximum counter-torque due to NRMF. | ||
| Resource Requirements: | Reverse Ip. I-coil and C-coil in n=3 configuration with maxiumum possible current capability. | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 361: Prompt torque and zonal flow damping | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | J.S. DeGrassie, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to determine the damping rate of the zero mean frequency zonal flow and the plasma poloidal rotation by periodically perturbing the plasma rotation using modulated co and counter neutral beam injection. The beam modulation will be fast compared to the fast ion slowing down time, so that the modulation will primarily be due to the prompt torque caused by fast ion orbit shift. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment is best done in QH-mode plasmas, because they are high temperature and low density, which leads to long ion-ion collision times. In addition, they have long steady periods, which allows significant averaging. Use the prompt torque from the beam orbit shift to apply periodic co and counter torques to the plasma by modulating the co and counter beams out of phase. Orbit shift calculations show that the 210LT and 330 RT beams give approximately equal prompt torque profiles out to rho=0.6. This allows 330 LT and 30LT to be run continuously to get CER data. Experimentally, what we are looking for is the evolution of the induced poloidal rotation (or radial electric field) after the initial jump which occurs when we add an extra co or counter beam. The beam modulation period will be chosen so that there are several ion collision times within one beam on time; this will be between 10 and 40 ms. CER will be set to a short integration time, something like 2 ms. We can average over multiple pulses to improve the quality of the rotation measurement. We will scan ion-ion collision time by changing the ion temperature using different power levels and by changing the core density by using ECH to induce density pumpout. The ECH will also provide extra electron heating to increase the fast ion slowing down time. | ||
| Background: | When neutral beams deposit toroidal angular momentum in the plasma, they do so on two time scales, one for the momentum deposited perpendicular to the magnetic field and another for the momentum deposited parallel. The parallel momentum couples to the background plasma on the time scale of the collisions between fast ions and the background ions. The perpendicular momentum is deposited much more quickly, through a process involving radial currents. When a beam neutral ionizes, the resulting D+ ion travels on a orbit whose guiding center is shifted from the ionization point. For D+ ions born outside the magnetic axis, this shift is outwards (towards larger minor radius) for counter injected neutrals and inwards (towards smaller major radius) for co-injected neutrals. This shift represents a radial current of fast ions. Processes in the background plasma produce on offsetting radial current, which then imposes a torque on the background plasma. However, this offsetting radial current grows up on the ion-ion collision time. During this time, the poloidal rotation and the radial electric field both evolve. If we use out of phase modulation of the counter and co beams, we can periodically reverse this torque, creating a square wave modulation. If the modulation period is fast compared to the fast ion slowing down, we only need to consider the prompt torque. For a plasma with 15 keV central temperature and 5 x 10^19 m^-3 density, the fast ion slowing down time is greater than 100 ms even for the 1/3 energy component. The damping of the overall plasma poloidal rotation is the same as the damping time of the plasma electric field. Accordingly, CER measurements of any impurity ion can be used to determine the overall poloidal rotation damping. More importantly, this damping time of the plasma electric field is the zonal flow damping time, which is crucial to turbulence behavior. Theory predicts that this damping time is of order the ion-ion collision time which is around 20 ms in our candidate plasmas. | ||
| Resource Requirements: | Reverse Ip. 7 NBI sources. 3-4 ECH gyrotrons | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 362: Bootstrap Current Change During RMP ELM Suppression | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | D. Thomas | ITPA Joint Experiment : | No |
| Description: | Using the Li Beam and MSE diagnostics, directly measure the change in the H-mode pedestal bootstrap current that is caused by the change in the pedestal gradients when the RMP is turned on. The most accurate measurement would be a RMP/non-RMP comparison. It would be good to also look at cases where q_95 is outside the resonance window to see if there are any differences. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This should be a "piggyback" or background experiment, in that the data should be obtainable in the course of the RMP ELM-suppression experiments in 2011. The only criteria is that the plasma boundary be located in a favorable place for the edge MSE and Li Beam. In addition, long analysis windows are desired to reduce the random errors in the pitch angle measurements. The basic dataset would contain four discharges: (1) RMP ELM-suppressed case inside the q_95 resonance window, (2) repeat without RMP, (3) RMP case outside the q_95 resonance window (ELMs not expected to be suppressed), and (4) repeat without RMP. | ||
| Background: | Our physics picture of RMP ELM-suppression is that the RMP reduces the H-mode pedestal gradients, thus reducing the pedestal bootstrap current. This makes the plasma stable to ballooning-peeling modes according to ELITE calculations. We would like to test this experimentally by using the Li Beam and edge MSE diagnostics to directly measure the changes in the magnetic field pitch angles. The edge current density is proportional to the channel-to-channel derivative of the pitch angles. The pitch angle data can either be used in a equilibrium reconstruction, or analyzed directly. | ||
| Resource Requirements: | I-coil in RMP configuration. | ||
| Diagnostic Requirements: | Li Beam and MSE are critical. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 363: Diagnostic spatial cross calibration using edge sweeps in QH-mode | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | C. Holcomb, G.R. McKee, W.M Solomon | ITPA Joint Experiment : | No |
| Description: | Perform spatial cross calibration of the CER, BES and MSE systems using edge sweeps in QH-mode discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Run QH-mode discharges like 128542 with edge sweeps which change Rmidout from 2.29 m to 2.16 m. Tune the CER system to look at the Doppler-shifted D-alpha from the neutral beams. (BES and MSE already view this wavelength). Modulate the beams to obtain the needed data. The various beam combinations typically take 6 shots to complete. | ||
| Background: | In order to successfully combine data from the CER, BES and MSE systems for edge plasma studies, we need to know the relative spatial calibration of these system to millimeter accuracy. This has been done before using edge sweeps in QH-mode plasmas. This calibration needs to be done again so that we can finally include the MSE views of the 210 beam. This MSE portion of the calibration is particularly important, since the relative location of the 210 system relative to the other MSE systems has never been established. Establishing this location is an essential first step in using the co plus counter MSE views to determine the edge current density. | ||
| Resource Requirements: | Reverse Ip. 7 NBI sources. | ||
| Diagnostic Requirements: | CER, MSE, BES are essential. Standard profile diagnostics are also needed. ECE-I for EHO studies. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 364: N=1 Optimization with TBM at Low Torque | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | N. Oyama, V. Pustovitov, J-K Park, TBD... | ITPA Joint Experiment : | No |
| Description: | Optimize n=1 compensation of deleterious TBM error field effects in plasmas having slow toroidal rotation. The plasmas will be more representative of expected ITER plasmas than were the all-co-injected DIII-D plasmas that dominated in the 2009 TBM campaign. The proposed low- rotation experiments should be done after n=1 compensation has first been developed for the more familiar co-injected plasmas (proposals 129, 192) that were used extensively in 2009. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop low-torque, H-mode target plasmas using combined co- and counter NBI injection, balanced injection, and ECH. Optimize the TBM n=1 error compensation using experience gained first from co- injected H-mode plasmas (proposals 129 and 192), including dynamic error field correction (DEFC). Compare effectiveness of n=1 compensation on these low-flow plasmas vs its effectiveness on high-flow plasmas (proposals 129, 192). Monitor plasma rotation and magnetic responses closely. | ||
| Background: | TBM effects on slowly rotating H-mode plasmas were not studied systematically, except for L-H power threshold, during the 2009 TBM campaign. This was mainly because we made little progress developing suitable low flow target plasmas and we moved on to fully co-injected plasmas. However, ITER plasmas will have low (normalized) torque injection. It is important to test TBM effects on confinement and locked modes at low torque. Also, the low torque experiments should include the best n=1 correction of TBM error effects that will be developed for co-injected beams (proposals 129, 192). | ||
| Resource Requirements: | TBM mock-up, both counter beams, cryo pumps, Dynamic error field correction | ||
| Diagnostic Requirements: | CER, MSE, Thomson scattering, MHD magnetics (plasma response) | ||
| Analysis Requirements: | EFIT, SURFMN, IPEC, MARS-F, NTV calculations | ||
| Other Requirements: | -- | ||
| Title | 365: High Beta Hybrids and Pressure Profile Broadening | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Use 5 MW of off-axis beam injection to broaden the total pressure profile compared to on-axis injection. Most of this will be due to a change in the fast ion pressure profile, but some broadening of the thermal pressure profile may also occur depending upon how stiff the transport dependence is. Determine whether the broader pressure profile allows a high beta_N to be obtained in steady-state hybrid plasmas, with the goal being beta_N=4. Will also calculate whether the ideal wall limit changes significantly with the broader pressure profile for these low q_min discharges. The off-axis beam will probably not effect the current drive profile much since the off-axis NBCD efficiency remains high, and the poloidal magnetic flux pumping inherent in hybrids tends to keep the total current profile constant regardless of the driven current profile. Since the noninductive current drive is not crucial to this experiment, it could be done with either positive or negative B_T values.
For steady-state considerations, the co-ECCD should be deposited inside the q=1.5 surface for this experiment. However, we could broaden the scope of this experiment by re-directing some of the ECH power (probably 4 gyrotrons) to deposit at the q=2 surface to see if we can suppress the 2/1 mode. This would be deem a success if the beta limit comes from a RWM rather than a 2/1 mode (the latter is the current situation). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Ideally we would already have in hand a high beta, steady-state hybrid case with 6 co-/on-axis beams that would serve as a fiducial. The beta_N will likely be 3.4-3.5 given previous results. (2) Repeat the fiducial case, but using the 150 beamline tilting fully downwards. (3) Use the 150 beams at full power, but scan the NBI power for the other co-beams to vary beta_N. Determine the limit for the 2/1 mode, the goal being beta_N=4. (4) If time permits, compare the stability limit for cases where all the co-ECCD is deposited inside the q=1.5 surface, and where a minimum of 4 gyrotrons are aimed at the q=2 surface. | ||
| Background: | High beta hybrids have been operated stably (to the 2/1 mode) up to beta_N=3.8 at high density, which is well above the ideal no-wall limit. At lower densities and with central co-ECCD, high beta hybrid plasmas have been created with beta_N=3.4 and nearly zero loop voltage (9 mV). TRANSP calculations show that these plasmas should be very close to fully noninductive. The ideal wall stability limit is calculated to be around beta_N=4 by DCON. The near term goal of high beta hybrid research is to obtain beta_N=4 with zero loop voltage for as long as the beams will run. To give some overhead between the beta_N=4 goal and the ideal wall limit, some broadening of the pressure profile may be desirable. This can be achieved using the off-axis beam. Since the NBCD efficiency remains high even for off-axis injection (especially with positive B_T), we do not have to give up on the steady-state goal to do this experiment. | ||
| Resource Requirements: | NBI: Tilted 150 beamline is critical. All 6 co-beams are needed.
ECH: 6 gyrotrons required. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 366: Mitigation of carbon deposition on diagnostic mirrors by D2 gas puff | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Requested |
| Co-Author(s): | A. Litnovsky (FZJ), V. Philipps (FZJ), M. Matveeva (FZJ), C. Wong, N. Brooks, W. Wampler, A. McLean, R. Boivin, J. Watkins | ITPA Joint Experiment : | Yes |
| Description: | This experiment is a continuation of mirror tests for ITER diagnostics performed in DIII-D in collaboration with FZJ in 2005-2008. Earlier experiments in DIII-D with D2 gas puff demonstrated about 10-fold reduction of carbon deposition rate on diagnostic mirrors mounted on a DiMES holder compared to non-mitigated case. In that experiment D2 gas was puffed in the divertor ~12 cm toroidally away from DiMES. Since the gas feed was not localized near the mirrors, a large flow rate (~70 Torr*l/s) had to be used to achieve the deposition mitigation. Such high D2 injection rate was perturbing the discharge. For 2011 campaign a dedicated gas puff line will be routed to DiMES opening in the divertor shelf, which should allow deposition mitigation at lower flow rates, making it more attractive for possible use in ITER. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Experimental approach will be similar to earlier mirror tests performed in DIII-D. DiMES holder with two Mo mirrors will be kept in the private flux region of high-density LSN H-mode discharges with detached OSP. D2 gas puff will be through the new capillary routed intro DiMES hole and opening directly on one of the mirrors. | ||
| Background: | All laser and optical diagnostic systems in ITER will implement metallic mirrors as their first plasma-viewing optical components. Diagnostic mirrors will suffer from erosion, deposition and particle implantation leading to a degradation of their properties and impacting the performance of entire respective diagnostic systems in ITER. A robust solution is needed to ensure the optimal performance of diagnostic mirrors in ITER. The development of the deposition mitigation techniques is of crucial importance on the way towards such a solution. The investigations on first mirrors are presently recognized as High Priority Task of the ITPA Topical Group on Diagnostics and are the subject of IEA-ITPA Joint Experiments Program (Task DIAG 2). The importance of the R&D program on first mirrors was outlined in the recent ITER Diagnostics review, carried out in Cadarache, France in July 2007. Presently, the Work Plan of the R&D program on diagnostic mirrors is under development and work packages on mitigation of deposition play an important role within research on diagnostic mirrors. The dedicated experiments with ITER-candidate mirror materials under ITER-relevant conditions delivered important information on the active deposition mitigation on divertor mirrors by elevated temperature and demonstrated a capability to prevent the carbon deposition and degradation of optical properties by heating the mirrors. Gas feeding in the vicinity of first mirrors is another attractive option to gain an active control over deposition. In experiments performed in TEXTOR, it was demonstrated that gas feed is capable to prevent the carbon deposition on the mirrors directly exposed in plasma. First experiments with a gas feed near the mirrors performed in DIII-D in 2008 showed significant mitigation of deposition. In the proposed experiment the optimized way of gas feed will be used: D2 will be fed directly onto the DiMES mirror holder to ease the evaluation of neutral pressure, and to provide advantages for possible modeling. It is anticipated that deposition mitigation may be achieved at a significantly lower flow rate than that used in 2008 experiment. | ||
| Resource Requirements: | 1/2 day experiment (~10 DiMES exposure shots). High-density moderate power LSN H-mode with detached divertor. LSN patch panel allowing placing OSP outboard of DiMES. Model discharge 132772. | ||
| Diagnostic Requirements: | Core and divertor Thomson, DiMES TV, tangential divertor TVs, MDS, filterscopes, floor Langmuir probes, CER, ITRV | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | D2 gas connected to the new DiMES capillary. | ||
| Title | 367: Comparison Mach probe and CER measurements of intrinsic rotation profile of main ions in helium plasmas | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | S. Mueller, J.S. DeGrassie, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure the edge main ion toroidal rotation profile in helium plasmas with the Mach probe with the CER system and then compare them. The goal is to verify that we see the same edge rotation structure on both diagnostics, including the localized peak seen previously (see background). The data will be used to compare with XGC0 calculations to see if we can achieve a physics understanding of this work. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize low power ECH H-mode plasmas like 140420-437 as the basic target. Measure edge main ion profile with Mach probe and CER at various times both before and after the L to H transition. Since the Mach probe plunges at one or two times and since the beam blip for CER seriously affects the rotation, multiple shot will be required to obtain a complete time history. Obtain complete edge profile data needed for the XGC0 modelling. | ||
| Background: | Experimental measurements of the edge main ion rotation profile in ECH H-modes (shot range 141444-141487) using the reciprocating Mach probe showed a localized peak in the deuteron rotation profile with the top of the peak on the separatrix. Data mining of helium H-mode plasmas from several years ago (shots 140420-437) demonstrated a similar structure in the main ion rotation. From the standpoint of angular momentum transport, this localized peak is quite surprising, since it is inconsistent with simple transport models. The structure may be a consequence of ion orbits crossing the separatrix; such effects have been predicted by the XGC0 code, although the spatial structure doesnot exactly match the experiment. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Mach probe. Standard profile diagnostics including CER. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 368: N=1 TBM Compensation in ITER Scenarios | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | E. Doyle, TBD... | ITPA Joint Experiment : | No |
| Description: | Demonstrate n=1 TBM error amelioration on at least one "ITER Demonstration" plasma at DIII-D. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce a chosen ITER Demonstration plasma. Optimize the TBM n=1 error compensation using experience gained from high-beta H-mode plasmas (proposals 129 and 192), including dynamic error field correction (DEFC). Compare effectiveness of n=1 compensation on this plasma vs its effectiveness on the high-flow, high-beta plasmas used in proposals 129 and 192. Monitor plasma rotation and magnetic responses closely. | ||
| Background: | During the 2009 TBM error field campaign, most experiments used co-injected DIII-D target plasmas that had ITER-similar shape (ISS), 3.3 < q95 < 3.6, and 2.1 < beta_N < 2.3. Over the past few years, various "ITER Demonstration" plasmas have been developed and characterized at DIII-D. It would be useful to ITER if the n=1 amelioration of TBM error effects was extended to one or more of these ITER Demonstration plasmas. | ||
| Resource Requirements: | TBM mock-up, 5 co-beams, cryo pumps, Dynamic error field correction | ||
| Diagnostic Requirements: | CER, MSE, Thomson scattering, MHD magnetics (plasma response) | ||
| Analysis Requirements: | EFIT, SURFMN, IPEC, MARS-F, NTV calculations | ||
| Other Requirements: | |||
| Title | 369: Introduction of pre-characterized dust and flakes in divertor and SOL | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Requested |
| Co-Author(s): | R. Smirnov, S. Krasheninnikov, J. Yu, N. Brooks, M. Fenstermacher, C. Lasnier, C. Wong, A. Litnovsky | ITPA Joint Experiment : | Yes |
| Description: | Experiments in support of the joint ITPA experiment DSOL-21 'Introduction of pre-characterized dust for dust transport studies in the divertor and SOL'. Dust of varying size and shape will be introduced in the lower divertor of DIII-D using DiMES and in the outboard SOL using reciprocating probe/MiMES. The objectives are: 1) characterization of core penetration efficiency and impact of dust of varying size and chemical composition on the core plasma performance in different conditions and geometries; 2) Characterization of dust mobilization from PFCs; 3) benchmarking of DustT modeling of dust transport and dynamics. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Dust will be introduced into the lower divertor using DiMES. Injection will be performed in LSN
configuration by sweeping strike points over DiMES holder loaded with dust. 3 separate experiments are proposed: 1) injection of large size graphite dust (50-100 micron diameter) from DiMES for tests of DustT modeling 2) tests of dust mobilization from ITER-relevant castellation gaps 3) mobilization and transport of tokamak-generated co-deposit flakes Dust and flakes will be mobilized by sweeping OSP over DiMES. Dust injections in the outboard SOL can be performed in piggyback during the above experiments. | ||
| Background: | Dust penetration of the core plasma in ITER can cause unacceptably high impurity concentration and degrade performance. Therefore, knowledge of the dust transport and dynamics is important. Studies on the contemporary machines are needed to benchmark modeling for extrapolations to ITER. Introduction of pre-characterized dust from a known location offers a way to benchmark modeling of dust dynamics and transport. Dust can be either actively injected or launched off a surface by plasma contact. Initial studies of the dust launch by plasma contact have been performed in DIII-D and TEXTOR. Contrary to expectations, core penetration efficiency of the dust in limiter configuration (TEXTOR) proved much lower than in the divertor configuration (DIII-D).
Joint ITPA experiment DSOL-21 was initiated in 2009 in order to improve understanding of dust mobilization and transport and to provide the basis for modeling. Participating machines include DIII-D, TEXTOR, MAST, NSTX, LHD, AUG, and T-10. | ||
| Resource Requirements: | Each experiment would require 2-3 shots including one setup shot and 1-2 exposure shots. LSN H-mode discharges with OSP sweeps over DiMES are required for the dust mobilization. May be performed in piggyback near the end of LSN discharges allowing OSP sweep to DiMES major radius. | ||
| Diagnostic Requirements: | Core and divertor Thomson, DiMES TV, tangential divertor TVs, MDS, filterscopes, fast UCSD camera, CER. Fast camera(s) coupled to DiMES TV and/or tangential TV views highly desirable. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 370: 2D Carbon Ion Parallel Flow Measurement Near the Divertor | ||
| Name: | Weber webert@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Divertor Detachment and Plasma Flows | Presentation time: | Not requested |
| Co-Author(s): | S. Allen, J. Howard | ITPA Joint Experiment : | No |
| Description: | Our goal is to measure carbon ion parallel flows near the lower divertor using a new technique in polarization interferometry. This diagnostic produces a series of time resolved, 2D carbon ion parallel flow measurements in the poloidal plane. Such detailed flow profiles will (a) be useful in understanding carbon transport from the limiter and (b) help validate UEDGE modeling of carbon emission and transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using two ports from the LLNL group with tangential views of the lower and upper divertor, we plan to measure carbon ion flow in this region by taking multiples images with a fast camera mounted behind a spatially multiplexed polarization spectrometer. We will look at a 465 nm emission line from CIII, and measure the Doppler shift to recover the flow velocity. Each image will then encode a 2D carbon ion parallel flow profile in the poloidal plane. We plan to vary the machine parameters to generate plasmas over a wide range of flows, and large CIII population. Possibilities include changing the toroidal magnetic field direction, and location of the divertor. We would like a sum total of one day to do these experiments. | ||
| Background: | Carbon transport from the limiter is critical in achieving a burning plasma. Unfortunately, this complex process is poorly understood due to a lack of adequate diagnostics, and modeling difficulties. This new technique offers a way to obtain a detailed 2D carbon ion flow profile in the poloidal plane. In the 2009-2010 campaign, we proved the utility of this flow diagnostic by successfully measuring a number of carbon flows near the lower divertor. One flow was shown to be in reasonable agreement with results from a UEDGE simulation. This preliminary exercise has sparked many new ideas for improvements and use of the diagnostic. Now we would like the opportunity to implement these ideas and further test the potential of this new diagnostic. | ||
| Resource Requirements: | 1 day | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 371: Dependence of upstream Te on target plate potential | ||
| Name: | Watkins watkins@fusion.gat.com | Affiliation: | Sandia National Lab |
| Research Area: | General PBI | Presentation time: | Not requested |
| Co-Author(s): | Stangeby | ITPA Joint Experiment : | No |
| Description: | In a carefully constructed magnetic field geometry, use the divertor Thomson to measure upstream Te as the target plate potential is changed at the Langmuir probe. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The divertor Thomson laser pulse would have to be phased with the Langmuir probe bias programming such that the upstream Te could be measured on the same field line with and without bias. To have a better chance of alignment, the current or toroidal field could be ramped slowly near the desired value with the strike point near the divertor Thomson radius of 1.49 m. | ||
| Background: | Target plate heat flux calculated from Langmuir probe particle flux and electron temperature agrees with the IR camera heat flux when using a theoretical value of 7 for the sheath power transmission factor in the far SOL but drops to ~1 near the strike point. One theory is that the electrons in the Langmuir probe flux tube are reflected back upstream by the negative bias resulting in a higher upstream temperature. | ||
| Resource Requirements: | very stable plasma shape | ||
| Diagnostic Requirements: | divertor Thomson
divertor Langmuir probes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 372: TBM Effects for All Coil Signs | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | D. Spong | ITPA Joint Experiment : | No |
| Description: | Apply magnetic perturbations to the plasma from the TBM mock-up having different geometries from a true TBM. There are 4 permutations of the current signs that can be used in principle in DIII-D experiment, but only the one sign combination that correctly represented magnetized steel in the local tokamak magnetic field was used previously in 2009. Each current sign permutation gives a different magnetic perturbation geometry and spatial Fourier harmonic spectrum. We expect a different set of consequences to the plasma for each geometry. We propose to apply one or more of the untried coil sign permutations and measure selected plasma responses. The results should provide for more complete validation tests of 3D physics and its modeling. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce a target plasma previously used for TBM experiments. Measure basic plasma responses to a selected mock-up coil current perturbation. | ||
| Background: | The TBM mockup has two magnet coil sets, one to mock up ferromagnetic toroidal magnetization and the other to mock up ferromagnetic poloidal magnetization. During the 2009 experiments the mock-up coils were operated only with their currents in the directions that corresponded to magnetized steel in the local tokamak magnetic field. There are 4 permutations of the current signs that can be used in principle in DIII-D experiments. The 3 untested permutations offer different ripple field spatial geometries and Fourier harmonics. Effects, from fast ion losses to locked modes to confinement changes, should be different for each of signs. Data from these 3 cases can provide new useful information for more complete validation tests of 3D physics and its modeling. NOTE: The mock-up was stress analyzed only for the current directions for ferromagnetic steel. Additional analysis will be required before the mock-up can be operated in the other direction permutations. | ||
| Resource Requirements: | TBM mock-up, 5 co-beams, cryo pumps, Dynamic error field correction | ||
| Diagnostic Requirements: | CER, MSE, Thomson scattering, MHD magnetics (plasma response), fast ion loss diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 373: 3D at qmin ~ 3 | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | D Spong, TBD... | ITPA Joint Experiment : | No |
| Description: | Run discharges in the qmin ~ 3 advanced tokamak regime. 3D equilibrium calculations generally show more significant 3D changes in flux surface shape for these regimes than for qmin ~ 1 cases. Also, ripple losses may be higher. Use results to test 3D equilibrium codes, braking calculations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment needs further definition.
Advice and help from people with qmin ~ 3 experience. Choice of non- axisymmetric test fields | ||
| Background: | 3D equilibrium calculations generally show more significant 3D changes in flux surface shape for plasma regimes that have qmin ~ 3 than for qmin ~ 1 cases. | ||
| Resource Requirements: | Target plasma with qmin ~ 3 | ||
| Diagnostic Requirements: | MHD magnetics (plasma response) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 374: xpt gas puff for ELM control | ||
| Name: | Watkins watkins@fusion.gat.com | Affiliation: | Sandia National Lab |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | By locally fueling the core plasma in a flux expanded region near the primary or secondary x-point, it may be possible to form a localized pressure gradient that will trigger rapid ELMs in the flux compressed region at the outer midplane. The key idea is that in the flux expanded region, the fueling is much more localized in flux space and would allow very narrow local pressure gradients at the outer midplane high enough to trigger small rapid ELMs and avoid the larger and more damaging type 1 ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | make a shape conforming to the upper baffle and puff small amounts of gas directly into the x-point. Reverse Bt may be necessary. | ||
| Background: | Rapid ELMs have been observed when placing the outer strike point on the upper dome which moved the neutral recycling source very near the xpt. The effect appeared to be an xpt fueling effect but no dedicated investigation has been performed. | ||
| Resource Requirements: | dedicated and calibrated gas controller for the dome xpt gas line | ||
| Diagnostic Requirements: | upper xpt camera | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 375: ELM heat flux profiles and energy deposition at the primary and secondary target plates | ||
| Name: | Watkins watkins@fusion.gat.com | Affiliation: | Sandia National Lab |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | lasnier, leonard, osborne, makowski, petrie | ITPA Joint Experiment : | No |
| Description: | Previous work in this area has shown general trends for elm heat flux and energy deposition profiles in the secondary divertor with magnetic balance and density. More detailed scans and more ELMs are needed to build up a more complete database. Improvements in target plate heat flux measurements and coverage is desired as well as more accurate core ELM energy loss measurements. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Alternately run primary and secondary strike points on shelf and measure dependence of ELM heat flux footprints on ELM size, shape, and q95. Measure primary and secondary heat flux simultaneously. | ||
| Background: | ITER is concerned about first wall ELM energy deposition including the secondary divertor. We performed some work on this topic last year but more data is needed to resolve some experimental issues and collect more ELMs in a wider range of conditions. We need better resolution of core ELM energy loss, simultaneous primary and secondary divertor heat flux, and outer and inner wall heat flux (if available). | ||
| Resource Requirements: | inside vessel diamagnetic loop desired for better core ELM energy loss | ||
| Diagnostic Requirements: | fast IRTV desired in both upper and lower divertors | ||
| Analysis Requirements: | fast EFITs | ||
| Other Requirements: | xpt probe | ||
| Title | 376: Advanced tungsten PMI coatings for erosion/re-deposition testing | ||
| Name: | Wong wongc@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | J.P. Allain, O. El-Atwani, S. Suslov, Purdue University, D. Rudakov, UCSD | ITPA Joint Experiment : | No |
| Description: | We are interested in namely studying erosion/re-deposition phenomena from various tungsten-based material coatings and systems that include: 1) core-shell nanocomposites of ultrafine grained W (using WC dispersoids) 2) multi-modal grained W, 3) nanocomposite Li-W and 4) control samples (e.g. commerical W) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Different W-based material coatings and systems to be exposed at the DIII-D lower divertor with the use of the DiMES sample exposure system. Samples would be exposed in piggyback mode and if possible with inclusion of OSP on DiMES. | ||
| Background: | This is the first exposure of innovative W-based coating and implanted material in a tokamak divertor. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Divertor Thompson Scattering (DTS) and othe nearby Langmuir probes are desired. One particular sample contains low-Z materials including small concentrations (< 20% lithium) and thus a VIS camera filtered for the 671-nm line is desired to examine erosion transport properties due to lithiumā??s ease of ionization. The other three samples will be tungsten and carbon based materials. | ||
| Analysis Requirements: | Analysis will be performed before and after the sample exposure. | ||
| Other Requirements: | |||
| Title | 377: Arc Experiments at DIMES | ||
| Name: | Wong wongc@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | V. Rohde, R. Neu, A. Kallenbach, M. Balden, IPP, D. Rudakov, UCSD | ITPA Joint Experiment : | Yes |
| Description: | The first wall material of ITER is under discussion. Based on the experience of ASDEX Upgrade (AUG) high-Z materials as tungsten are proposed. These material offer low erosion and low hydrogen retention. As chemical erosion is absent and physical sputtering is quite low, other erosion processes as arcing has to be taken into account again. These processes were discussed in the 80ties [1], and recently in AUG [2,3].
To extent the database on arc erosion tungsten probes should be exposed to DIII-D plasma discharges using the DIMES probe system. DIII-D offers a carbon environment and ITER relevant H-mode discharges. [1] McCracken G M 1980 J. Nucl. Mater. 93-94 3 [2] Rohde V. et al 2009, Physica Scripta, T138, 014024 [3] Rohde V et al, in press, J.Nucl.Mat, NUMA45187 |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | A full tungsten sample will be prepared at IPP and mounted at DIMES using a modified probe head. The size of the sample is 12*44 mm. Microscopic marker on the sample will allow determining the tungsten erosion. This can be modified to fix onto the DiMES sample module.
The probe should be exposed to 20 sec H-mode divertor plasma conditions. | ||
| Background: | Investigations at AUG show that the amount of material eroded by an arc depends strongly on the surface properties. Strong erosion is observed, if non conducting layers cover the surface. To investigate this phenomenon also in DIII-D the full tungsten sample will be partly coated with Si02. | ||
| Resource Requirements: | Sample to be prepared at IPP | ||
| Diagnostic Requirements: | TBD | ||
| Analysis Requirements: | The sample will be measured before and after exposition at IPP, using the Helios device and a confocal scanning microscope at IPP. | ||
| Other Requirements: | |||
| Title | 378: Systematically Sweep Rotation To Test, Reduce RMP Screening | ||
| Name: | Callen jdcallen@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | T. Evans, F. Waelbroeck, R. Groebner, R. Moyer, A. Cole, C. Hegna | ITPA Joint Experiment : | No |
| Description: | The major open and most striking issue concerning RMP effects on edge plasmas is the degree to which plasma toroidal flow causes the 3D RMP fields to be "screened" in the edge plasma. As discussed in the Background section below, theory predicts flow screening is strong and impedes the penetration of RMP fields into the edge of H-mode plasmas for large flows; but it should become negligible at a small toroidal plasma rotation value at the relevant rational surface. When flow screening effects are reduced, 3D fields are predicted to penetrate and resonant (locked) magnetic islands become more likely; with RMP fields multiple overlapping islands and hence local magnetic field stochasticity become possible. When magnetic stochasticity occurs, Rechester-Rosenbluth type plasma transport should occur.
This ROF proposal seeks to explore the degree to which flow screening of RMP fields occurs in ELM-suppressed H-mode plasmas by systematically sweeping the toroidal rotation from large co through to large counter plasma rotation. For large co or counter rotation the RMP field penetration and its effects are predicted to be small. At small rotation the screening effects should become negligible or significantly reduced. The physics objective would be to look for resonant locked magnetic islands and/or evidence of changes in stochasticity-induced plasma transport produced by near zero toroidal plasma flow, perhaps at concomitantly reduced I-coil currents. Elucidating these RMP flow screening effects is a critical physics element that needs to be resolved to understand how RMPs suppress ELMs. It could also provide critical information for the present consideration of including dynamic, internal coils in ITER for ELM suppression. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To experimentally test if flow screening of RMP-applied 3D fields is really important in ELM suppression, one needs to systematically sweep the plasma toroidal flow in a relevant radial region across near zero values. The dominant RMP-induced resonant and possibly stochastic behavior apparently occurs in DIII-D at the top of the pedestal (defined here as 0.83 < rho_N < 0.97, 0.85 < Psi_N < 0.98); this is thus the key region for examining flow screening effects. The proposed experiments would carry out a systematic scan (within one shot or in a sequence of shots?) of the toroidal plasma flow at say Psi_N ~ 0.92 from the co to counter I_p direction (via varying the mix of co and counter NBI-induced torques) in ELM-suppressed RMP H-mode plasmas. Theory predicts the least flow-induced screening and maximum propensity for resonant island formation or magnetic stochasticity occurs near zero toroidal rotation, perhaps at Omega_t ~ 5 to 15 krads/s. The physics objective would be to look for the possible formation of a resonant magnetic island (with m/n = 8/3 to 11/3) at the top of the pedestal and/or the degree to which the electron temperature and density gradients are reduced there [via Rechester-Rosenbluth (RR) induced stochastic transport] as a function of the RMP-inducing I-coil current and relative to a corresponding ELM case without the RMPs. | ||
| Background: | It has become increasingly apparent over the past few years that rotational screening plays a critical role in the penetration of RMP fields into the edge of H-mode plasmas and hence in suppression of ELMs in H-mode plasmas -- see for example Section 9 of Callen OV/4-3 talk at 2010 Daejeon IAEA FEC (available as UW-CPTC 10-8 via http://www.cptc.wisc.edu), Waelbroeck's 2008 IAEA FEC review talk [Nuclear Fusion 49, 104025 (2009)] and references cited in these theory-based review papers. At large toroidal plasma rotation the plasma is effectively a superconductor at the rational surface and hence causes an externally applied resonant perturbation to vanish there and at smaller radii. However, when the relevant plasma toroidal rotation at the rational surface is reduced or vanishes, resonant fields can penetrate into the plasma and allow magnetic reconnection into a resonant magnetic island topology. With many resonant magnetic perturbations (RMPs) multiple overlapping islands and hence local magnetic field stochasticity becomes possible if the resultant magnetic islands overlap (Chirikov criterion).
Recent numerical modeling of these important effects in H-mode pedestals using resistive singular layer physics were presented at the recent IAEA Daejeon meeting -- by Chu et al. (THS/P5-04) and Liu et al. (THS/P5-10). Two-fluid sheared slab theory predicts the relevant toroidal flow at which reconnection occurs is where the total (ExB plus diamagnetic) electron flow vanishes. In a near-axisymmetric tokamak the (~ neoclassical?) ion poloidal flow should be added. The precise toroidal rotation speed for maximum 3D field penetration is thus a bit uncertain. However, it is anticipated to be of order the electron diamagnetic flow velocity and hence of order 5 to 15 krads/s at the top of a DIII-D H-mode pedestal. Both the original paper on ELM suppression via RMPs in DIII-D plasmas [Evans et al, Nature Physics 2, 419 (2006)] and recent studies [Moyer at al., paper CO4 12 at the 2010 Chicago DPP-APS meeting] have shown that increasing I-coil currents induce ever smaller electron temperature (and to a lessor extent density) gradients at the top (Psi_N ~ 0.92) of the H-mode pedestals. Presumably this flattening of the T_e gradient is induced by the formation of resonant magnetic islands and/or magnetic-stochasticity-induced RR-type plasma transport there. Theoretically, resonant island widths and/or induced magnetic stochasticity should be enhanced at low plasma toroidal rotation, with the effects peaked around Omega_t ~ 5 to 15 krad/s. The objective of this proposed experiment is to vary the plasma toroidal rotation at Psi_N ~ 0.92 to explore the degree to which the T_e (and n_e) gradients in this region are influenced by toroidal rotation screening of the penetration of RMPs there. The theory outlined above implies that for lower toroidal plasma flows smaller I-coil currents may be able to suppress ELMs at the top of the pedestal. | ||
| Resource Requirements: | The main requirements are for a varying mix of co/ctr NBI torques (in a single shot or on successive shots?) applied to H-mode plasmas with q_95 adjusted for maximum RMP ELM-suppression effects induced at a variety of I-coil currents. At least a half day would be required to set up the desired plasmas and test out sweeping the NBI-induced torque. Then a day or more is needed to systematically explore the effects of changes at the top of the pedestal in response to varying the toroidal torque there, with different I-coil currents. Two full days of experiments would be optimal. | ||
| Diagnostic Requirements: | CER measurements of the carbon (and possibly deuterium?) ion toroidal flow is critical both for monitoring the induced toroidal flow and perhaps as a control parameter in a NBI-induced co/ctr torque sweep. In addition, careful Thomson measurements of the n_e and T_e profiles and small changes in their gradients at the top of the pedestal are needed. ECE measurements at the top of the pedestal could also be critical in determining if resonant islands form at the top of the pedestal and/or the degree to which T_e gradients become very small there. | ||
| Analysis Requirements: | MARS-F calculations of the penetration of RMP fields for a variety of relevant toroidal plasma rotation values at the top of the pedestal would be useful -- both before this proposed experiment is carried out and afterwards for analyzing the results. | ||
| Other Requirements: | Previous attempts at adding some counter NBI have encountered MHD modes (mainly core 2/1 NTMs) that led to mode locking and an L --> H transition -- see Evans et al., Nuclear Fusion 48, 024002 (2008). In this proposed experiment low rotation 2/1 NTMs leading to locked modes need to be avoided -- maybe they can be prevented by using the q_95 ~ 7.4 RMP resonant window, lower beta_N, high magnetic shear at the 2/1 surface or other tricks used in the rotation scans exploring the peak NTV at small, counter rotation (Cole et al., UW-CPTC 10-1)? | ||
| Title | 379: Systematically Sweep Rotation To Test, Reduce RMP Screening (dup #378) | ||
| Name: | Callen jdcallen@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Requested |
| Co-Author(s): | T. Evans, F. Waelbroeck, R. Groebner, R. Moyer, A. Cole, C. Hegna | ITPA Joint Experiment : | No |
| Description: | The major open and most striking issue concerning RMP effects on edge plasmas is the degree to which plasma toroidal flow causes the 3D RMP fields to be "screened" in the edge plasma. As discussed in the Background section below, theory predicts flow screening is strong and impedes the penetration of RMP fields into the edge of H-mode plasmas for large flows; but it should become negligible at a small toroidal plasma rotation value at the relevant rational surface. When flow screening effects are reduced, 3D fields are predicted to penetrate and resonant (locked) magnetic islands become more likely; with RMP fields multiple overlapping islands and hence local magnetic field stochasticity become possible. When magnetic stochasticity occurs, Rechester-Rosenbluth type plasma transport should occur.
This ROF proposal seeks to explore the degree to which flow screening of RMP fields occurs in ELM-suppressed H-mode plasmas by systematically sweeping the toroidal rotation from large co through to large counter plasma rotation. For large co or counter rotation the RMP field penetration and its effects are predicted to be small. At small rotation the screening effects should become negligible or significantly reduced. The physics objective would be to look for resonant locked magnetic islands and/or evidence of changes in stochasticity-induced plasma transport produced by near zero toroidal plasma flow, perhaps at concomitantly reduced I-coil currents. Elucidating these RMP flow screening effects is a critical physics element that needs to be resolved to understand how RMPs suppress ELMs. It could also provide critical information for the present consideration of including dynamic, internal coils in ITER for ELM suppression. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To experimentally test if flow screening of RMP-applied 3D fields is really important in ELM suppression, one needs to systematically sweep the plasma toroidal flow in a relevant radial region across near zero values. The dominant RMP-induced resonant and possibly stochastic behavior apparently occurs in DIII-D at the top of the pedestal (defined here as 0.83 < rho_N < 0.97, 0.85 < Psi_N < 0.98); this is thus the key region for examining flow screening effects. The proposed experiments would carry out a systematic scan (within one shot or in a sequence of shots?) of the toroidal plasma flow at say Psi_N ~ 0.92 from the co to counter I_p direction (via varying the mix of co and counter NBI-induced torques) in ELM-suppressed RMP H-mode plasmas. Theory predicts the least flow-induced screening and maximum propensity for resonant island formation or magnetic stochasticity occurs near zero toroidal rotation, perhaps at Omega_t ~ 5 to 15 krads/s. The physics objective would be to look for the possible formation of a resonant magnetic island (with m/n = 8/3 to 11/3) at the top of the pedestal and/or the degree to which the electron temperature and density gradients are reduced there [via Rechester-Rosenbluth (RR)induced stochastic transport] as a function of the RMP-inducing I-coil current and relative to a corresponding ELM case without the RMPs. | ||
| Background: | It has become increasingly apparent over the past few years that rotational screening plays a critical role in the penetration of RMP fields into the edge of H-mode plasmas and hence in suppression of ELMs in H-mode plasmas -- see for example Section 9 of Callen OV/4-3 talk at 2010 Daejeon IAEA FEC (available as UW-CPTC 10-8 via http://www.cptc.wisc.edu), Waelbroeck's 2008 IAEA FEC review talk [Nuclear Fusion 49, 104025 (2009)] and references cited in these theory-based review papers. At large toroidal plasma rotation the plasma is effectively a superconductor at the rational surface and hence causes an externally applied resonant perturbation to vanish there and at smaller radii. However, when the relevant plasma toroidal rotation at the rational surface is reduced or vanishes, resonant fields can penetrate into the plasma and allow magnetic reconnection into a resonant magnetic island topology. With many resonant magnetic perturbations (RMPs) multiple overlapping islands and hence local magnetic field stochasticity becomes possible if the resultant magnetic islands overlap (Chirikov criterion).
Recent numerical modeling of these important effects in H-mode pedestals using resistive singular layer physics were presented at the recent IAEA Daejeon meeting -- by Chu et al. (THS/P5-04) and Liu et al. (THS/P5-10). Two-fluid sheared slab theory predicts the relevant toroidal flow at which reconnection occurs is where the total (ExB plus diamagnetic) electron flow vanishes. In a near-axisymmetric tokamak the (~ neoclassical?) ion poloidal flow should be added. The precise toroidal rotation speed for maximum 3D field penetration is thus a bit uncertain. However, it is anticipated to be of order the electron diamagnetic flow velocity and hence of order 5 to 15 krads/s at the top of a DIII-D H-mode pedestal. Both the original paper on ELM suppression via RMPs in DIII-D plasmas [Evans et al, Nature Physics 2, 419 (2006)] and recent studies [Moyer at al., paper CO4 12 at the 2010 Chicago DPP-APS meeting] have shown that increasing I-coil currents induce ever smaller electron temperature (and to a lessor extent density) gradients at the top (Psi_N ~ 0.92) of the H-mode pedestals. Presumably this flattening of the T_e gradient is induced by the formation of resonant magnetic islands and/or magnetic-stochasticity-induced RR-type plasma transport there. Theoretically, resonant island widths and/or induced magnetic stochasticity should be enhanced at low plasma toroidal rotation, with the effects peaked around Omega_t ~ 5 to 15 krad/s. The objective of this proposed experiment is to vary the plasma toroidal rotation at Psi_N ~ 0.92 to explore the degree to which the T_e (and n_e) gradients in this region are influenced by toroidal rotation screening of the penetration of RMPs there. The theory outlined above implies that for lower toroidal plasma flows smaller I-coil currents may be able to suppress ELMs at the top of the pedestal. | ||
| Resource Requirements: | The main requirements are for a varying mix of co/ctr NBI torques (in a single shot or on successive shots?) applied to H-mode plasmas with q_95 adjusted for maximum RMP ELM-suppression effects induced at a variety of I-coil currents. At least a half day would be required to set up the desired plasmas and test out sweeping the NBI-induced torque. Then a day or more is needed to systematically explore the effects of changes at the top of the pedestal in response to varying the toroidal torque there, with different I-coil currents. Two full days of experiments would be optimal. | ||
| Diagnostic Requirements: | CER measurements of the carbon (and possibly deuterium?) ion toroidal flow is critical both for monitoring the induced toroidal flow and perhaps as a control parameter in a NBI-induced co/ctr torque sweep. In addition, careful Thomson measurements of the n_e and T_e profiles and small changes in their gradients at the top of the pedestal are needed. ECE measurements at the top of the pedestal could also be critical in determining if resonant islands form at the top of the pedestal and/or the degree to which T_e gradients become very small there. | ||
| Analysis Requirements: | MARS-F calculations of the penetration of RMP fields for a variety of relevant toroidal plasma rotation values at the top of the pedestal would be useful -- both before this proposed experiment is carried out and afterwards for analyzing the results. | ||
| Other Requirements: | Previous attempts at adding some counter NBI have encountered MHD modes (mainly core 2/1 NTMs) that led to mode locking and an H --> L transition -- see Evans et al., Nuclear Fusion 48, 024002 (2008). In this proposed experiment low rotation 2/1 NTMs leading to locked modes need to be avoided -- maybe they can be prevented by using the q_95 ~ 7.4 RMP resonant window, lower beta_N, high magnetic shear at the q = 2 surface or other tricks used in the rotation scans exploring the peak NTV at small, counter rotation (Cole et al, UUW-CPTC 10-1)? | ||
| Title | 380: Magnetic Barriers | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans | ITPA Joint Experiment : | No |
| Description: | Make two large islands (either NTMs, or, better, EF- or RMP-generated). Scan their widths or distance. Theory predicts that, for adequate widths and distance, a laminar surface called "magnetic barrier" will form in the middle of the stochastic region. Experimentally confirming its existence has fundamental and applicative implications, as it might concur to explain the q95 dependence of ELM control by RMPs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Even though n=3 barriers are more interesting, for their potential relevance to ELM control, initial experimental efforts will concentrate on n=1 barriers, as these are numerically predicted to be easier to excite (they require weaker RMPs) and to observe (they are "perfect" barrier, i.e. perfectly laminar, whereas the predicted n=3 barriers are semi-permeable chains of thin, elongated islands).
The main requirement is an intense n=1 error field. The starting point will be a discharge used for the assessment of the EF by the standard low density locked mode technique. This choice is motivated by the experience in coping with large deliberate EFs in such discharges, but changes will be required; for instance the density ramp-down will be truncated. The best EF correction will be deployed to start from an initially axisymmetric configuration. An n=1 field corresponding to 1.5kA of I-coil currents will be superimposed. Centrally deposited ECH will be modulated in time to generate heat pulses that will be measured by ECE and are expected to temporarily slow down in correspondence of the barrier. The discharge will be repeated for higher I-coil currents, or the I-coil currents will be slowly ramped during the shot. This is because more than one laminar region may be present in the stochastic region. The magnetic barrier is the last laminar surface to survive in a RMP ramp in which the surrounding field becomes more and more stochastic. Once the marginal I-coil current has been established, q95 will be ramped, to confirm that barriers depend on it. | ||
| Background: | The overlap of large magnetic islands generates chaotic fields. These generally enhance transport, deteriorate confinement, possibly cause or contribute to the "density pump-out". These effects are well-known from the RMP control of ELMs.
However, theory suggests that invariant manifolds can develop within chaotic fields, with the effect of reducing, rather than enhancing, transport. In particular, barriers should form at some special, "noble" irrational values of q in between two rational surfaces such as the 3/2 and 2/1 surfaces, or the 3/1 and 4/1. Barriers can also form at "mediant" irrational q's. Here it is proposed to experimentally test modelling by H. Ali and A. Punjabi [PPCF 2007] and by J. Kessler, F. Volpe et al. [MHD Control Workshop 2010]. If successful, it would represent the first realization and demonstration of a new kind of transport barriers. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 381: ECH effects on pedestal and ELMs | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Understand/characterize density pump-out, its balance with heating effects and their combined effect on the height of the pedestal shoulder and on the steepness of the pressure gradient. Mix these effects in a stabilizing/destabilizing way on marginally ballooning-unstable/stable plasmas, respectively. Repeat for type-I, type-II and type-III ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Deposit ECH on low field side (LFS). Scan rho~0.85-1.05 in small steps of Drho=0.02-0.05. Effect on ELM frequency and amplitude should be immediately visible. For its interpretation in terms of heating, pump-out ad ballooning stability, diagnose edge density, temperature and current. | ||
| Background: | A similar experiment was carried out with success at DIII-D by J.Lohr et al. in the late 80's and early 90's (Stambaugh et al., PPCF 1988; T.Luce et al., IAEA 1990). 1.2MW of ECH at 60GHz were launched from the high field side (HFS) and deposited on the LFS. Deposition inside/outside the separatrix was observed to halve/double the ELM period; no ECCD was attempted at that time. The idea would be to repeat the experiment with the new 110GHz system and with the improved edge diagnostics (improved TS, to measure the edge pressure, and MSE, to measure the edge current, among others) and codes (ELITE) that became available in the meantime. These offer the perspective of a deeper understanding of peeling-ballooning physics and of density pump-out.
From the point of view of MHD control, the minimum goal would be to reproduce and possibly improve past results, thanks to the increased ECRH power, the improved focusing and the fact that gyrotron beams would be launched from the LFS and absorbed on the LFS, thus they would broaden less. Also, ELMs develop predominantly on the LFS, so intuitively they should be attacked on the LFS. Additional reasons to prefer LFS deposition come from the scenario development: the scenario for LFS deposition was found to be more resilient to mode locking and shape control issues, in a more recent attempt at DIII-D, on April 4, 2008. Those ECH experiments at rho=0.82-0.94 saw increased ELM frequency but also increased energy-loss per ELM. The balance between pump-out and heating at various radii might reconcile these results with earlier Lohrā??s experiments, but need further experimental tests. There are 4 radii of interest for a scan of ECH deposition: the pedestal shoulder, the centre of pedestal region, the separatrix and the ā??originā?? of ELMs. Hence, time permitting, 12 radii should be considered for a comprehensive scan: the 4 mentioned before, 4 positions shifted upstream, and 4 downstream. TORAY calculations suggest that the deposition will be sufficiently localized for this fine scan to make sense. Fast diagnostics should monitor what happens to the ELM as it passes through the ECH deposition region. Finally, April 2008 experiments serendipitously found a QH-mode in full co-injection that needs better characterization and optimization and, at the same time, might represent an easier, more quiescent ELM-free target where to isolate and contrast heating and pump-out effects of ECH without the complication of ELMs and associated changes of plasma parameters. | ||
| Resource Requirements: | 6 gyrotrons | ||
| Diagnostic Requirements: | Edge TS, MSE | ||
| Analysis Requirements: | ELITE | ||
| Other Requirements: | |||
| Title | 382: ECCD and Ohkawa CD stabilization of marginally peeling-unstable ELMs | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Drive small edge currents both via the conventional Fisch-Boozer mechanism and the less exploited, for ECCD, Ohkawa effect, which is expected to dominate at rho>0.95. Demonstrate and characterize the effect of edge currents (both co- and ctr-) on the pedestal and on ELMs. Exploit small edge CD as a fine knob to investigate peeling stability and marginal stability. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The shot plan and diagnostic requirements are similar to ECH Proposal No.381, except that the launch will be oblique, for ECCD. Compare co/ctr, expected destabilizing/stabilizing, respectively.
This experiment will also be an excellent test-bed for a novel reflectometric diagnostic of the edge q-profile, proposed by F. Volpe, supported by T. Luce, and expected to reach microsecond and millimeter time and space resolutions, respectively. | ||
| Background: | Unlike ECH, the use of ECCD for controlling ELMs has never been systematically experimented, except for sporadic attempts at AUG and JT-60U. Reasons include unfavorable CD estimates. At DIII-D, for example, it is estimated that 2MW can drive 1-10A/cm2 at rho>0.9. Typical bootstrap (BS) current densities are much higher. However, plasmas were identified with edge BS current densities of 50A/cm2, in which the ECCD might act as a dwarf on giant's shoulders, the giant being the BS current and, we hope to demonstrate, ECH control. ECCD control could add fine adjustments at zero cost: once ECH is used, a toroidal tilt of launch would result in ECCD too, with no deterioration of ECH heating and, presumably, pump-out.
Note that although ECH effects on ne and Te at the edge are significant, their product -the effect on pressure- is modest. For this reason it will be worth assessing ECCD effects on ELMs: they might play a role comparable with or not-much-smaller than ECH. Besides control, ECCD might find application as a tool for fundamental studies of peeling stability and marginal stability. For instance, for strong plasma shaping the peeling-ballooning stability boundary should "bulge" and give rise to a bifurcation and a second instability region at low currents and high pressure-gradients which has not been proved experimentally yet, even because the two requirements are in contradiction: high pressure gradient is coupled with high (BS) currents. Ctr-ECCD, however, could be used to reduce the latter and provide 2D experimental flexibility. The experiment should begin with a marginally (un)stable plasma, that a small EC-driven current can (de)stabilize. For this reason, the experiment could be combined with experiments aimed at tracing the peeling-ballooning stability boundary. Finally, it is expected that significant currents can be driven at extreme radii, rho>0.95, by the Ohkawa CD mechanism, based on asymmetric de-trapping in the velocity space. Its experimental identification is made possible by the fact that Ohkawa CD's direction is opposite to conventional Fisch-Boozer ECCD. The (de) stabilization of ELMs would be a simple and spectacular evidence of this change of CD direction. | ||
| Resource Requirements: | 6 gyrotrons | ||
| Diagnostic Requirements: | MSE; new reflectometric diagnostic of edge q, if ready | ||
| Analysis Requirements: | ELITE | ||
| Other Requirements: | -- | ||
| Title | 383: Filling with ECCD an unstable ELM current-hole | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | J. Callen | ITPA Joint Experiment : | No |
| Description: | Use ECCD to modify a special feature of the edge current profile and make it less ELM-unstable. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Restore an ELMy shot with modest bootstrap (BS) current at the edge. Use MSE data and control-room analysis to radially localize the "hole" in the current profile between the bulk core current and the edge BS peak. Repeat with co-ECCD at that location, and scan around it. Repeat with ctr-ECCD for comparison. Look for changes in the inter-ELM period in D_alpha. | ||
| Background: | Li-beam measurements, diagnostic-constrained EFIT reconstructions and ELITE modelling of the current profile exhibit a "hole" between the core and the edge peak. A rapid transition to a smoother profile is observed at ELM crashes. Later, the previous profile recovers, slowly, and eventually crashes again. It is proposed here that co-ECCD at rho=0.8-0.9 might fill or reduce such a hole, and the current profile be more stable as a result. The increased stability should manifest itself as a delayed crash, or possibly complete ELM avoidance. | ||
| Resource Requirements: | 5-6 gyrotrons | ||
| Diagnostic Requirements: | MSE, Li-beam? | ||
| Analysis Requirements: | ELITE | ||
| Other Requirements: | -- | ||
| Title | 384: ELM-pacing by pre-programmed modulated ECH/ECCD | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Demonstrate that modulated ECH/ECCD can periodically destabilize the edge pressure and gradient, and so trigger ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Begin by setting a value of rho which is known or expected to maximize ECH/ECCD effect on ELMs. Modulate ECH/ECCD 10% slower/faster and a factor of 2 slower/faster than the natural ELM frequency. Compare perpendicular, co- and ctr- launch.
Initial tests can piggyback on RMP control of ELMs. ECH pulses of 200ms towards the end of the discharges are brief enough not to damage diagnostics and other equipment in case absorption is only partial and stray is high. At the same time, they are long enough to see effects on ELMs, if any, including slow changes following transport and relaxation phenomena. Dedicated run-time could be shared with continuous ECH/ECCD proposals no.381 and 382, by adding modulation at the end of the continuous pulses. The advantage of exclusively dedicated time, not shared with others, is that different ECH frequencies can be tested in the same discharge, at different times. Finally, if ELM-frequency control succeeds, the other important parameter to control is the energy loss per ELM. The knobs in this case are the radius of ECH deposition and the amount of driven current. This is why, as a final experiment we propose to fix the ECH frequency and scan, on a shot-to-shot basis, the radius of deposition and the toroidal direction of launch. | ||
| Background: | In 2004, AUG modulated ECH at 100Hz at the edge of an ELM-ing plasma. The ELM frequency, initially of 150Hz, changed accordingly. Also, as expected, the ELMs slightly intensified. The opposite change, i.e. making the ELMs smaller and more frequent, has not been demonstrated yet, and would be highly desirable for ITER. Moreover, in AUG the ECH effects were suspected to dominate over ECCD, but the two were never really disentangled. Although seminal, AUG results leave much room for improvement and for the first demonstration of ELM-pacing by modulated ECH/ECCD at HIGHER frequencies. | ||
| Resource Requirements: | 5-6 gyrotrons | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | ELITE | ||
| Other Requirements: | -- | ||
| Title | 385: ELM-suppression by ECH/ECCD modulated in the rotating ELM filament, in f/back with focused D_alpha | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Modulate ECH/ECCD in phase and in synch with the rotating ELM filament, in search for enhanced stabilization. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Consider a D_alpha or other diagnostic of ELMs. If necessary, change optics to narrow its view and resolve one ELM filament at the time. As the ELM filament rotates, it modulates this D_alpha signal. The latter can be used as a driver for ECH/ECCD modulation in synch and in phase with the ELM, similar to oblique ECE for NTMs. The same electronics which interfaced the oblique ECE to the gyrotrons can be adapted to this purpose. | ||
| Background: | ECCD modulated by Mirnov probes at AUG and by oblique ECE at DIII-D in phase and in synch with a rotating islands has been effective in stabilizing 3/2 NTMs.
On the other hand, continuous ECH has been shown to affect, and in some cases completely stabilize ELMs at DIII-D, AUG and JT-60. Fusing these results, the present proposal intends to investigate the possible benefits of modulating the ECH/ECCD in synch with the rotating ELM filament. The scope is to selectively pump-down the ELM filament, or drive a current in it, or heat the space in between two filaments. The idea is that, by doing so, one might apply a perturbation equal and opposite to the ELM, and directly suppress it, similar to the ECCD compensating for the missing bootstrap current in a neoclassical island. The cw ECH/ECCD approach, instead, aims at making the plasma less unstable. In other words, it moves j_par and/or grad P away from the peeling-ballooning stability boundary. It removes the unstable condition, it does not suppress the instability. The downside is a cost in plasma performances. Conversely, active control of the instability enables operation in a nominally unstable, possibly higher performance region. Modulated ECH is more likely to have an effect, but modulated co- and ctr-ECCD should be tried too, especially on considering that ELM filaments have been demonstrated to carry current. ECCD might enhance or reduce these currents, much like it compensates for the bootstrap current deficit in neoclassical islands. Finally, ECCD at extreme radii or even outside the separatrix, might affect, possibly cancel the SOL currents. As a bonus, the method also has a potential as an indirect, comparative diagnostic of SOL currents, provided ne and Te in the SOL are known and ECCD can be calculated. | ||
| Resource Requirements: | 5-6 gyrotrons | ||
| Diagnostic Requirements: | Change optics in front of a D_alpha filterscope to narrow its view and resolve single ELM filaments. Filaments are clearly visible in UCSD Phantom camera, which, however, buffers data and cannot transfer them in real-time. Acquire D_alpha in real-time and pass signals to oblique-ECE ļæ½??boxļæ½?ļæ½ in the annex, with modified voltage thresholds and pass-bands. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 386: Modulate I-coils to induce edge currents and affect/study ELMs | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use ac currents in the I-coils to induce currents in the plasma edge and so perturb the edge current above/below the peeling boundary. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Starting with a marginally peeling-stable plasma, modulate current in the I-coils on a time-scale comparable with the current-diffusion time. Several kA of current might be necessary. Compare sine- and triangular waves. | ||
| Background: | External coils were used in COMPASS-D to induce currents in the plasma edge. Indirectly, they also affected the edge pressure gradient. Although not measured directly (COMPASS-D was not equipped with MSE, Li-beam or edge Thomson scattering), these perturbations were strong enough, according to EFIT, to cause a modulation across the peeling limit, which, in fact resulted in recursive stabilization/destabilization of ELMs [S.J. Fielding et al., EPS 2001, P5.014, Sec.2].
The earlier COMPASS results and the newly proposed experiment might or might not be related to recent DIII-D results by W. Solomon and NSTX results by Canik [PRL 2010]. There are important analogies and differences to discuss. For example, NSTX results were ascribed to changes in the pressure profile (ballooning limit), rather than to electromagnetic effects (peeling limit). At DIII-D, ELMs are triggered when the applied field is the strongest, whereas, if inductive, the effect should be strongest when the derivative is largest, i.e. at zeros of sine-waves and extremes of triangular waves. | ||
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| Title | 387: Modulate I_p to modulate edge current above/below peeling limit | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Modulate the plasma current Ip to indirectly modulate the edge current above/below the peeling limit. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Pre-program oscillating plasma current. | ||
| Background: | Similar to proposal #386, except that the modulation in the edge current is not induced by the I-coils, but by the E-coil. This will cause a modulation of Ip and of the whole current profile. The shape of the current profile will also be modulated. The effect at the edge will be an oscillation of the local current density. Probably the edge pressure gradient will fluctuate too. | ||
| Resource Requirements: | |||
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| Title | 388: Use a large 4/3 mode (locked or rotating) to stochastize the edge | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Check whether ā??internalā?? n=3 perturbations exerted by a 4/3 NTM are as effective as external n=3 RMPs in suppressing ELMs, in certain ranges of q95. If it works, exciting a 4/3 NTM and tuning q95 might represent a new, relatively simple recipe for an ELM-free or small-ELMs scenario, with no need for external coils. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For the sake of comparison, start with a discharge normally used for RMP control of ELMs, including the q95 ramp. Turn off the external n=3 RMPs and try to excite a 4/3 NTM, e.g. by increasing beta. The simultaneous presence of 2/1 or 3/2 NTMs is permitted, but is preferable to suppress them by means of ECCD in order to make free energy available and ease the excitation of the 4/3.
In case of difficulties in adapting ELM-control shots to 4/3 NTMs, try the opposite approach: begin with a discharge which is known to develop 4/3 NTMs, ad a q95 ramp-down and gradually make the shot as similar to ELM-control discharges as possible, but without RMPs. If difficulties persist, follow ECH/ECCD strategy of proposals 139 and 140, ā??NTMs on demandā??. | ||
| Background: | n=3 externally generated RMPs are known to suppress ELMs in DIII-D. Our aim is to test whether ā??internalā?? n=3 perturbations, exerted for instance by a 4/3 NTM, can be equally effective in stabilizing ELMs. | ||
| Resource Requirements: | |||
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| Title | 389: Extend QH-mode operation to high performance AT plasmas | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to demonstrate sustained ELM-stable AT plasmas with very high beta and confinement at near-zero toroidal rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | - Use 10 kA I-coil operation to apply a large counter torque to the plasma (TNRMFā??Ī“B2).
- Double-null shape with applied ECH could provide target plasma with zero or counter-Ip intrinsic rotation. - Combine NRMF-assisted QH-mode recipe with BT-ramp AT scenario | ||
| Background: | -- | ||
| Resource Requirements: | Needs engineering review to allow I-coil operation at 10 kA, at reduced toroidal field. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 390: Operate at optimal q95 and simply turn n=3 RMPs on/off to pace ELMs | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Operate at optimal q95 and simply turn n=3 RMPs on/off to pace ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Recreate standard ELM control discharge. Find broadest q95 window and repeat shot with q95 fixed in the middle of that window. Simply modulate RMPs on/off. Repeat for faster and faster modulation, to cause shorter and shorter ELM-ing and ELM-free periods. If/when arrived to a single ELM or only a few ELMs per period, change duty cycle to change the inter-ELM period.
Apart from representing a potential new ELM-pacing technique, this experiment would improve our understanding of how rapidly the plasma edge gradients, currents and transport respond to the RMPs. Modulation will allow high-quality measurements by lock-in and coherent averaging analysis techniques. Finally, note modulating the edge gradients is interesting for transport studies in general, regardless of ELM control, in that it would test transport models, TGLF23 simulations and ETG and ITG theory. For this reason, in addition to the discharges described above, that would alternate ELM-ing and ELM-free periods, it is proposed to take some full-modulation shots at non-optimal q95 (always ELM-ing) and some discharges at optimal q95 in which the RMPs are not turned off completely, but only modulated in amplitude (ELM-free). Differentiating between two ELM-ing periods or two ELM-free periods will give the transport coefficients in the ELM-ing and the ELM-free plasma, respectively. | ||
| Background: | ELMs are not always undesirable: small, frequent ā??grassyā?? ELMs might help control the plasma density and Helium ashes without affecting too severely the tokamak walls. Learning how to ā??paceā?? ELMs would be sufficient in this respect, given the inverse proportionality observed at AUG and other machines between the ELM frequency and the energy that they carry.
Modulation of RMP control is not new, but the proposal here is to force it to the ā??naturalā?? inter-ELM period and below, for faster, smaller ELMs. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | BES for n~, CECE for T~ and fast magnetics for B~. They all showed correlation or anti-correlation with the RMPs. | ||
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| Title | 391: Modulated central ECH for heat transport studies during ELMs and ELM-control | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Modulate central ECH to generate heat waves. Use ECE to study their propagation across the plasma, in particular across the edge. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate the ECH at 8-80Hz in: 1) an ELM-suppressed discharge at optimal q95, 2) an ELMy one at non-optimal q95, 3) an ELMy discharge with the RMPs turned off, 4) a q95 ramp and, if time, 5) an ELM-free H-mode without RMPs. In the latter, use 5-6 gyrotrons to trigger the ELM-free H-mode and 1-2 to generate the heat waves. | ||
| Background: | The Pulse-height analysis of ā??heat wavesā?? generated by time-modulated ECH is a well-known transport analysis technique providing the electron heat diffusivity profile and other transport coefficients. It is proposed to use it to document the heat transport in the stochastic edge of DIII-D during in ELM-suppressed discharges s well as, for comparison, in ELM-ing ones. It will be interesting to diagnose how q95 affects the heat transport, and compare with particle transport (density pump-out). | ||
| Resource Requirements: | 5-6 gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Height Pulse Analysis (HPA) software [T. Luce, C. Petty]. | ||
| Other Requirements: | |||
| Title | 392: ELM-pacing by Periodic Plasma Compression | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Periodically compress the plasma and thus its pressure and current profiles, in order to repetitively exceed the peeling-ballooning limit of edge pressure-gradient or edge current-density and so destabilize an ELM. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Restore ELMy H-mode discharge in which the ELM frequency was 10-100Hz. Program PCS to oscillate between two plasma shapes at a slightly different frequency and try to ā??lockā?? the ELM frequency. In case of difficulties, reduce NBI or ECH power to reduce the edge pressure gradient and current, to approach the peeling-ballooning limit. Even better, try to operate slightly below it, under conditions of marginal stability, so that small changes destabilize an ELM. | ||
| Background: | DIII-D has one of the best plasma control systems in the wall. Oscillations of individual plasma shape parameters (e.g. triangularity, elongation, etc.) are performed routinely, to test new features of the PCS. This experiment would involve a similar oscillation, but of several parameters at the same time. Basically the plasma would oscillate between two shapes, one smaller, the other larger, resulting in modulation of the pressure and current profiles. Indirectly, these will modulate the edge pressure gradient and current density. If the peeling-ballooning stability boundary is crossed in the process, an ELM will be triggered. | ||
| Resource Requirements: | |||
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| Title | 393: Horizontal and/or vertical plasma wobbling | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use controlled vertical oscillations of the plasma to make it temporarily ELM-free, and fast oscillations for ELM-pacing. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Starting with a DND H-mode, pre-program in the PCS vertical oscillations by +/-1cm, initially with a period of 500ms. This should result in alternate ELM-ing and ELM-free periods. Then try faster oscillations. When the period becomes comparable with or faster than the ā??naturalā?? inter-ELM period, ELM pacing should be obtained. Repeat for horizontal wobbling. Finally, combine the two, in and out-of phase. | ||
| Background: | Controlled vertical oscillations of the plasma have been observed to trigger/pace ELMs in TCV, AUG and NSTX. This was attributed to currents induced in the wall. Horizontal jogs, never used before, should in principle have the same effect. Note that both jogs move the plasma column parallel to some surfaces, and perpendicular to others. Both effects induce currents. Comparing horizontal and vertical jogs will tell which is most effective.
As a by-product, combining horizontal and vertical oscillations out-of-phase by 90deg will poloidally rotate the plasma column. This will be equivalent to poloidally rotate the wall, which is known to stabilize kinks and increase the resilience to mode locking. | ||
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| Title | 394: Dependence of Stiffness on Rotation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | Scan the temperature gradient at fixed density and fixed pedestal height using a tilt scan of the 150 beam plus an ECH deposition scan to determine the stiffness of transport. Repeat this at various values of plasma toroidal rotation, keeping the pedestal height fixed. Compare results with GYRO, TGLF, and GLF23, and determine if the changes in transport are more related to changes in stiffness or the critical gradient.
If it is more desirable to study low-k turbulence, then this experiment should be done in modest beta_N ELMy H-mode plasmas. However, if there is greater interest to study high-k turbulence, then high beta_N hybrid discharges should be utilized. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish H-mode plasmas with density control using the 150 beamline and 6 gyrotrons. Inject additional co-beam sources consistent with the level on beta_N desired. (2) In steps, scan the deposition of the heating power by tilting the 150 beamline and ECH launcher angles. If the 150 beam can be only tilted overnight, then do an ECH deposition scan with one beam tilt angle in the afternoon, and then the following morning repeat the ECH deposition scan for the other beam tilt angle. (3) Try to maintain constant ExB shear, Ti/Te ratio and pedestal height. This may be obtained maturally, but if not use the counter beams to control the rotation, and the overall heating power can be adjusted to keep the pedestal height fixed. (4) Replace the co-beam sources, with the except ion of the 150 beamline, with counter-beam sources. Make the toroidal rotation as small as possible without locking, matching the density and pedestal height from the high toroidal rotation cases. Repeat steps (1)-(3) with this low rotation discharge. | ||
| Background: | Analysis of hybrid discharges in which the NBI direction was switched from co to balanced at fixed beta_N demonstrated an interesting behavior. While the (thermal) pedestal height remained fixed, the co-beam case had a higher core pressure gradient than the balanced-beam case. (The balanced-beam case had high fast ion pressure so that the total beta_N was the same as the co-beam case despite the lower thermal pressure.) While the lower core transport for the co-beam case is clearly related to higher ExB shear, this could either be due to lower stiffness (i.e., slope) or a higher critical gradient (i.e., offset). The proposed experiment will determine which effect is more important by tracing out the heat flux vs. gradient curve for both the low-ExB shear and high-ExB shear conditions. Note that it is important to keep the (thermal) pedestal height fixed in this experiment to eliminate the sensitivity to the boundary condition. | ||
| Resource Requirements: | NBI: 150 beamline and 210 beamline are critical; need two additional co sources.
EC: 6 gyrotrons required. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 395: Helium Transport Measurements | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | This proposal is a request that we measure the helium impurity transport for interesting conditions. This should be considered as part of the "documentation" process. All that is needed is a repeat shot with the CER system (the vertical views being the most critical) tuned to helium and a small (5%) helium gas puff added. The evolution of the helium ion density profile will allow us to determine the diffusion and convection of the helium particle transport. The data will be used for an ITPA TC joint experiment on impurity transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | When the "transport documentation" of a particular condition is desired, one shot should be devoted to measuring the helium particle transport by (1) tuning the CER systems to helium (the vertical views are the most critical since the tangential views suffer from the plume effect), and (2) adding a short helium puff equal to about 5% of the electron density. The CER beams should be modulated on/off 10 ms/10 ms, or possibly 7 ms/8 ms. | ||
| Background: | This type of helium particle transport measurement has been done many times in the 1990's and 2000's. | ||
| Resource Requirements: | CER beams. Helium gas puffing. | ||
| Diagnostic Requirements: | CER system tuned to helium. | ||
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| Title | 396: Investigate rotational screening (copy: 2) | ||
| Name: | Mordijck smordijck@wm.edu | Affiliation: | The college of William and Mary |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Using the off-axis Neutral beam and n=3 I-coil perturbation, investigate the influence of rotation screening in L and H-mode plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use the off axis neutral beam to drive edge rotation. For L-mode plasmas we can look at the measured plasma response, the density pump-out, rotation profile and the footprint structure. For H-mode plasmas we can investigate if this changes the amount of I-coil current required to suppress/mitigate ELMs. | ||
| Background: | Up till now to investigate rotation screening, the balance of co- and counter beam injection has been altered. This changes especially the core rotation, makes the discharges prone to error-field penetration ( with lots of early disruptions). With the n=3 perturbation we are mostly interested what happens at the plasma edge, which makes the off-axis neutral beam an interesting candidate to change edge rotation. | ||
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| Title | 397: Direct Measurement of Transport Stiffness in the Electron Channel | ||
| Name: | Gentle k.gentle@utexas.edu | Affiliation: | U of Texas, Austin |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | Max Austin | ITPA Joint Experiment : | No |
| Description: | Stiffness in the electron thermal transport channel can be measured directly as a function of radius using modulated ECH. Measured heat flux as a function of measured temperature gradient may be found over the modulation cycle. The availability of 4 MW of ECH will allow determination of the Q(dT/dr) function over a greater range of flux and gradient.
Although simulation codes cannot treated modulated power cases directly for comparison, the two end points (0 and 4 MW) should be sufficiently different that code runs for each would be usefully different. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The full available ECH power will be applied near the center and 100% modulated to obtain the maximum range of flux variation over the maximum range of radii of the target plasma. This will determine a Q(dT/dr) function, but some of the variation in Q could be caused by changes in other parameters, e.g. Te, Te/Ti, etc. As a partial check on this, the experiment will be repeated using 1 MW of modulated ECH and 0, 1, 2, and 3 MW of CW ECH.
The target discharges for these experiments include L-mode (for comparison and extension of previous results) and particular discharges of special interest to the fusion program to which this approach could be effectively applied: the QH mode and L-modes with low and negative central shear. Other targets of interest could be considered, but the analysis technique cannot be applied to discharges with strong ELMs or other intrinsic temporal variations on the time scale of the modulation. | ||
| Background: | Transport stiffness is best defined by the shape of the curve of flux as a function of gradient. A linear proportionality is ordinary diffusion (Fick's Law). A steeper slope with an (apparent) offset in intercept gradient characterizes stiff transport and possibly a critical gradient.
Flux as a function of gradient has been measured in DIII-D for the electron channel (for a limited range of variation) using sawteeth and modulated ECH for low-power L modes. Transport was found to be quite stiff in the edge (rho>0.7) and tends to become diffusive in the core. | ||
| Resource Requirements: | 3-4 MW ECH
QH mode for measurements on that target | ||
| Diagnostic Requirements: | ECE Te(r,t) | ||
| Analysis Requirements: | ONETWO for equilibrium analysis and existing specific codes of analysis of modulated data. | ||
| Other Requirements: | |||
| Title | 398: Interaction of off-axis fast-ion population with sawteeth | ||
| Name: | Muscatello muscatello@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | Heidbrink | ITPA Joint Experiment : | No |
| Description: | An understanding of energetic ion interaction with sawteeth is an important issue for tokamaks operating with q less than or about unity. Sawteeth are well-known to cause deleterious effects on fast-ion populations peaked on axis. However, it is unclear how sawteeth affect an enhanced density of fast-ions near and outside the q=1 surface (most readily produced by off-axis neutral beam injection). On the other hand, an off-axis fast-ion population outside the q=1 surface has been observed to destabilize the mode, decreasing the sawtooth period (Chapman et al, Phys. Plasmas 16, 072506 (2009)). The goal of the proposed experiment is to utilize the new off-axis neutral beam in conjunction with the confined fast-ion diagnostics to test the interaction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce ITER-relevant shape but keep density low and plasma in L-mode to optimize FIDA. A NBI energy scan should be attempted for two types of shots: 1.) the control: only on-axis beams 2.) replacing one of the on-axis co-injected beams with the off-axis beam. In addition, if hardware permits, an angle scan of the neutral beam (at least the endpoints) would provide information on the dependence of the separation between the mode location and peak of the off-axis fast-ion population. | ||
| Background: | -- | ||
| Resource Requirements: | Operation of 330L, 210R, 30L, 150R and either 330R or 30R is required. All beam lines would be ideal. | ||
| Diagnostic Requirements: | FIDA, MSE, neutrons, ECE | ||
| Analysis Requirements: | EFIT, TRANSP, FIDAsim | ||
| Other Requirements: | -- | ||
| Title | 399: AC-compensation for high-Ī² plasmas | ||
| Name: | Piron lidia.piron@igi.cnr.it | Affiliation: | Consorzio RFX |
| Research Area: | General SSI | Presentation time: | Not requested |
| Co-Author(s): | L. Piron, G. Marchiori, L. Marrelli, P. Martin, P. Piovesan, A. Soppelsa for RFX-mod team and J. Hanson, Y. In, M. Okabayashi, E.J. Strait, H. Reimerdes for DIII-D team | ITPA Joint Experiment : | No |
| Description: | Interesting results have been obtained testing in real-time the AC decoupler in Ohmic plasma experiments, in terms of reduction of the n=1 amplitude. Unfortunately no current driven RWM was destabilized in these experiments, and the above evidence is due to AC effects introduced only by the field shaping coils. The importance of AC effects due to feedback coils has not been investigated yet, for this reason we propose to test AC compensation in Ohmic plasmas, in presence of unstable RWM.
We propose to test the algorithm also in high-Ī² plasmas. The possible relevance of AC compensation for magnetic feedback in high-Ī² scenarios has been evaluated analyzing past experiments with DC compensation, in which the high-Ī² phase had been terminated by a RWM. The experiment discussed at the APS conference (P9.00064) and MHD workshop considers a high-Ī² plasma, in which a fishbone drives a RWM. In particular, the I-coils react and reduce significantly the mode amplitude. Nonetheless a residual oscillation remains, growing on a time scale much longer than that typical of an unstable RWM. This causes the I-coil currents to increase in time up to the point when they saturate and the Ī² collapse occurs. The n=1 residual oscillation can be interpreted as an error field that the feedback is producing. Based on this analysis, it may be argued that an improved compensation of the sensor signals, which accounts also for the AC response of the wall, may be useful in conditions where the plasma is not resilient to error fields such as at high-Ī² or low-rotation. We propose another method to study the feasibility of the AC decoupler. This consists in inducing in high-Ī² plasmas an n=1 perturbation produced by the C-coil rotating at frequencies relevant for AC compensation and try to compensate it with the I-coil. These experiments can be done in piggyback. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | - AC compensation in Ohmic plasmas in presence of an unstable RWM.
- Fast feedback and DEFC with AC compensation in high-Ī² plasmas, especially when RWMs driven by fast MHD events (fishbones and ELMs) are expected. - Piggyback tests in high-Ī² experiments e.g. cancel with the I-coils an n=1 perturbations produced with the C-coils rotating at frequencies relevant for AC compensation. | ||
| Background: | Time-varying magnetic fields induce eddy currents in the wall, whose pattern is modified by 3D structures. The effects depend on frequency and if not taken into account, they could produce magnetic field errors that can be amplified, especially in high-Ī² plasmas. This aspect has been investigated in the framework of a collaboration between DIII-D and RFX-mod team.
Offline analyses suggested that the frequency response of DIII-D wall could be relevant for RWM control in low and high-Ī² plasmas. A scheme for the compensation of magnetic sensors from spurious n = 1 fields due to the coupling with the feedback and axisymmetric coils has been implemented in real time and tested in Ohmic plasmas. The main finding of these experiments is that taking into account the AC effects allows to improve the DEFC part since no RWM activity appeared in the performed tests. | ||
| Resource Requirements: | |||
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| Title | 400: Correlation length of electron temperature turbulence and comparison with gyrokinetic simulat | ||
| Name: | Wang wangg@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Rhodes, Peebles | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to investigate radial correlation length of electron temperature turbulence and compare with gyrokinetic simulations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Generate 2 steady-state plasma phases with ECH and ECH+NBI L-mode in one target shot. At 3 mid-radius locations take CECE measurements from shot to shot: fix one channel location and program to scan the location of another channel around it to collect correlation data sets. To ensure signal-to-noise ratio in correlation analysis, slow enough scan should be applied. | ||
| Background: | The turbulence correlation length provides a robust alternative to fluctuation levels to compare with turbulence modeling predictions. Due to the much larger thermal noise superimposed in the electron temperature fluctuations, it has been challenging to obtain a radial correlation length of electron temperature turbulence using the CECE measurement. This experiment will allow for a detailed investigation. | ||
| Resource Requirements: | most gyrotrons and NBI sources | ||
| Diagnostic Requirements: | CECE, DBS, BES, PCI, profile reflectometer and routine profile and other diagnostics | ||
| Analysis Requirements: | ONETWO, TGLF, GYRO | ||
| Other Requirements: | -- | ||
| Title | 401: Density turbulence correlation length and n-T cross phase in ITG and/or TEM dominant L-mode plasmas | ||
| Name: | Wang wangg@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Peebles, Rhodes, DeBoo, Staebler | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to study density turbulence correlation length and the phase angle between density and electron temperature turbulence in ITG and TEM dominant L-mode plasmas respectively, and compare with TGLF/GYRO predictions | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize the ECH modulation method (DeBoo, PoP 2010) to modulate local grad-Te to generate two plasma phases with dominant ITG and TEM turbulence respectively. Measure Density turbulence correlation length with 5-channel Doppler backscattering system, and the phase angle between density and electron temperature turbulence by the coupled CECE-reflectometer systems. | ||
| Background: | It has been successful generating ITG or TEM dominant plasmas with ECH alternately applied at two spatial locations in an L-mide limiter discharge at low collisionality (DeBoo, PoP 2010). Density correlation length has been shown to scale as (5-8)*rho_s for ITG turbulence (Rhodes, PoP 2001). The phase angle between density and electron temperature turbulence has been shown to increase with enhanced TEM turbulence, consistent with TGLF predictions (Wang, APS 2010). | ||
| Resource Requirements: | all gyrotrons, NB for CER and BES diagnostics | ||
| Diagnostic Requirements: | DBS5, CECE, DBS8, BES and other profile and turbulence diagnostics | ||
| Analysis Requirements: | ONETWO, TGLF, GYRO, etc | ||
| Other Requirements: | |||
| Title | 402: Improved noninductive plasma operation at high betap | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Develop high betap, stationary, fully noninductive discharges that are more relevant to fusion reactor operation. Previous DIII-D experiments have operated under fully noninductive conditions (no transformer), at high beta (betaN ā?? betap ā?? 3.2), and at high f_bs (ā?„ 85%) for well over a current relaxation time. However, these plasmas were run at q_95 ~9-10 making them uninteresting for fusion applications. The objective of the proposed experiment is to extend the operating range of these discharges to lower, more relevant values of q_95. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Reduce q_95 by operation at lower Bt. Prior work was done with Bt ā?„ 2 T. As has been seen in other work on noninductive plasmas, operation at lower field does not carry a severe penalty. Current drive with ECCD is useful down to ~1.6 T (2nd harmonic) or 1.1 T (3rd harmonic).
b. Reduce MHD and improve confinement. In the prior experiments the plasmas evolves from state with broadband MHD (possibly some variety of AE modes) toward conditions of low MHD activity and improved confinement. Variation of the initial conditions, including broadening of the initial current profile using the off-axis beams can yield higher noninductive current drive efficiency. c. Broaden the noninductive current profile. As noted, previous discharges showed improved performance as the current profile evolved to lower li. Using the off-axis beams to reach this state early in the discharge, and to maintain a more stable current profile will further improve performance. d. Improve discharge control. Prior experiments without inductive regulation of Ip showed that equilibrium changes, particularly the collapse of ITBs, have a significant effect on Ip and on confinement. Development of control methods using EC and NB feedback will improve the performance of these self-organized plasmas. | ||
| Background: | |||
| Resource Requirements: | |||
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| Other Requirements: | |||
| Title | 403: Collisionality scaling of confinement and transport in Advanced Inductive plasmas | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, T. Luce, C. Challis (JET), E. Joffrin (JET) | ITPA Joint Experiment : | Yes |
| Description: | Determine the scaling of confinement and transport with collisionality (nu*) in AI plasmas. [ITPA ā?? IOS4.3] | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The ITPA database of AI parameters indicates that there is a strong dependence of the H98y2 confinement multiplier on nu*. However, the database includes discharges with a wide range of parameters. A nu* scan holding other dimensionless parameters fixed will clarify this dependence. This study is important for extrapolation of AI scenario performance to ITER and other tokamaks, as the collisionality is reduced by a large factor between present tokamaks and ITER.
As this is proposed as a joint experiment, a portion of this experiment should be done under conditions (particularly shape) accessible in JET and AUG. This will also allow combination of the nu* scaling result with prior gyroradius scaling results. | ||
| Background: | |||
| Resource Requirements: | |||
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| Title | 404: Fusion burn control simulation | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | General SSI | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Study the evolution and stationary state of a plasma with power input dependent on plasma parameters (e.g., beta^2 to simulate an alpha particle heat source). Develop methods to control the operating point. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Regulate a portion of the power input to the DIII-D plasma to be proportional to the equivalent fusion power, P_alpha ~ n^2*f(T_i). The remainder of the power is used in part for steady auxiliary heating and in part for feedback control of beta and of the operating point (i.e., maintain constant P_alpha) in the presence of perturbations such as ELMs, sawteeth, MHD instabilities, etc.
b. Also examine burn control using other possible actuators: fueling and pumping to modify the fuel ion mix, impurity radiation, and average density. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | also submitted to General PCO | ||
| Title | 405: Study of H-L back transition | ||
| Name: | Wang wangg@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | Burrell, Gohil, Thomas | ITPA Joint Experiment : | No |
| Description: | The purpose is to characterize plasma response to the H-L back transition to understand the power hysteresis in the H-mode transition power, which could help understand the physics of the L-H transition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | During the H-mode phase ramp down power (but keep torque unchanged) to obtain the H-L back transition and characterize edge plasma response. Repeat this process at different H-mode density levels and USN/LSN configurations. Document edge plasma profile and turbulence response. | ||
| Background: | The hysteresis effect in H-mode power has been observed in DIII-D (Thomas, PPCF 1998). A characterization of the plasma response to the H-L back transition and understanding the power hysteresis could help understand the physics of the L-H transition. | ||
| Resource Requirements: | all NB sources | ||
| Diagnostic Requirements: | all profile and turbulence diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 406: Fusion burn control simulation [DUP 404] | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | General Plasma Control and Operation | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Study the evolution and stationary state of a plasma with power input dependent on plasma parameters (e.g., beta^2 to simulate an alpha particle heat source). Develop methods to control the operating point. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Regulate a portion of the power input to the DIII-D plasma to be proportional to the equivalent fusion power, P_alpha ~ n^2*f(T_i). The remainder of the power is used in part for steady auxiliary heating and in part for feedback control of beta and of the operating point (i.e., maintain constant P_alpha) in the presence of perturbations such as ELMs, sawteeth, MHD instabilities, etc.
b. Also examine burn control using other possible actuators: fueling and pumping to modify the fuel ion mix, impurity radiation, and average density. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | also submitted to General SSI | ||
| Title | 407: Two possible ELM pacing methods: modulated loop voltage and modulated ECH | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Study the effects of modulated loop voltage and of ECH at the plasma edge on ELM frequency and amplitude. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. ELMs are a nonlinear evolution of the peeling-ballooning mode at the H-mode pedestal. Local modulation of the pedestal pressure and pressure gradient with modulated ECH should change the stability conditions and thus entrain the ELMs, at least over some range in applied EC power, location, and modulation rate.
Apply maximum EC heating power to the pedestal. Scan the modulation frequency and duty cycle. Also scan the deposition location across the pedestal to the separatrix. Look for synchronous modification of the ELM frequency and/or amplitude. b. The steady applied toroidal electric field (via the E-coil) causes an inward radial drift (pinch) of the plasma. Ordinarily this is balanced by outward transport and quasi-stationary profiles are observed. Controlled modulation of the pinch by modulating the loop voltage should have observable effects on the pedestal profiles and may provide an additional actuator for modification and control of ELMs. Varying the loop voltage also modulates the current density and current density gradient at the edge. The effect of voltage on ELMs is supported by the observation that the ELM frequency (and the H-L transition power) varies as the current is ramped up or down (via changes in loop voltage). | ||
| Background: | |||
| Resource Requirements: | |||
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| Other Requirements: | |||
| Title | 408: q95 scan of steady-state ITER demonstration discharges | ||
| Name: | Park parkjm@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Not requested |
| Co-Author(s): | M. Murakami, E. Doyle, J. Ferron, et al. | ITPA Joint Experiment : | No |
| Description: | Document noninductive fraction, fusion performance, and edge pedestal as a function of q95. Demonstrate Q=5 and fully-noninductive condition at lower q95 and higher qmin using off-axis beam. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dedicated q95 scan in ITER shape at fixed betaN=3.2 (or 2.8) with better match to ITER shape and RWM control. Find optimum q95 for ITER steady-state (SS) scenario development and modeling validation for ITER projection. Document differences between ITER SS demonstration discharges and "DIII-D standard" SS scenario in double null (DN) shape. Develop higher qmin scenario with a larger radius of qmin, specifically at lower q95 (determined from the q95 scan) using off-axis beam towards simultaneously achieving the Q=5 and non-inductive goals. Focus on improving plasma confinement at ITER target of betaN~3.2 rather than trying to increase betaN. | ||
| Background: | In 2008, fully noninductive condition was demonstrated at higher q95 but with a relatively low fusion performance (G~0.15, ITER target 0.3) [E. Doyle, NF 2010]. The discharges were not stationary and revealed significant differences from steady-state discharges in DN shape (confinement, edge pedestal, stability, fast ion confinement, ...). Experiment and modeling show a strong dependency of confinement, stability, pedestal, and noninductive fraction (fNI) on q95 [Park, IAEA2010]. Theory-based projection of such discharges to ITER shows a tradeoff between fusion performance and fNI with variation in q95, as observed in DIII-D [Murakami, IAEA2010], indicating that optimization of q95 is critical to simultaneously achieve the fNI=1 and Q=5 goals. The dedicated q95 scan will provide crucial information on modeling validation for ITER SS scenario development. TGLF simulation suggests that a larger radius for the minimum of q helps to increase both the fusion performance and fNI by maximally utilizing the benefits of low magnetic shear. New off-axis beam will allow to develop higher qmin scenario with a larger radius of qmin to simultaneously achieve the Q=5 and non-inductive goals. | ||
| Resource Requirements: | All neutral beam sources except 210 LT with 150 beams at maximum tilting angle. All available gyrotrons | ||
| Diagnostic Requirements: | All sets of DIII-D diagnostics system | ||
| Analysis Requirements: | ONETWO, TRANSP, NVLOOP, FIDASIM analysis | ||
| Other Requirements: | |||
| Title | 409: ICRF beat-wave control of low-frequency modes and turbulence in DIII-D | ||
| Name: | Carter tcarter@physics.ucla.edu | Affiliation: | UC, Los Angeles |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | W.A. Peebles, L. Schmitz, T. Rhodes, V. Chan, R. Pinsker | ITPA Joint Experiment : | No |
| Description: | This proposal seeks run time to study the nonlinear response to the
beating between two ICRF waves with closely-spaced ($Delta f lesssim 100$kHz) frequencies, and the influence of this response on low-frequency modes. ICRF beat waves have been successfully driven and used to excite Alfven Eigenmodes in JET [A. Fasoli, et al., Nucl. Fusion 36, 258 (1996)]. We propose to establish the capability to drive ICRF beat waves in DIII-D. We additionally propose a novel application for ICRF beat waves: for interacting with and controlling gradient-driven instabilities such as the ion temperature gradient mode. Based on results from experiments in basic plasma physics facilities and theoretical predictions, it might be possible to modify and perhaps control these instabilities and the associated transport using this technique. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A single ICRF antenna would be used to simultaneously broadcast at two
closely spaced frequencies (equivalent to feeding it with a modulated RF signal). The antenna located at 285-300 degrees would be used, and two frequencies, e.g. 60 and 60.1MHz, would be feed into the amplifier chain driving the antenna. An important part of the proposed experiment would be establishing that a nonlinear response of non-neglible amplitude can be driven in DIII-D using two ICRF waves with closely-spaced frequencies. The frequency of separation would be scanned, looking for response in turbulence diagnostics (BES, DBS, Reflectometry, FIR scattering, etc) at the beat frequency. The size of the modulation in the ICRF could be varied as well as its frequency. At the same time, these diagnostics would be monitoring any changes in the background turbulence. It is likely that any nonlinear response to the modulation/beating would be spatially localized, and care would be taken to develop an experimental plan to allow variation in the localization of the effect. For example, the response might be enhanced near locations where the ICRF is being absorbed. On the other hand, strong absorption could generate fast ions that could complicate the experiment, and weak damping would enhance the effective cavity Q for the fast waves. In planning the experiment, having the ability to vary the degree and location of absorption is therefore important. In addition to looking at the effect on drift waves by this process, one could imagine interacting with other low-frequency modes, including Alfven eigenmodes (as was demonstrated on JET) and perhaps tearing modes. Resonant drive of these modes, or nonlinear modification of stability could be possible. | ||
| Background: | This proposal is motivated by recent experiments in LAPD. It was
found that co-propagating kinetic shear Alfv'{e}n waves can drive a surprisingly large nonlinear response at their beat frequency; this response in turn scatters the pump waves to generate an extended series of sidebands [T.A. Carter, et al., PRL 100, 155001 (2006)]. Recently we have explored interaction of this beat response with drift waves. Field-aligned density depletions are created in LAPD by selectively blocking plasma production using an obstruction; unstable drift waves grow around these depletions. An experiment was conducted in which two co-propagating KAWs are broadcast along this striation with varying frequency separation. The surprising observation in this experiment is that the drift instability is suppressed when the beat-driven mode is nearby (above or below) the natural instability frequency: all that remains is the beat response, generally at much lower amplitude. [D.W. Auerbach, et al., PRL 105, 135005 (2010)]. Initial experiments with broadband turbulent spectra (which result from using larger obstacles) show a dramatic change in the spatial structure of the turbulence and a reduction in the broadband turbulent amplitude, replaced by the coherent beat-driven response. Similar effects have been observed by using an external array of electrodes to directly excite drift-waves in a laboratory plasma [Klinger, et al., PRL 86, 5711 (2001)]. The effects of single frequency ICRF on ITG modes has been explored theoretically [Chiu, Chan, Bhadra, PPCF 31, 1095 (1989)]. In this work it was found that modifications to stability do occur, but that they are generally small except through a sideband interaction: an RF sideband develops through beating between the ITG mode and the ICRF pump; this sideband can then interact with the pump to generate a response at the ITG frequency. We are proposing to stimulate this potentially strong sideband interaction by simultaneously launching two closely spaced ICRF waves. | ||
| Resource Requirements: | ICRF | ||
| Diagnostic Requirements: | Turbulence diagnostics (Reflectometry, Scattering, DBS, BES, etc) | ||
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| Title | 410: Advanced Inductive plasmas at low and counter rotation | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, T. Luce | ITPA Joint Experiment : | No |
| Description: | Develop AI operating scenarios that operate with low torque and at low rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Startup at low rotation. Use RF only (or with balanced beams if necessary) to establish AI conditions. Thus far all DIII-D work on AI plasmas has used co-current NBI to initiate and develop the discharge. Previous work has indicated that approaching zero torque after establishing the discharge is limited by error field penetration and locking. Possibly this can be avoided with ab initio operation at low torque.
b. Approach AI operation starting with counter-current torque. The behavior of AI plasmas with counter rotation has not been explored. As indicated in standard H-mode studies, plasma behavior is not symmetric with respect to direction of rotation. Also, initiation with counter-current torque may lead to better access to low rotation conditions. c. Use ECCD for 2/1 mode suppression. The establishment of high performance AI conditions has been shown to be sensitive to development of 2/1 NTMs. In addition to allowing operation of low torque AI plasmas, ECCD following the q=2 surface can be used for 2/1 mode suppression. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 411: Access to and improvement of the robustness of Advanced Inductive scenario operation | ||
| Name: | Politzer pete.politzer@gmail.com | Affiliation: | Retired |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, T. Luce, E. Joffrin (JET), C. Challis (JET) | ITPA Joint Experiment : | Yes |
| Description: | Understand the several approaches that have been used around the world to reach AI conditions and develop a robust access scenario. DIII-D has used early heating and early high beta operation to establish AI conditions, generally with a 3/2 NTM present. AI discharges in AUG generally have fishbone instabilities instead of the NTM. JET uses an Ip overshoot to establish a transient AI condition without MHD. In all cases 2/1 modes can interfere with establishment of the scenario. [ITPA ā?? IOS 4.1] | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Develop and test startup trajectories that avoid the 2/1 NTM. Use the steerable ECCD to help suppress this mode.
b. Under similar conditions, compare the behavior of AI discharges formed with varying collisionality and timing of high power heating (NTM and fishbones) and with/without current overshoot. c. Use the off-axis beams to broaden the current profile. It is thought that a broad current profile is an important feature of AI operation. All access techniques lead to broader profiles. The off-axis beams can be used to quantify this effect and possibly to establish stationary AI performance without MHD. d. Determine whether the NTM mode number is influenced by plasma shape. In double null, high triangularity shapes in DIII-D the 3/2 mode is dominant. However, in single null, ITER-like shapes the 4/3 mode (which has a weaker effect on confinement) is often seen. | ||
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| Title | 412: Understand tearing limit in low torque high beta scenarios like advanced inductive/hybrid (Dup 240 | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | TER like baselines at low torque have been found to be highly susceptible to error fields, and even with optimal error correction, they encounter tearing modes at low betan, ~2.2. This raises questions for higher beta scenarios like advanced inductive or hybrid, not least because plasma response to error fields is well established to rise with betan - leading to increased braking and more likely triggering of tearing modes. Also the higher beta will increase bootstrap currents which potentially makes it easier for small islands to bifurcate to large amplitude. Current profile is likely a key parameter governing the whole process, and an important factor in establishing scenario viability. Therefore it becomes particularly important to evaluate the prevalence and sensitivities of 2/1 mode threshold in advanced inductive, to assess: (i) whether the changes in current profile for the more advanced regimes improves stability (and how to improve further); (ii) to establish viability and limits for the regime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish low torque variant of advanced inductive plasma. Test prevalence of modes by varying current profile formation recipe (eg early hearing timing) and betan, between standard advanced inductive values and relaxed (later heating start, lower betan) ITER baseline like regimes. Test 3-D field role with n=1 I coil ramps in some cases. Key goals are to determine how stability and 3-D field sensitivity vary with current profile and beta at low rotation. | ||
| Background: | |||
| Resource Requirements: | Varies from quick checks (few shots) on the back of low rotation regime development, to dedicated scans to achieve complete goals. | ||
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| Title | 413: High Beta Hybrids at Low Rotation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | This experiment will integrate a high beta hybrid plasma with the reactor relevance of Te~Ti and low torque heating from ECH and balanced NBI. We will examine the effect of toroidal rotation on the confinement and stability properties of these plasmas by comparing balanced- and co-NBI (only the co-beam case has steady-state capabilities). We will also explore if broadening of the fast ion pressure profile using the off-axis co-beamline, rather than an on-axis co-beamline, will result in a higher stability limit. The combination of the off-axis co-beamline and the counter-beamline should result in very broad pressure profiles.
Owing to the high efficiency for on-axis current drive, which will offset the modest bootstrap current fraction, this high beta hybrid scenario will satisfy the requirements for FNSF as well as (or better than) the high q_min scenario with strong off-axis current drive.The poloidal magnetic flux pumping that is self-generated in hybrids will suppress the sawteeth despite the strong on-axis ECCD, which is important for avoiding the 2/1 mode. The magnetic flux pumping will also keep the current profile nearly constant during the switch from balanced- to co-NBI. There will be a low rotation limit when the counter NBI torque near the plasma edge results in no pedestal rotation; it may be possible to mitigate this limit by optimizing the error field correction to reduce the drag on the edge rotation. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Start by repeating shot 133881. (2) Inject all six gyrotrons with central current drive. For the six co-NBI sources, increase the injection voltages as much as possible while maintaining a plasma pulse length of at least 5 seconds. (3) Optimize the dynamic error correction (may use broadband feedback), adjust the plasma shape for optimal pumping. (4) Reduce plasma rotation by injecting counter-beams. Evaluate the effect on the transport and stability properties. Is there a decrease in the stability limit for the 2/1 mode? Can this limit be increased by aiming some of the ECCD (4 gyrotrons) at the q=2 surface? (5) Use perturbations in the NBI power to measure the intrinsic rotation for these high beta hybrids. (6) Substitute the off-axis beamline for a (co) on-axis beamline to see if the resulting broadening of the fast ion pressure profile increases the stability limit for the 2/1 mode. | ||
| Background: | The current proposal for FNSF envisions a high q_min advanced tokamak scenario with 70% bootstrap current fraction. While this is compatible with the US view of DEMO, the physics of the high q_min AT scenario is still being developed. There is also an issue regarding the high off-axis current drive efficiency needed for a FNSF in this proposal.
Here I propose that the low q_min hybrid scenario is compatible with the requirements of FNSF, and it has several advantages. First, the physics basis is well advanced. Long duration hybrid discharge with high beta and high confinement are routinely achieved. Second, because q_min=1 in the hybrid scenario, all of the external current drive can be deposited near the plasma center where the current drive efficiency is the highest (because of the lack of trapped particles and the high electron temperature). While the bootstrap current fraction will be lower in this low q_min hybrid scenario (50% rather than 70%), the increase in the current drive efficiency for central deposition more than makes up for this. Experiments on DIII-D have come very close to demonstrating this scenario using five co-beams and five gyrotrons. Hybrid plasmas with beta_N=3.4 were stably produced with a loop voltage of 9 mV. The loop voltage was a strongly decreasing function of heating power. While the ion and electron temperature were nearly the same outside of rho=0.2, the H-mode confinement factor remained high, H_98=1.4. This result is better than for the typical hybrid regime on DIII-D and is correlated with better than usual electron thermal transport in this LSN plasma shape. Therefore, this proposal will likely lead to the development of a high beta, high confinement, steady state scenario based on the hybrid regime. The remaining question regarding reactor relevance will the be role of high toroidal rotation. While we cannot separate the roles of torque injection and current drive in the NBI system, we can evaluate the role of rotation in the plasmas by using the counter beams. I have long predicted that the thermal pressure profile should be broader for low rotation discharges because strong core ExB forms an ITB. Indeed, profile analysis of high rotation and low rotation hybrids at the same beta_N shows that the thermal pressure profile is considerably broader for the low rotation case. However, a broader total pressure profile has not been evident in low rotation hybrid discharges because (1) the reduction in confinement at low rotation means that the NBI power has to be increased to keep beta_N constant, (2) the fast ion pressure profile is very peaked, (3) the summation of the larger fast ion pressure profile and the smaller thermal pressure profile leaves the total pressure profile nearly unchanged. We can overcome this by using the off-axis beamline to heat the low rotation hybrid with a broad fast ion pressure profile. The resulting broad total pressure profile may have higher stability limits. | ||
| Resource Requirements: | NBI: All 8 beam sources are needed.
ECH: 6 gyrotrons with 4 MW of injected power. FW: It is desirable to couple 1 MW or more, but core absorption needs to be demonstrated. I-coils: Dynamic error field correction will be used (possibly broadband feedback). | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | TRANSP for current drive and transport, DCON for stability. | ||
| Other Requirements: | -- | ||
| Title | 414: Assess optimal Error Field Correction by modulating I-coils at incommensurable frequencies | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Perform several non-destructive Error Field Correction (EFC) tests within a single discharge, including non-resonant components. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Feed AC currents of incommensurable frequencies to each pair of I-coils in order to generate error fields that are different at every instant. Infer from plasma rotation or other indicator what set of currents gives best error correction. | ||
| Background: | Most of the times, optimal EFCs are assessed by ā??trial-and-errorā??, with a new EFC being tested in each shot. One of the reasons for this is that the current indicator of optimal EFC is the lowest density at which the plasma can be ramped down without locking. This usually results in a disruption and can thus be considered a ā??destructive testā?? (destructive of the plasma). Utilizing a non-destructive indicator such as the plasma rotation (faster or slower, depending on how strong the magnetic braking from the residual error field is) would allow multiple EFC tests within a single discharge. The limit on how many EFC configurations can be tested is set by the plasma rotation response time. At this point, there are various choices on how to conduct the EFC scan during the discharge. For example, one can fix the phases between the I-coil circuits and scan the currents, while keeping their ratios fixed. This would fix the 3D geometry of the EFC field and scan the overall strength of the correction. Alternatively, one can fix the strength and vary the phases so as to ā??rigidlyā?? rotate the EFC. In reality, to maximize the number of configurations, one can change the amplitude and the phases, as well as the topology, including non-resonant components. The latter can be achieved by individually modulating the I-coils, which is technically possible. In particular, modulating them at different frequencies would permit to test various strengths, topology and directions of the resulting EFC. To maximise the number of configurations, distinct coils should be modulated at incommensurable frequencies (incommensurable over the duration of a discharge, or a number of discharges). | ||
| Resource Requirements: | |||
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| Title | 415: EFC by ramping beta up, or q95 down, or density up (instead of density down) in 3-4 discharges | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Revise a robust and well-established Error Field Correction (EFC) ā??classicā?? by ramping other quantities that cause locking. This extends the technique from low-density to more ITER-relevant high-density, high-beta and low q95. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Prepare 4 different I-coil or C-coil settings as for standard EF-assessment discharges. Run 12-16 good discharges: 4 in which the NBI power is ramped up, 4 in which the density is ramped up, 4 in which q95 is ramped down and, for reference and if time, 4 conventional EFC discharges in which the density is ramped down. Continue all ramps until locking but not beyond, to prevent disruptions. Use the ā??dud detectorā?? to stop the ramp at locking, or to order NBI decrease to make the mode less disruptive, or magnetic perturbations and ECCD to suppress it. | ||
| Background: | Standard Error Field Correction (EFC) utilizes low-density locked modes (LDLMs) generated by error-field (EF) penetration at low-density. Note that the EF in question is the total EF resulting from the machine EF and an applied magnetic perturbation (MP). By applying different MPs in 3-4 discharges and ramping down the density, the lowest density achievable and the higher or lower susceptibility to locking are assessed. These translates into larger or smaller total EFs that, because the applied MPs are known, allow to indirectly infer the machine EF.
This EFC however, only applies to low-density plasmas. Its extension to high density or beta would require taking the Resonant Field Amplification (RFA) into account. Instead, here it is proposed to directly measure the EFC at high beta, high density, or low q95. The key point is that all of these conditions, like low density, make the plasma more susceptible to locking. Hence, just like lower density, it is proposed to use the achievement of higher beta, higher density or lower q95 without locking as an indicator of better EFC. | ||
| Resource Requirements: | |||
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| Title | 416: Variation of turbulence and transport with RMP amplitude | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | S. Mordijck; acknowledge useful discussions with Jim Callen. | ITPA Joint Experiment : | No |
| Description: | One possible mechanism for the RMP-induced transport changes that lead to ELM suppression is modification of the fluctuation-driven transport by the RMP. In CY2010, we completed a series of discharges with I-coil modulations between approximately 4 kA and 2 kA with duration short enough to maintain and ELM-free plasma. The intent was to turn the RMP "off" (although the power supplies were known to not come on again once turned off so a low current was chosen - initially targeted to be 500 A). Operationally, however, the power supplies tripped for any lower I-coil current below 2 kA. Analysis of this experiment suggests that covering the range of modulation in smaller steps from the ELM-suppressed level would be very useful for isolating the turbulent transport response to the RMP. We also ran out of time and were forced to choose between moving the BES 2D array further in or out; we would document the fluctuation behavior in the region from 0.85 to 1 in rho which was missed in CY2010. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a "standard" ISS shape, ITER pedestal collisionality ELMing H-mode with co-NBI; modulate the I-coil in small (0.5 kA) steps down from the level needed for suppression while measuring turbulence, transport (profile), rotation and Er changes (the latter with fast CER). Previously, the neutral beams were modulated during the fast CER time, complicating the analysis and quality of the data: this time, keep all NB programming constant during the fast CER window to improve the quality of the fast rotation and Er data. Because comparing fluctuation data between ELMing and ELM-free times makes interpretation difficult (due to the ELM-induced effects being convoluted with the RMP effects), this experiment should be done with co-NBI and requires ELM suppression to be achieved. Once a range of I-coil modulation steps are completed, repeat while raising pedestal collisionality for validation of plasma response models. | ||
| Background: | As shown in Moyer APS Contributed Oral 2010, the inverse scale lengths for ne and Te show that the change in transport occurs in the region psi_n = 0.9 to 0.98. At 0.93, a/Lne drops from 5 to 3 (less than a factor of 2), while at the same radius, a/LTe drops from 10 to 1. This behavior is similar to expectations for stochastic transport (much stronger reduction in electron thermal transport than in particle transport). Existing data also shows that modulation of the RMP results in a prompt response to the toroidal and vertical rotation in the pedestal on the same timescale as the I-coil current ramp [see Moyer APS Contributed Oral, 2010]. In the pedestal, there is an equivalently fast change in the density fluctuation level and the pedestal density profile. However, the large step (4 kA to 2 kA) resulted in a transition from ELM-suppressed H-mode to ELM-free H-mode. This experiment will use this modulated I-coil technique in the ELM-suppressed state to document the dependence of the pedestal rotation, Er, fluctuations and transport on the RMP using smaller I-coil steps that should not trigger a transition to ELM-free H-mode. | ||
| Resource Requirements: | I-coil, 5 co NBI, fluctuation, profile, and CER diagnostics; fast CER measurements. Fast imaging of the edge with the UCD ECE imaging, and the UCSD fast cameras. Attempt phase coherent detection of the RMP structure using the fast camera and image intensifier. | ||
| Diagnostic Requirements: | Full fluctuation, profile, and CER diagnostics. Fast CER for a portion of the I-coil modulation. ECE imaging. UCSD fast camera with fast image intensifier to perform coherent detection of the edge structure during the modulations. | ||
| Analysis Requirements: | 2D kinetic equilibirum analysis; TRIP3D runs; development of phase-coherent detection of the RMP-induced edge structures imaged with the UCSD fast camera and the fast image intensifier.
Fluctuation, profile, rotation, and Er analysis. | ||
| Other Requirements: | -- | ||
| Title | 417: Dust generation from deposited layers and leading edges | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | C. Wong, J. Yu, M. Groth, N. Brooks, M. Fenstermacher, S. Krasheninnikov, C. Lasnier, R. Smirnov, W. Solomon, J. Watkins | ITPA Joint Experiment : | No |
| Description: | Characterize dust generation from DiMES samples with
pre-deposited hydrocarbon films and specially machined leading edges. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | DiMES samples with pre-deposited hydrocarbon films and specially machined leading edges
will be exposed to known particle/heat fluxes at the strike point in LSN configuration. Dust generation will be characterized by available diagnostics (visible cameras, IR TV, MDS) and postmortem analysis of the samples. | ||
| Background: | Dust production and accumulation present potential safety and operational issues for ITER by contributing to tritium inventory rise and leading to radiological and explosion hazards. In
addition, dust penetration of the core plasma can cause undesirably high impurity concentration and degrade performance. Projections of dust production rates based on experience from existing devices are needed. ITER physics work programme for 2009-2011 calls for ā??Exposure of tokamak generated deposits (carbon in the short termā?¦) to ITER relevant transient heat loads and analysis of generated dust". | ||
| Resource Requirements: | Two experiments, one with pre-deposited layers, one with a leading edge. 1-2 setup shots and 2-3 exposure shots requested for each experiment. LSN patch panel, OSP sweep on DiMES. | ||
| Diagnostic Requirements: | DiMES, DiMES TV, lower divertor tangential TVs, UCSD fast camera, CER, Thomson
(divertor and core), filterscopes, MDS, lower divertor Langmuir probes, SPRED. IR TV and fast visible camera with view of DiMES are highly desirable. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 418: Far off-axis NBCD using vertically shifted small plasma | ||
| Name: | Park parkjm@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | M. Murakami, B. Heidbrink, M. Van Zeeland, M. Wade, et al. | ITPA Joint Experiment : | No |
| Description: | Evaluation of off-axis NBCD physics at rho>>0.5 | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use upshifted small plasma developed in 2008 for off-axis NBCD measurement. Scan vertical position of plasma at maximum tilting angle of off-axis beam and/or scan tilting angle at fixed plasma position. Start with +BT/+IP and repeat with +BT/-IP and -BT/+Ip. Measure OANB profile, beam ion density/energy distribution, and fast ion loss as a function of NBCD location. | ||
| Background: | This experiment is multi-purposes. For example, far off-axis NBI will allow a wide range of variations in or direct control of the rotation and radial electric field profiles near the edge pedestal as well as fast ion orbit loss to test its impact on the edge pedestal structure. We may also have better chance to study the effects of micro-turbulence on fast ion confinement since the thermal ion diffusion is in general larger when moving to outer radius region. | ||
| Resource Requirements: | All neutral beam sources with 150 beams | ||
| Diagnostic Requirements: | MSE, Neutrons, FIDA spectrometers & cameras, FILD, Core and edge reflectometer | ||
| Analysis Requirements: | NUBEAM, ONETWO, TRANSP, NVLOOP, FIDASIM analysis | ||
| Other Requirements: | |||
| Title | 419: Aerogel targets to study velocity, size and composition of dust particles in DIII-D SOL | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | S. Ratynskaia (Royal Institute of Technology, Stockholm), C. Castaldo (Euratom ENEA Association, Frascati, Italy) | ITPA Joint Experiment : | No |
| Description: | Targets made of silica Aerogel highly porous material composed of clusters of 2-5 nm solid silica spheres with up to 95 % empty space will be used to study velocity, size and composition of dust particles present in the outboard SOL of DIII-D during plasma discharges. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Aerogel target will be installed in Midplane Material Evaluation Station (MiMES) and kept just
outside of the wall tile radius in 240R0 port during plasma discharges for a few days. Then the target will be removed and analyzed for captured dust. | ||
| Background: | Dust penetration of the core plasma in ITER can cause unacceptably high impurity concentration and degrade performance. Therefore, knowledge of the dust transport and dynamics is important. Detection of dust present in the plasma during discharges is non-trivial. The main parameters of interest, apart from the particle material, are dust velocity, size and number density. Due to the uncertainties in the present estimates of the dust parameters it is important that diagnostics cover the maximum possible range of these parameters in order not to overlook some dust populations. A new method for dust collection has recently been proposed and is based on the use of aerogel - a highly porous, very low density material. Aerogel collectors can capture dust grains without destroying them, even in the high velocity range. Analysis of the tracks of captured particles allows to evaluate the dust velocity and the dust composition can be deduced upon particle extraction. | ||
| Resource Requirements: | MiMES with a slot for an Aerogel target. Piggyback experiments, no machine time requested. May be performed during plasma startup following LTOA. | ||
| Diagnostic Requirements: | Fast USCD camera coupled to 135T0 port is highly desirable. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 420: Pair formation during disruptions | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Not requested |
| Co-Author(s): | Alex James (UCSD) | ITPA Joint Experiment : | No |
| Description: | Provide first evidence of disruption-generated positrons. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform microwave and gamma-ray measurements during disruptions. Cause of disruption is not important. Can be combined with any experiment with intentional or probable disruptions, especially if copious runaway electrons are expected.
1. Microwaves: discriminate between clockwise and counter-clockwise elliptical polarization at the X2 harmonic in the oblique ECE radiometer. Because positrons gyrate opposite to electrons, polarization also change. Counter-clockwise polarization would be a signature of "Positron Cyclotron Emission". 2. Gamma-rays: Use scintillators to compare forward and backward emission either within the same discharge, with two scintillators looking in the co- and ctr-Ip direction respectively, or with the same scintillator in two discharges of opposite Ip and BT. Emission will be in one case the sum of backward emission from electrons (e-) and forward emission from positrons (e+). Because forward and backward emission of each species are related to each other, it will be easy to recognize anomalies in the comparison with the forward emission from e- summed to the backward emission from e+. 3. Finally, coincidence counters of 511keV photons, of the type used in accelerators or in positron emission tomography might also be utilized. | ||
| Background: | P. Helander (IPP Greifswald) predicted that, during disruptions, collisions between runaway electrons and thermal ions generate electron-positron pairs [PRL 2003]. Here it is proposed to provide the first experimental evidence of this theoretical prediction.
This will be the first step toward learning how to maximize pair production: at present, the predicted pair-production rate is small and expected to have a small effect in the energy economy of the disruption, but it is not excluded that it can be increased by tailoring the runaway electron beam. Note also that, although the predicted positron population is small, each pair removes more than 1MeV from the disruption energy balance. Channelling disruption energy in these electron-positron pairs might constitute an innovative mitigation scheme: pairs eventually annihilate in 511keV photons, relatively innocuous for the tokamak walls, certainly much less harmful than massive particles. First, however, we need to prove the formation of positrons in a tokamak for the first time. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Scintillators, Oblique ECE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 421: SOL width in top-limited discharges | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | R. Pitts (ITER), J. Boedo, A. Leonard, C. Lasnier, G. Jackson, P. Stangeby, R. Moyer, J. Watkins | ITPA Joint Experiment : | No |
| Description: | Experiments were performed on DIII-D in 2009 to benchmark the assumed ITER SOL power width scaling for startup/ramdown limiter phases. Both the high field side (HFS) and low (LFS) field side startup options are considered for ITER. In DIII-D a good data base of HFS-limited discharges was obtained and used for comparison with the scaling. However, only one good discharge was obtained in the top-limited configuration - the best proxy to a toroidally symmetric LFS-limited configuration available at DIII-D. We propose a 1/2 day experiment designed to complete the LFS-limited part of the data base. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Experimental approach will be similar to that used in 2009 experiments. The main diagnostic will be mid-plane reciprocating probe that will be plunged twice in every discharge. Shape and parameters of shot 136595 should be restored, then the shot will be repeated with NBI power going from 0 to 1.1 MW around 3 seconds into the discharge. Plasma current and density will be varied from shot to shot. Some LSN discharges may be run for reference. | ||
| Background: | The ITER first wall(FW) is being designed to allow start-up on the actively cooled beryllium panels on both the high (HFS) and low (LFS) field sides, and plasma scenarios have been developed. Power handling is determined by the parallel heat flux density, and the panel shaping. The former is characterized by the SOL power flux density e-folding length lambda_q. ITER presently assumes a modified divertor scaling based mainly on data from JT-60U and JET for lambda_q in the limited phase. Experiments performed on DIII-D did not confirm the functional dependencies on the plasma parameters assumed in the scaling, but most measured values of lambda_q in HFS-limited configuration agreed with the scaling within the assumed uncertainty (a factor of 2). For LFS-limited configuration only one good shot was available, so more data are needed for a meaningful comparison. | ||
| Resource Requirements: | 1/2 day experiment (~10 documentation discharges). Top-limided Ohmic and L-mode with up to 1.1 MW of NBI. | ||
| Diagnostic Requirements: | Mid-plane reciprocating probe, IRTV (if LSN discharges are run), core Thomson, CER, fast UCSD camera, tangential TVs, mid-plane filterscopes, profile reflectometry. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 422: Image RMP-induced structures using fast cameras and coherent detection | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | A critical step in developing predictive understanding of RMP ELM -control for ITER is a direct validation of the resulting magnetic field topology in the pedestal region of the plasma. The new UCSD fast framing camera (higher sensitivity and bit resolution), in combination with the fast image intensifier, can be used to image structures in the edge of the plasma (along the centerpost and possibly the outer midplane) in conjunction with a.c. modulated I-coil currents to allow coherent detection of the RMP-induced structures, similar to the measurements of internal MHD mode structure performed previously by J. Yu and M. Van Zeeland. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish RMP ELM-suppressed discharges. Add an a.c. modulation to the I-coil current level needed for suppression. Image the plasma boundary (structures have been previously seen on the high field side near the centerpost) with the fast framing camera and image intensifier. Use the modulation to perform coherent detection of the induced structures (if present). Scale the RMP current, collisionality, and momentum input to validate rotational screening models. | ||
| Background: | In order to develop predictive understanding of how RMP ELM-control will work (or not?) in ITER, we need to understand the nature of the plasma response to the applied RMP - we need to know the resulting magnetic field in the plasma. One approach which has proven successful in previous stochastic layer experiments has been imaging of the remnant island structures generated by the applied RMP. Initial measurements in DIII-D with the fast cameras have captured structures along the centerpost, but have not yet resolved structure on the outer midplane (where resolving such structures requires high spatial resolution). IN CY2010, UCSD acquired a new fast framing camera with improved resolution and bit depth, enabling most complex analysis of the resulting data. We propose to use this camera, in conjunction with an image intensifier, to image the plasma boundary in "conventional" ISS RMP ELM-suppressed discharges to directly compare with magnetic topologies computed with existing plasma response models, including NIMROD, M3D, XGC0-M3D, M3D-C1 etc. | ||
| Resource Requirements: | 5 co beam sources; I coil with modulation capability. | ||
| Diagnostic Requirements: | fast framing camera and image intensifier. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 423: Investigation of Plasma Response at Zero net torque input | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of this experiment is to provide an optimized set of conditions for testing of plasma response models which incorporate plasma rotation effects. Previous efforts to maintain RMP ELM suppression at zero net torque input from co and counter neutral beams have had little success due to the development and locking of internal MHD modes (tearing modes) when the core rotation dropped below about 40 km/s. The idea of this proposal is to first establish an ELMing H-mode at zero net torque using ECH only (with limited beam blips for diagnostics). The existing plasma response models (NIMROD and M3D extended MHD codes; XGC0-M3D; etc.) are relatively limited in the length of the "real" discharge which can be simulated with a finite number of CPU cycles (XGC0 simulations are typically 4-6 ms while M3D cases can be much shorter than this). Further, the goal of this experiment is to test models of plasma response, which doesn't require that ELM suppression is obtained provided the time around the I-coil on phase is well diagnosed. The focus on zero net torque, while ITER relevant, will also make changes in the pedestal rotation (especially the critical electron diamagnetic and E x B rotation) more apparent. This zero net torque focus makes this proposal similar to one from Callen et al. which chooses instead to use balanced NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use ISS shape with cryo-pumping. Establish an ELMing H-mode with ECH (4-5 gyrotrons?). Use short beam blips for ion profile, rotation and Er diagnostics. Apply I-coil RMP (even and odd parity) and document changes to pedestal rotation, Er, profiles, and transport. Use imaging (ECE imaging, UCSD fast cameras, SXR X-point camera if available) to document changes to the structure of the discharge. | ||
| Background: | Rotational screening is one of the leading models for plasma response to the RMP. Waelbroeck's slab model predicts that islands will be suppressed by currents induced in response to the RMP. Islands will only open (the RMP will "penetrate") only on those surfaces where the electrons are unable to carry this screening current, i.e. where the electron perpendicular velocity is zero. Several MHD models (JOREK, M3D, NIMROD) have been used to model this fluid plasma response in more realistic geometries, and in CY10, the CPES group incorporated the MHD equations necessary for modeling the plasma response into a version of XGC0 (called "XGC0-M3D" although this isn't a coupled code: the "M3D" refers to use of the kernal from M3D to solve the MHD equations in the XGC0 model.). The XGC0 team (C.S Chang, G. Park, et al.) | ||
| Resource Requirements: | ISS discharges with cryo-pumping and enough ECH to reliably operate in ELMing H-mode (4-5 gyrotrons desired). Use beam blips to provide ion channel measurements. | ||
| Diagnostic Requirements: | The diagnostic set will be limited due to use of ECH (no mm wave diagnostics?) and limited beams (no BES) so the focus will be on the rotation and MHD response, not transport per se. | ||
| Analysis Requirements: | The XGC0 group has expressed interest in modeling these discharges as part of validation of the plasma response model in XGC0-M3D, and might be of interest to the MHD simulation groups as well. | ||
| Other Requirements: | |||
| Title | 424: Evaluation of Soft X-Ray Imaging System Resolution & RMP Penetration; n=1, L-mode | ||
| Name: | Shafer shafer@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans, E. Unterberg, D. Battaglia, J. Canik, J. Harris | ITPA Joint Experiment : | No |
| Description: | Application of RMP field may or may not generate island structure given plasma response. The new tangential SXR camera is poised to make new measurements of island structure. However,the DIII-D SXRIS needs dedicated time to scan through operating space. In principle, this space includes: I-coil amplitude, applied mode number (1 vs 3), phasing, q95, H vs L, Rotation. We propose a limit set where some discharges can contain multiple scans. We plan to start with low-rotating (balanced injection) L-mode plasma, with theoretically low screening and vary the resonant island size and location. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start at a balanced injection, with intent to keep in L-mode. Use n=1 I-coils, with the goal of imaging 3/1 and 4/1 islands. Ideally, the low rotation and lower temperature pedestal will screen fields less, allowing greater penetration. The I-coil n=1 phasing is needed to control the placement of the 3/1 resonance in the proper field of view of the camera. q95 can be varied to change the location of the 3/1 & 4/1 surface w.r.t. to psi_norm. Power can be increased later in the discharge to get n=1 H-mode, where screening is stronger. Repeat at different I-coil amplitude to change the island size. | ||
| Background: | A new tangential Soft X-Ray Imaging System (SXRIS) will be installed for this run campaign. This system is designed to examine magnetic the magnetic topology at and inside the pedestal by examining the X-point region, where flux expansion is large. The formation of islands due to RMPs is an open issue and the resolution of the SXRIS system needs to be verified. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | SXRIS | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 425: Impact of triangularity and pedestal rotation on RMP penetration | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of this experiment is to document the variation in pedestal rotation and transport as a function of q95 (outside and inside the ELM suppression resonant window(s) ) in low and high triangularity discharges. Previous ISS discharges have shown 1 to several narrow windows in q95 in which ELM suppression is obtained, in contrast to the much broader q95 region for which density pumpout is obtained. The differences in these two behaviors suggests that it may be possible to further optimize the ELM suppression while reducing or limiting the extent of particle transport induced by the RMP. Time-dependent analysis of high triangularity, ISS discharges correlated the narrow ELM suppression windows with reductions in the pedestal Te (still below levels anticipated from Rechester-Rosenbluth). In this experiment, we will explore the existance of such multiple, narrow resonances in q95 for ELM suppression, document the role of increased electron thermal transport during these windows, and finally document the dependence of this "stochastic layer-like" response on plasma shaping (triangularity) by making measurements in low triangularity as well. These comparisons should also shed light on the role of neoclassical effects on the transport response by varying the trapped particle fraction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop an RMP ELM-suppressed H-mode with low traiangularity: this is a non-trivial task since it requires operation on the shelf, reducing the effectiveness of the pumping and leading most likely to higher colllisionalities. However, this might provide a better configuration for exploring the collisionality dependence of the plasma response (for bridging the gap between the ISS pedestal collisionality ~ 0.1 and the ITER shape pedestal collisionality > 1 datasets). Scan q95 (using, time permitting: Bt ramp at constant Ip, and Ip ramp at constant Bt) and document windows for ELM suppression. There are documented differences in the pedestal rotation profiles at low and high triangularity (and in the impact of the RMP on the core confinement and pedestal profiles) which should allow for validation of plasma response models based on rotation. | ||
| Background: | |||
| Resource Requirements: | development of a LSN low triangularity discharge with relatively low pedestal collisionality and reliable RMP ELM suppression. | ||
| Diagnostic Requirements: | "Standard" rotation, Er, profile, and fluctuation diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 426: Electron Bernstein Wave Heating and Emission Studies | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate Electron Bernstein Wave (EBW) emission, heating and current drive, for the first time in a big tokamak, utilizing the Ordinary-eXtraordinay-Bernstein (OXB) mode conversion. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) angular scan of EBW emission with the oblique ECE 108-112GHz radiometer, as piggyback on pellet or other high density discharges; 2) dedicated H&CD attempt at 110GHz, assisted by pellet or other means of achieving ne>1.5e20m-3 in the core: use max ECH power but modulated with low duty-cycle, to avoid excessive stray and damages to diagnostics. Perform a shot-to-shot angular scan in steps of 2deg around calculated optimum. Heat-pulse-analyze SXR and stored diamagnetic energy (ECE will be in cutoff) looking for evidence of EBW-generated heat waves. | ||
| Background: | There are two main reasons of interest in EBW studies at DIII-D.
The first one is that two new concepts have been recently proposed to measure the q-profile at the DIII-D edge with unprecedented microsecond, millimetre resolution (ask details to F. Volpe and T. Luce). Both concepts are based on the OX mode conversion and on its dependence on the line of sight. Additionally, one of the two concepts, to be tested here, also requires that the plasma emits EBWs. Obviously, before applying it to the proposed edge q-profile diagnostic, it is important to provide evidence of EBW emission at DIII-D in the first place. Another reason of interest, more historical, is that the density range of operation of ECE and ECH would be extended. Ideally, this would require new gyrotrons and radiometers operating at the first EC harmonic (~55GHz) rather than at the second (110GHz). This would guarantee a range of densities at which conventional ECE and ECH can coexist with (and transition to) their EBW counterparts. However, several tests can be performed with the existing equipment. At DIII-D, conventional 2nd harmonic X-mode (X2) ECH and ECCD at 110GHz, as well as a considerable number of channels of the ECE radiometer, go in cut-off at ne=7.5e19m-3. If the local density could non-disruptively, e.g. by pellets, be raised up to ne>1.5e20m-3, an additional microwave heating, CD and diagnostic scheme would become available. This makes use of OXB-converted EBWs, requiring the presence of the O-mode cut-off in the plasma, steep ne gradients (making the ne scale-length comparable with the wave-length) and a special view/launch direction, accessible to the DIII-D launchers. EBWs were used with success in stellarators and spherical tokamaks but, so far, only in one conventional tokamak: TCV. It is the aim of the present proposal to extend it to larger tokamaks such as DIII-D, also in view of testing the feasibility of EBW-based divertor diagnostics being considered for ITER. The high densities required for this experiment are a consequence of exploiting the existing hardware at 110GHz, i.e. at the second harmonic. In case of encouraging results, one might consider radiometric measurements and eventually heating at 60GHz, i.e. at the first harmonic. The requested density in that case would be 4.5e19 m-3, which is routinely accessed at DIII-D. | ||
| Resource Requirements: | pellet, 5 gyrotrons | ||
| Diagnostic Requirements: | Oblique ECE radiometer | ||
| Analysis Requirements: | toray, with modifications | ||
| Other Requirements: | |||
| Title | 427: Torque Waves - Transport of Angular Momentum | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Inject torque at specific radial locations. See how the CER rotation profile evolves. Repeat in a pulsed, repetitive way, to generate "torque waves". From their propagation, infer transport of momentum. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a discharge in which a rotating n=1 magnetic perturbation (MP) was successfully entrained to an initially locked 2/1 mode, or to an initially rotating 2/1 NTM (see proposals #134 and #133 respectively). Repeat but turn the MP off after 2-3 momentum confinement times (assumed of the order of the local energy confinement time). After a comparable time, turn on again. For this technique, it is important for the mode to always be present -and, at the same time, not to grow too large-. If the mode or the plasma do not survive to this first on-off-on sequence, either because the mode gets too small or too disruptive, change the beams and therefore beta accordingly, or add ECH for mode stabilization. When successful for 1-2 pulses, repeat for several pulses within the same shot. Acquire CER. | ||
| Background: | Rotating MPs have been successfully used to unlock initially locked modes and force their rotation. The coupling of rotating MPs with rotating NTMs is now under investigation (#133).
Either way, MPs impart momentum to tearing modes. Indirectly, they also impart momentum to the plasma, because islands are nearly "frozen" in the plasma, therefore the rotation of the island nearly coincides with the rotation of the plasma, apart from a small offset. Here it is proposed to inject torque in the plasma at specific radial locations, by means of rotating MPs: if there are NTMs in the plasma, the MPs will impart momentum to them and, indirectly, to the plasma, at the rational surface locations. It will be interesting to measure, for example by means of the CER diagnostic, how the toroidal rotation profile evolves/relaxes after the perturbation. In particular, it will be interesting to time-modulate this space-localized injection of torque. "Torque waves" will be generated as a result, similar to "heat waves"" generated by modulated, localized heating. Temperature diagnostics such as ECE allow to study the propagation of these heat waves and, from them, infer the profiles of heat transport coefficients. In the same way, it is proposed to diagnose "torque waves" by CER, measure how they propagate (viscously or by other means) i.e. how momentum is transported. | ||
| Resource Requirements: | SPAs, I-coils | ||
| Diagnostic Requirements: | CER | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 428: Coil multi-tasking: simultaneous EFC, ELM, RWM and Locked Mode control | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | General IP | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Simultaneously use I-coils for EF compensation, RMP control of ELMs and for RWM control, with transition to locked-mode control if necessary. Look for conflicts or synergies between various controls. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Simple superposition of AC, <1kA RWM control waveform, whatever algorithm or method it is based on (e.g. dynamical error field correction, DEFC) on top of a DC, 4-5kA baseline for ELM control. Ramp down q95 in initial discharges, then continue at optimal values for ELM control. Try to obtain first ELM suppression above no-wall limit (and, for comparison, shots with ELM control only, RWM control only, and no control at all).
Enable dud detector to trigger a change from RWM to locked mode control, to cause the mode to lock with the O-point in the ECCD view. If time, repeat the experiment with the C-coils and compare with I-coil results. The possibly superior capability of the internal coils to simultaneously cope with multi-mode control of multiple instabilities might affect the ITER decision with re: internal coils. | ||
| Background: | I-coils have been used with success at DIII-D to correct the error field (EF), control ELMs, RWMs and, more recently, locked modes (LMs). While EFC is always in place, typical experiments only aim at controlling ELMs, RWMs or LMs in every single discharge. In ITER and in a fusion power plant, however, all of these instabilities will have to be controlled at the same time, possibly by the same set of internal or external coils, which will thus have to be multi-tasked.
On the way to an integrated coil system for ITER, it is important to assess conflicts or synergies between various control techniques. In this first proposal we propose to explore conflicts and synergies between ELM and RWM control, and to switch if necessary from RWM to locked mode control (in particular, control of the toroidal phase of locking, for ECCD in the island O-point). The concern is that optimal current and helicity settings for RWMs might not be optimal for ELMs and viceversa. For example, the n=1 dynamic correction for the RWMs might change the optimal q95 window for ELM control. On the other hand, from a RWM perspective, ELM-suppressing RMPs are a static, typically n=3 error field affecting the plasma rotation and thus the RWM stability. Finally, the n=3 ELM control field will have an effect on locking: it will brake the plasma, thus making locking more rapid, which in general is undesirable, but will not be a worry in ITER, where all time-scales are expanded. The n=3 field will also slightly change the locking position. Hints of simultaneous RWM suppression and ELM mitigation can be found in A. Garofaloā??s shots 122591-594, where an n=3 magnetic braking field was applied with the C-coils, during n=1 RWM feed-back with the I-coils. Although the real goal of those experiments was to study RWM control in the presence of n=3 magnetic braking, they provide useful information on the interplay with RMPs for ELMs. For instance, they seem to confirm the above speculation on the modified q95 window. Although encouraging, the experiments will need to be repeated at increased n=3 current (6kA, as 3kA were marginal for ELM suppression). Moreover, the I-coils can be wired as to simultaneously apply an n=1 and n=3 perturbation. Finally it will be beneficial to drastically reduce the gas puff and move the strike point to improve the pumping and limit the collisionality, and to reduce the triangularity (was 0.6). All these modifications go in the direction of facilitating the ELM suppression. It will also be important to avoid ELM-free H-modes or temporary losses of H-mode, in order to isolate real ELM-suppression evidence. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 429: Poloidal rotation of plasma column to simulate rotating wall | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Stabilize RWMs by making the plasma column rotate poloidally, which is equivalent to make the wall rotate. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In spite of the different context and goal, the plan is similar to the ELM-control proposal #393. Vertical oscillations of +/1cm will be pre-programmed in the PCS. These will be made faster and faster, from shot to shot or within the same discharge. Frequencies >200Hz, comparable with the inverse resistive-wall time, are necessary to see the stabilizing effect of the ā??fakeā?? rotating wall. Horizontal oscillations of similar amplitude and frequency will first be studied separately and then super-imposed to the vertical ones. They will be out-of-phase by 90deg, to make the plasma column rotate poloidally.
Note that at high oscillation or rotation frequencies the plasma is not expected to move ā??rigidlyā??. EFIT and Thomson Scattering will probably be too slow, but ECE will allow to diagnose the deformation of the plasma as it moves. | ||
| Background: | It is predicted by theory, and has been recently verified experimentally by the Rotating Wall Machine at UW-Madison, that if two conducting walls surround the plasma and if one of the walls is moving poloidally with respect to the other, the Resistive Wall Mode can be stabilized [http://plasma.physics.wisc.edu/rwe-mission]. In the long term, liquid Lithium walls might one day serve this and other purposes in a large tokamak. Meanwhile, it is proposed to study this effect by mimicking the effect of the rotating wall, by making the plasma column rotate instead. This would be equivalent to a single wall, poloidally rotating, instead of a static and a rotating one as in the Madison experiment, but is expected to be stabilizing anyway, in the same way as toroidal rotation stabilizes RWMs at DIII-D. | ||
| Resource Requirements: | |||
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| Other Requirements: | |||
| Title | 430: Completion of heat flux width data | ||
| Name: | Lasnier Lasnier@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Thermal Transport in the Boundry | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Complete remaining needed heat flux width data. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We have a range of data, but it may become apparent with ongoing analysis that more data is needed. | ||
| Background: | A wide selection of heat flux width data has been obtained already. So may need to be supplemented or repeated. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 431: Plasma current ramps to test KBM physics | ||
| Name: | Groebner groebner@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | J. Callen, T. Osborne, P. Snyder, D. Thomas | ITPA Joint Experiment : | No |
| Description: | Perform fast plasma current ramps to vary the magnetic shear in the pedestal. Measure corresponding changes in total pedestal pressure gradient. Determine if pressure gradient varies with magnetic shear as expected from theory. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce plasmas with steady current and then ramp current down to a new steady level. Also perform upward ramps. Perform ramps in a time fast relative to the time in which the pedestal current density changes. Measure the temporal evolution of the total pedestal pressure gradient during the ramps. If possible, make measurements to help infer temporal behavior of edge magnetic shear. LIBEAM or edge MSE would be very useful. Also, consider using CER system to measure mod-B profiles out to pedestal. Plasmas with large pedestals, regular ELMs and long ELM periods are good candidates for this experiment. The ITER demo discharges fill these requirements. | ||
| Background: | The EPED1 model, using a combination of peeling-ballooning theory and kinetic ballooning mode theory has successfully predicted the pedestal height and width in a wide range of discharges in DIII-D and also in other tokamaks. While the peeling-ballooning theory is generally accepted as providing the ultimate limits to the pedestal pressure profile, the physics of the KBM is much less tested. The goal of this experiment is to vary the major controlling parameter for KBM physics and to determine if the plasma shows the predicted behavior. Specifically, the KBM model predicts that the pedestal pressure gradient varies roughly as the inverse of the square root of the magnetic shear (for the shear not too small and for positive shear). Better predictions can be obtained with BALOO. The goal of this experiment is to perform a rapid variation of the edge magnetic shear and to determine if the pressure gradient varies as predicted. The execution of this experiment will be similar to current ramps performed in JT-60U (Urano, Nucl. Fusion 49 (2009) 095006). | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 432: Pedestal density profile with an opaque SOL | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | R. Groebner, J. Callen | ITPA Joint Experiment : | No |
| Description: | Create a very opaque SOL to reduce recycled neutral fueling and compare the resulting pedestal density profile to a more strongly fueled pedestal. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The approach to creating an opaque SOL is to run at the highest density possible that maintains a divertor and SOL hot enough to ionize all recycling neutrals. The high density can be obtained by operating at high plasma current, at least 2.0 MA. High power will be needed to maintain the plasma confinement at high density. Typically at high density the inboard divertor becomes detached at high density. Operating in reversed toroidal field symmetrizes the inboard and outboard divertors allowing a much higher density before the inboard divertor detaches.
The experiment will essentially be a density scan. It will be carried out with reversed toroidal field, 2.0 MA, LSN and relatively high upper triangularity. Use NBI heating with feedback set to maintain BetanN just over 2.0. Start with the lowest attainable density and then raise the density in a series of discharges with good edge diagnostics and monitor the divertor plasma to see that it remains attached. After the highest density has been obtained drop the plasma current to 1.0 MA and run at low density, but high beta, and document the resulting pedestal density profile. Interpretation of this data will require modeling to determine how the fueling of the pedestal was varied as a function of density and how the density profile responded to the change in fueling. | ||
| Background: | A question remains concerning the H-mode pedestal density and if it is set primarily by neutral ionization fueling balanced by diffusion. Or is a particle pinch involved in building the pedestal density profile. One method to examine this is to reduce fueling within the pedestal to a very low level. This can be done by creating a very dense SOL and divertor that remains hot enough to ionize all recycling neutrals before they can cross the separatrix. This experiment will attempt this by operating at high current for a high density SOL and divertor and in reversed toroidal field to keep the inboard divertor attached and hot enough to ionize neutral deuterium. | ||
| Resource Requirements: | Reversed Bt, 2.0 MA in LSN. 6 NBI sources | ||
| Diagnostic Requirements: | All boundary and pedestal diagnostics | ||
| Analysis Requirements: | Neutral fueling modeling | ||
| Other Requirements: | |||
| Title | 433: Measurements of Off-Axis NBCD | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Accurately measure the difference in the NBCD profiles for cases of co-off-axis NBI (i.e., tilted 150 beamline) and co-on-axis NBI (i.e., 330 beamline). This should first be done for the 150 beamline at its maximum tilt angle; the results will determine whether it is necessary to repeat at additional tilt angles. The 30LT and 210RT beams will be pulsed for all cases to acquire MSE data.
The highest priority is to make this measurement in H-mode plasmas. These discharges should be MHD quiescent, so beta_N should be low enough to inhibit NTMs and early beam heating should be used to delay sawteeth. The NBCD analysis window is between the start of current flattop and the first sawtooth. The highest priority is to measure the NBCD profiles for both forward and reverse BT. These measurements will need to be made on separate days. Another important variable to scan is the electron temperature using central ECH. Otherwise the plasma density and beam injection energy can be varied of time permits. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Using early NBI (the 30LT and 330 sources), establish a H-mode discharge with beta_N~2 and delayed sawteeth. Use cyropumping to keep the density around 3E19 m^-3. (2) Document a long duration discharge using the 150 beamline at maximum tilt angle. The 30LT and 210RT beams should be pulsed to collect MSE data. (3) Repeat with the 330 beamline replacing the 150 beamline. If time permits, compare individual left and right sources in addition to both sources simultaneously. When making this comparison, keep the density and temperature profiles as fixed a possible. (4) Repeat steps 2-3 with core ECH to raise the electron temperature. (5) On a different day, repeat steps 1-4 with the other sign of BT. (6) If this is a full day experiment rather than a 1/2 day experiment, scan the plasma density and beam injection energy as well. | ||
| Background: | In principle the measurement of the NBCD profile for off-axis deposition is straightforward owing to the local measurements of the magnetic field pitch angle from the MSE diagnostic. These pitch angles are converted to the (flux-surface-average) current density and parallel electric field, either using equilibrium reconstruction or a more direct analysis [PPCF 47. 1077 (2005)]. However, the inductive current is often large, and the Zeff measurement typically has significant uncertainty. Therefore, the determination of the non-inductive current profile for any given discharge has relatively large error bars. These error bars can be reduced considerably if we concentrate on measuring the difference between two current drive sources. This is because the systematic errors, which are usually the dominant source of uncertainty, tend to cancel out. The random error can be made very small by using long analysis time windows.
The question then becomes what kind of current drive fiducial case should be used to compare with the off-axis beam? In previous cases of FW, EC or NB current drive measurements on DIII-D, a counter-injection case was used as the fiducial. However, for the off-axis beam measurement this does not appear to be the best fiducial for two reasons: (1) what we really care about is whether the off-axis beam drive a different current profile than an on-axis beam, i.e., the experimental choice for the SSI group is co-off-axis vs. co-on-axis, not co-off-axis vs. counter-on-axis. (2) There is evidence that the large change in the ExB shear between co- and counter-NBI affects the thermal pressure profiles, which alters the bootstrap current profile, etc. Therefore, it is best to compare the co-off-axis NBCD profile with the co-on-axis NBCD profile in plasmas with similar amounts of co-beam torque. | ||
| Resource Requirements: | NBI: 30LT, 150LT, 150RT, 210RT, 330LT, 330RT. The 150 beamline should initially be tilted the maximum amount.
ECH: 6 gyrotrons required. This experiment is in two parts: one with forward BT and one with reverse BT. The two parts can be executed on different days. | ||
| Diagnostic Requirements: | MSE (co and counter) is critical. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 434: L-H transition dependence on divertor detachment | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the H-mode power threshold as a function of density and the resulting inboard divertor detachment and SOL flow. Test the conjecture that the increase in power threshold at lower density is due to inboard divertor reattachment and a resulting drop in SOL flow. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure the H-mode power threshold as a function of density in an USN plasma with the Grad B drift towards the upper divertor. Monitor the SOL flow with the X-point Mach probe and upper inboard divertor detachment with the upper fixed languir probes, filterscopes, divertor spectrometer and divertor tangential cameras. Correlate the increase in threshold at low density with changes to the upper inboard divertor plasma and resulting SOL flow. | ||
| Background: | The H-mode power threshold is observed to increase at lower density. This trend is important for designing ITER operational scenarios and for determining the mix of power required for ITER to access H-mode. This experiment tests the conjecture that SOL flow from the outboard divertor towards the inboard side aids the H-mode transition. Also conjectured is this SOL flow is enabled by a detached inboard divertor plasma that allows easy access to the core plasma for neutrals born in the inboard divertor. The final part of this conjecture is that at low density the inboard divertor reattaches shutting off the SOL flow and thus inhibiting the H-mode transition. If this conjecture is born out, then the optimal density for ITER to achieve H-mode can be determined by accurate modeling of the ITERā??s inboard divertor detachment. | ||
| Resource Requirements: | USN and NBI heating ramps for determining H-mode power threshold | ||
| Diagnostic Requirements: | X-point and midplane Mach probes, all upper divertor diagnostics, particularly target langmuir probes and tangential TV | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 435: In/Out ELM heat flux asymmetry with counter injection | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the in/out asymmetry of ELM divertor heat flux deposition comparing co-NBI to counter-NBI injection. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Set up standard LSN ELMing H-mode discharges. Measure the ELM divertor heat flux with the fast IR camera. Additional useful diagnostics include the X-point probe for measuring Te on a fast timescale. Carry out an input torque scan to vary toroidal rotation from positive to negative with respect to the plasma current. Goal is to measure the ratio of inboard to outboard ELM heat flux as a function of toroidal rotation of the pedestal region. Fast Te measurements in the divertor will also help with understanding the ELM dynamics. | ||
| Background: | The in/out balance of ELM divertor heat flux is an important factor for determining a tolerable ELM size for ITER. Previous studies have typically found the ELM heat flux is greater to the inboard divertor by a factor 2-3 compared to the outboard divertor. This asymmetry changes direction, greater outboard ELM heat flux, with change in the toroidal field direction. A proposed explanation for these observations is that the ELM heat flux primarily flows with the ion convection. Co-NBI injection with the resulting pedestal rotation in the plasma current direction would result in preferential ion convection of ELM flux towards the inboard divertor for the case with the toroidal field direction with the GradB drift towards the X-point, and the opposite asymmetry for reversed toroidal field. This model also implies that the ELM asymmetry should also reverse if the pedestal toroidal rotation changes direction. This experiment is designed to test this conjecture. | ||
| Resource Requirements: | Co and Counter injection beams, LSN with both strike points in view of IR camera | ||
| Diagnostic Requirements: | Lower divertor IR camera | ||
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| Other Requirements: | |||
| Title | 436: Initial test of Super-X divertor configuration in DIII-D | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | General PBI | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Compare divertor operation in two configurations with large change in major radius of strike-point with the X-point location held fixed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop a divertor configuration that has many aspects of the proposed Super-X divertor. This configuration will require development of very high X-point, > 50 cm, with a long divertor leg. In the first configuration the outboard divertor separatrix will run vertically from the X-point to the target near Rmaj= 120 cm. In the second configuration the core plasma shape is the same as the first, but now the outboard divertor leg angles outward in major radius to near Rmaj= 165 cm. Once both configurations have been developed increase density in both configurations in H-mode until outboard divertor detachment onsets. Document divertor conditions and particularly detachment onset. | ||
| Background: | The Super-X divertor configuration has been proposed as a method to produce divertor detachment and heat flux dissipation at a lower upstream separatrix and core density. It is also suggested that this configuration may be less susceptible to a detachment instability where the detachment front quickly moves to the X-point. This configuration mainly consists of an outboard divertor strikepoint at much larger major radius than the X-point. This experiment will be a first test of producing such a configuration and DIII-D and initial investigation on how this new configuration affects divertor operation. | ||
| Resource Requirements: | Significant plasma control resources will be required to develop the divertor configurations needed in this experiment | ||
| Diagnostic Requirements: | Lower divertor diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 437: Error field correction for DIII-D and ITER (3): Feedback control of locked mode | ||
| Name: | Strait strait@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | M. Chu, A. Garofalo, R. La Haye, H. Reimerdes, M. Schaffer, R. Buttery | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to develop a possible new approach to error field correction at low beta, based on direct feedback control of a locked mode. It is important to note that the primary goal of this experiment is error field correction, but locked mode suppression should also occur as a side benefit.
The experiment is based on the assumption that the phase of a locked mode will be roughly aligned with the error field. The saturated mode thus provides an input to the feedback system that contains information about the phase of the error field. Integral gain is used to ramp up the error correction, and then hold it at a constant value after the locked mode becomes either unlocked or stabilized. If successful, this experiment would yield a simple, automated way of determining the optimum error field correction with a single shot. This would be very favorable for ITER, where operating time will be at a premium. It also has the potential to provide quick verification and improvement of DIII-Dā??s error correction algorithms. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiment will use the I-coils in feedback mode. The C-coils will be used for approximate error correction, and to apply a deliberate ā??error fieldā??.
With I-coil feedback enabled in an ohmic plasma, a locked mode is induced by ramping down the density or by ramping up an n=1 ā??error fieldā?? from the C-coil. Because the goal is error field correction and not active mode stabilization, the time response of the feedback need not be rapid. In fact, a significant integral gain is a key feature. The locked mode provides an input signal to the feedback system that has a constant amplitude and a toroidal phase aligned with the error field. Integral gain will be used, so that the feedback response continuously increases in amplitude. As the net error field decreases, eventually the mode will either disappear or begin to rotate, at which point the integral gain should stop accumulating and hold the current error correction. Optimization of this scheme will include adjustment of the gains, and the possible addition of a fixed toroidal phase shift to the feedback, if needed to compensate for plasma rotation effects. | ||
| Background: | Previous experiments have observed cases where feedback forced a locked tearing mode to rotate (shot 127927, for example), but there are several significant features that distinguish the experiment proposed here. Previous observations typically occurred under conditions of strong neutral beam heating and torque, while the present proposal is for an ohmic plasma with no NBI torque. The most important difference is the use of integral gain in the present proposal. Integral gain should avoid the behavior where the feedback response ā??chasesā?? the mode as it begins to rotate; instead, the proposed scheme should simply hold the present output value as an approximation to error field correction. | ||
| Resource Requirements: | I-coils with SPAs.
C-coils with SPAs or C-supplies | ||
| Diagnostic Requirements: | Magnetic diagnostics, including RWM pairs for locked mode measurements. | ||
| Analysis Requirements: | Simple feedback modeling should be done before the experiment to estimate the initial values of proportional and integral gain. | ||
| Other Requirements: | |||
| Title | 438: Compare edge particle transport in ELMing H-mode and QH-mode with and without NRMF | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure the edge impurity particle transport in ELMing H-mode plasmas and contrast it with the edge particle transport in QH-mode both with and without n=3 nonresonant magnetic fields (NRMF). This will allow us to determine separately determine the net edge loss due to ELMs, EHO and EHO plus NRMF | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create a QH-mode plasma with moderately low toroidal rotation which can be run both with and without NRMF. Inject LiF pellets into QH-mode phases both with and without NRMF. Turn off NRMF and raise density until ELMs return; perform same measurements in ELMing H-mode. These measurements can be used as part of other QH-mode parameter scans to map out particle transport as a function of those parameters. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. A key part of developing QH-mode with NRMF as an operating scenario for future devices is developing a predictive understanding of the edge particle transport. This is the main goal of the task force on ELM-control: 3-D Field Induced Transport. To develop a predictive understanding, we need to be able to measure the edge particle transport. Previous experiments in QH-mode without NRMF have used injection of pellets doped with LiF [K.H. Burrell et al, Phys. Plasmas 12, 056121 (2005)]. The essential features of LiF are 1) lithium and fluorine do not recycle, thus allowing direct measurement of the edge loss rate from the decrease of core density and 2) LiF contains a low and a moderate Z element, allowing determination of the loss rate as a function of Z. | ||
| Resource Requirements: | The lithium pellet injector must be brought back into operation before this experiment can be done. All standard profile and fluctation diagnostics, especially edge BES and ECE-I for EHO studies. | ||
| Diagnostic Requirements: | CER tuned to Li or F lines for impurity transport study | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 439: Magnetic + ECCD and magnetic-only control of "born-locked" modes | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye, M. Lanctot, S. Mao, R. Prater, E.J. Strait, A. Welander | ITPA Joint Experiment : | No |
| Description: | Use Magnetic Perturbations (MPs) in real-time to either cancel the Error Field (EF) that penetrated and caused a Locked-mode (LM), or to bring it in view of the ECCD, which will stabilize it. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat the January-February 2010 LM-control experiment at even lower rotation. This can be obtained with reduced, zero or slightly negative NBI torque. The lower rotation will reduce the rotational shield and promote the EF penetration. If not sufficient, apply a deliberately imperfect EF-correction to operate in presence of a finite EF, or reduce q95 to move the q=2 surface closer to the edge, where it will experience reduced shielding and be more prone to EF penetration. The reduced q95 will also make the experiment more relevant to ITER and disruptions.
Once the born-locked forms, a threshold in the DUSBRADIAL radial field amplitude will be exceeded and cause the "dud detector" to trip and trigger changes in the ECCD and MPs similar to the past experiments. The main difference will be that the MPs will obviously be applied after locking. In the rotating precursor case, the dud detects the rotating precursor approximately 30ms before the actual locking, so that the mode is made lock directly in the desired position. Here, first the mode forms (locked, already, in the "wrong" position), then the MPs are applied. This opens up the choice of whether to apply a static MP or a rotating one, that drags the mode from one position to the other. The two will be compared experimentally, and various rotation velocities will be attempted in the dynamic approach: the rotation needs to be rapid, in order to rapidly start the ECCD stabilization, but not too rapid, otherwise the mode will "slip". Finally, among possible MPs to be applied at the dud trip, we suggest applying a MP equal and opposite to the n=1 EF detected in real-time. This might result in unlocking or stabilization. Most signals (e.g., the LM amplitude) are already available in real-time. Some work might be required to analyze in real time the LM phase from the internal saddle loops, and some logic needs to be implemented in the PCS to translate the desired MP amplitude and phase into the appropriate I-coil currents. | ||
| Background: | In January-February 2010, magnetic + ECCD experiments resulted in the complete stabilization of locked modes (LMs) for the first time. Those experiments concentrated on LMs with rotating precursors. Earlier experiments (2006-08) showed that a slow (0.66Hz) rotating n=1 field can successfully control the LM position, both in the case with and without precursor. At that time, however, the ECCD power was not sufficient for complete stabilization.
In brief, the 2006-08 results suggest good position control and the 2010 results suggest that we have sufficient ECCD power for the success of the proposed experiment. It is important to learn how to stabilize "born-locked" modes at DIII-D, where they represent approximately half of the LM population, and it will be even more important in ITER where, due to slow rotation, the fraction is expected to be even higher. | ||
| Resource Requirements: | 6 gyrotrons | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 440: Pellet in a locked-mode | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Vary from shot to shot the toroidal position of a locked mode and drop D and impurity pellets to evaluate their effect -through resistivity and radiative losses- on the O- and the X-point. Possible mitigation/stabilization. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Either deliberately cause locking with a certain toroidal phase by ramping the error field, or generate a rotating precursor, make it slow down and, slightly before locking, apply the desired magnetic perturbation, as in ECCD control of Locked Modes (LMs). At locking, in lieu of the ECCD, drop pellet. Repeat for various phases. The pellet will cause different effects, depending whether it is injected in the O- or the X-point and whether it is a deuterium or impurity pellet.
Deuterium will increase n_e, drop T_e, increase the resistivity, reduce the local current. If done in the X-point, this would equalize the (higher) local current with the (smaller) current in the O-point, which suffers from a bootstrap (BS) current deficit. Re-establishing the poloidal (and, de facto, toroidal) symmetry of current would suppress the mode. A possible limit is represented by the effect of resistivity on reconnection, which might make the mode worse rather than milder. The other approach is to launch an impurity pellet at the radial location of the locked mode, and cause it to radiatively dissipate its magnetic energy. It is not clear yet -and it will be interesting to determine experimentally- whether the O-point will radiate or "leak" more than the X-point, and whether and how the magnetic topology in the vicinity of the X-point might be altered and possibly ameliorated by radiative losses. Might require some q profile tailoring to adjust size and position of q=2 surface underneath the pellet dropper. Try dropping the pellet both tangentially, for maximum interaction length, and orthogonally, for comparison. | ||
| Background: | There is not much background. To author's knowledge, controlled injection of pellets in locked islands (or in islands, in general), has never been tried. However, there are hints of impurity effects on island stability both at DIII-D and NSTX, e.g. in relation with wall conditioning. | ||
| Resource Requirements: | Pellet, I-coils | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 441: Off-axis NBI H&CD in a locked mode | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Magnetically, toroidally steer an initially locked mode to align the off-axis NBI to its O- or X-point. This will have several effects on the island, via the increased beta (destabilizing), torque (unlocking), rotation shear (stabilizing), localized heating (stabilizing), current density (stabilizing) and fast particles (stabilizing [C. Hegna, PRL 1989]). It is proposed to study whether the resultant of these effects will be stabilizing, destabilizing or unlocking. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Similar to ECH and ECCD experiments on an initially locked 2/1 island slowly dragged in the toroidal direction by a rotating n=1 field exerted by an I-coil traveling wave. However, use NBI from the new off-axis injector as a source of heating and current drive (H&CD). Repeat with static n=1 field, for NBI deposition in the island O-point, in the X-point and in between. For comparison repeat without off-axis NBI, or with a different co/ctr mixture of on-axis NBI. | ||
| Background: | Central NBI deposition is too broad and too central, but the new off-axis NBI has the potential to locally affect the locked mode (LM). Besides, it will have a reduced global effect on beta. For these reasons, it might represent a new tool for LM stabilization. | ||
| Resource Requirements: | I-coils | ||
| Diagnostic Requirements: | FIDA and other fast particle diagnostics? MSE. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 442: Test M3D calculations of EHO structure against measurements in QH-mode | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to compare the structure of the edge harmonic oscillation (EHO) with the structure calculated by the M3D code. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Run low rotation QH-mode plasmas with n=3 nonresonant magnetic perturbations. Use edge ECE-I and BES to investigate the 2D structure of the EHO. Perform these measurements under a variety of conditions as part of other QH-mode experiments. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. A key feature of QH-mode is the edge particle transport induced by the edge harmonic oscillation (EHO). As part of the analysis work under this Task Force, we will be using the M3D code to see if it can calculate the structure of the EHO and the particle transport that it induces. In order to check this calculation, we will investigate the 2 D edge structure of the EHO using edge ECE-I and BES. The BES measurements will give the density perturbation while the ECE-I will give a mix of density and electron temperature perturbation, with the temperature part dominant towards the top of the edge pedestal. | ||
| Resource Requirements: | Reverse Ip. I-coil in n=3 nonresonant configuration. | ||
| Diagnostic Requirements: | ECE-I and BES mandatory. Standard profile and fluctuation diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 443: Off-axis NBI H&CD in a locked mode (dupl. 441) | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Magnetically, toroidally steer an initially locked mode to align the off-axis NBI to its O- or X-point. This will have several effects on the island, via the increased beta (destabilizing), torque (unlocking), rotation shear (stabilizing), localized heating (stabilizing), current density (stabilizing) and fast particles (stabilizing [C. Hegna, PRL 1989]). It is proposed to study whether the resultant of these effects will be stabilizing, destabilizing or unlocking. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Similar to ECH and ECCD experiments on an initially locked 2/1 island slowly dragged in the toroidal direction by a rotating n=1 field exerted by an I-coil travelling wave. However, use NBI from the new off-axis injector as a source of heating and current drive (H&CD). Repeat with static n=1 field, for NBI deposition in the island O-point, in the X-point and in between. For comparison repeat without off-axis NBI, or with a different co/ctr mixture of on-axis NBI. | ||
| Background: | Central NBI deposition is too broad and too central, but the new off-axis NBI has the potential to locally affect the locked mode (LM). Besides, it will have a reduced global effect on beta. For these reasons, it might represent a new tool for LM stabilization. | ||
| Resource Requirements: | I-coils | ||
| Diagnostic Requirements: | FIDA and other fast particle diagnostics? MSE. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 444: Unlocking by NBI Torque and Study of the Locking-Unlocking Hysteresis | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye | ITPA Joint Experiment : | No |
| Description: | At locking, increase the NBI torque. Repeat for fixed NBI torque but higher NBI power, to assess the highest beta at which the mode can be unlocked. Generalize to a 2D scan of NBI torque and power. An economical way to perform such scan would consist in prescribing to the NBI feedback a fixed torque input and decreasing beta in every discharge, then scan the torque from shot-to-shot. Similarly, one can prescribe a fixed beta and increasing torque in every discharge, then scan beta from shot-to-shot. Either way, the mode will unlock for a certain combination of beta and torque. Later in the same discharges, perform the reverse ramps of beta and torque and yet again the original ramps (basically, pre-program triangular waves of beta and torque). This will allow to identify the thresholds for unlocking and relocking. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce 141501-502. Repeat with more or less NBI torque. The mode is expected to unlock sooner (and possibly not relock) or later (or not unlock at all).
Perform a 2D scan of torque and beta as indicated above. Compare the NBI torque at which the mode locks with the minimum torque to apply in order to unlock it. These should be asymmetric (in particular, in absolute terms, more torque should be necessary for unlocking than for locking), due a difference between dynamic and static friction. | ||
| Background: | The ā??dud detectorā?? detects locked modes and their rotating precursors in real time. The PCS is usually instructed to respond by dropping the NBI power and thus beta, to make the locked mode less disruptive. Different real-time changes are proposed here.
This work will take advantage of J. Ferronā??s simultaneous and independent control of beta and torque by acting on the total NBI power and co/ctr mix. This work will also provide useful input to ā??calibrateā?? the torque balance equation solver TORBA under development at UW-Madison. The latter will help to translate mode unlocking requirements (the torque to apply to the mode) into an NBI torque request (the torque to be imparted to the plasma). Mode unlocking by a change of co/ctr mix has already been obtained in shots 128903, 141501 and 141502 during other locked mode control experiments. The scope of this proposal is to assess the pros and the limits of this technique and to ā??automateā?? it, for future dial-in. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 445: Integrated Disruption Control | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Not requested |
| Co-Author(s): | D. Humphreys, E.J. Strait, M. Walker | ITPA Joint Experiment : | No |
| Description: | Develop and test an integrated system for disruption avoidance/mitigation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Technical tests and ā??calibrationsā?? (e.g. of the dud detectors) in 2-hour slots on Thursdays. Physical tests initially as piggybacks on shots at risk of disruptions, then in dedicated experiments where disruptions are generated by diverse methods such as n_e ramp-down, EF ramp-up, beta ramp-up for mode onset and locking, laser blow-off or other impurity seeding. | ||
| Background: | See block diagram at slide 21 of the presentation given by F. Volpe at the Friday Science Meeting on Feb.1, 2008.
A new PCS disruption-specific piece of software will consist of 4 parts: 1) STABILITY BOUNDARIES: monitor (a) q95 from EFIT, (b) n_e from the interferometer, (c) betaN from EFIT. At the same time, compute their stability limits with DCON in real time. If needed and if requested, act on Ip, density control or beams. 2) PRECURSORS: analyze in real-time magnetics, optical and edge diagnostics in order to promptly identify (a) locked mode or its rotating precursor, (b) MARFE, (c) detachment. 3) AVOIDANCE: (a) solve torque balance equation. Calculate whatā??s most efficient among the following: (i) drop NBI to make mode less disruptive, (ii) maximize NBI torque to unlock the mode, (iii) apply static n=1 Magnetic Perturbations (MPs) and continuous ECCD or (iv) rotating MPs and modulated ECCD, for mode suppression. (b) increase NBI ( c ) ? not clear how to respond to detachment 4) MITIGATION: keep monitoring signals of 1) and 2). If 3) has failed, drop NBI and deploy one of the following: MGI, killer pellet, ECH (ā??a la FTU and AUG). | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Real-time DCON | ||
| Other Requirements: | Changes to the PCS | ||
| Title | 446: Extend zero NBI torque QH-mode operation toward ITER baseline objectives (Dup. 341) | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Scenario Development at Low Torque/Rotation for FNSF and ITER | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell | ITPA Joint Experiment : | No |
| Description: | The goals of this experiment are:
- Demonstrate zero-torque QH-mode using NRMFs at q95 ~ 3.0 in ISS discharges. - Develop zero-torque ELM-free path to zero-torque QH-mode state of DIII-D discharge 141439 (3 |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | - Starting from discharge 141439, find NBI torque requirements for QH-mode operation at q95~3. Use Bt and Ip ramps to get to target q95 with minimal perturbations of the early phase of the discharge.
- Reduce NBI torque in early phase, 1 | ||
| Background: | The ELM-stable regime of QH-mode is sustained in DIII-D discharge 141439 for ~1 s with zero NBI torque and at or above the ITER baseline target values of H89 and betan: H89ā?„2 and betanā?„2. Important steps to improve the parameter match of ITER baseline scenario are to:
- change the value of q95 from ~5 to ~3, - extend the zero NBI torque operation to the entire discharge duration, - reduce Zeff from ~4 to ~2. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 447: Extend zero NBI torque QH-mode operation toward ITER baseline objectives (Dup. 341) | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell | ITPA Joint Experiment : | No |
| Description: | The goals of this experiment are:
- Demonstrate zero-torque QH-mode using NRMFs at q95 ~ 3.0 in ISS discharges. - Develop zero-torque ELM-free path to zero-torque QH-mode state of DIII-D discharge 141439 (3 |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | - Starting from discharge 141439, find NBI torque requirements for QH-mode operation at q95~3. Use Bt and Ip ramps to get to target q95 with minimal perturbations of the early phase of the discharge.
- Reduce NBI torque in early phase, 1 | ||
| Background: | The ELM-stable regime of QH-mode is sustained in DIII-D discharge 141439 for ~1 s with zero NBI torque and at or above the ITER baseline target values of H89 and betan: H89ā?„2 and betanā?„2. Important steps to improve the parameter match of ITER baseline scenario are to:
- change the value of q95 from ~5 to ~3, - extend the zero NBI torque operation to the entire discharge duration, - reduce Zeff from ~4 to ~2. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 448: The effect of fueling efficiency on carbon wall retention (split with ROF#102) | ||
| Name: | Unterberg unterbergea@ornl.gov | Affiliation: | ORNL |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux, L. Baylor, T. Jernigan, G. Jackson | ITPA Joint Experiment : | No |
| Description: | Recent particle balance experiments on DIII-D at the natural H-mode density have confirmed that during the startup phase of the discharge has high time-dependent global wall retention. This phase is usually the initial ohmic/L-mode phase. All past experiments have used the gas puff fueling, which has a low fueling efficiency. This experiment will test the possible correlation between high wall retention rate by varying the fueling source between gas puff and pellet fueling. HFS pellet fueling has been shown to be up to 90% efficient. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Discharges with f_GW ~ 0.3 in the H-mode will be generated similar to the discharges used in the most recent particle balance experiments (i.e., quickly divertored, and USN with strong inboard/outboard cryopumping). Prefill gas puffing will remain unchanged throughout the experiment but the fueling in the ohmic/L-mode (i.e., startup) phase will be varied and the effects on the global particle balance will be monitored. Ideally the puffing/pellet fueling ratio would vary from 100/0 to 0/100 in a systematic way.
Specifically, we will be looking for the global particle inventory before H-mode. The goal will be to try and reduce this inventory to minimal levels. | ||
| Background: | Most tokamak global particle balance experiments show high wall retention rates in the startup phase. This is true for graphite and metal walls. This characteristic has been seen on Tore Supra, JET, ASDEX-U, ALCATOR C-MOD, JT-60U --- and earlier DIII-D discharges. | ||
| Resource Requirements: | HFS pellets. 1 co-NBI beamline (2 sources). Cryopuming in the upper divertor. | ||
| Diagnostic Requirements: | ASDEX gauges, filterscopes, Core & tang. thomson. Typical CER. Upper divertor floor langmuir probes. | ||
| Analysis Requirements: | D2BAL. | ||
| Other Requirements: | This proposal is synchronous with ROF# 102 | ||
| Title | 449: H-mode Wall Retention Rates at High Greenwald Fractions (f_GW ~>0.65) | ||
| Name: | Unterberg unterbergea@ornl.gov | Affiliation: | ORNL |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux, L. Baylor, T. Jernigan | ITPA Joint Experiment : | No |
| Description: | Most particle balance experiments at low Greenwald fractions (f_GW ~ < 0.3) and moderate to strong divertor cryopuming show negative retention rates in the steady-state phase of the discharge. This low f_GW is at the natural H-mode density and hence low source fueling rates are needed in this phase. The goal of this experiment is to test retention rates in the H-mode phase of discharges that go to f_GW ~ 0.7 and higher. In this regime, high fueling rates will be needed -- most likely from pellet fueling. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To Be continued.... | ||
| Background: | To Be continued.... | ||
| Resource Requirements: | To Be continued.... | ||
| Diagnostic Requirements: | To Be continued.... | ||
| Analysis Requirements: | D2BAL; SOLPS | ||
| Other Requirements: | -- | ||
| Title | 450: Effect of different 3D coils configuration on mode rigidity and RWM control | ||
| Name: | Baruzzo none | Affiliation: | Consorzio RFX |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | T. Bolzonella (Consorzio RFX), M. Takechi (JAEA), M. Okabayashi (Princeton U) | ITPA Joint Experiment : | No |
| Description: | we propose to study the control of current and pressure driven RWMs using different sets of actuators, i.e. different configurations of active coils and/or amplifiers. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | n=1 RWM could be controlled with a reduced set of coils, exploiting asymmetries in poloidal and toroidal directions in order to reduce the risk of mode ā??slippingā?? and to minimize the amplitude of unwanted sidebands.
The connections of active coils to the available amplifiers could also be changed in order to estimate the impact of higher degrees of freedom opened by the reduced number of coil used. The 3D structure of rotating control sidebands should be characterized in vacuum, and then studied in their radial eigenfunction using thermal measurements. This exercise can be performed using one set of coils (internal-external) to maintain plasma stability, while the other is used to produce error fields with a reconfigured set, in a sort of MHD spectroscopy technique. The experiment is envisaged to begin with current driven RWM study, given the mode drive similarity with RFX, where only current driven RWMs are present. Successive experiments can repeat the same control configuration to High beta RWMs, to check the effect of the different mode drive on mode reaction to reconfigured control. | ||
| Background: | RWM control experiments with different coil sets have been already successfully performed on RFX-mod. The control seemed to be effective even using a small amount of active saddle coils, provided the control was switched on since the mode appearance and had a suitable topology.
Control from a large mode amplitude value implied the production of large unwanted feedback sidebands, which terminated the discharge. Further experiments in RFX-mod working in tokamak configuration are proposed, and could give the opportunity to match the conductive boundary between tokamak and RFP configuration. Control with a single array of non toroidally symmetric coils also highlighted the possibility of the mode to change its phase to grow in the gaps between active coils, ignoring the effect of the active control. These experiments could help the understanding of mode rigidity under the action of external control, and can be important for the design of coils systems in future devices. The experiment could be improved and completed with the possibility to study the same effect on high beta pressure driven RWM in DIIID. The extension of the present experiments to other RWM control equipped tokamaks is also under evaluation. | ||
| Resource Requirements: | 1 day experiment with current driven RWM configuration.
1 days experiment with NBs to exceed no-wall beta limit. CCD for NTM suppression | ||
| Diagnostic Requirements: | SXR and ECE diagnostics can be useful to see the impact of produced sidebands on thermal profiles, and measure the radial eigenfunction of sidebands produced error fields. Coherence method can be used with respect to the field produced by active coils. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 451: Ip scan for ITER steady state scenario development | ||
| Name: | Murakami masanori_murakami@att.com | Affiliation: | Retired |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | J.M. Park, E. Doyle, , T. Luce, J. Ferron, et al. | ITPA Joint Experiment : | Yes |
| Description: | (1) Ip scan for optimization of fNI and fusion performance G(Q)
(2) Good documentation of the "boundary" profiles (Ļ? = 0.8 ā?? 1.0) including pedestal and further inside (no-man's land) |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Reproduce shot like 133103
2. Scan Ip 3. At each step, min-scan with Ī²N and density, and see control-room evaluation of fNI, fBS and G= Ī²NH/q2,etc. 4. Document the best discharge, in particular, the "boundary" profiles (ELM average, ā??breathingā??) | ||
| Background: | ā?¢In 2008, 2 successful ITER SS Demo shots: equiv. Ip=8.5 MA and Ip=13 MA, but without edge CER breathing
ā?¢ITER steady state scenario modeling carried out using scaled edge from DIII-D ITER demo discharge (#134372) ==> Comes short in simultaneously achieving fNI=100% and QDT=5 ==> Used in H&CD mixes discussion (e.g., ITPA IAEA paper) ā?¢IP scan is important for ā??More credible ITER SS scenario development ā??Addressing the H&CD mixes / Upgrade questions ā?¢ Initial Ip scan (with 2-D SS simulations with the scaled DIII-D edge) shows promise at Ip=8MA with Day-1 H&CD=> fNI>100%, Q=3.4 - 4.5, fBS = 0.6, but lack ICD=1 - 2MA for Ip = 9 MA ā?¢ EPED prediction ~25% below bN(r); Upper limit of experimental uncertainty ā?¢ So we seek for : (1) Ip scan for optimization of fNI and fusion performance G; (2) reliable boundary (r = 0.8 ā?? 1.0) profiles | ||
| Resource Requirements: | Use NBCD like ITER (intermediate + on-axis) NBI: >5 co sources + 210RT for balanced MSE
EC: >4 gyros FW: (90 MHz and 60 MHz) >2 MW desirable | ||
| Diagnostic Requirements: | Fast ion diagnostic (UC, Irvine), MSE (LLNL), edge reflectometer (UCLA) | ||
| Analysis Requirements: | Scenario modeling with GLF23 FASTRAN//ONETWO, TRANSP/GLF23, ONETWO analysis; TORAT/CQL3D, CURRAY
Other | ||
| Other Requirements: | Multiple days | ||
| Title | 452: Simultaneous measure of midplane and X-point neutral profiles | ||
| Name: | Unterberg unterbergea@ornl.gov | Affiliation: | ORNL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | RJ Groebner (GA) , JM Canik (ORNL), LW Owen (ORNL), J Lore (ORNL) | ITPA Joint Experiment : | No |
| Description: | TBD | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | TBD | ||
| Background: | TBD | ||
| Resource Requirements: | TBD | ||
| Diagnostic Requirements: | TBD | ||
| Analysis Requirements: | TBD | ||
| Other Requirements: | |||
| Title | 453: High-Z gas mitigation of mature REs | ||
| Name: | Wesley wesley@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | PoP experiments with neon gas injection in 2010 demonstrate efficacy of 'late' high-Z gas injection in rapidly dissipating 'mature' avalanche-equilibriated plateau-phase runaway electron current (number). Comparison of the limited data with similar D2 and He gas injection strongly suggests that radiative collisional dissipation at high RE energies (~10 MeV) is the mechanism. Open issues include comparative effect of higher-Z species (eg Ar, Kr and Xe), assimilation and retention of injected gas in the RE channel and relation of current/number loss rates to gas species, injected quantity and injection time relative to development of the RE plateau phase; also gas ionization/retention vs. initial RE current magnitude and subsequent configuration evolution and 'control'. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use 'standard' argon-pellet RE generation method + OH drive to produce quasi-stationary 100-400 kA RE plateau discharges; inject candidate gases in various quantities or slightly before plateau onset plateau onset. Compile database of current/number decay rate vs. species, current magnitude, configuration (position, elongation) and surface voltage/E-field. | ||
| Background: | ITER needs an ex-post-facto RE mitigation scheme. Foreseen limitations on 'natural' RE duration suggest that 'strong' high-Z gas injection provides the best option for in-situ energy/momentum dissipation (rather than direct loss to the wall). | ||
| Resource Requirements: | Standard low-elongation EC_heated RE target (4 Gyrop), Ar cryopellet injector, Medusa gas injector for neon, argon, krypton, xenon (ā?¤ 1000 torr-liters per pulse). 10-20 RE plateau 'shots' per gas to allow for quantity and/or timing variations. Ca 40 shots = 2 run days. | ||
| Diagnostic Requirements: | Standard magnetics + RTEFIT, CO2 + late TS (feasibility?); fast camera monitoring of synch emission; BGO and non-saturated fplastic. Spectro, etc TBD (impurity and background content) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 454: Short-pulse high-flow MGI by IRD | ||
| Name: | Wesley wesley@fusion.gat.com | Affiliation: | GA |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | PoP test of short-pulse MGI using one-shot IRD injector . IRD = inverse rupture disk injector aka Parks tube aka Ludweigsrohr. Short pulse = ā?¤ 1 ms main pulse duration. High-flow (Q) = > 3000 Torr-l/ms. Into high-Wth beam-heated ā?„ 1MJ target. Test gases in my priority order = D2, He (NBI threat!) and Ne. Main objectives: validate injector method and [lack of] deleterious effect(s); look for enhance initial assimilation/retention (compared to M-I, M-II and SPI data) | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Standard 1.5-MA LSN high-Wth 'MGI' target; IRD injection at > 2000 ms (likely with asynch timing). | ||
| Background: | MEDUSA-I vs. MEDUSA-II results 2008-10 with low-Z gases and ASDEX-U in-vessel valve results 2008-2010 show added density/assimilation gain for short pulse, close-coupled MGI. A 'standard' 0.5-m IRD could provide a 1 ms duration D2 or He pulse for ca 3000 Torr-liter quantity. A 'short' 0.25-m IRD could provide ditto for Ne (need larger diameter). D2 is the most benign gas re NBI after effects and directly comparable to D2 SPI. Neon is ITER favorite and may be comparable to Ne SPI. He comparison is only to past MEDUSA-I + -II or new (2011) MEDUSA-II | ||
| Resource Requirements: | IRD cartridge(s) + insertion/replacement means/schedule (likely clean vent); standard 4-6 beam 1.5 MA LSN target; D2, He and/or neon fill + ops capability. | ||
| Diagnostic Requirements: | Standard high-ne MGI diagnostics (note asynch trigger). | ||
| Analysis Requirements: | |||
| Other Requirements: | Availability of an electrical discharge trigger IRD would provide synchronous timing for diagnostics | ||
| Title | 455: Disruption Avoidance Demo | ||
| Name: | Wesley wesley@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Do a simple disruption prediction and avoidance demo using 'real-time control. Identify one or more long-lead predictors of pending disruption onset (TM, NTM, LM or ....); apply ECH or ECCD + possibly rotating/phase controlled RMP or NBI reduction to forestall what would otherwise end up as a disruption. Possibly take some kind of further active 'repair' or soft-landing action(s). If promising, test in combination with a 'real-world' disruption-provoking run day. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Develop prediction strategies and RT control off-line and/or in background; 0.5-1 runday tests of intervention and response method(s); possible further deployment test in conjunction with other experiments. | ||
| Background: | Extension of 'dud' detector, but with 'active' + intelligent intervention. Test case of real-time predict/act/recover sequence via PCS or Finite State Machine (FSM) | ||
| Resource Requirements: | Prediction algorithms, PCS and/or FSM; 'on-demand' ECH or ECCD and/or phase-controlled RMP | ||
| Diagnostic Requirements: | Support prediction | ||
| Analysis Requirements: | Predictor development | ||
| Other Requirements: | Comment: may be beneficial to do something simple/easy [physics-wise] and focus on RT implementation and deployment demo test. Alternate would be more physics/methods focus and less RT control + demo. Rhetorical question: what's the best way to show that this type of approach is benefical/helpful in avoiding 'unnecessary' disruptions. Need to discuss. | ||
| Title | 456: Extend fNI, stability & transport vs. q-profile scans to qmin>2 using OANBI | ||
| Name: | Holcomb holcomb@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Off-axis neutral beam injection should improve our ability to sustain elevated qmin. Extend the q-profile scans done in 2009 to qmin>2 and evaluate the transport, bootstrap current, noninductive current and stability limits at these higher values of qmin. Compare to 2009 discharges at lower qmin and see how the new data matches the trends identified by previous analysis. Off-axis beams may help to broaden the pressure profile and raise the ideal-wall betaN limit, so this will continue to test the trade-off between pressure peaking factor & high betaN vs. maximizing bootstrap current. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | First scan should be at betaN=2.8, BT=2 Tesla. Make 3 discharges with qmin>2 for ~1 sec at q95=4.5, 5.5, 6.8. In a subsequent scan, push each of these regimes to the maximum achievable betaN. The 2009 scans were done in normal BT. For a direct comparison, BT should be the normal direction in these experiments as well, but this may not be required (or preferred) if we can achieved similar pumping and density profiles in reverse BT. (In normal BT the off-axis beams will not drive as much off-axis current, but at least the on-axis CD that contributes to the q0 evolution will be reduced). | ||
| Background: | In 2009 we scanned the q-profile by adjusting q95 and qmin to identify the dependence of transport, bootstrap current, and noninductive current fraction on the q-profile. This was done both at betaN=2.8, and the maximum injected power. After thorough kinetic efit analysis, is appears that the maximum average qmin for the betaN=2.8 shots was limited to less than or equal to 2, and only a single ļæ½??maximum powerļæ½?ļæ½ discharge was produced with qmin~1.75 at q95~6.8, with no high qmin discharges at lower q95. The achieved betaN & calculated ideal-wall betaN limits were also relatively low for the high power qmin~1.75 shot. This may have been due to pressure peaking and/or insufficient off-axis current density. Two papers have been written so far on the analysis of these results, so new data at qmin>2 would provide additional checks on the trends and scalings identified in these papers. | ||
| Resource Requirements: | Off-axis neutral beams, ~3.5 MW EC | ||
| Diagnostic Requirements: | All profile diagnostics | ||
| Analysis Requirements: | kinetic efits, ONETWO analysis | ||
| Other Requirements: | -- | ||
| Title | 457: Reduce fast ion loss from OANBI in q95~6.3, qmin~1.5, fNI~1, betaN~3.8 scenario by increasing B & I | ||
| Name: | Holcomb holcomb@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The scenario with fNI~1 and almost zero inductive current density everywhere has betaN~3.8, qmin~1.5, q95~6.3, and BT=1.75 T. These discharges (133103, 140695) are very close to the predicted n=1 ideal wall limit, and they were produced before the 30rt neutral beam became available. Ideally reproducing these discharges but with 5 MW of off-axis NBI will drive additional current off-axis where it is needed to bring j_inductive = 0 everywhere. However, ONETWO analysis of 133103 suggested 1 m2/s of anomalous fast ion diffusion. Other BT scans in 2010 AT discharges suggest more fast ion diffusion at lower BT as well. If this is true then the off-axis NBCD may be low. If q95 is held fixed, but Ip and BT are increased, then anomalous fast ion loss may be decreased and the off-axis NBCD maximized. At higher BT the beam power required to maintain betaN~3.8 will go up (we have additional power from 30rt). This will also cause the on-axis NBCD to increase, which may be undesirable for maintaining higher qmin. In this case, we may be able to adjust the amount of shear reversal inside of rhoqmin to encourage greater fast ion diffusion near the axis without affecting it near mid-radius. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | off-axis beams | ||
| Diagnostic Requirements: | All profile diagnostics, Energetic particle diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 458: Test betaN limit at qmin>2, rhoqmin>0.5 using OANBI, more ECCD, density control, and B & I ramps | ||
| Name: | Holcomb holcomb@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Previous work on DIII-D (Garofalo POP 2006) used BT and Ip ramps to make non-steady discharges with qmin>2 and more current density near rho~0.7-0.9 than we are able to drive using ECCD or NBCD. These had rhoqmin~0.5, internal transport barriers, and betaN~4. Stability analysis and modeling suggests the ideal wall betaN limits increase with qmin and the high off-axis current density driven near rho~0.7-0.9 by the BT ramp. We can push further in this direction to test the stability modeling predictions by applying the new off-axis NBā??s to similar BT- & Ip-ramp discharges. Since the initial experiments, we also now have more ECCD power that can be applied off-axis and a new divertor geometry that will improve density control and therefore current drive. ITBā??s may be allowed in these discharges if they do not cause too high a pressure peaking factor, but they may not be essential and the focus would be on broadening the current profile beyond what has been done before. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | Off-axis beams, good RWM stabilization and error field control | ||
| Diagnostic Requirements: | All profile diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 459: Ip scan for ITER steady state scenario development | ||
| Name: | Murakami masanori_murakami@att.com | Affiliation: | Retired |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Not requested |
| Co-Author(s): | J.M. Park, E. Doyle, , T. Luce, J. Ferron, T. Osborne, et al | ITPA Joint Experiment : | Yes |
| Description: | (1) Ip scan for optimization of fNI and fusion performance G(Q)
(2) Good documentation of the "boundary" profiles (Ļ? = 0.8 ā?? 1.0) including pedestal and further inside (no-man's land) |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Reproduce shot like 133103
2. Scan Ip 3. At each step, min-scan with Ī²N and density, and see control-room evaluation of fNI, fBS and G= Ī²NH/q2,etc. 4. Document the best discharge, in particular, the "boundary" profiles (ELM average, ā??breathingā??) | ||
| Background: | ā?¢In 2008, 2 successful ITER SS Demo shots: equiv. Ip=8.5 MA and Ip=13 MA, but without edge CER breathing
ā?¢ITER steady state scenario modeling carried out using scaled edge from DIII-D ITER demo discharge (#134372) ==> Comes short in simultaneously achieving fNI=100% and QDT=5 ==> Used in H&CD mixes discussion (e.g., ITPA IAEA paper) ā?¢IP scan is important for ā??More credible ITER SS scenario development ā??Addressing the H&CD mixes / Upgrade questions ā?¢ Initial Ip scan (with 2-D SS simulations with the scaled DIII-D edge) shows promise at Ip=8MA with Day-1 H&CD=> fNI>100%, Q=3.4 - 4.5, fBS = 0.6, but lack ICD=1 - 2MA for Ip = 9 MA ā?¢ EPED prediction ~25% below bN(r); Upper limit of experimental uncertainty ā?¢ So we seek for : (1) Ip scan for optimization of fNI and fusion performance G; (2) reliable boundary (r = 0.8 ā?? 1.0) profiles | ||
| Resource Requirements: | Use NBCD like ITER (intermediate + on-axis) NBI: >5 co sources + 210RT for balanced MSE
EC: >4 gyros FW: (90 MHz and 60 MHz) >2 MW desirable | ||
| Diagnostic Requirements: | Fast ion diagnostic (UC, Irvine), MSE (LLNL), edge reflectometer (UCLA) | ||
| Analysis Requirements: | Scenario modeling with GLF23 FASTRAN//ONETWO, TRANSP/GLF23, ONETWO analysis; TORAT/CQL3D, CURRAY | ||
| Other Requirements: | Multiple days | ||
| Title | 460: High beta Operation with Broad Current Profile | ||
| Name: | Wade wademr1@ornl.gov | Affiliation: | ORNL |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo | ITPA Joint Experiment : | No |
| Description: | Utilize off-axis NBI and Bt ramp to achieve extremely broad current profile and achieve high beta operation | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Couple off-axis NBI with BT ramp discharges used in previous experiments to achieve broad current profiles at betaN ~ 4. | ||
| Background: | |||
| Resource Requirements: | Off-axis NBI | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 461: D stability characterization for Off-axis NBI in Steady State type discharges | ||
| Name: | Murakami masanori_murakami@att.com | Affiliation: | Retired |
| Research Area: | Fully Non-inductive Scenarios with Off-Axis Neutral Beam Injection | Presentation time: | Not requested |
| Co-Author(s): | JM Park, T. Luce, J. Ferron, C. Petty, W. Heidbrink,, M. Van Zeeland, T. Suzuki, et al. | ITPA Joint Experiment : | No |
| Description: | ā?¢ MHD stabilities for off-axis NBI for variations of +/- BT direction; NBI steering variation
ā?¢This will help for steady state experiments and also ITER planning ā?¢ It is important to characterize MHD stability in discharged reasonably close to fully non-inductive plasmas at early stage to find problem if exist. ā?¢ These have important implication to ITER planning |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | ā?¢Experiments (3 days(?) with shared other experiments) for:
1)off-axis LT and RT beam 2)2) Variation of BT direction 3) Steering angle scan 4) m/n=1/1; 3/2; 2/1 (shared perhaps 1/1 and 3/2 with Fusion science: and 3/2 and 2/1 with SSI Group) 5) Various flatness of q(r) | ||
| Background: | ā?¢Fast-ion driven instability, such as fish-bone instability is predicted to be unstable when toroidal wave velocity is resonant with the toridal drift experienced by trapped beam ions. Since the requirement for a particle to be trapped is the velocity sufficiently perpendicular to magnetic field, therefore this instability would not occur for injection paralllel to the magnetic field. The sensitivity to the magnetic pitch is similar to off-axis NBCD.
ā?¢In the DIII-D NB steering configuration, reverse BT has much better CD efficiency than normal BT. So typical NBCD experiment could lead to MHD-quite operation. ā?¢ITER has the wrong BT direction for the off-axis NB steering, so it is a problem if off-axis NBI configuration prunes to fishbone instability. ā?¢Certain results of off-axis NBI experiments in other devices could be confused by this MHD instability effect ā?¢There may be some data available in: ā??MAST ā??ASDEX-U (M. Van Zeeland) ā??PBX | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | MSE; magnetics;
FIDA / neutrons; neutral particle analyzer, Loss detector; soft x-ray; | ||
| Analysis Requirements: | NUBEAM; TRANSP; MHD codes | ||
| Other Requirements: | |||
| Title | 462: Suppression of runaway electrons with dust injection | ||
| Name: | Smirnov none | Affiliation: | UCSD |
| Research Area: | Runaway Electron Dissipation and Control for ITER | Presentation time: | Not requested |
| Co-Author(s): | E.M. Hollmann, S.I. Krasheninnikov, D.L. Rudakov, T. Jernigan | ITPA Joint Experiment : | No |
| Description: | Test injection of carbon dust of micron size during the disruption CQ phase as a method for suppression of RE. The estimates show that injection of ~1g of carbon dust in DIII-D may provide number of bound electrons sufficient to reach the Rosenbluth density. Also, the dust, when injected in the relatively cold plasma after TQ, may provide better material assimilation then MGI due to large dust inertia. Survival time of the dust in such plasma is estimated to be ~100ms. However, dust delivery time should be in a few millisecond range to provide adequate RE control during the CQ that requires dust injection speeds ~100m/s. To achieve the desired speed and quantity of dust, a pressurized rupture disk tube is proposed for the dust injection with the dust placed inside the tube close to the disk. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Initiate plasma shutdown with argon pellet (discharge parameters TBD). Fire the dust using the rupture disk tube within a few ms into the CQ phase to suppress RE generation. Alternatively, dust can be injected into RE beam during the CQ plateau phase to assess dust ability to dissipate already generated RE. | ||
| Background: | Preventing of generation or suppression of RE during disruption is an important issue for ITER. Recent DIII-D experiments on disruption mitigation with MGI demonstrate that gas assimilation efficiency in the central plasma is very low during the CQ phase, when RE can be effectively generated via the avalanche amplification mechanism in ITER. Controlled injection of solid dust in the CQ phase can be one possibility to achieve sufficient assimilation of injected material for complete collisional RE suppression. | ||
| Resource Requirements: | 1-2 dedicated or time-shared discharges, rupture disk tube | ||
| Diagnostic Requirements: | fast camera, SPRED, SXR, interferometers | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 463: High Beta, Steady State Hybrids (Dup 333) | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This experiment will integrate a high beta hybrid plasma with the reactor relevance of Te~Ti and full noninductive current drive. In 2011, the addition of a sixth gyrotron and optimization of the six co-beams will allow us to eliminate the residual 9 mV loop voltage of our best previous case, and hopefully lower q_95 from 5.85 to 5.0 at the same B_T. Additionally, the higher heating power should allow us to increase beta closer to the ideal wall limit, which is around beta_N=4.
This experiment will demonstrate that H-mode (hybrid) discharges with q_min~1 are capable of high beta (beta_N~4) operation with >50% bootstrap current fraction. The remaining noninductive current will will be supplied by on-axis sources at high efficiency. The poloidal magnetic flux pumping that is self-generated in hybrid will suppress the sawteeth despite the strong on-axis current drive, which is important for avoiding the 2/1 mode. The higher efficiency for on-axis current drive will offset the modest bootstrap current fraction such that this scenario will satisfy the requirements for FDF as well as (or better than) the high q_min scenario with strong off-axis current drive. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Start by repeating shot 133881. (2) Inject all six gyrotrons with central current drive. For the six co-NBI sources, increase the injection voltages as much as possible while maintaining a plasma pulse length of at least 5 seconds. (3) Optimize the dynamic error correction (may use broadband feedback), adjust the plasma shape for optimal pumping. (4) Attempt to increase beta_N using the full heating power. If plasma current is overdriven (i.e. negative loop voltage), then increase plasma current to compensate. | ||
| Background: | The current proposal for FDF envisions a high q_min advanced tokamak scenario with 70% bootstrap current fraction. While this is compatible with the US view of DEMO, the physics of the high q_min AT scenario is still being developed. There is also an issue regarding the high off-axis current drive efficiency needed for FDF in this proposal.
Here I propose that the low q_min hybrid scenario is compatible with the requirements of FDF, and it has several advantages. First, the physics basis is well advanced. Long duration hybrid discharge with high beta and high confinement are routinely achieved. Second, because q_min=1 in the hybrid scenario, all of the external current drive can be deposited near the plasma center where the current drive efficiency is the highest (because of the lack of trapped particles and the high electron temperature). While the bootstrap current fraction will be lower in this low q_min hybrid scenario (50% rather than 70%), the increase in the current drive efficiency for central deposition more than makes up for this. Experiments on DIII-D have come very close to demonstrating this scenario using five co-beams and five gyrotrons. Hybrid plasmas with beta_N=3.4 were stably produced with a loop voltage of 9 mV. The loop voltage was a strongly decreasing function of heating power. While the ion and electron temperature were nearly the same outside of rho=0.2, the H-mode confinement factor remained high, H_98=1.4. This result is better than for the typical hybrid regime on DIII-D and is correlated with better than usual electron thermal transport in this LSN plasma shape. Therefore, this proposal will likely lead to the development of a high beta, high confinement, steady state scenario based on the hybrid regime. A half-day experiment in 2010 did not result in improved parameters despite the additional of a sixth co beam source because of 2/1 NTM issues. My hypothesis is that the 2.1 mode onset in hybrids, at least for cases well below the ideal wall limit, is related to having a too peaked pressure profile. This could explain several facets of the 2/1 mode onset, such as the dependence on the current evolution and the dependence on the confinement factor. We will need to pay close attention to the peakness of the pressure profile and find ways to decrease it if necessary, such as using the off-axis beam, changing the wall conditions or gas pre-fill levels. | ||
| Resource Requirements: | NBI: 6 co sources are needed. 210RT may be used to collect MSE data.
ECH: 6 gyrotrons with 4 MW of injected power. FW: It is desirable to couple 1 MW or more, but core absorption needs to be demonstrated. I-coils: Dynamic error field correction will be used (possibly broadband feedback). | ||
| Diagnostic Requirements: | MSE is critical. | ||
| Analysis Requirements: | TRANSP for current drive and transport, DCON for stability. | ||
| Other Requirements: | |||
| Title | 464: High Beta Hybrids and Pressure Profile Broadening (Dup 365) | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Core Integration (Other Advanced Inductive) | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use 5 MW of off-axis beam injection to broaden the total pressure profile compared to on-axis injection. Most of this will be due to a change in the fast ion pressure profile, but some broadening of the thermal pressure profile may also occur depending upon how stiff the transport dependence is. Determine whether the broader pressure profile allows a high beta_N to be obtained in steady-state hybrid plasmas, with the goal being beta_N=4. Will also calculate whether the ideal wall limit changes significantly with the broader pressure profile for these low q_min discharges. The off-axis beam will probably not effect the current drive profile much since the off-axis NBCD efficiency remains high, and the poloidal magnetic flux pumping inherent in hybrids tends to keep the total current profile constant regardless of the driven current profile. Since the noninductive current drive is not crucial to this experiment, it could be done with either positive or negative B_T values.
For steady-state considerations, the co-ECCD should be deposited inside the q=1.5 surface for this experiment. However, we could broaden the scope of this experiment by re-directing some of the ECH power (probably 4 gyrotrons) to deposit at the q=2 surface to see if we can suppress the 2/1 mode. This would be deem a success if the beta limit comes from a RWM rather than a 2/1 mode (the latter is the current situation). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Ideally we would already have in hand a high beta, steady-state hybrid case with 6 co-/on-axis beams that would serve as a fiducial. The beta_N will likely be 3.4-3.5 given previous results. (2) Repeat the fiducial case, but using the 150 beamline tilting fully downwards. (3) Use the 150 beams at full power, but scan the NBI power for the other co-beams to vary beta_N. Determine the limit for the 2/1 mode, the goal being beta_N=4. (4) If time permits, compare the stability limit for cases where all the co-ECCD is deposited inside the q=1.5 surface, and where a minimum of 4 gyrotrons are aimed at the q=2 surface. | ||
| Background: | High beta hybrids have been operated stably (to the 2/1 mode) up to beta_N=3.8 at high density, which is well above the ideal no-wall limit. At lower densities and with central co-ECCD, high beta hybrid plasmas have been created with beta_N=3.4 and nearly zero loop voltage (9 mV). TRANSP calculations show that these plasmas should be very close to fully noninductive. The ideal wall stability limit is calculated to be around beta_N=4 by DCON. The near term goal of high beta hybrid research is to obtain beta_N=4 with zero loop voltage for as long as the beams will run. To give some overhead between the beta_N=4 goal and the ideal wall limit, some broadening of the pressure profile may be desirable. This can be achieved using the off-axis beam. Since the NBCD efficiency remains high even for off-axis injection (especially with positive B_T), we do not have to give up on the steady-state goal to do this experiment. | ||
| Resource Requirements: | NBI: Tilted 150 beamline is critical. All 6 co-beams are needed.
ECH: 6 gyrotrons required. | ||
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| Title | 465: ELM mitigation by magnetic-surface-preserving RMPs at q_max | ||
| Name: | Zheng none | Affiliation: | IFS/EURATOM-CIEMAT |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Requested |
| Co-Author(s): | M. Kotschenreuther, P. J. Morrison, S. Mahajan, and P. Valanju (IFS), and E. R. Solano (EURATOM-CIEMAT ) | ITPA Joint Experiment : | No |
| Description: | Conventional resonance magnetic perturbation (RMP) experiments aim to generate a chaotic field line structure at the plasma edge in order to release plasma energy to avoid ELMs. However, magnetic islands and chaotic field line structures can, considerably, downgrade H-mode confinement. Here, we propose an RMP ELM mitigation approach, referred to as magnetic-surface-preserving RMPs (MSP-RMPs), that may be much less damaging to the magnetic surface structure, thereby allowing ELM mitigation without seriously downgrading H-mode confinement.
In the H-mode pedestal region, the safety factor is non-monotonic. The bootstrap current induces a q_max at the pedestal top and q_min closer to the edge. To achieve MSP-RMPs, one just needs to tailor the RMP helicity to resonate at q_max. For a rotating plasma, the pitch matching is arranged with respect to the rotating frame at q_max. Since magnetic shear vanishes at q_max, the total field line reconnection (i.e., island formation) is minimized. On the other hand, the minimum shear stabilization pertinent at extremal q, tends to favor the excitation of infernal modes (of kink type). In our preliminary studies, the excitation of infernal modes is verified by our AEGIS/DCON numerical computations. Note that q_max defines the position of the transport barrier [see P. J. Morrison et al., Scholarpedia, 4 (9), 3551 (2009); also Fig. 14, L. J. Zheng et al, Phys. Plasmas 17, 052508 (2010)]. Infernal modes, therefore, can allow the release of transport barrier energy in a controlled manner. Here, we note that the kink type of perturbations is much less harmful than the tearing type. The latter can reconnect the pedestal with the scrape-off-layer and, likely, lead to ELMs through a possible positive feedback process [L. J. Zheng, et al, PRL 100, 115001 (2008)]. Acting via the excitation of infernal modes at q_max, ELM mitigation employing the MSP-RMPs technique, may be more benign to confinement than the standard RMP. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | measure q_max and local rotation, and apply RFP accordingly. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | We have linear stability code: AEGIS_R | ||
| Other Requirements: | |||
| Title | 466: Quiescent H-modes with an externally driven EHO (duplicate of #166 in TJA) | ||
| Name: | Reimerdes reimerdes@fusion.gat.com | Affiliation: | CRPP-EPFL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | K.H. Burrell, A. Fasoli, A.M. Garofalo, J.M. Hanson, M.J. Lanctot, P.B. Snyder, W.M. Solomon, D. Testa | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to extend the parameter regime of quiescent H-mode discharges by driving a perturbation similar to the EHO in discharges where the edge transport usually results in ELMs. Even if an MHD mode is stable, it can be driven to a finite amplitude by applying a suitable non-axisymmetric magnetic perturbation with external coils (antennas) [A. Fasoli, et al., Phys. Rev. Lett. 75, 645 (1995), A.M. Garofalo, et al., Phys. Plasmas 10, 4776 (2003)]. Since the EHO has a dominant low n structure and rotates in the 5-10kHz frequency range, the DIII-D I-coil could apply a suitable external field. Driving a stable perturbation has the advantage that the perturbation amplitude can be controlled by the amplitude of the driving field and, therefore, does not only dependent on plasma parameters. In addition any resonant magnetic braking torque generated by the external field would pull the plasma towards the rotation frequency of the external field rather than zero, thereby avoiding the locking in the case the external field is too large. While an extended parameter regime could possibly result in an attractive ELM suppression technique for ITER, the response to the external field also yields information about the stability of the EHO. An external control of the EHO would also enable studies of the transport enhancement as a function of the mode amplitude at otherwise similar plasma parameters. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In this experiment the I-coil is used to apply a rotating n=1 field with a kink mode helicity (240-300Deg I-coil phasing with the exact phasing to be determined by modeling) at 5-10kHz. In order to access this frequency range and maximize the external field amplitude the I-coil will be connected in toroidally opposed anti-symmetrical pairs and powered by parallel audio-amplifier pairs. It is estimated that this configuration will result in approximately 200A of current in the I-coil. Possible target plasmas are:
1) Counter-rotating H-modes that are close to the QH-mode operating regime. 2) ELM-free H-modes, where a small enhancement of the particle transport should have a large effect on the density evolution. 3) Standard ELMing H-mode. Since the EHO stability is strongly affected by plasma rotation, the experiment includes frequency sweeps in order to find a frequency, where the external field couples best to an edge mode. Measurements will include magnetic measurements of the plasma response and measurements of the density and temperature profiles. Of interest are the modification of the time averaged kinetic profiles, which indicates transport changes, as well as the component that oscillates at the applied frequency, which indicates the perturbation structure. Since the rotation period of the external field is short compared to confinement time scales the oscillating component can be interpreted as a displacement of flux surfaces. | ||
| Background: | Quiescent H-modes, i.e ELM-free discharge at constant density and radiated power, but with improved energy confinement given by an edge pedestal, have been observed in various machines, such as the QH-mode in DIII-D [C.M. Greenfield, et al, Phys. Rev. Lett. 86, 4544 (2001)] and the Enhanced D-alpha H-modes in C-Mod [Y. Takese, et al., Phys Plasmas 4, 1647 (1997)]. In these operating regimes MHD fluctuations, namely the edge harmonic oscillation (EHO) in DIII-D and the quasi-coherent mode (QCM) in C-mod, are thought to be responsible for an enhanced particle transport that avoids the onset of ELMs. In DIII-D the EHO is observed in a limited parameter regime. The discharges typically exhibit a relatively low pedestal density and high pedestal temperature as well as a large edge rotation shear [K.H. Burrell, et al., Phys. Rev. Lett. 102, 155003 (2009), A.M. Garofalo, et al., 23rd IAEA FEC, EXS/1-2]. It is thought that the EHO is a low n peeling mode that is driven unstable by rotation shear at edge conditions slightly below the ELM stability limit and which saturates due to a change of rotation shear at finite mode amplitude [P.B. Snyder, et al., Nucl. Fusion 47, 961 (2007)]. This interpretation is consistent with the observed strong dependence of the EHO amplitude and the ensuing transport enhancement on plasma rotation. Limits of the QH-mode operating regime are encountered when the transport enhancement is too weak and cannot avoid the onset of ELMs or when the EHO amplitude is too large and causes locking. The operating regime could be greatly extended, if the drive of the mode by rotation shear could be replaced with an external field and the resulting transport controlled by the amplitude of the external field. First attempts to drive MHD instabilities in order to control transport have been carried out on JET [D. Testa, et al., 28th EPS conference on Controlled Fusion and Plasma Physics (2001)], where internal saddle coils were used to drive global Alfven waves in the range from 30-70kHz. | ||
| Resource Requirements: | I-coil on audio-amplifiers (preferably I-coils connected as odd pairs) | ||
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| Title | 467: Main Chamber Recycling Effects On Pedestal n_e, T_e Profile Offset And Toroidal Rotation Profile | ||
| Name: | Callen jdcallen@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | R. Groebner, A. Leonard, T. Osborne, E. Unterberg | ITPA Joint Experiment : | No |
| Description: | As described in the Background section below, a new paleoclassical-based model (UW-CPTC 10-6, August 30, 2010) predicts that edge neutral recycling has two main effects on the pedestal structure. Specifically, the model predicts that an increase in the neutral recycling source in the pedestal should simultaneously: 1) increase the outward "offset" of the n_e profile relative to the T_e profile and 2) reduce the plasma toroidal rotation at the top of the pedestal relative to its value at separatrix.
The decreased neutral penetration depth in higher electron density DIII-D H-mode plasmas has been identified as a significant factor in reducing the pedestal density profile width -- in M.A. Mahdavi et al., Nucl. Fusion 42, 52 (2002); R.J. Groebner et al., Plasma Phys. Control. Fusion 44, A265 (2002); M.A. Mahdavi et al., Phys. Plasmas 10, 3984 (2003). Also, an outward shift or "offset" of the n_e profile relative to the T_e profile has been observed in the high NBI power (and high toroidal field, low rho_*, high n_e) DIII-D data in the JET/DIII-D comparison experiments -- in M.N.A. Beurkens, T.H. Osborne et al., Plasma Phys. Control. Fusion 51, 124051 (2009). With regard to toroidal plasma rotation in the pedestal, it has been found that it decreases approximately linearly with distance in from its value on the separatrix -- in DIII-D ECH H-mode plasmas in J.S. deGrassie et al., Phys. Plasmas 14, 056115 (2007) and in high density AUG pedestals in T. Putterich et al., "Evidence for Strong Inversed Shear of Toroidal Rotation at the Edge-Transport Barrier in the ASDEX Upgrade," Phys. Rev. Lett. 102, 025001 (2009). This proposal seeks to determine if these neutral recycling effects on the n_e, T_e profile offset and toroidal rotation profile are as predicted by the new pedestal structure model. The main hypothesis to be tested is if main chamber recycling neutrals penetrate more easily into the narrower (in physical space) mid-plane pedestal compared to the X-point region where the flux expansion makes the pedestal thicker in physical space. Thus, the proposal is to systematically vary (increase) the degree of main chamber recycling and observe the effects on the n_e, T_e profile offset and the toroidal flow profile in the pedestal. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Methods to vary (increase) main chamber recycling relative to the fueling from the divertor X-point region need to be identified and explored. Some possibilities include: 1) variation from divertor pumping to strong gas puffing, 2) decreasing the gap in some configuration to induce more main chamber fueling, 3) a scan from LSN to DND coupled with variations in dRsep and 4) contrast low and high triangularity discharges to explore shape effects. After one or more of these procedures has been shown to have significant effects on the n_e, T_e profile offset and/or the pedestal toroidal flow profile, a systematic scan of the degree of main chamber recycling relative to divertor fueling should be performed. | ||
| Background: | Recently, predictions have been developed for the structure of plasma parameter profiles of transport quasi-equilibrium H-mode pedestals -- in "A Model Of Pedestal Structure," report UW-CPTC 10-6, which is available via http://www.cptc.wisc.edu. The predictions are based on assuming paleoclassical radial plasma transport processes dominate throughout the pedestal. Model predictions have been given for the profiles and magnitudes of the electron density and temperature, and plasma toroidal rotation in the pedestal. All the predictions have been shown to agree quantitatively (within a factor of about two) with properties of the recently studied low density 98889 DIII-D pedestal [J.D. Callen et al., Nucl. Fusion 50, 064004 (2010)]. Also, recent SOLPS modeling by John Canik has shown that (as reported in his invited talk JI2.1 talk at the recent Chicago DPP-APS meeting) these model predictions for transport quasi-equilibrium electron heat transport and density profile shape "are consistent with" NSTX pedestal data both with and without Lithium wall coatings. Further, John has estimated the Z_eff variation in the 98889 pedestal and also obtained "pretty good" agreement between the new pedestal structure model predictions and the SOLPS modeling results for that pedestal's chi_e(rho) and n_e(rho). The additional pedestal structure model predictions that increasing the neutral fueling source in the pedestal increases the "offset" of the n_e profile relative to the T_e profile [in Eq. (29) or (32)] and decreases the plasma toroidal rotation as one moves into the pedestal from the separatrix [Eq. (47)] have not yet been tested.
In the UW-CPTC 10-6 report 4 fundamental tests, 4 secondary tests and 4 improvement scenarios are identified for this new pedestal structure model. Secondary tests #2 and #4 involve determining if the experimentally observed neutral recycling and resultant charge exchange effects on the n_e profile outward offset relative to the T_e profile and positive gradient of the plasma toroidal plasma rotation in the pedestal are as predicted by the new paleoclassical-based pedestal structure model. | ||
| Resource Requirements: | Tests are needed of one or more of the possibilities (see Experimental Approach/Plan above) for increasing the main chamber recycling relative to the divertor fueling source. Then, for a selected method, changes in the n_e, T_e profile offset and toroidal flow profile in the pedestal induced by systematic increases in the degree of main chamber recycling need to be measured. | ||
| Diagnostic Requirements: | The new edge Thomson system is needed to better measure the electron temperature and density profiles in the pedestal, and particularly their "offset" at high electron density. Also, the radial profile of the carbon (and hydrogen?) toroidal rotation needs to be measured via the CER system as the mix of main chamber to divrtor recycling is increased. | ||
| Analysis Requirements: | It would be desirable to see if the n_e, T_e offsets in the DIII-D part of the DIII-D/JET similarity experiments agree at least roughly with the prediction of Eqs. (29) or (32) in UW-CPTC 10-6. Also, maybe the collisionality scan data from December 3, 2009 (shots 140411 - 140441) could be analyzed for both the profile offsets and the spatial variation of the toroidal rotation profile in the pedestal. | ||
| Other Requirements: | Some more direct measure of the neutral density in the (poloidal) region where the main chamber recycling dominantly occurs (near baffles?, up inner leg of X-point?) would be useful to quantify the neutral ionization fueling source that results from main chamber recycling. | ||
| Title | 468: Investigate elongation limit in low li discharges with/without applied 3D magnetic fields (Dup. 356) | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Not requested |
| Co-Author(s): | L. Lao, R. Buttery, M. Chu, N. Ferraro, A. Reiman, A. Turnbull | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to produce high kappa ~ 2.7 discharge with low li to provide target for high beta + 3D perturbation study | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using the formation technique employed to produce low-li DIII-D discharge #122976 (early beam heating, Bt and Ip ramps), attempt to reproduce the profiles of 122976 in a discharge of smaller minor radius (a~50 cm).
At li~0.5, the elongation is predicted stable even beyond k~2.6. Look for n=0 stability limit. Apply maximum n=3 I-coil perturbation, look for an effect. Compare with MHD calculations. | ||
| Background: | A possible upgrade of DIII-D, currently under study, consists in the installation of nonaxisymmetric coils above and below the midplane, capable of applying a field large enough to improve the vertical stability of highly elongated plasmas [Rieman, PRL 99, 135007 (2007)].
High elongation is expected to be beneficial for both confinement quality and stability. FNSF and DEMO studies rely on high elongation to reach very high fusion performance. Stable highly elongated plasmas can also be produced by using very low li (~0.5). Recent calculations by Lang Lao show that a low li DIII-D discharge like 122976 would be stable to n=0 with kappa up to 2.66 with the DIII-D plasma-wall distance. This proposed experiment would provide us low li plasmas with the very high elongation that 3D fields could enable in moderate li as well. | ||
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| Title | 469: Extend QH-mode operation to high performance AT plasmas (Dup. 389) | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award for Innovative Research | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to demonstrate sustained ELM-stable AT plasmas with very high beta and confinement at near-zero toroidal rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | - Use 10 kA I-coil operation to apply a large counter torque to the plasma (TNRMFā??Ī“B2).
- Double-null shape with applied ECH could provide target plasma with zero or counter-Ip intrinsic rotation. - Combine NRMF-assisted QH-mode recipe with BT-ramp AT scenario | ||
| Background: | |||
| Resource Requirements: | Needs engineering review to allow I-coil operation at 10 kA, at reduced toroidal field. | ||
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| Title | 470: Investigating the presence of runaway ions in runaway plateaus | ||
| Name: | James jamesan@fusion.gat.com | Affiliation: | |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | By injecting pellets infused with Helium-3 into a runaway plateau or another reactant, the presence of runaway deuterium ions could be diagnosed by the emitted high energy by-products. | ||
| Background: | While runaway ions are entirely neglected in most runaway theories, experiments exhibit soft x-ray bursts at the inner strike point during rapid-shutdowns: the direction runaway ions would be accelerated. | ||
| Resource Requirements: | Pellet injector | ||
| Diagnostic Requirements: | Fast ion loss detector?, Gamma spectroscopy? | ||
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| Title | 471: Reoptimize DIII-D Breakdown | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Not requested |
| Co-Author(s): | A. Hyatt | ITPA Joint Experiment : | No |
| Description: | Reoptimize the DIII-D Ohmic breakdown phase to restore previous best breakdown results | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This work can be done piggyback (if session leader allows changing the front end of the discharge), or on startup days. Optimize effect on breakdown by minimizing VloopB when first light is detected. Parameter Scans:
1. Vary null position 2. F6/F7 scan (wvsp1p) 3. Obtain broadest null by varying ??? 3. Prefill scan 4. Glow duration scan (Lazarus [1998] found machine condition was important) | ||
| Background: | Reproducible discharges require a robust breakdown. This was optimized by Lazarus (NF 1998), but the breakdown is no longer optimum. Using the same methodology as Lazarus, current discharges have VloopB=4-5 V at first light (using the VB array, see 140502 or 142400). Best previously was VloopB=0.75V with same analysis (see 88470). | ||
| Resource Requirements: | |||
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| Analysis Requirements: | Modify the breakdown database (Jackson) to include V_bkdn and monitor this parameter on a shot-by-shot basis. Repopulate the current database with previous data. | ||
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| Title | 472: Eliminating some Analysis Bugs in EFIT and magnetics | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Plasma Control and Operation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Fix some known long term problems in EFIT and magnetics | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | No dedicated shots are needed.
1. Develop a more robust snap file for the Ip rampup phase (may possibly use 2 snap files: an early one, then switching to the default later in time) 2. Jump in magnetics data at t=10 ms may be a known timing problem switching from one time domain to another. 3. Ip_probes. Either a better algorithm is needed, or the signal can be set to zero by calculating the first few samples after t=-50ms when Ip=0. Some other data use this technique. | ||
| Background: | 1. Many automatic EFIT slices (EFIT01) do not converge during rampup leaving large gaps in the data. This can lead to confusion when plotting quantities such as li, and also when calculating quantities like psibdy and v_bdy.
2. There is a jump in many magnetics signals at t=+10ms. Ted Strait thinks this is due to switching time domains when ptdata is one sample off, but this needs to be verified and a patch applied. 3. ip_probes has always had an offset, typically 4-6 kA. Offset is probably due to BT pickup, since integrators start during ip ramp. | ||
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| Title | 473: Improved modeling of edge ionization sources in pedestal matching experiments | ||
| Name: | Hughes jwhughes@psfc.mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | R.J. Groebner | ITPA Joint Experiment : | No |
| Description: | We wish to characterize the differences in ionization source distribution that are obtained between DIII-D and other devices, when non-dimensional plasma parameters are matched in H-mode.
This requires the use of 2D neutral modeling, heavily constrained by experimental measurements. Knowing the source distribution enables improved tests of pedestal models. Developing the technique on existing discharges would help in the planning of future joint experiments. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Initially we would examine the data set on C-Mod and DIII-D, in which we matched non-dimensional parameters at the top of the pedestal, and see whether sufficient edge/divertor data exist to perform accurate integrated modeling of the neutral source (for example with OEDGE and other codes). We could start with discharges run in 2002 as part of the first C-Mod/DIII-D pedestal collaboration, and also with identity discharges attempted as a part of the FY10 JRT on SOL widths. Once the ionization source is determined in each machine, we determine whether it is consistent with the a density pedestal width set by poloidally averaged neutral penetration length. More generally, we would calculate particle transport in the pedestal and determine how it scales between machines with matched non-dimensional parameters. If more identity experiments are run in the future, such a technique should be integrated into the planning. If the older data sets are insufficient to adequately constrain the neutral modeling, then we should work to develop some test discharges with a full component of data from divertor probes, spectroscopy, IR imaging, etc. | ||
| Background: | Characterization of pedestal particle transport is subject to uncertainties in the edge neutral fueling source---both its magnitude and poloidal distribution. This has in particular impacted the interpretation of experiments which match non-dimensional pedestal parameters in two devices of distinct size. For example, in the case of the high-collisionality identity study between DIII-D and C-Mod [e.g., Mossessian, PoP 10(3) 689], the width of the density pedestal on the two machines matched very well, suggesting that plasma transport processes, and not neutral penetration characteristics, set the pedestal width. However, it was possible, using reasonable assumptions for the neutral source distribution on either machine, to attain the matched width using a model for pedestal structure dominated by neutral physics. 2D modeling of neutral sources, constrained by experimental data, ought to help resolve outstanding questions in the C-Mod/DIII-D case, as well as other similarity studies. | ||
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| Title | 474: Power scan in plasmas with fixed pedestal pressure | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | J. Kinsey | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to test the TGLF transport model under conditions similar to those used in the ITER modelling. See background for details. The essential scan is to vary the NBI and ECH input power and see how much the temperatures change. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a balanced beam QH-mode plasma in the ITER similar shape, using shot 141439 as a prototype. Determine the lowest NBI power that can be used to maintain QH-mode; this is the low power end of the power scan. Increase power to maximum NBI power possible with balanced injection. Add ECH to adjust density profile to match the profile from the low power case. This is the high end of the power scan. If time permits, obtain cases at intermediate power levels. | ||
| Background: | ITER modelling using TGLF presented at the IAEA 2010 meeting revealed a fairly rapid decrease of fusion gain with auxiliary heating power. This showed a variation in Q that was approximately P_aux^-0.8 for a case done with fixed pedestal beta. This calculation was done for a case without significant E x B shear and density peaking; it also ignored finite beta effects. Because the stiffness in this plasma seems so extreme, it would worth doing an experiment to validate the TGLF predictions for a case like this. The challenge in doing this is to run a real plasma where the various factors reducing transport (E x B shear, density peaking, finite beta) are eliminated or minimized and the pedestal beta is held constant. By combining features of previously run QH-mode plasma, we can realize these conditions in actual discharges. By using nonresonant magnetic fields (NRMF), we can run QH-mode plasmas with zero net input torque and very low toroidal rotation [Garofalo, IAEA 2010]. This eliminates E x B shear. Density peaking can be controlled in QH-mode by adding various levels of ECH around rho=0.3. Finite beta effects can be minimized by restricting the power range; 10 MW NBI input is the most we can use for balanced beam conditions. Finally, for reasons that are not yet understood, QH-mode plasma seem to run with pedestal beta which remains approximately constant as the neutral beam heating power is increased from six to over 9 MW. [Burrell, PoP 12, 056121 (2005)]. In these plasma, the nu* was quite low and the 9.2 MW level, beta_N reached 1.9, which is fairly close to the ITER baseline value of 1.8. By performing a power scan in balanced beam QH-mode plasmas with NRMF, we can mimic the plasma conditions used in the TGLF modelling of ITER. | ||
| Resource Requirements: | Reverse Ip. 7 NBI sources including both 210 sources. 5 to 6 ECH gyrotrons. | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics,
especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | Modelling with TGLF and simulations with GYRO | ||
| Other Requirements: | |||
| Title | 475: Comparison of pedestal and ELMs in matched discharges on C-Mod and DIII-D | ||
| Name: | Hughes jwhughes@psfc.mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | P.B. Snyder | ITPA Joint Experiment : | No |
| Description: | Produce ELMy H-modes in DIII-D using the atypical shaping that is required in C-Mod to obtain Type I ELMs. Use matched discharges to better characterize peeling-ballooning stability and extend the validation of EPED. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Assess feasibility of a non-dimensional match of a good C-Mod ELMy H-mode case. Likely DIII-D target parameters:
kappa=1.45, delta_l=0.80, delta_u=0.15 Bt=1.7T, Ip=0.7MA, neped~3e19, Teped~0.4--0.6keV. If this condition can be successfully run, perform density and power scans to assess effects in ELMs and pedestal structure. Collect enough data to fully assess peeling-ballooning stability in this shape, then repeat with a more moderate elongation. Compare to a case with kappa~1.8 as well. | ||
| Background: | For the purposes of validating models such as EPED, it would be beneficial to have companion discharges on different devices, which are as closely matched (in non-dimensional parameters) as possible. On Alcator C-Mod, access to ELMy H-mode is greatly enhanced by operating in an atypical shape, in which kappa~1.5, delta_u~0.2 and delta_l~0.8. Operating with this shape is conducive to obtaining a low-collisionality H-mode edge, with regular ELMs, at relatively low input power. Because regular Type I ELMs are not observed with more typical H-mode configurations on C-Mod, tests of models for ELMs and the pedestal structure in ELMy H-mode (e.g. from Snyder's EPED) use this shape exclusively, whereas DIII-D accesses ELMs over a much wider variation in shape. Reproducing a non-dimensional match, shape included, of the C-Mod ELMy H-mode in DIII-D would provide an important test case for a joint benchmarking of EPED across machines, and provide an interesting data set for comparing and contrasting ELM behavior on a medium-sized tokamak and a compact, high-field device. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 476: WITHDRAWN | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | I made a dumb mistake. The ELM control experiment I thought of cannot be done on DIII-D. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 477: Transport modification induced by triggered ELMs in otherwise ELM suppressed discharges | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Investigate the modification to profiles, turbulence and transport resulting from the external triggering of an ELM (for example, by n=3 NRMF ELM pacing) in otherwise ELM suppressed discharges (QH-mode and/or RMP ELM suppressed). In particular, what can we learn from the interplay between the 3D structures such as the EHO that keeps the plasma below the PB stability limit, and the externally applied non-axisymmetric field that apparently has the opposite effect. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Begin with standard QH-mode plasma, with low collisionality and relatively high beta, conducive to ELM pacing with modulated n=3 fields. Add modulated n=3 at moderate frequency (between 20-50 Hz) of different amplitude and look at the effect on the EHO as the ELM is externally triggered (change in EHO amplitude, mode structure, frequency etc), as well as the plasma profiles and fluctuations. Since we will know the timing of when an ELM will occur, we can utilize high time resolution CER measurements and other "burst" mode diagnostics. If possible, repeat in RMP ELM suppressed plasmas. Due to the limitation of 4.5 kA for the SPAs, this would only work for conditions amenable to low current (~3 kA) ELM suppression. | ||
| Background: | External 3D fields from RMP apparently modify the transport in such a way as to stabilize ELMs. Similarly, internal 3D structures such as the EHO in QH-mode also have a clear impact on edge transport allowing operation below the PB stability limit. On the other hand, external n=3 fields have also been used to destabilize ELMs (eg n=3 NRMF ELM pacing). It would be instructive to look at the interaction between these two competing processes, originating from fundamentally similar 3d physics effects. | ||
| Resource Requirements: | I-coil, reverse Ip | ||
| Diagnostic Requirements: | Full profile diagnostics, particularly edge and fluctuations. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 478: Investigate in-out density asymmetry at large toroidal rotation | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | C. Chrystal | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure the in-out density variation of various impurities on a flux surface and, from that, to infer the poloidal variation of the electrostatic potential. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Run a plasma which is capable of operating at both low and high rotation speeds and which runs at relatively low density. For example, QH-mode in the ITER shape with the nonresonant magnetic fields would be a good candidate, since it has run at both low and high rotation. Do a rotation scan and use the CER system to investigate the in-out asymmetry in the carbon density. If possible, keep the density profile the same during the rotation scan. Use the low rotation points to cross calibrate the CER chords inside and outside the magnetic axis, since carbon density is expected to be a flux function at zero rotation. Make measurements also with helium and argon since the in-out variation is expected to change strongly with charge. The potential variation determined with all three impurities should be the same for constant plasma conditions; use this as a cross check. To insure that the plasma is the same for all impurity measurements, inject He and Ar on all shots so that plasma composition does not change. | ||
| Background: | Lowest order parallel force balance in a rapidly rotating tokamak plasma leads to the prediction that the ion and electron density is not constant on a flux surface because of centrifugal effects. The rapid plasma rotation causes the ions to bunch up on the large major radius side of a flux surface. A poloidal electric field develops to insure charge neutrality. The ultimate poloidal variation of the densities of the various species is due to a balance of the electric field and centrifugal forces. For the 2011 campaign, the CER system has been expanded to include measurements both inside and outside the magnetic axis. Using this, we can measure the in-out asymmetry in the carbon density and, from that, infer the poloidal variation of the electrostatic potential. Low density plasmas are preferred for this work both because the neutral beams penetrate better to the high field side of the plasma and because the rotation speeds of low density plasmas are higher. Proper beam modulation is essential to insure good signal from the chords at small major radius because the chords which view the 30 beam at small major radius pass through the 330 beam in the plasma edge. | ||
| Resource Requirements: | Proper beam modulation to get best measurements inside magnetic axis. If QH-mode is chosen for experiment, need reverse Ip. | ||
| Diagnostic Requirements: | CER chords viewing points inside magnetic axis. All profile diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 479: ELM Suppression by Combined n=2,3,4 | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | TBD | ITPA Joint Experiment : | No |
| Description: | Broaden the understanding of RMP ELM suppression at q95 ~ 3.5 by taking advantage of the DIII-D hardware capability to apply n=2 and n=3 RMP harmonics simultaneously. The main physics to be learned is whether such multiple harmonics work synergistically (i.e. more than the sum of the parts) for or against ELM suppression. The plasma MHD response to multiple-n harmonics ought to be very interesting, too. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Take a standard n=3 RMP ELM suppression reference shot sometime during the experiment; this needs an I-coil patch panel change. Then, follow the usual technique to find if there is an RMP ELM suppression window as a function of q95. Ramp q95 upward slowly while a steady mixed-n RMP field is applied, first at maximum amplitude. If ELM suppression is not found, change the toroidal phase of the currents in one row of coils (requires patch panel change). If ELM suppression is found, then then search for the minimum RMP amplitude that still suppresses ELMs. Collect data to characterize the plasma state and compare with conventional n=3 suppression. The plasma MHD response to multiple-n harmonics ought to be very interesting, so MHD magnetics to measure plasma MHD response are requested. | ||
| Background: | "Sideband" toroidal harmonics are generated when a sinusoidal current is approximated by a finite number of perturbation coils, except for special integer relations between the fundamental toroidal harmonic number n and the number N of identical perturbation coils equally spaced around the torus. For example, the proposed ITER "ELM" perturbation array has N=9 coils and may be operated to make n=4, which makes a substantial n=5 sideband harmonic. It is not known how either unintentional sideband harmonics or deliberately applied multiple-n harmonics will affect RMP ELM suppression. SURFMN analysis shows that there is little advantage from multiple n to the Chirikov overlap layer width, when the summed coil Amp*turns are kept constant. However, the validity of the overlap criterion is questionable, and it is worth an experiment to learn if different physics rules. The DIII-D I-coils can be connected to make a fundamental n=2 harmonic, in which case they also make a significant n=4 harmonic. It is possible (two valid patch panel concepts) to apply n=2,3,4 simultaneously from the DIII-D I-coils by connecting one I-coil row for n=3 and the other row for n=2. | ||
| Resource Requirements: | 2 C-supplies connected for 6.5 kA to I-coils, with special patch panel for n=3 to one coil row and n=2 to the other row. | ||
| Diagnostic Requirements: | Request MHD magnetics to measure plasma MHD response. | ||
| Analysis Requirements: | Compare results from this multiple-n experiment with conventional n=3 RMP ELM suppression. | ||
| Other Requirements: | |||
| Title | 480: Error Correction vs. I-coil Phasing | ||
| Name: | Schaffer schaffer411@gmail.com | Affiliation: | Self-Employed |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | TBD | ITPA Joint Experiment : | No |
| Description: | Guided by the IPEC "dominant mode" theory, test low-beta error correction in DIII-D using I-coils connected for 300deg top-bottom phasing difference. The 300deg phasing should couple more effectively to the dominant mode at high q95 (4-7) than the conventionally used 240deg phasing. 300deg also applies less non-resonant magnetic energy to the plasma. This experiment should be run at q95 = 4.5 to make contact with the high-q end of previous 240deg phasing experiments. If the 300deg phasing corrects better than 240deg at q95 = 4.5, then continue the experiment with 300deg phasing to a higher q95 = 6 or 7. The additional data will be used to make a new high-q "standard" correction algorithm for PCS and routine DIII-D use. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will be done with standard Ohmic, locked-mode, error test plasmas. Use standard DIII-D magnetic field directions. Connect the I-coil for n=1 in three 300deg quartet circuits. Empirically optimize locked mode error avoidance using the standard 4-quadrant technique. In the first test of 300deg phasing, set q95 = 4.5, the same as the "high point" q95 in previous 240deg phasing empirical correction experiments. Compare the lowest plasma density without disruption at the shared q95 = 4.5. If the 300deg phasing corrects better at this q95, and if there is time, find the empirical correction at a higher q95, say 6 or 7, so that an empirical correction for high q can be programmed into PCS. | ||
| Background: | The I-coil with 240deg top-bottom phasing presently provides good error correction to left-handed Ohmic low-beta plasmas at DIII-D. The complementary phasing, 120deg, is used for right-handed plasmas. Insight from the IPEC dominant n=1 mode theory suggests that 300deg and 60deg phasings, respectively, have better field geometries to couple to the dominant mode while also applying less non-resonant magnetic energy, at all but the lowest q95 values run at DIII-D. The outcome of empirical error correction with 300deg phasing will be compared with IPEC analysis, to further validate or repudiate the IPEC model. It will add good poloidal spectrum data to the Buttery proposals 201 and 207. The outcome of this experiment will also be of practical value for DIII-D error correction, especially for high-q experiments. | ||
| Resource Requirements: | Standard DIII-D Bt and Ip directions. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | IPEC, SURFMN | ||
| Other Requirements: | |||
| Title | 481: Investigations of fuel accumulation and impurity transport in the gaps of ITER like castellation | ||
| Name: | Litnovsky a.litnovsky@fz-juelich.de | Affiliation: | Juelich |
| Research Area: | ITER First Wall Issues | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov (UCSD), V. Philipps (FZJ), C. Wong (GA),R. Boivin(GA), N. Brooks (GA), P. Wienhold (FZJ), O. Schmitz (FZJ), R. Bastasz (SNL), J. Whaley (SNL), W. Wampler (SNL), J. Watkins (SNL), J. Brooks (ANL), T. Evans (GA), D. Whyte (MIT), P. Stangeby (Univ. of Toronto), A.Mclean (Univ. of Toronto), J. Boedo (UCSD), R. Moyer (UCSD), D. Matveev (FZJ) | ITPA Joint Experiment : | Yes |
| Description: | The aim of this experiment is to mitigate carbon transport and fuel deposition in the gaps of ITER-like castellated structures and to study the processes in the plasma-shadowed areas of gaps. Castellation cells having roof-like shape will be used in this experiment. Shaping of castellation cells should provide significant difficulties for impurities and fuel particles to penetrate and accumulate inside the gaps. Metallic plates will be installed below the castellation to investigate the deposition at the bottom of the gaps. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The shaping of castellated structures represents a natural way for reduction of the impurity deposition and fuel accumulation in the gaps of castellation. The dedicated experiment had indeed demonstrated the positive effect of castellation shaping. However, because of limited exposure time, the detailed characterization and quantification of the deposition in the gaps is accompanied with significant difficulties and uncertainties. Recent studies revealed the formation of rather thick deposits at the bottom of the gaps which are yet to be understood and reliably modeled.
It is planned to elaborate the experiment with castellation shaping by making the long-term piggyback exposure using DiMES system. Castellation samples of different shapes will be exposed simultaneously with ones having the conventional (rectangular) cells to allow for a direct comparison. Tungsten castellation will be used for the experiment, since tungsten is planned to be used in ITER divertor. Piggyback exposure during approximately 1 week of plasma operations, preferably in LSN configuration, is requested for castellation to obtain the representative deposition patterns in the gaps. The instrumented plates will be installed below the castellation to collect the deposited material at the bottom of castellation. The surfaces of the gaps along with instrumented plates will be analyzed in the American and European laboratories. | ||
| Background: | In ITER, the castellated armor of the first wall and divertor will be used to maintain the durability of the machine under the thermal excursions during plasma operation and to avoid the forces by induced currents. There are concerns about the impurity deposition and fuel accumulation in the gaps of castellated structures, representing safety issue for ITER operation. Past and present research demonstrated that the fuel inventory in the gaps of castellated structures is significant and there are essential difficulties in fuel removal.
The mitigation of the fuel accumulation in the gaps by the gap shaping and the investigation of material migration in the plasma-shadowed part the gap, including gaps bottom are among the main topics for a new focused task of the IEA-ITPA Joint Experiments Program (Task DSOL 27). To address these topics, dedicated investigations are ongoing on several tokamaks worldwide: DIII-D, TEXTOR, Tore Supra, EAST and ASDEX-Upgrade. The same design of a castellation will be used in DIII-D, ASDEX-Upgrade, EAST and TEXTOR. Flexible design allows for a direct comparison of conventional and optimized shaping within the same experiment along with an easy access to the bottom of the gaps. There is a possibility to study both poloidal and toroidal gaps under ITER-relevant conditions. An essential advantage of this experiment is that the exposure of castellated samples will be performed at shallow angle with respect to magnetic field, similarly as expected in ITER. The possibility of the direct comparison of experimental results from the major tokamaks worldwide is another advantage of this experiment. | ||
| Resource Requirements: | Machine Time: One week piggyback exposure using DiMES manipulator system, preferably LSN operation, NBI-heated ELMy H-mode. | ||
| Diagnostic Requirements: | DiMES TV, floor Langmuir probes, in particular the probe at the DiMES radial location, MDS chord looking at DiMES. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 482: Investigate angular momentum diffusion and pinch using off-axis torque | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Evaluation of Off-axis NBI Physics | Presentation time: | Not requested |
| Co-Author(s): | W.M. Solomon, B.A. Grierson, C. Chrystal | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to determine the angular momentum diffusivity and pinch velocity from the toroidal rotation change caused by off-axis injection of angular momentum using the 150 beams. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This work requires a specific combination of neutral beams but otherwise can be done in a whole range of plasmas. The key is to have the 30LT and 210RT beams on continuously and to modulate the 150 beams in a situation where the 150 beam is tilted to give the maximum off-axis injection. The 30LT and 210 RT beams provide CER data for both the carbon ions and the main ions. We will to use the prompt torque from the 150 beams to do a modulated angular momentum transport. By analyzing the transient response of the toroidal velocity to the modulation, we can extract the angular momentum diffusivity and pinch velocity across most of the minor radius. | ||
| Background: | Modulated momentum transport work has been done recently on D III-D [Solomon et al, Nuclear Fusion 49, 085005 (2009)], JET [Tardini et al, Nuclear Fusion 49,
085010 (2009)] and JT-60U [Yoshida et al, Nuclear Fusion 47,856 (2007)]. The latter work is particularly elegant, since the modulated beam was far off axis and the analysis of the rotation response could be done using a source-free transport equation. With the advent of the off-axis neutral beam on D III-D, we can now perform similar experiments. In addition, with the new, main ion CER system, we can extend that work to study both the impurity and the main ion rotation. | ||
| Resource Requirements: | Off-axis setting of the 150 beam. 30LT and 210RT beams for CER measurements. | ||
| Diagnostic Requirements: | Main ion and carbon CER systems | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 483: Determine if inside-launch pellet injection is compatible with low rotation QH-mode with NRMF | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | L.R. Baylor, N. Commaux, T.C. Jernigan, A. Garofalo | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to test whether pellet fueling using the inside launch pellet injector is compatible with low rotation QH-mode produced using nonresonant magnetic fields (NRMF) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment should be done as part of the overall investigation of the low rotation QH-mode described in ROF proposals 82 and 83. When good QH-mode conditions are obtained, pellet injection should be used in repeat shots to see its influence on the plasma, especially whether the discharges continues to operate without ELMs. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. Further development of low rotation QH-mode is planned in the 2011 campaign. | ||
| Resource Requirements: | Reverse Ip. | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies. Pellet diagnostics are needed. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 484: Investigate change in plasma turbulence when temperature gradient crosses critical gradient for TEM | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | J.C. DeBoo | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the changes in plasma turbulence when the trapped electron mode (TEM) is destablized and to investigate the effect of collisinality on the TEM. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish L-mode plasma with 0.4 MA current and full toroidal field, similar to that used in previous D III-D and ASDEX-U experiments. Plasma line averaged density was about 2.2 x 10^19 m^-3. Scan electron heat flux through the surface at rho = 0.45 by changing the ratio of the ECH at rho = 0.35 and rho = 0.55 while keeping the total ECH power constant. The constant total ECH power insures that the edge plasma does not change. Investigate density and electron temperature fluctuations using all the D III-D fluctuation diagnostics. Scan the region between the two ECH deposition locations. Use short beam blips to allow ion temperature measurements with CER. On some shots, modulate a portion of the ECH power to allow determination of the heat pulse propagation. Finally, scan density up to about 4 x 10^19 m^-3 to demonstrate the reduction
of TEM at higher density. | ||
| Background: | Experiments in ASDEX-U [Ryter et al, Phys. Rev. Lett. 95, 085001 (2005)] have established a method of scanning the electron heat flux with ECH that allows detection of the critical gradient for TEM turbulence. Clear evidence of a change in the slope of the heat flux versus temperature gradient curve was produced with this method. However, no turbulence measurements were available to see how the plasma turbulence changed when the experimental gradient crossed the critical gradient. L-mode plasmas similar to those used in ASDEX-U have been run in D III-D; however, the Ohmic current was too high and the Ohmic power alone pushed the electron temperature gradient above the TEM gradient. The proposed experiment will use 0.4 MA plasma current to reduce the Ohmic heating, as was done in ASDEX-U. | ||
| Resource Requirements: | ECH, 6 gyrotrons needed | ||
| Diagnostic Requirements: | All profile and fluctuation diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 485: Measure Island Structure of a QSM | ||
| Name: | La Haye rbrtlahaye@gmail.com | Affiliation: | Retired from GA |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | C. Petty, M. Austin, B. Bray, D. Brennan, T. Carter, R. Groebner, J. King, M. Makowski, Y. Shankar, B. Tobias, M. Van Zeeland, J. Yu | ITPA Joint Experiment : | No |
| Description: | Produce a very slowly rotating m/n=2/1 quasi-stationary mode (QSM) and measure the radial structure of all perturbed quantities across it. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | An initially rotating 2/1 mode tends to slow down and lock. If torque applied is sufficient this does not lock but rotates at a few Hz, a QSM. To insure no locking, the application of a rotating n=1 I240 field can force the rotation at a prescribed low value of "constant" frequency which might make analysis by Fourier averaging better. | ||
| Background: | D3DMP 2010-54-01 run on March 4, 2010 tried to get a 2/1 mode at around 1 kHz for CER to be able to get perturbed Ti and Vphi. This was successful in only a very few shots as either the mode locked or it was too fast. The MSE could not resolve perturbed Btheta on any of the good shots. Neither could the DBS. Francesco Volpe's locked mode control was earlier (2009) successful in catching a mode that was going to lock and entraining it to the I-coil n=1 rotating field. The very low frequency rotating islands from a natural QSM or with entrainment should be much easier to make and diagnose. The data at around 1kHz looks nice (except for MSE) but is of short duration steady state and sparse radially. Still needs further analysis. | ||
| Resource Requirements: | Usual beams in H-mode, no ECCD/ECH, usual I-coil. | ||
| Diagnostic Requirements: | All diagnostics looking at q=2 region etc. DBS will have to be optimized for the situation expected. | ||
| Analysis Requirements: | Usual ECE, ECEI, TV and SXR analysis and imaging, with eventual comparison of measured radial profiles to reconstruction by NIMROD. Nothing special on Thomson, CER etc as a slow mode should present no difficulties. | ||
| Other Requirements: | None. | ||
| Title | 486: Effect of Static Resonant m/n=1/1 Field on Sawteeth Instability | ||
| Name: | La Haye rbrtlahaye@gmail.com | Affiliation: | Retired from GA |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | I. Chapman, C. Petty, R. Pinsker, B. Tobias | ITPA Joint Experiment : | Yes |
| Description: | Apply a large 1/1 static field to an H-mode plasma with rotating 1/1 resistive sawteeth precursors. Look at the effect of 1/1 static reconnection on the sawteeth instability, i.e. precursors, crash time, period, etc. Measure the static and time dependent 1/1 structure. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To an H-mode plasma with rotating resistive sawteeth precursors, using I240 n=1 EFC, apply as large as possible static C-coil field (largest component is 1/1) without inducing rotating 2/1 modes or locking. Look at the effect on the periodic sawteeth instability and image the 1/1. | ||
| Background: | In 1992, published as R. Snider NF 1994, it was observed that application of a large 1/1 dominated field from the "n=1 coil" slowed plasma rotation AND stabilized resistive sawteeth. The q(0) went further below 1 destabilizing the ideal sawteeth with a longer period, no precursor and a fast crash. The stabilization was attributed to "jitter" stabilization. Removing the I240 n=1 EFC slowly to zero in more recent plasmas reproduced the effect (this uncovers the intrinsic 1/1 EF). Very much better diagnostics are of course in hand now than in 1992. | ||
| Resource Requirements: | H-mode plasma with sawteeth. No ECCD/ECRH. | ||
| Diagnostic Requirements: | Imaging of the core with TV, ECEI, DBS(?), rest standard. | ||
| Analysis Requirements: | Stability analysis of kinetic EFITs by PEST3 for example of cases with and without large static 1/1 field applied. | ||
| Other Requirements: | None. | ||
| Title | 487: Investigation of the combined effect of rotational and magnetic shear on ion stiffness | ||
| Name: | Mantica none | Affiliation: | IFP ENEA-CNR Milano Italy |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | C.Petty, G. McKee | ITPA Joint Experiment : | Yes |
| Description: | A series of experiments to reproduce the JET results that ion stiffness is significantly reduced by concomitant high rotational shear and low magnetic shear is proposed for DIII-D in view of the new availability of off-axis NBI. This experiment is also part of the ITPA joint experiment TC-13 on ion and electron critical gradient and profile stiffness, which includes also JET, C-MOD and AUG. The experiment consists in a core ion heat flux scan at constant total ion heat flux, obtained by comparing on- and off-axis co-NBI, both in a situation of low rotation plasma ((where low means a rotation gradient less than 15 krad/sm) and high rotation plasma (where high means a rotation gradient larger than 70 krad/sm). The ideal q profile for such experiment features a core region with small magnetic shear, such as at early time of a discharge or with hybrid-like conditions. Conceptually it would be easier to perform the experiment in L-mode and only in a second phase to repeat it in high power Hybrid plasma. NBi modulation could also provide good Ti modulation data on top of the profile observations. After having performed the above, the same flux scan should be repeated in plasmas with different q profiles, to check that the effect is only present when s is small. This can be obtained by exploiting the natural current profile peaking within the discharge, or also with current ramps-up and -down. The target plasma should be without sawteeth or big ELMS (high q95 L mode was used in JET) and at low collisionality, to minimize coupling with electrons. The absence of pedestal helps also increasing the range of normalized heat flux scanned, since the Gyro-Bohm normalization contains a factor 1/T^5/2. It is very important that the rotation profile stays the same between on- and off-axis NBI, otherwise the ion stiffness determination will be heavily affected. This can be obtained by carefully using co- and counter beams, although the absence of counter off-axis NBI may complicate the recipe to be used, especially in the low rotation case. ECH can if necessary be used as a way to reduce the rotation level. Perform fluctuation measurements at low/mid k. The analysis should be at two radial positions: the main one in the core at rho_tor~0.35, but it is useful to keep an eye on what happens also at an outer position rho_tor~0.7. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) establish a low nu*, high q95 but hybrid like q profile shape L-mode plasma with off-axis NBI and low rotation (by balancing with on-axis co-NBi or using ECH). This should yield the linear threshold at low rotation.
2) Repeat with on-axis NBI with balanced injection to preserve the low rotation. This should yield a high ion stiffness level typical of a non-rotating plasma. 3) Increase the amount of rotation to the maximum allowed by the 5 MW off-axis co beams. It should be peaked because of the momentum pinch. This should yield the threshold at high rotation. 4) Use the amount of on-axis co+counter beams that provides the same rotation as in 3) for same power. This would yield the stiffness at high rotation, hopefully smaller than in 1-2). 5) repeat 1-2-3-4 in a plasma with more peaked core q profile. This could well be a by-product of the time evolution of the discharges of 1) to 4). 6) repeat 1-2-3-4 in a plasma with Ip ramp-up or down 7) repeat 1-2-3-4 in a plasma in H-mode (such as with added ECH power or some additional on-axis balanced beam) In all shots it should be monitored that Te/Ti does not vary by more than 10-15%, otherwise try to compensate using ECH. To add modulation, at least in the on-axis cases, it should be discussed if all shots should have a small on-axis balanced NBI modulation on top, or if it is preferred to run separate shots with modulation but otherwise identical parameters. It would be helpful to have assessed in advance if Ti modulation using (balanced) NBI modulation is a useful technique for perturbative ion transport studies. | ||
| Background: | JET experiments on ion stiffness have revealed a significant reduction in ion stiffness level in presence of high rotational shear and low magnetic shear (Mantica, IAEA 2010). This observation has implications on the physics understanding of improved ion core confinement, such as observed in Hybrid or ion ITBs, which may be seen as the same physical phenomenon of low ion stiffness in presence of rotation and flat q profile. As such, the results also imply that both high rotational shear and low magnetic shear are necessary to achieve such low levels of ion heat transport, required for AT scenarios both in present machines and in ITER. It would be extremely useful that experimental evidence of such ion stiffness mitigation by rotation and low magnetic shear is found also in machines different from JET. Further comparison with TGLF and GYRO predictions are also necessary for validation of the transport models. In DIII-D in particular detailed experimental information on fluctuations will be available, which at JET was missing. | ||
| Resource Requirements: | NBI co- and counter, on-axis and off-axis
Possibly ECH | ||
| Diagnostic Requirements: | standard for transport analysis, including ECE, CX, MSE. Fluctuation diagnostics | ||
| Analysis Requirements: | Interpretative transport runs, power deposition calculations, profile analysis, FFT analysis, transport simulations with empirical critical gradient model, TGLF and GYRO simulations | ||
| Other Requirements: | |||
| Title | 488: Core and Pedestal Stiffness | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | J. Kinsey | ITPA Joint Experiment : | No |
| Description: | Explore the "global" stiffness of H-mode plasmas via investigation of both core and pedestal stiffness in power scans in multiple shapes. Test models of both core (eg TGLF) and pedestal (eg EPED) stiffness. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a target H-mode discharge with a saturated (Type I) pedestal, and ramp power to explore core and pedestal stiffness.
Starting target (minimum power value) will be all co-beam, in a configuration in which core barriers are not triggered, and relatively low power is required for access to T1 ELMing H-mode. Scan power up from base value using balanced beam injection, to maintain approximately constant beam torque profile across the power scan. Scan power in both a weakly shaped discharge (eg triangularity~0), where the EPED model would expect the pedestal to be relatively insensitive to core beta (ie global Shafranov shift), and in a strongly shaped discharge where the pedestal stability is expected to be more sensitive to global Shafranov shift (yielding a pedestal that appears to be less stiff). | ||
| Background: | Pedestal height, critical gradients and profile stiffness all play an important role in determining fusion performance. Recent studies with TGLF have found that the core of ITER (and DIII-D) is expected to be relatively stiff at fixed pedestal height. Pedestal models such as EPED predict that the pedestal height in Type I ELMing H-mode does not depend _directly_ on power, but can depend on it indirectly via global Shafranov shift effects on edge stability and through altering parameter space trajectories. Here we wish to study both core stiffness at approximately fixed pedestal conditions, and pedestal stiffness as a function of core beta, and test both the TGLF and EPED models. | ||
| Resource Requirements: | High availability NBI | ||
| Diagnostic Requirements: | profile diagnostics essential, turbulence diagnostics highly desirable | ||
| Analysis Requirements: | TGLF and EPED studies before the expt | ||
| Other Requirements: | |||
| Title | 489: Super H-Mode with NTV torque | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, A. Garofalo | ITPA Joint Experiment : | No |
| Description: | Probe the limits of H-mode confinement and global performance in an ELM-free discharge. Continue development of the "Super H-Mode" (high density QH mode), moving to higher density and pressure. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Begin with previous Super H-Mode discharges, and continue to increase density via fuelling, increased target density, and reduced NBI torque. Use NTV torque to minimize required counter-NBI torque to maintain a QH edge, and optimize the combined NBI/NTV system to maximize edge density (and pressure) and global performance. | ||
| Background: | Understanding of the physics of the pedestal width and height is improving (though still incomplete). In particular, peeling-ballooning stability (and the EPED1 model) predicts constraints on the pedestal that have been tested in some detail.
That understanding allows us to calculate stability diagrams that are usually presented in a j-alpha space. The stability boundary in that space increases rather dramatically shape. In fact, for very optimized shapes (high triangularity, low squareness...), a "nose" on the upper-right is pulled out and extremely high pedestals are predicted to be stable. However, that region is not accessible for any fixed pedestal density. At low density one hits the low-n peeling bound, and at high density the high-n ballooning bound. However, by starting at low density, moving up to the peeling bound, and only then increasing density (avoiding large ELMs), it is possible to access this space. This has been partially achieved in QH mode discharges in 2008, but in principle can be carried to much higher levels if QH mode can be maintained. | ||
| Resource Requirements: | Counter Ip most likely, high NBI availability | ||
| Diagnostic Requirements: | All pedestal diagnostics | ||
| Analysis Requirements: | EPED and peeling-ballooning stability studies prior to expt | ||
| Other Requirements: | |||
| Title | 490: Tests of the EPED pedestal model | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | R. Groebner, T. Osborne | ITPA Joint Experiment : | No |
| Description: | Thoroughly test all aspects of the EPED pedestal model using the enhanced edge Thomson system and possible edge current measurements. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Vary the key parameters of the EPED model (Ip, Bt, shape, density, global beta) over a wide range in ELMy H-mode, in discharges optimized for best possible diagnostic coverage. | ||
| Background: | The EPED model predicts the pedestal height and width by combining peeling-ballooning and KBM constraints. Each of these constraints separately, and the model as a whole, have complex dependencies on magnetic shear, plasma shape, collisionality etc. The model can be used to predict the expected changes beforehand, and then test them thoroughly over a broad range. Extending to ELMing discharges at very low and current, and possibly going to higher Bt (if allowed) would provide particularly good tests. | ||
| Resource Requirements: | 2 days | ||
| Diagnostic Requirements: | Enhanced edge Thomson system, and if possible, edge current measurements | ||
| Analysis Requirements: | EPED analysis before the expt | ||
| Other Requirements: | |||
| Title | 491: ELM suppression with n=2 RMP | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | many | ITPA Joint Experiment : | No |
| Description: | Test models of RMP ELM control, and acquire an improved data set, by developing ELM control using n=2 perturbations from the I-coil set. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Conduct an extensive set of q, density (and if necessary, shape) scans to develop an n=2 RMP ELM-control scenario. Once achieved, rotate the n=2 perturbation to improve diagnostic coverage. | ||
| Background: | Extending RMP ELM control to n=2 perturbations would:
1) Demonstrate flexibility and extrapolability of RMP ELM control beyond n=3 2) Strenuously test models of RMP transport and ELM control 3) Allow much better diagnosis of 3D edge effects via rotation of the n=2 perturbations | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | enhanced Thomson system | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 492: Modulated ECCD for 2/1 NTM suppression | ||
| Name: | Welander welander@fusion.gat.com | Affiliation: | GA |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | Rob LaHaye, Ron Prater, Ted Strait, John Lohr, Ben Penaflor | ITPA Joint Experiment : | No |
| Description: | The purpose is to evaluate benefit of modulated ECCD for suppression of the 2/1 NTM and to scan how suppression depends on the phase between the ECCD and the island O-point. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The shots will use a combination of co- and counter beams to produce a 2/1 NTM with a frequency of about 5 kHz. A plan for how to trigger NTMs should be tested before this experiment.
Modulation of gyrotrons will be used. This requires: (1) at least two gyrotrons on the same launcher, (2) aim at q = 2/1, (3) launch angles for Ī“eccd about three times as wide as usual, (4) spread launcher angles so they straddle q = 2/1 by dĻ? = 0.6Ī“eccd for further total width, where Ī“eccd is FWHM. The goal is to suppress a 2/1 NTM using modulated ECCD with the best guess of phase for the modulation. In the next step suppression will be attempted with cw-ECCD for comparison. After that modulation in the X point of the island will be tested for comparison with modulation in the O-point. After that a shot with a continuous sweep of the phase between O and X will be done. | ||
| Background: | Continuous wave ECCD has proven effective in completely suppressing NTMs. It can also prevent NTMs, when preemptively applied in the correct radial location. The cw-ECCD requires good alignment and a narrow deposition region with respect to the island width. In ITER the ECCD will have a relatively wide deposition region which will make cw-ECCD less effective since the destabilizing effect from current driven in the X-point will nearly cancel the stabilizing effect from the current driven in the O-point. This problem can be solved by switching the gyrotrons on when the O-point passes by their respective line-of-sight and off when the X-point passes by. Predictions made by F.W. Perkins suggest that a modulation scheme using a square pulse train with a 50% duty cycle will give close to maximum feasible suppression rate. Previous work on ASDEX-UG has demonstrated the efficacy of modulated ECCD. The present experiment seeks to study the suppression of the 2/1 NTM with modulated ECCD, and demonstrate the use of modulated ECCD with real-time frequency/phase detection along with sustained synchronization with the mode. The system to be demonstrated constitutes a general solution for mode frequency/phase detection that can be easily extended to simultaneous suppression of multiple modes using launcher steering in future DIII-D upgrades. | ||
| Resource Requirements: | Plasma current: 1.06 MA (standard direction) Toroidal field: -1.61 to -1.69 T (standard direction) Shape: Lower single null, as in 129330
Cryo-pumps: Lower pump (upper pumps desirable), He cooled Error Field Coils: I240-Coil for n = 1 error field correction. Neutral Beams All sources required including 30L and 330L for diagnostics (MSE, CER) and both 210 sources for counter injection and to control rotation. I coils in I240 configuration, powered by the SPAs. ECRH Two gyrotrons on the same launcher with modulation capability are the minimum required. Injected power > 1 MW for > 2s is needed. Four on two launchers (two on each desired). | ||
| Diagnostic Requirements: | Magnetic (fast and slow) CER on 30L and 330L (for V4,5,6) alternating with 10 ms off/10 ms on from 0.9 to 5.0 s except for continuous 0.274 ms resolution for 160 ms around 4.5 s for a total of 994 pulses. MSE Thomson CO2 interferometers ECE radiometer
oblique ECE SXR diodes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 493: Oblique ECE for radial alignment during NTM suppression | ||
| Name: | Welander welander@fusion.gat.com | Affiliation: | GA |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | Dinh Truong, Francesco Volpe, Max Austin | ITPA Joint Experiment : | No |
| Description: | Implement PCS code for use of oblique ECE for swift correct radial alignment of ECCD during NTM suppression and for phase adjustment during modulation of ECCD. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | First step is to write the PCS software and test the real-time analysis without affecting the plasma in a calculation-only mode.
Second step is to radially align the plasma in a piggy back experiment with a rotating NTM, no ECH is needed. Third step is to use the new methods during an NTM suppression experiment. | ||
| Background: | The oblique ECE is a diagnostic that looks at radiation coming out from the plasma from the same direction that the ECH beam is injected and at frequencies both above and below that of the ECH. This allows detection of the temperature fluctuations both inside and outside the surface where ECH is absorbed.
If a rotating NTM is present in the plasma the fluctuations on the oblique ECE channels will be out of phase when the ECH is centered on the NTM. If both channels are viewing one side of the island then a comparison with Mirnow signals can reveal if they are viewing inside or outside the island. With this information the alignment can quickly be corrected. An efficient real-time analysis of the phase between the fluctuations together with an appropriate algorithm holds the promise of making radial alignment of NTMs to ECCD an automatic standard procedure. The first goal of the experiment is to develop and implement this technique. The second goal is to implement a PCS code that will use a combination of oblique ECE and Mirnov data to determine the correct phase and frequency for the modulation of ECCD. The advantage of the oblique ECE for phase detection is that the phase mapping from the diagnostic to the deposition point is simply a slight difference in toroidal angle. The advantages of including Mirnov data are that the correct phase can still be found when all ECE channels are on one side of the island. In previous years the oblique ECE had a large noise but during this year upgrades are planned to produce less noise and increase the channels to 16-20. Some of these new channels could be digitized on ACQ216 and used in the experiment. | ||
| Resource Requirements: | For the third step of the experiment all the usual resources for NTM suppression experiments are required.
Plasma current: 1.06 MA (standard direction) Toroidal field: -1.61 to -1.69 T (standard direction) Shape: Lower single null, as in 129330 Cryo-pumps: Lower pump (upper pumps desirable), He cooled Error Field Coils: I240-Coil for n = 1 error field correction. Neutral Beams All sources required including 30L and 330L for diagnostics (MSE, CER) and both 210 sources for counter injection and to control rotation. I coils in I240 configuration, powered by the SPAs. ECRH Two gyrotrons on the same launcher with modulation capability are the minimum required. Injected power > 1 MW for > 2s is needed. Four on two launchers (two on each desired). | ||
| Diagnostic Requirements: | It is desirable to improve the signal-to-noise-ratio of the oblique ECE diagnostic and this work is under way. Some oblique ECE channels need to be plugged into ACQ 216 for the experiment.
Magnetic (fast and slow) CER on 30L and 330L (for V4,5,6) alternating with 10 ms off/10 ms on from 0.9 to 5.0 s except for continuous 0.274 ms resolution for 160 ms around 4.5 s for a total of 994 pulses. MSE Thomson CO2 interferometers ECE radiometer oblique ECE SXR diodes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 494: NTM control with real-time mirror steering | ||
| Name: | Welander welander@fusion.gat.com | Affiliation: | GA |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | Rob LaHaye, John Lohr, Ben Penaflor, Ron Prater | ITPA Joint Experiment : | No |
| Description: | Steer the ECH mirrors in real-time to maintain alignment between ECCD and NTM-resonant q-surface and study the resulting NTM dynamics compares to shots where the shape or toroidal field are changed to maintain alignment. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The mirror steering using the NTM control algorithm can be tested first without plasma, then with plasma but no ECH, and finally with ECH in a piggy back experiment.
When this commissioning is done the new control technique can be used to repeat shots that used vertical shifting of the plasma to maintain alignment in response to changes in both q-profile and refraction of the ECH beam. | ||
| Background: | Real-time steering of ECH mirrors has recently become available and the code in the PCS is ready.
In the past NTM alignment was made by adjusting the toroidal field or the major radius of the surface or the vertical position. Adjusting the aim of the mirrors is less intrusive. | ||
| Resource Requirements: | Cryo-pumps: Lower pump (upper pumps desirable), He cooled Error Field Coils: I240-Coil for n = 1 error field correction.
Neutral Beams All sources required including 30L and 330L for diagnostics (MSE, CER) and both 210 sources for counter injection and to control rotation. I coils in I240 configuration, powered by the SPAs. ECRH Two gyrotrons on the same launcher with modulation capability are the minimum required. Injected power > 1 MW for > 2s is needed. Four on two launchers (two on each desired). | ||
| Diagnostic Requirements: | Magnetic (fast and slow) CER on 30L and 330L (for V4,5,6) alternating with 10 ms off/10 ms on from 0.9 to 5.0 s except for continuous 0.274 ms resolution for 160 ms around 4.5 s for a total of 994 pulses. MSE Thomson CO2 interferometers ECE radiometer
oblique ECE SXR diodes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 495: Compare ECCD and ECH control of disruptions | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Test/confirm in other types of disruptions (density-limit, high-beta, low q95, laser blow-off) what has been found in locked-mode disruptions: that ECCD is more effective than ECH at avoiding disruptions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Cause disruptions of 2-4 types: 1) by ramping ne and hitting the Greenwald density limit, for comparison with AUG and FTU, 2) by ramping beta up, 3) by ramping q95 down, 4) by injecting impurities (for example by laser blow-off, if available, for comparison with FTU), which increases the resistivity and cools the plasma, resulting in current and thermal quench. Approaches 2) and 3) and, by radiative destabilization, approach 4), are likely to lead to the formation of an island. If so, operate at high rotation to prevent locking, in order to differentiate from the high-beta locked-mode disruptions studied so far.
Dud on Vloop (disruption detector, as in FTU and earlier AUG experiments) or on DUSBRADIAL etc. (locked mode detector, as at DIII-D and, more recently, AUG), whichever trips first. For comparison, also dud on Vloop only. On average, this will react ~400ms later but, unlike the locked mode detector, will also pick disruptions neither caused nor aggravated by LMs. Apply ECH. Disruption should be delayed or avoided compared to no-ECH reference. Similar to AUG, scan radial location of ECH from shot to shot. Maximum effect is expected for deposition at q=2. A smaller peak is expected at q=3. Repeat with ECCD. | ||
| Background: | EFFECT of ECCD
It has been already demonstrated experimentally at DIII-D [Volpe 2010] that ECCD is more effective than ECH at stabilizing locked NTMs and thus at avoiding locked-mode disruptions. This is not surprising, because it is the same physics of rotating NTM stabilization, where the superior stabilization efficiency of ECCD over ECH is well-known. In fact, stabilization in ITER is expected to be entirely dominated by ECCD. EFFECT of ECH For efficient stabilization it is required that ECCD is deposited in the island O-point. In the case of locked islands, this is easily and reproducibly achieved at DIII-D by means static n=1 magnetic perturbations. Thus, the simplicity of ECH, of not requiring magnetic control, is only a minor asset. Besides, although not as much as ECCD, ECH would also benefit from the magnetic control of the toroidal phase of locking. The real potential asset of ECH is that by heating the plasma it can prevent, counteract or reduce the thermal quench. Heating, however, is always present, even during ECCD. In fact, ECCD is a consequence of heating and its asymmetry in the velocity space. Besides, FTU and AUG scans show the maximum effect for ECH at q=2, not in the core. This suggests that ECH stabilizes an MHD instability or other q-resonant phenomenon, and that the effect of compensating for the thermal quench is secondary. In summary, ECH is suspected to have minor effects. All these ECH effects are also present during ECCD, which offers additional advantages. On the other hand, FTU avoided more complicated disruptions than at DIII-D, with a locked 3/1 mode followed by a rotating 2/1 which eventually locked too. MARFEs were also observed in those disruptions. LM and DISRUPTION STATISTICS According to JET statistics, nearly 100% of disruptions originate from or develop a LM. If a similar statistics is confirmed for DIII-D, controlling LMs would virtually be equivalent to controlling disruptions. A preliminary result, however, suggests that at DIII-D only 30% of LMs cause disruptions, and only 10% of disruptions are caused by LMs. If this is confirmed, the extension of ECH/ECCD to non-LM disruptions would earn high priority, and so would the comparison between pure ECH and ECCD. | ||
| Resource Requirements: | 5-6 gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 496: Locked mode avoidance by magnetic f/back on the saddle loops | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Variant of mode locking avoidance ā??a la Okabayashi. The substantial difference is feeding back on the saddle loops when rotation gets too slow for Bp probes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce #127927, feeding back on Mirnov. Repeat feeding back on external saddle loops. Repeat for internal saddle loops. Pick internal or external. Repeat with both Mirnov and saddle loops enabled, for f> and <100Hz, respectively.
Scan phase and gains. Find conditions such that the mode, after avoiding locking, spins up rapidly (which should also rotationally mitigate it) or keeps rotating at frequencies, f<1kHz, amenable by BES, CER and MSE diagnostics, or f<5kHz, optimal for modulated ECCD. For large enough gains, the ā??phase flip instabilityā?? predicted by E. Lazzaro and R. Coelho should be observed. | ||
| Background: | Although initially conceived for RWM control, M. Okabayashiā??s magnetic feedback was successful in avoiding locking of a rotating, slowing down NTM in #127927. In that case, the I-coils were feeding back on magnetic probe measurements.
The probes worked remarkably well even at the low rotation frequencies (tens of Hz) reached by the mode before spinning up again. As a variant of M. Okabayashiā??s proposal 224, here it is proposed to feed back on Bp probes when f>100Hz and on BR loops when f<100Hz. Also related to E.J. Straitā??s proposal 437. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Real-time versions of M. Lanctotā??s rwmplus3 or T. Scovilleā??s rwm are desirable but not indispensable. | ||
| Other Requirements: | |||
| Title | 497: Add modulated ECCD to magnetic feedback control of locked modes | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Completely suppress by means of modulated ECCD a locked mode that is only unlocked or prevented to lock, and forced to rotate, by magnetic feedback. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Add modulated ECCD to M. Okabayashiā??s discharges for #224. Modulation can be pre-programmed, if mode rotation is uniform and reproducible. Simple repetitions of the discharge and small shot-to-shot differences would give an automatic scan of the relative phase between the modulated ECCD and the rotating island. Alternatively, modulation can feed back on magnetics. In fact, it can adopt the I-coil real-time waveforms. For this purpose, connect magnetic-feedback real-time computer to ECH opto-isolators in the annex: apart from imposing 0-10V, the waveforms are the same! Choosing the real-time waveform for one set of I-coils or the other to feed the gyrotrons will also scan the relative phase. | ||
| Background: | Previous results by M. Okabayashi showed that magnetic f/back can either unlock a 2/1 mode initially locked to the EF, or prevent locking altogether. In both cases the mode rotated at 15-40 Hz and had a finite amplitude, comparable with a perfectly locked (0 Hz) mode. Hence in many respects, such as confinement and beta, this slowly rotating mode is as bad as a locked one. ECH/ECCD has the potential to fully suppress the mode. For best results, ECCD should be modulated in phase with the transit of the island O-point in the ECCD deposition region. Otherwise, for continuous ECCD and such a slow mode rotation, the X-point might be exposed to ECCD for too long, resulting in destabilization. | ||
| Resource Requirements: | 5-6 gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | Connection between magnetic f/back computer and ECH opto-isolators in the annex. | ||
| Title | 498: Magnetically assisted ECCD stabilization of locked modes at ITER-like q95 and with external C-coils | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | ITER Demonstration Discharges | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Demonstrate magnetically-assisted ECCD stabilization of locked-modes under two ITER conditions for the first time. The technique has been already demonstrated at high beta, but needs to be extended to: 1) low q95 and to 2) external coils, should the internal ones not be approved/installed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat 141492 but with the C-coils, same toroidal phase. Might require adjusting the C-coil currents. Repeat lowering q95 from shot to shot in steps of 0.3 and changing BT consistently to keep the q=2 location approximately fixed. Repeat the q95 scan again, but without consistently varying BT. As a result, the q=2 location will move towards colder regions, and ECCD be less efficient. Hence, we will increase the current drive density by switching to narrow ECCD (in 141492 it was deliberately broad, for ease of alignment). This will make the alignment more challenging, but we will use the Search and Suppress or Active Tracking to adjust the plasma position, BT or steer the mirrors in real time, if ready. | ||
| Background: | ECCD stabilization of disruptive locked modes assisted by static n=1 magnetic perturbations succeeded at q95=4.5 but should be tested at low, ITER-like values of q95=3-3.5. As a result the mode will lie at outer radii, closer to the wall and error field, and thus interact more intensively with them, i.e. will lock more rapidly. It will also be less shielded by rotation. After locking, the main differences will be proximity to the edge (thus, stronger perturbation from the EF, and bigger island) and the fact that locally the plasma will be colder, and ECCD less efficient.
The other rationale for this experiment is that so far the technique used the internal I-coils, but to date the installation of internal coils in ITER is still uncertain. Because the technique only needs static n=1 perturbations comparable in strength with the machine EF, it is likely to work with external coils as well, but it is important to confirm it experimentally. As a by-product, this experiment will also test the efficacy of ECCD control of locked modes in preventing or avoiding low q95 disruptions (see #495). | ||
| Resource Requirements: | 5-6 gyrotrons | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 499: ELM pacing using squareness modulation | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | G.L. Jackson | ITPA Joint Experiment : | No |
| Description: | Modulate the outer squareness of the LCFS to induce ELMing | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | * Establish a target discharge with "low enough" ELM frequency. A DND may be desired for up/down symmetry. If successful, a LSND ITER shape can be tested also.
* Modulate the outer squareness to promote more rapid ELMing. * If well-controlled ELM frequency conditions can be achieved, this will allow a careful measurement of the ELM heat load versus ELM frequency. | ||
| Background: | * It is well established that greater outer squareness increases the ELM frequency. John Ferron did experiments on this a decade ago. His changes in squareness were relatively large. Hopefully an ELM can be induced with smaller squareness steps. | ||
| Resource Requirements: | Standard | ||
| Diagnostic Requirements: | Standard. IR cameras for heat loads. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 500: Test and Map the ion Polarization Term | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | Waelbroeck, La Haye | ITPA Joint Experiment : | Yes |
| Description: | Critical to resolving the threshold and the stabilization requirements for NTMs is to understand the role of small island stabilization terms that set the threshold for island growth/complete suppression. One of the key theorized terms arises from ion polarization currents, which has been very hard to measure and distinguish from other effects. A new idea is proposed which will readily get at this term and enable us to map it out. This is important in determining how the role of this physics, the extrapolation of NTM issues, and potentially determining new better methods of NTM avoidance or control by manipulating this effect. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The key challenge to test the ion polarization term is to achieve variation in how the island propagates relative to the background plasma. The new concept to get at this term is to use the RMP to hold an island with respect to the background plasma. Propagation could then be varied, either by varying the background plasma with NBI torque, or island with RMP. In fact the torque variation method is best, as varying RMP rotation frequency is limited and leads to high levels of wall shielding if high rotation (~kHz) is attempted. THEREFORE PROPOSAL IS: generate an island in a relatively benign regime (q=5); lock the island to a rotating RMP (10-500Hz)**; explore effect of varying NBI torque on island size. | ||
| Background: | **NOTES: rotating RMP is desirable as it allows island to be rotated past diagnostics to see width and other structural elements. Rotation also help avoid error field drive to island, and may help provide a stabilizing wall contribution (if higher frequencies can be accessed). Experiment should be performed at high enough Bt for good ECE diagnosis. THIS EXPERIMENT IS A KEY ELEMENT OF ITPA MDC-14 FLAGGED FOR PRIORITY IN 2011 | ||
| Resource Requirements: | 0.5 days. ECE, I coils, 2 counter + 3 co beams | ||
| Diagnostic Requirements: | ECE + usual MHD | ||
| Analysis Requirements: | Results should be obvious, but top theorist is participating (!) to provide interpretation in terms of cutting edge theory! | ||
| Other Requirements: | |||
| Title | 501: Inner divertor carbon source | ||
| Name: | Groth groth@fusion.gat.com | Affiliation: | Aalto U |
| Research Area: | Divertor Detachment and Plasma Flows | Presentation time: | Not requested |
| Co-Author(s): | R. Groebner, N. Brooks, A. Leonard, PBI group | ITPA Joint Experiment : | No |
| Description: | Measure the core carbon content (nC4+ and nC6+) as a function of inner divertor detachment | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish either lower or upper single null L-mode discharges with both inner and outer divertor legs attached to the target plates; in LSN, utilize upper outer baffle to limit SOL; in USN, utilize divertor shelf; achieve attachment of inner leg by either divertor pumping or BT operation in unfavorable direction; conduct a density scan: two repeat discharges for CER spectrometer setting | ||
| Background: | Analyses of a series of L-mode plasmas showed that the measured nC4+ and nC6+ in the core is invariant to detachment of the outer divertor leg (and proximity of upstream separatrix to the upper outer baffle within 1-2 SOL fall-off lengths). The hypothesis on the core carbon source is, therefore, the inner divertor region. In this experiment we propose varying the degree of inner leg detachment to determine its effect on the dependence of core carbon. | ||
| Resource Requirements: | L-mode plasma with < half a beam source (330R) in favorable Bt/1 beam in unfavorable Bt configuration; setup of suitable window frame configuration to upper outer baffle and/or lower divertor shelf; experimental time 1/2-1 day | ||
| Diagnostic Requirements: | CER, MDS, FS, TS and DTS, LPs, outer midplane and x-point RCP, profile reflectometer, flow imaging diagnostic, high-field side RCP | ||
| Analysis Requirements: | CER analysis for nC4+ and nC6+, SOL profile analysis at outer midplane and divertor, including carbon poloidal emission profiles | ||
| Other Requirements: | -- | ||
| Title | 502: Main chamber carbon source | ||
| Name: | Groth groth@fusion.gat.com | Affiliation: | Aalto U |
| Research Area: | Fuel Retention and Carbon Erosion | Presentation time: | Not requested |
| Co-Author(s): | R. Groebner, N. Brooks, A. Leonard, PBI group | ITPA Joint Experiment : | No |
| Description: | Measure the core carbon content (nC4+ and nC6+) as a function of proximity of separatrix to limiting surfaces in the main chamber: outer limiter, upper outer baffle (if LSN), divertor shelf (if USN), inner wall | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Set up reference L-mode plasma with large gaps to all limiting wall surfaces: e.g., outer midplane 8 cm, upper outer baffle 6 cm, inner wall 10 cm. Step-wise lower distance to each of these surfaces, and measure its effect on core carbon content; repeat gap scans in high-recycling and detached plasmas (outer divertor leg). | ||
| Background: | Analyses of a series of L-mode plasmas showed that the measured nC4+ and nC6+ in the core is invariant to the proximity of upstream separatrix to the upper outer baffle within 1-2 SOL fall-off lengths. This experimental proposals aims to investigate the minimum gaps size to the primary limiting surfaces in the main chamber for significant increase of the core carbon content. | ||
| Resource Requirements: | L-mode plasma with < half a beam source (330R) in favorable Bt/1 beam in unfavorable Bt configuration; setup of suitable window frame configuration to primary limiting surfaces in the mainchamber: outer and inner limiter, upper outer baffle (LSN) or lower divertor shelf (USN), experimental time: 1 day | ||
| Diagnostic Requirements: | CER, MDS, FS, TS and DTS, LPs, outer midplane and x-point RCP, profile reflectometer, flow imaging diagnostic, high-field side RCP | ||
| Analysis Requirements: | CER analysis for nC4+ and nC6+, SOL profile analysis at outer midplane and divertor, including carbon poloidal emission profiles | ||
| Other Requirements: | -- | ||
| Title | 503: Study of density and ion temperature fluctuations dependence on the ratio of the scale lengths(Eta) | ||
| Name: | Uzun-Kaymak iuzun@metu.edu.tr | Affiliation: | METU - Middle East Technical U |
| Research Area: | Transport: Profile Stiffness and Critical Gradients | Presentation time: | Not requested |
| Co-Author(s): | C. Holland, G. McKee, T. Rhodes, G. Wang, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Investigate the Ī·_i-dependence of localized density and thermal fluctuations by varying the ion temperature gradient scale length via beam heating at nearly constant density profile. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use low density, low current (0.8 MA) L-mode plasmas, and run upper-single-null or inner wall limited shape discharges to keep the LH threshold high. Vary the ion temperature (and gradient) using beam power/voltage scans and maintain a constant relatively low (n<2e19 m^-3) density and rotation. Utilize a solid 150L beam for diagnostics purposes (0-degree rotation), and vary injected balanced power to vary Ti while utilizing co-counter beams to keep the rotation velocity nearly constant over a range of Ti. | ||
| Background: | It is expected that as the normalized ion temperature gradient scale length (R/L_T,i) is increased, heat flux will increase rapidly as the critical gradient is exceeded. Fluctuations are likely to increase as well, though temperature and density fluctuations should respond differently. Furthermore, linear eigenmode calculations predict that as the gradient scale length ratio is reduced, the nature of the modes may change from ITG dominated to TEM. A highly upgraded UF-CHERS diagnostic will be deployed for the 2011 D3D campaign to measure ion temperature and toroidal velocity fluctuations. This experiment will test and exploit this new diagnostic capability, along with the 2D BES and CECE diagnostics to more fully diagnose the long-wavelength turbulence characteristics. | ||
| Resource Requirements: | co and counter beams | ||
| Diagnostic Requirements: | UF-CHERS, BES, CECE and other fluctuation diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 504: Locked mode control "trophy shots" at high beta, high confinement | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Extend Locked Mode (LM) control to even higher betaN and learn how to suppress the LM without affecting confinement. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make Locked Mode (LM) control more efficient by narrowing the ECCD deposition of 6 gyrotrons (which was deliberately broad in previous experiments, for ease of alignment), taking advantage of new and pre-existing alignment capabilities at DIII-D (Search and Suppress or Active Tracking algorithms to adjust the plasma position, BT or steer the mirrors in real time, if ready). As a result of the increased EC-driven current DENSITY, stabilization should become possible under more challenging conditions, e.g. at even higher betaN (was 2.6).
Make magnetic control of the toroidal phase of locking less perturbative by applying just as much perturbation as needed, and by improving the correction of the machine error field, for example by #171-172. Pre-program or f/back fuelling to restore, after locking and its control, the same density as before. Finally, deliberately misalign ECCD (3 gyrotrons too far in, 3 too far out) as reference shot, instead of the 'No ECH/ECCD' or 'Pure ECH' shots. This should just remove the stabilizing effect of the ECCD, but not the effect of ECH on stability and confinement. | ||
| Background: | Suppressing the LM at DIII-D also suppressed most of the undesired consequences of locking, such as the degradation or loss of confinement. BetaN and, to a lesser extent, confinement, partly recover to pre-locking values.
The H-mode is recovered or never lost, but the quality of confinement after stabilization is slightly worse than before locking. This is suspected to indirectly degrade betaN, that amounts to ~2.5 before locking and <2.2 after stabilization (higher values of betaN are obtained by increasing the NBI power). The effect on confinement is a consequence of suppressing the LM (that degrades confinement) by means of ECH/ECCD and, in fact, of an error field (both degrading confinement). The confinement-degradation associated with ECH/ECCD is unavoidable, but the magnetic perturbation (and its effect on confinement) can be minimized. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 505: Effect of vertical field on locked and rotating modes | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Study the stabilizing effect of vertical fields on islands. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Straightforward modification of any experiment investigating (not necessarily controlling) NTMs or locked modes: repeat best shot 2-3 times at increased/reduced vertical field Bz. The dipole mu will tend to align to the total field B_tot. If that is "more vertical", so might become mu.
(2) Oscillate Bz. Can this "pumped swing" act as a humming spinning top and spin up the mode? (3) Zen approach to unlocking: in presence of a LM of magnetic dipole mu, apply B in the same direction -> mu x B =0! Here B is the total field B_tot, that includes the EF. | ||
| Background: | Islands can be thought of as current-carrying wires of complicated shape. Their magnetic dipole mu is locked or rotating with the island, and is mostly but not perfectly vertical. Most if not all control experiments concentrate on the small horizontal component of mu: radial fields BR are applied to either steer mu or "cancel" its horizontal component. Here we propose to think in 3D: at least three new effects might arise from the interaction with a vertical field Bz. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 506: Make locked mode and disruption control more routinely available by real-time steering of ECH/ECCD | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make locked mode and disruption control more routinely available by programming the dud and responding with magnetic perturbations and with ECH/ECCD steered in real-time. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In every discharge featuring ECH/ECCD, set the dud detector to detect a locked mode and/or disruption (from Vloop), whichever comes first and, instead of the usual drop in NBI power, respond to the dud by steering the ECH/ECCD towards the q=2 location, as tracked by the ā??Search and suppressā?? or ā??Active trackingā?? algorithms. Also, at the dud, apply an n=1 static perturbation (with the I-coils or the C-coils, whichever was used; in the I-coil case, apply an n=1 regardless of the relative phase between the upper and lower row).
In shots where ECH/ECCD is not required, we can turn it on later in the discharge at a ā??stand-by (10% of maximum) power level. ECH/ECCD can only be modulated (for example from 10 to 100%) but not turned on (from 0 to 100%) by the PCS. NOTE: permission will be asked to individual session leaders every day, but some form of co-ordination and authorization at an upper level would be appreciated. | ||
| Background: | Locked mode and disruption control was demonstrated at DIII-D in discharges where the NBI was ramped to generate an NTM and rotation was kept low, to let it lock. These high-beta, highly disruptive discharges are sometimes tagged as "idealized", but should rather be referred to as "worst case scenarios". Regardless, it is agreed that the automatic extension of control to a wider class of locked modes and disruptions is highly desirable to test its applicability and versatility in view of ITER. | ||
| Resource Requirements: | We are not asking for dedicated time, but for permission to program the dud detector and, if tripped, use magnetic perturbations and steer the ECH/ECCD in as many experiments as possible, especially if they foresee a risk of locking or disruptions and if they plan to use ECH/ECCD anyway. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 507: Control of the sawtooth heat pulse propagation with a strong toroidal shear flow | ||
| Name: | Tobias tobias@lanl.gov | Affiliation: | Los Alamos National Laboratory |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | H.K. Park, G.S. Yun, H.K. Park, B. Tobias | ITPA Joint Experiment : | No |
| Description: | Simultaneous 2D images of the core and edge plasma rotation were captured by the KSTAR ECEI system during the 2010 KSTAR campaign. It is observed that the edge becomes less affected by the core MHD activities (sawtooth) compare to that of the circular and L-mode phase and develops filamentary structures (m~25) as the plasma evolves into deeper H-mode. A strong shear in the poloidal flow has been observed during the appearance of filaments (peeling ballooning mode). In fully developed H-mode state, the ELM bursts are often preceded by the filaments, i.e., ELM-precursor. In addition, while the toroidal rotation of the core is driven by the NBI (co-current in KSTAR) as expected, the toroidal flow in the outer edge appears to be in the opposite direction. It is our interest to study the role of the layer where the rotation is sheared the most and the layer can be used to break up the coherent radial heat pulse propagation which may trigger NTM mode. The shear layer can be controlled by NBI deposition profile using density profile control.
Similar study on DIII-D for 2D visualization of sawtooth heat pulse propagation physics with the controlled rotation shear layer can be performed with the more versatile heating and current drive system on DIII-D. A synergy of the counter-/co-current and on-/off-axis ECH combined with NBI system will be interesting because ECH directly affects the core sawtooth activities through the localized current flow. Furthermore, co- and counter-NBI capability can be used to study the effect of rotation shear on the confinement, NTM control, and edge instability via control of the sawtooth crash heat pulse to the outer region of q~1 surface. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish double-null L-mode plasmas (kappa~1.8, delta~0.6, NBI~2MW co- & counter-, varying ECH injection angle: ref.shot.???).
(2) Obtain the 2D ECEI images across the poloidal cross-section from the q~1 surface through the scrape-off layer as follows. This will require 3 reproducible shots. (3) Turn on ECH and Repeat for varying injection angle. | ||
| Background: | The study of sawtooth-initiated H-mode transitions helped to understand that the improved conļ¬?nement is caused by the development of a transport barrier right at the edge but inside the separatrix [1,2]. Although the physics behind the H-mode transition has not been clariļ¬?ed yet, there is substantial experimental and theoretical evidence that turbulent ļ¬?ows, which enhance transport and limit the conļ¬?nement, are diminished by sheared poloidal ļ¬?ows residing at the plasma edge. Rather reciprocally, this argument led to a theory that ļ¬?uctuations themselves induce the ļ¬?ows via Reynolds stress, which then act back and annihilate the turbulence. Also the shear of toroidal rotation is considered important for edge instability and confinement of the inner region although the role of the toroidal shear is not clear yet.
In this context, it is important to understand how the sawtooth crash and the subsequent heat flow towards the outside of the q~1 surface affect the NTM, ELMs and edge confinement. If these changes are associated with the generation of shear flow, we can systematically control the toroidal shear flow which will be combination of the intrinsic rotation modified by NBI momentum profile. The ECEI system [3] has a unique capability to measure 2D Te fluctuations with simultaneous coverage of both core and edge regions in 1~2cm spatial resolution and ļ?sec time resolution, suitable for detailed measurement of 2D structure of heat and plasma flows. Although it may be difficult to resolve the Reynolds process and the reduction of turbulence directly, the ECEI system will provide substantial information for (or against) the conceived H-transition process. [1] F. Wagner et al., Phys. Rev. Lett. 53 (1984) [2] F. Wagner, Plasma Phys. Control. Fusion 49 (2007) [3] B. Tobias et al., Rev. Sci. Instr. 81 (2010) | ||
| Resource Requirements: | Heating Requirements:
NBI, ECH | ||
| Diagnostic Requirements: | ECEI, ECE radiometry, MSE, BES, CXR spectroscopy, CO2 interferometers, SXR | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 508: Determine local effect of flow shear on NTM stability | ||
| Name: | La Haye rbrtlahaye@gmail.com | Affiliation: | Retired from GA |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | R. Buttery | ITPA Joint Experiment : | Yes |
| Description: | Use reverse Ip counter beams and/or new off-axis NBI to investigate if 3/2 mode size continues to grow and 2/1 onset beta continues to fall as a "sign effect" of reversing the local flow shear at q=3/2 (existing mode) or at q=2/1 (onset mode). Or does the stability turn up again with an offset minimum in counter side. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish ELMing H-mode target regime at intermediate betaN and COUNTER ROTATION but such that the plasma does not become ELM-free with an uncontrollable density rise. Slowly raise heating power for betaN onset of 2/1 mode in presence of 3/2 mode. Use off-axis NBI for further variation of rotation shear. | ||
| Background: | This is an extension of previous scans to help understand a key physics effect. | ||
| Resource Requirements: | Reverse IP, 5 counter beams, 1 co beam, 2 off-axis beams. | ||
| Diagnostic Requirements: | CER. MSE, usual. | ||
| Analysis Requirements: | GAPROFILE analysis as a basis along with NEWSPEC and ECE. | ||
| Other Requirements: | None | ||
| Title | 509: Rotate n=2 Complete ELM Suppression for Diagnosis of Physics | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | One of the key challenges of n=3 ELm control is that we cannot rotate the field with 6 I coils - so we can't see rotate the perturbation to see the structure of the changes made inside the plasma. However, complete ELM suppression with n=2 fields was demonstrated in 2007 (shot 128963), and n=2 fields can be rotated - so we could use this to diagnose structure of plasma response with RMP ELM suppression - looking for island, fingers, flat spots or gradients. As this was done at 1.94 T, ECE measurements may be of particular value. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | q95 sweep with n=2 coils as in 128963. Then freeze at target q point and apply rotating n=2 field - with high or low frequency (100 to a few Hz) to get good data from various diagnostics - ECE, TS, CER, reflectometer, etc. Might also look at turbulence diagnostics? | ||
| Background: | |||
| Resource Requirements: | n=2 RMPs. Can we apply rotating waveform at sufficient amplitude? Even one revolution would meet main goals. | ||
| Diagnostic Requirements: | Full profile diagnostics | ||
| Analysis Requirements: | Analyse diagnostics and feed into ongoing physics udnerstandinga ctivity | ||
| Other Requirements: | |||
| Title | 510: Understand tearing limit in low torque high beta scenarios like advanced inductive/hybrid (Dup 240) | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | ASKED TO ADD THIS TO A 4TH GROUP! | ITPA Joint Experiment : | Yes |
| Description: | ITER like baselines at low torque have been found to be highly susceptible to error fields, and even with optimal error correction, they encounter tearing modes at low betan, ~2.2. This raises questions for higher beta scenarios like advanced inductive or hybrid, not least because plasma response to error fields is well established to rise with betan - leading to increased braking and more likely triggering of tearing modes. Also the higher beta will increase bootstrap currents which potentially makes it easier for small islands to bifurcate to large amplitude. Current profile is likely a key parameter governing the whole process, and an important factor in establishing scenario viability. Therefore it becomes particularly important to evaluate the prevalence and sensitivities of 2/1 mode threshold in advanced inductive, to assess: (i) whether the changes in current profile for the more advanced regimes improves stability (and how to improve further); (ii) to establish viability and limits for the regime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish low torque variant of advanced inductive plasma. Test prevalence of modes by varying current profile formation recipe (eg early hearing timing) and betan, between standard advanced inductive values and relaxed (later heating start, lower betan) ITER baseline like regimes. Test 3-D field role with n=1 I coil ramps in some cases. Key goals are to determine how stability and 3-D field sensitivity vary with current profile and beta at low rotation. | ||
| Background: | |||
| Resource Requirements: | aries from quick checks (few shots) on the back of low rotation regime development, to dedicated scans to achieve complete goals. | ||
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| Other Requirements: | |||
| Title | 511: Delaying TQ onset during disruption mitigation using ECH (Dup# 73) | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Characterization and Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Use ECH heating of the q=2 surface during pellet injection to delay the onset of the thermal quench and enable impurity deposition in the core. | ITPA Joint Experiment : | No |
| Description: | Use ECH heating of the q=2 surface during pellet injection to delay the onset of the thermal quench and enable impurity deposition in the core | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Standard disruption LSN target. Turn on ECH at q=2 surface shortly before injection of Argon killer pellet. Measure the time from the pellet launch to TQ with and without ECH. Vary applied ECH power and timing relative to pellet launch. | ||
| Background: | A basic problem for almost all massive impurity injection techniques is that the TQ often occurs before the impurity payload can be delivered to the plasma core (< q=2). This results in greater heat conduction to the divertor and inhibits density coalitional suppression of runaway electrons. Delaying the TQ onset time even 1-2 ms would greatly enhance the ability of the various mitigation methods to penetrate to the core. ECH on the q=2 has been shown to delay/avoid numerous types of disruptions on numerous tokamaks. | ||
| Resource Requirements: | 6x ECH, Ar killer pellets, 30L (MSE), 0.5 DAY | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, scintillators, fast camera | ||
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| Other Requirements: | |||
| Title | 512: Use of magnetic feedback to avoid locking of a rotating mode | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, Y. In, H. Reimerdes, T. Strait | ITPA Joint Experiment : | No |
| Description: | - NTMs onset and mode locking is anticipated to occur at various circumstances from low plasma density to beta-collapse in the ITER operational scenarios. The avoidance of tearing mode locking is a critical issue in every step of discharge developments.
- Here, it is proposed to demonstrate a unique usage of internal coils for avoiding mode locking, if successful, this would present an important driving force for ITER to put in such coils. - As we reported in YER in 2010 and 2008, application of feedback can robustly synchronized tearing mode with a rotating external field in the limited piggyback experiments. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Approach
- Apply the feedback with relatively slow bandwidth ( tau~ 10ms) and to decelerate the NTM rotation from a few kHz down to the order of tens Hz and sustain the mode rotation. - In these previous piggyback experiments, we decelerated to 15 Hz ( tau= 40 ms) and 40 Hz (tau= 10 ms) and avoided the mode locking when the mode was rotating in the Ip direction. Goal - To answer several critical issues, - Capability to controlling the rotation frequency - Applicability range of betan - Minimum mode amplitude for sustaining the NTM - Process of mode synchronization and desynchronization (if occurs) - To assess the applicability to the coil and plasma surface separation like in ITER | ||
| Background: | |||
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| Title | 513: High frequency TBM operation | ||
| Name: | Leuer leuer@fusion.gat.com | Affiliation: | GA |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | M. Schaffer, T. Evans, G. Jackson | ITPA Joint Experiment : | No |
| Description: | Goal of experiment is to determine plasma response to high frequency, high m,n fields created by the Test Blanked Module (TBM). The TBM represents the highest m,n order of disturbance that can be injected and controlled in the DIII-D device and it is quite different from that of the I-coils The TBM coils have ~10 times the dipole moment of one I-coil, and they can apply local magnetic field to the plasma. Thus, they should produce substantial local oscillating distortion of the plasma. This would be expected to impact pedestal properties and ELM behavior. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish Elm'ing H-mode with outer plasma edge located as close to the outer midplane TBM location as possible (REDGE ~ 2.3m). Using the maximum current and voltage capabilities of the SPA into the TBM coil loads, sweep the frequency from 10Hz to 1 kHz individually exciting the Toroidal and Poloidal TBM coil. This is expected to be an exploratory campaign to investigate edge behavior in terms of confinement, density pump-out, rotation, pedestal profile properties and ELM behavior and can be executed as a piggy-back experiment. Impacts of a single, high frequency, 3-D source will be quantified. If significant impact to edge properties is detected, plasma parameter scans and plasma shapes variations will be explored. We would also want to operate a single I-coil in similar conditions, to investigate the differences associated with the characteristics of each coil. | ||
| Background: | The TBM has been used extensively to emulate the static fields expected in the ITER test blanket modules [Schaffer, IAEA10]. Its magnetic field represents a very localized B-field in the plasma edge and interacts with the plasma over a very different and smaller footprint relative to other coils (i.e. individual I-coil). Its characteristics, which are associated with tangential and vertical orientated coils are very different from the I and C-coils, which are primarily radial directed. The area ratio of the TBM Toroidal coil to a single I-coil is ~10 and, neglecting the orientation differences, the m*n operating point of the TBM will be an order-of-magnitude higher than that of an individual I-coil. Operating the TBM's two coils over a wide range of frequencies should provide varying plasma penetration depths and offer the capability of modifying the edge properties in a very localized way, and quite different from an individual I-coil. This will also provide valuable 3-D information for our ongoing ELM control by 3-D transport Task Force. Comparisons with similar experiments using an individual I-coil could provide valuable insight into plasma response to localized 3-D fields in the plasma edge. | ||
| Resource Requirements: | TBM insitu, ELM'ing H-mode, preferably with outside edge located near the outer midplane limiter. Machine time: A) Commissioning Outside Machine Testing: 1Hz-0.5kHz at Max SPA voltage; B) Startup Testing: 10Hz-1kHz at Max SPA Voltage (with/without plasma); C) Piggyback: Several piggybacks on experiments during H-mode days; D) one-half day dedicated experiment | ||
| Diagnostic Requirements: | Edge diagnostics will be important, Fast magnetics, MSE, Thompson Scattering, CER, CO2 Interferometer, SPREAD, Visible camera view of bumper limiter, IR camera, ECE, SXR, Bolometers | ||
| Analysis Requirements: | EFIT/Transport | ||
| Other Requirements: | PRE-INSERTION Testing of TBM with SPA at maximum voltage/current with sweeping frequency from 1Hz to 0.5kHz prior to installation on the machine to insure integrity of the system and determine any resonance issues. (note: included in resource requirements above) | ||
| Title | 514: Excitation and characterization of Quasi-stationary Modes (QSMs) | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Understand the physics of QSMs, their excitation and their rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Restore both shots with "spontaneous" QSMs (see S. Mao's database) and shots which initially featured a LM that later evolved into a QSM (F. Volpe's experiments, e.g. 141058-60, 64 etc.). In both types, scan the NBI torque from slightly co- to slightly counter, the n=1 EF (thus, the associated torque) and possibly, super-imposed, the n=3 magnetic braking.
Repeat with ECCD, that was observed to reduce the QSM frequency, probably as an indirect result of affecting its amplitude (141058-59). For comparison repeat with ECH only, which has a weaker dependence on toroidal phase. If time, try to trigger a QSM by pellet injection, similar to J. Snipes at JET [NF 28, 1085 (1988)]. Snipes also observed QSMs being destabilized by large sawteeth. This however, is more easily reproducible as a piggyback on a sawtooth or giant sawtooth experiment than as part of the dedicated experiment. Finally, apply a magnetic perturbation (MP) rotating opposite to the QSM. Apply different rotation frequencies, to confirm that the rapid oscillations and slow growth of the QSM amplitude observed in 141064 and other discharges are beating phenomena occurring at the small difference and large sum of the QSM and MP frequencies. Note that the slow growth was periodically reset by rapid sawtooth-like amplitude crashes, flashes of light and bursts of particle and heat losses. | ||
| Background: | Quasi-stationary modes (QSMs) are naturally slowly rotating modes (10-30 Hz), intermediate in nature between locked modes (LMs, 0 Hz) and rapidly rotating NTMs (5-30 kHz). Their slow rotation is probably a stable solution of the torque balance equation, just like LMs and NTMsā?? are. However, QSMs are observed much less frequently (and reproducibly?) than LMs and NTMs. This proposal seeks to understand why.
Apart from their fundamental interest, QSMs are appealing for two main reasons: they are ā??imperfectā?? locked modes, not really locked but slowly rotating, thus easier to stabilize (e.g. by ECCD only, with no need for magnetic control of their phase); they are ā??imperfectā?? NTMs, rotating at unusually slow frequencies, which makes them easy to diagnose in detail even with diagnostics, such as CER and MSE, that normally lack the necessary time resolution. For this reasons, it is important to understand how to convert LMs and NTMs in QSMs. Specific questions to address are: is the QSM an energetically less-favorable solution of the torque balance equation? Or it only appears under certain conditions, for example depending on the competition of distinct torques (from the wall, from the error field (EF), viscous etc.) having different dependences upon rotation frequency? It is well known that the magnetic friction from the wall and the torque from the EF counteract and partly cancel each other at frequencies of the order of the inverse resistive wall time, consistent with QSMs. It is also theoretically predicted that stronger/weaker EFs can move this energetically favorable torque minimum to lower/higher frequencies. In fact, very strong/weak EF might prevent the QSM from being observed by letting it ā??degenerateā?? in a LM or rapid NTM. This is confirmed by various LM control experiments where QSMs were prevented by brute force (large MPs). By contrast, small (but not too small) MP led to QSM, for certain phases (shot 141052) but not for others, for which probably it was partly canceling the EF (141051). It is speculated that typical EFs are high enough to observe LMs but too small for the QSM solution to exist or be observable. An alternative interpretation is that the PLASMA rotation ā??transitsā?? too rapidly through the range of frequencies at which the MODE likes to rotate. This is analogous to a mechanical system having several modes of vibration. Different eigenmodes will have different eigenfrequencies. Hence, if the frequency of a perturbation (for example a motor, or a piston) is ramped up (e.g. when a car is accelerating), the system will pass through several resonances. However, if the acceleration is rapid enough (=if the system spends at a certain frequency less than a period) the system will rapidly go out of resonance, and a certain mode will not be observed, or be barely observed. Talking of timescales, QSMs often strike 0.3-1s after locking, suggesting that some slow, global resistive effect might be involved in their onset | ||
| Resource Requirements: | -- | ||
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| Other Requirements: | -- | ||
| Title | 515: Enhancement of the Bootstrap Current in a Banana Regime Pedestal | ||
| Name: | Kagan none | Affiliation: | LANL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the bootstrap current near the electric field maximum in a low collisionality pedestal, as well as the electric field itself. By looking at different spatial locations/shots find this current dependence on the radial electric field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Lithium beam diagnostics similar to the one described in D. Thomas et al Phys. Rev. Lett. 93, 065003 (2004). | ||
| Background: | The strong radial electric field, inherent to a subsonic tokamak pedestal, cannot modify drift orbits of electrons as it does for ions, because the poloidal gyroradius of the former is much less than that of the latter. Indirectly, however, electrons do feel the electric field through their friction with ions, whose net flow is substantially modified by this field. A revised expression for the bootstrap current including the effect of the electric field predicts that in a banana regime pedestal this current is larger than it is given by conventional neoclassical formulae. Due to indubitable practical importance of the bootstrap current direct observation of the described effect could strongly impact major tokamak experiments. | ||
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| Title | 516: The Effect of the Radial Electric Field on Neoclassical Flows in a Tokamak Pedestal | ||
| Name: | Kagan none | Affiliation: | LANL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the main ion species poloidal flow near the electric field maximum in a low collisionality pedestal, as well as the electric field itself. By looking at different spatial locations/shots find the net poloidal velocity dependence on the radial electric field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Of course, it is highly desirable to measure the net poloidal velocity of background ions directly. However, even if it is only the toroidal component of the main ion flow that can be measured directly, but at the same time both toroidal and poloidal flow components can be measured for impurities, the main ion poloidal velocity can be recovered quite robustly through the pressure balance equation. | ||
| Background: | Poloidal flow of background ions is neoclassical in nature. In other words, it is due to the drift motion that ion gyrocenters undergo in the tokamak magnetic field line geometry. In a subsonic pedestal of a width comparable to the poloidal ion gyroradius a strong radial electric field arises to maintain pressure balance, making the corresponding gyrocenter orbits substantially different from their core counterparts. As a result, in banana and plateau regime pedestals, neoclassical phenomena become dependent upon the electric field. Most interestingly, a recent first principle study predicts that in the banana regime pedestal the poloidal flow of background ions is reduced in magnitude, or even reversed, compared to what is seen in the core. This prediction was indirectly confirmed by comparing the net poloidal velocity of boron impurities observed in the C-Mod pedestal with the first-principle based expression accounting for the electric field. Either measuring the main ion poloidal flow directly or deducing it through the pressure balance as described in the previous section should provide a more accurate knowledge of this flow. Hence, the proposed experiment would allow further verifying of the electric field effect on neoclassical flows in the pedestal. | ||
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| Title | 517: The Role of Neoclassical Orbits in Establishing the Background Ion Temperature Profile in a Pedestal | ||
| Name: | Kagan none | Affiliation: | LANL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Perform direct measurements of the background ion temperature profiles for a wide range of the pedestal width to the poloidal ion gyroradius ratios. Determine if the discrepancy between the ion temperature and plasma density scales grows as this ratio goes from ~5 to ~0.5. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | A first-principle based analysis finds that in a banana regime pedestal the main ion temperature profile must be much wider than the drift ion orbit; i.e. its characteristic scale must be noticeably greater than the poloidal ion gyroradius. Plasma density does not have such a limitation and, in fact, is found to have a scale comparable to the poloidal ion gyroradius in many experiments. Hence, when the pedestal width to rho_pol ratio is small the ion temperature profile must be much wider than that of the plasma density, whereas once this ratio becomes larger the two profiles are allowed to have similar scales. Direct measurements of the main ion temperature by deGrassie supports this point in the pedestal as wide as (1/2)rho_pol, but comparing the two profiles in the series of shots with pedestal width to rho_pol ratio ranging from ~5 to 0.5 would provide a more solid evidence for the mechanism underlying temperature equilibration in a banana regime pedestal. Clarification of this mechanism is necessary for adequate theoretical description of pedestals. Currently the ion temperature profile is often taken to be as narrow as the pedestal itself when modeling H-Mode, which is justified by impurity ion temperature measurements. However, impurity ion species is more collisional than the main one, making physics behind establishing its temperature profile quite different. It is therefore crucial to measure the temperature of main ions directly rather than to deduce it from that of impurities to elucidate the issue. | ||
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| Title | 518: Plasma response measurements in advanced inductive scenarios | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Core-Edge Integration | Presentation time: | Not requested |
| Co-Author(s): | J. Hanson, H. Reimerdes, T. Petrie | ITPA Joint Experiment : | No |
| Description: | Measure the dependence of the n=3 plasma response on betan and q95 in advanced inductive scenarios with and without ELM suppression. Obtain plasma response measurements to odd and even parity I-coil. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In each discharge, fix betan/li using feedback control of the NBI power. Ramp the plasma current to vary q95. Apply a n=3 I-coil field; modulate the current amplitude (~10-40 Hz) in order to use coherent detection of the plasma response. Use both odd and even parity I-coil configurations. Probe the region q95~7.1 using odd parity and q95~3.6 using even parity. | ||
| Background: | See also proposal #260. It is desirable to identify the plasma parameters and external field structures where ELM suppression by non-axisymmetric fields can be obtained in configurations other than the ITER-similar shaped scenario. Recent MARS-F calculations suggest a possible correlation between the structure of the plasma response and ELM suppression in ISS equilibria. This experiment will obtain data that can be used to compare the measured plasma response in AI discharges with measurements in ISS plasmas. The results would be used to inform future ELM suppression experiments and to validate the perturbed equilibrium model in MARS-F. | ||
| Resource Requirements: | 8 NBI sources, I-coil (n=3); odd and even parity | ||
| Diagnostic Requirements: | Equilibrium diagnostics | ||
| Analysis Requirements: | EFIT, MARS-F | ||
| Other Requirements: | Must first demonstrate n=3 plasma response measurements using modulated fields can be obtained during application of a large DC RMP field. | ||
| Title | 519: RMP ELM suppression in balanced DN plasmas | ||
| Name: | Evans evans@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | E. Lazarus, et al., | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to obtain ELM suppression in balanced DN plasmas. Well controlled, balance double null plasmas will be produced using plasma control algorithms developed in 2010 (ref. 4/1/10 experimental run day). Data from these discharges will be used to validate 3D plasma response models using stellarator codes such as PIES & SIESTA (resistive equilibrium), BOOZ_XFORM, DKES & NEO (neoclassical transport), BOOTSJ bootstrap current, TERPSICHORE (peeling), and COBRA (ballooning). Also, these cases will serve as a starting point for modifying these codes to handle the general 3D case (no stellarator symmetry). At the present time, these codes require stellarator symmetric discharges so they need to be changed to analyze ITER relevant plasma shapes. It is likely these code changes will not be trivial and will introduce numerical instabilities, as in VMEC, that will need to be remedied. Thus, DN RMP plasmas are needed to motivate and validate changes required to ITER simulations. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Using discharge 142599 as a starting point identify conditions suitable for good ELM suppression with n=3 RMP fields (vary q95, Pinj, BT and outer gap). | ||
| Background: | Previous attempts at ELM suppression with RMP fields in DN plasmas have suffered from spontaneous ELM-free periods prior to the application of the n=3 field that resulted in large beta normal excursions and the destabilization of MHD modes. In addition, relatively small n=3 fields caused large density drops. These conditions need to be avoided by reducing the particle exhaust and increasing the injected power. | ||
| Resource Requirements: | I-coil configured for n=3 even parity, upper and lower cryopumps and all 8 NBI sources. | ||
| Diagnostic Requirements: | Pedestal, fluctuation and divertor diagnostics. | ||
| Analysis Requirements: | Kinetic EFITs, pedestal profile analysis, ELITE analysis and modeling with the codes listed in the description section above. | ||
| Other Requirements: | -- | ||
| Title | 520: Mutual alignment of gyrotrons: a new technique based on steering and modulation | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Exploit the fact that two gyrotrons aiming at the exact same target and modulated out-of-phase are equivalent to a single gyrotron operated in cw, to solve the problem of how to align 24 beams in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate 2 gyrotrons at 100 Hz and out-of-phase. Keep one beam fixed and steer the other. When both deposit at the same rho, the situation is indistinguishable from continuous ECH from a single gyrotrons. Hence, no heat-waves will propagate in the plasma. The corresponding signature will be a zero or minimum of heat-pulse height in the ECE.
Repeat for the other two pairs of launchers. Finally, align each pair to the other. To assess the benefits of the improved alignment, aim all gyrotrons at the same target and modulate them in phase with each other. A broader or narrower deposition width is expected in the height pulse analysis of the ECE, depending whether the old or the new angular calibration is used. | ||
| Background: | ITER will be equipped with at least 24 gyrotrons delivering at least 20MW to the plasma. An earlier remote-steering design raised some concern on the ECH/ECCD deposition being too broad and the driven current density too low. This concern propelled a large, international modelling effort, that culminated in a new front-steering design. The new design was more than satisfactory as far as the deposition width of an INDIVIDUAL beam was concerned. ITER, however, will feature 24 gyrotrons. Failing to correctly align them to each other will result in a broadening of the TOTAL driven current that might make the improved design vain, unless an extremely high angular precision in the gyrotrons aiming (0.2deg) is achieved [F. Volpe, J. Phys.: Conf. Series 74, 1409 (2003)].
DIII-D has the highest number of gyrotrons in operation (6), and has been recently equipped with remote steering launchers, hence it is the best environment where to test the mutual alignment of gyrotrons under conditions close to ITER. | ||
| Resource Requirements: | 2-hour block on Thursday night and/or piggyback, e.g. on #307. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | toray, heat-pulse analysis | ||
| Other Requirements: | |||
| Title | 521: Dispersion of Heat Waves | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate that plasma is dispersive for heat waves, i.e. that energy disturbances of different frequencies propagate at different speeds. Measure the dispersion relation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Square-wave-modulate ECH at 30, 100, 300, 1000 and 3000 Hz and analyze heat-waves in ECE. | ||
| Background: | Heat transfer, energy transport, is frequency- and wavelength-dependent. Different frequencies are expected to propagate at different phase and group velocities. This dependence can be studied by studying the ā??dispersionā?? of heat waves generated by modulated ECH. Here by dispersion we mean their w-k relation, which is more easily extracted from their w-v relation, where v is the phase velocity at which a heat wave propagates. This frequency-dependence is also expected to manifest itself in the distortion of a square or triangular wave, as its sine-wave components propagate at different speeds. | ||
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| Title | 522: Dispersion of Heat Waves (Dupl. 521) | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate that plasma is dispersive for heat waves, i.e. that energy disturbances of different frequencies propagate at different speeds. Measure the dispersion relation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Square-wave-modulate ECH at 30, 100, 300, 1000 and 3000 Hz and analyze heat-waves in ECE. | ||
| Background: | Heat transfer, energy transport, is frequency- and wavelength-dependent. Different frequencies are expected to propagate at different phase and group velocities. This dependence can be studied by studying the ā??dispersionā?? of heat waves generated by modulated ECH. Here by dispersion we mean their w-k relation, which is more easily extracted from their w-v relation, where v is the phase velocity at which a heat wave propagates. This frequency-dependence is also expected to manifest itself in the distortion of a square or triangular wave, as its sine-wave components propagate at different speeds. | ||
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| Title | 523: Energetic NBI losses due to n=3 RMP fields | ||
| Name: | Evans evans@fusion.gat.com | Affiliation: | GA |
| Research Area: | General IP | Presentation time: | Not requested |
| Co-Author(s): | T.B.D. | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment asses the confinement of energetic ions injected during neutral beam heating during the application of n=3 RMP fields. Parameter scans will include a range of I-coil currents, parities and phases along with several values of q95, density and NBI configurations (e.g., beam voltage and co versus counter-Ip beam injection). | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Using a standard ITER Similar Shaped plasma apply n=3 I-coil currents steps ranging from 1 to 6 KA. Monitor the outer equatorial wall (R0) between phi=75 and 180 deg. with the LLNL IR periscope camera for changes in the prompt heat flux with each I-coil current step. Start with all co-Ip beams at full voltage then repeat with only counter-Ip beams with full voltage. Using either co- or counter- beams (depending on the the looses observed) repeat with shot-to-shot a variation of q95 (using Ip as a control parameter with fixed BT) between 3.4 and 5.5. Using the q95 with the largest NBI particle losses vary the plasma density from L-mode (with one modulated source) to low density H-mode (with one solid source) to high density H-modes with as many solid NBI sources as possible and gas puffing. | ||
| Background: | Calculations with various energetic particle transport codes in DIII-D discharge 126006 with n=3 RMP fields (G. Kramer, preliminary SPIRAL results) and in ITER discharges with n=4 RMP fields (K. Tani, et al., 2010 Seoul ITPA) indicate the a significant fraction (~5% in ITER) of the energetic NBI particles are lost from the plasma prior to thermalizing. This may have implications for first wall heat loading and L-H power thresholds in ITER. It is important to understand the characteristics of these prompt losses and how they scale with the application of various RMP fields from ELM coils, field-errors and field-error correction coils. | ||
| Resource Requirements: | I-coil n=3 RMPs and 8 NBI sources desired. | ||
| Diagnostic Requirements: | R0 periscope IR camera, pedestal profile diagnostics, energetic particle diagnostics (midplane FILD2 at 165R0), fast plasma cameras, divertor diagnostics and thermocouples. | ||
| Analysis Requirements: | Analysis with fast particle transport codes (e.g., SPIRAL, etc.). Profile analyses, kinetic EFITs, IR heat flux analysis. | ||
| Other Requirements: | -- | ||
| Title | 524: TBM-generated Trapped Electrons and their Effect on ECCD | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Error Field and TBM Mockup Effects | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Confirm existence of stellarator-like trapped electrons due to toroidal asymmetry introduced by TBM. Measure their effect on reduced ECCD efficiency. Could it be a concern for ECCD in ITER? | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Compare co/ctr-ECCD on axis for TBM coil energized at full, half or zero current (6 good shots). Repeat for mid-radius deposition. | ||
| Background: | The ITER TBM will break the toroidal symmetry of the magnetic field. As a result, it will generate new families of trapped particles. TBM-generated trapped ions have already have already raised concern and attracted interest, but electrons will also be trapped. The present proposal aims at investigating peculiarities of TBM-induced trapping as opposed to conventional trapping in a tokamak, with special emphasis on electrons.
Note that, unlike conventional tokamak trapping in the non-uniform BT~1/R field, this trapping is due to a toroidal mirror, as in stellarators. While ā??tokamak trappingā?? tends exactly to 0 at rho --> 0, maps of the TBM magnetic perturbations display finite non-axisymmetries (30-900 G) at all values of rho. Therefore, unlike standard trapping in a tokamak, particles will also be trapped near the magnetic axis. If confirmed, a similar transport, confinement, current drive (and bootstrap?) physics as for standard trapping in a tokamak might apply near the magnetic axis. More general consequences of trapping at all radii (i.e. with no restriction to the axis) are that 1) the ECCD efficiency will be reduced and 2) stellarator effects might be observed, due to the fact that the new trapped populations are similar to those encountered in stellarators. Finally, the study has the promise to test new current drive effects predicted by N. Fisch and named "ponderomotive one-way walls" [Phys. Plasmas 2007]. | ||
| Resource Requirements: | 0.5 day experiment | ||
| Diagnostic Requirements: | MSE | ||
| Analysis Requirements: | Stellarator codes: TRAVIS ray tracing, VMEC equilibrium, bounce-averaged Fokker-Planck code. | ||
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| Title | 525: Modulated ECCD (not in island) to study current diffusion by MSE | ||
| Name: | Volpe fv2168@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Heating & Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Modulate ECCD and study propagation of current perturbations by means of MSE. This is similar to (and inspired by) the well-known technique of modulating the ECH and inferring the propagation of heat waves from ECE. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate ECCD at 31 and 100 Hz (needs to be incommensurable with 60Hz) and analyze MSE to infer propagation speed of current perturbations as a function of the radius. Compare modulated co-ECCD, modulated ctr-ECCD, and a combination of co- and ctr-ECCD modulated out-of-phase in such a way that ECH is constant, in order to better separate current diffusion from heat transport (or, more precisely, its indirect effects on the current profile). | ||
| Background: | As current density j is a sum and integral of contributions of type nqv, transport of current is a combination of particle transport (through n), energy transport (through v), induction and electromagnetic forces on the current filaments. Emphasis in previous measurements at DIII-D was laid on how islands affect/modify the driven current and its diffusion. The present proposal has the more general and basic goal of investigating how current diffuses, and how well (or badly) it can be predicted from known transport coefficients. A good understanding of current transport would allow better predictions of current profiles and thus, indirectly, of q profiles, for example in ITER. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | MSE | ||
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| Title | 526: Fast mirror development for NTM control | ||
| Name: | La Haye rbrtlahaye@gmail.com | Affiliation: | Retired from GA |
| Research Area: | NTM Stabilization | Presentation time: | Not requested |
| Co-Author(s): | J. Lohr, B. Penaflor, R. Prater, A. Welander | ITPA Joint Experiment : | Yes |
| Description: | Show that the PCS can control the ECCD mirror aiming fast and accurately enough to prevent the onset of an m/n=2/1 NTM or lock onto it and suppress it before it grows to large amplitude. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Needs successful checkout of the real-time steerable mirror control of gyrotron aiming, and implementation into the PCS of mirror versions of active tracking (without a mode) and search and suppress (with a mode). Show that an otherwise unstable mode can be raised to higher beta stably with active ECCD tracking by mirror alignment. Show that with ECCD initially off, the PCS can detect the growing mode (capability already exists), can turn on the gyrotrons (capability already exists), use a logic of both active tracking and search and suppress to stop the mode growth and then completely suppress it. | ||
| Background: | ITER relies on 2/1 NTM control by ECCD using mirror steering. DIII-D has pioneered techniques of search and suppress and active tracking but only by promptly changing the plasma position for alignment (not possible in ITER) or by less promptly changing BT (also not possible in ITER). We now have a real-time mirror steering capability that needs to be exploited, particularly to deliver a mode controller to DIII-D AT plasmas. The logic for ITER control described in the above exp approach/plan is described in Figure 9 of La Haye et al. Nucl. Fusion 2009 but has never been demonstrated on any device with any alignment technique. | ||
| Resource Requirements: | Authorization for real-time steering which will allow basic checkout (see 307) at a minimum to be done.
6 gyrotrons. Previous thorough checkout of PCS mirror control. Implementation of PCS mirror versions of active tracking and search and suppress. Faster mirror steering by new fast mirrors (by PPPL in Jan-Feb) and work on ETHERNET/PCS link to reduce latency (concept in hand). | ||
| Diagnostic Requirements: | Standard. ECE and ECEI for island location. | ||
| Analysis Requirements: | Standard. | ||
| Other Requirements: | |||
| Title | 527: Ion temperature fluctuation studies with UF-CHERS | ||
| Name: | Uzun-Kaymak iuzun@metu.edu.tr | Affiliation: | METU - Middle East Technical U |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | R. Fonck, G. McKee, Z. Yan | ITPA Joint Experiment : | No |
| Description: | We will be developing experimental conditions likely to maximize the Ti-fluctuation signal levels for ITG turbulence studies with UF-CHERS diagnostic. Higher signal levels are expected during modest power ECH-heated L-mode discharges, moreover, deuterated methane injection is also a viable option to increase the signal levels during ECH pump-out. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish low density, low plasma current (0.7-0.8 MA), L-mode discharges, and apply moderate power ECH and FW heating to achieve higher signal levels while rotation velocity is kept nominally constant. Repeat shots and scan across the plasma radius. Repeat same procedure at a high density, L-mode discharges to vary the collisionality. These discharges will be long-pulsed, slowly evolving and sawtooth free. | ||
| Background: | In this experiment, the primary goal will be evaluating experimental conditions to measure ion thermal fluctuations with UF-CHERS and assessing turbulence characteristics and transport levels. Our new UF-CHERS diagnostic will provide fast Ti fluctuation measurements as well as fast Carbon density measurements allowing us to quantify the low -k turbulence and n-Ti phase correlations obtained using UF-CHERS and BES across the plasma mid-radii.
Turbulence driven transport models predict enhanced confinement with Ti >>Te. Additionally, Te/Ti ratio is predicted to play a significant role in relative balance between ITG and TEM turbulence. Previous experiments on Te/Ti ratio dependence show a dramatic change in electron and ion transport as a function of Te/Ti, but a little change in long-wavelength density fluctuations. Electron temperature fluctuations have been shown to increase with Te/Ti from A. Whiteā??s and otherā??s CECE studies. In this experiment, in addition to Te/Ti factor, we propose to incorporate a collisionality (high density) variation which might also change the long-wavelength instability mix and potentially affect Ti-fluctuations. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | UF-CHERS, BES, CECE and other fluctuation diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 533: Detachment onset dependence on input power | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor Detachment and Plasma Flows | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the upstream separatrix density where detachment onsets at the inboard and outboard divertor strikepoints. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In LSN configuration slowly raise the density on a shot to shot basis while fully characterizing both the inboard and outboard divertors at each density. The upper density value should have fully detached divertors, both inboard and outboard. Carry out this density scan at several different power levels. The upper power level should be the highest that can reasonably be run. Characterization of the divertors should involve all of the divertor diagnostics, particularly the divertor Thomson scattering diagnostic. This will require development of a divertor sweeping configuration that modifies the divetor plasma as little as possible. Other important diagnostics will include the tangential TV and flow diagnostics. | ||
| Background: | Divertor detachment is very dependent on input power. A 1D model would indicate that the upstream density at detachment onset should be proportional to power flow to the divertor. Some data indicates a weaker dependence. Also fluid modeling has been unable to match upstream density at detachment onset. This experiment would test both 1D models and the more complete fluid codes. Extensive diagnostic documentation would be extremely valuable for code validation. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | All divertor diagnostics, particularly divertor Thomson should be working well | ||
| Analysis Requirements: | Extensive divertor modeling | ||
| Other Requirements: | |||
| Title | 534: Dependence of pedestal pressure on heating source; ECH vs. NBI | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the pedestal parameters at fixed input power, comparing ECH to NBI heating | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In a standard LSN configuration at relatively high triangularity set up an ELMing H-mode discharge with regular ELMs and modest input NBI power that can be matched by ECH power. Regular ELMs are preferred in order to follow the pedestal evolution between ELMs. Over several shots trade off the NBI power for ECH, one ECH gyrotron at a time. Fully document the changes to the pedestal profile and the evolution of the pedestal between ELMs. The ECH power deposition should match the NBI deposition profile as best as possible. The discharge parameters, field and current, should be chosen to make this match as optimal as possible. | ||
| Background: | There is some indication from previous experiments that ECH driven H-mode has a lower pedestal pressure and lower confinement than NBI H-mode at the same power. This effect could have significant implications for ITER and reactor operation. The trade-off between ECH and NBI may not be linear in that only a little NBI power is required to achieve the full NBI parameters. This experiment would look at that dependence. Some care should be taken to match the power deposition profile as well as possible. | ||
| Resource Requirements: | All ECH gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 535: Confinement dependence of H-mode plasmas with impurity-seeding | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | C. Holland, T. Rhodes, L. Schmitz, S. Smith, G. Wang, A. White, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Inject low-Z to medium-Z impurities into standard (or hybrid) ELM'ing H-mode plasmas and examine the response of global energy and particle confinement, local transport and turbulence. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a low-current (~1 MA) hybrid or standard H-mode discharge. Hybrids are desirable for their long duration and lack of sawteeth (142019 could be a reference). Inject neon, argon and/or nitrogen in progressively increasing quantities and examine the turbulence, transport, confinement, and neutron rate response with the fluctuation and profile diagnostics.
These experiments will also support validation efforts by comparing measured turbulence/transport response with predictions from TGLF, GYRO and other codes. | ||
| Background: | Recent experiments in ASDEX have demonstrated improved confinement in discharges that utilize nitrogen seeding to radiatively cool the plasma edge, thereby mitigating damage to the tungsten first wall (Kallenbach et al. PPCF 52 (055002 (2010); IAEA-2010, OV/3-1.) This is suggested to possibly result from a change in critical gradient as a result of increased Zeff. Confinement factors were increased from H(98,y-2)=0.9 to 1.1, and stored energy and neutron rates increased accordingly. No fluctuation measurements were presented and the mechanism is not identified. These results reminiscent of the RI-mode experiments performed on TEXTOR and DIII-D in L-mode conditions, where a significant confinement improvement with injected neon is correlated with a large reduction in turbulence (McKee-PRL-2000). Given the importance of radiative cooling for burning plasma experiments, it will be very important to understand the impacts of impurity seeding on turbulence and transport, along with the potentially beneficial increase in confinement. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | BES (8x8), UF-CHERS, DBS, CECE, FIR, PCI, etc., full profile (CER, TS, MSE) | ||
| Analysis Requirements: | Transport, TGLF, GYRO, GEM, etc. | ||
| Other Requirements: | |||
| Title | 536: Validation of TokSys a priori Simulations of DIII-D Plasma Control | ||
| Name: | Humphreys none | Affiliation: | GA |
| Research Area: | Integrated and Model-Based Control | Presentation time: | Not requested |
| Co-Author(s): | M. Walker, M. Wade | ITPA Joint Experiment : | No |
| Description: | The goals of this several hour to 1/2-day experiment are to produce experimental validations of a priori TokSys simulations against DIII-D plasma responses. Various position/shape control perturbations will be performed, and the results will be compared with a priori simulations using TokSys simservers. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Identify a standard control reference equilibrium from previous data in which the plasma resistivity can be determined accurately. Reproduce this equilibrium in 2-hour Thursday evening window to confirm resistivity and transport characteristics. Run Simserver(s) to predict plasma response to various programmed position and shape perturbations (R, Z triangle waveforms; X-point perturbationsā?¦). Execute experiment in second 2-hour period or Ā½ day and program waveforms as in simulation. Compare with predictions. | ||
| Background: | Simserver simulations constructed from TokSys objects and modeling tools have been extensively validated against experimental responses and used for control design over the last decade. However, the user base for these tools is expanding, and is expected to increasingly include DIII-D physics operators this year. Updated demonstration of the usefulness and validity of these simulations will support the general use, as well as the further development and maintenance of the tools. | ||
| Resource Requirements: | 0-4 beams (co) | ||
| Diagnostic Requirements: | MSE, 2-5 kHz magnetics sampling, Thomson | ||
| Analysis Requirements: | standard EFITs, TokSys/Simserver with GSPERT nonrigid linear model, Corsica (or equivalent nonlinear model/simulation) | ||
| Other Requirements: | |||
| Title | 537: Disturbance Characterization and Control for ITER | ||
| Name: | Winter none | Affiliation: | ITER IO |
| Research Area: | General Plasma Control and Operation | Presentation time: | Not requested |
| Co-Author(s): | D. Humphreys, J. Leuer, M. Walker, A. Welander | ITPA Joint Experiment : | No |
| Description: | The goals of this 1-day experiment are to study key disturbances being used as fiducials to specify performance of the ITER control system. ELMā??s, H-L transitions, and locked modes will be characterized and their controllability assessed with varying feedback gains. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Using standard ITER target, produce different disturbances (type I ELMā??s, H-L transitions, and locked modes) with varying vertical control characteristics (with and without F2A/B, varying gains on outer coils). Increase elongation in steps and repeat disturbances with largest vertical displacements and best reproducibility (likely ELMs) with and without F2A/B to determine controllability boundaries. I-coils may be used for braking in order to lock moderate amplitude NTMā??s with low disruptivity. | ||
| Background: | Design of the ITER PF control system must make use of fiducial disturbance models to specify performance. For example, the ITER type I ELM model being used to specify control performance consists of a sudden drop in poloidal beta of up to 0.1, with a rise in internal inductance of up to 0.05. The ITER PF control system must preserve stability, preserve strike point geometry, and prevent the plasma from contacting the first wall in the presence of such disturbances. The reliability of ITER control designs depend on the accuracy of these disturbance scenarios, which have not been validated on many devices. Locked modes in particular have not been characterized in validated models for ITER designs, and are not yet routinely included in PF control designs. Preliminary analysis of DIII-D locked modes suggests that these may produce vertical stability disturbances among the most challenging likely to occur in ITER. This experiment is needed to quantify and validate controllability of locked mode and other key disturbances to help guide ITER designs. | ||
| Resource Requirements: | 2-4 co-beams, I-coils | ||
| Diagnostic Requirements: | 5 kHz magnetics, Thomson, MSE | ||
| Analysis Requirements: | Simservers for control development, EFITs | ||
| Other Requirements: | |||
| Title | 538: Collisionality and q95 scaling of Zonal flow in L mode plasma | ||
| Name: | Yan yanz@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | G. McKee, George Tynan and P. Diamond | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to vary q95 and density in L-mode plasma just below the LH transition power threshold and investigate how the GAM and ZMF zonal flow scales with density and q95. This will help identify and characterize GAM/ZMF zonal flow properties and their potential roles in triggering the L-H transition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main goal of the experiment is to perform a density and current scan (to vary q95 at constant B_t) in L mode with the injected neutral beam power just below the L-H transition power threshold to investigate the GAM/zonal flow scaling with these parameters. Input power will be kept constant during the scan and tuned to be just slightly below the threshold. The plasma will be operated in upper and lower single null configurations, USN to allow for a higher L-H transition power threshold, and LSN to more closely represent ITER like conditions. The 8x8 2D BES array will be used to measure density fluctuations and turbulence flows (GAM, ZMFZF), and can be scanned radially during repeat shots near the edge and SOL regions. DBS and ECE will also used to provide turbulence measurements. | ||
| Background: | The GAM has clearly been identified experimentally in tokamak plasmas [1]. There are also experimental indications of the existence of the ZMF zonal flow [2]. Previous experiments have observed a q95 dependence of the GAM at constant density and input power [3], qualitatively consistent with theoretical predictions and some simulations. It is also predicted that increasing the ion-ion collisionality (by varying density) will damp zonal flows and/or GAMs, potentially leading to higher ambient turbulence and transport levels [4]. The existence of GAM/ZMF zonal may play an important role in the LH transition, as suggested by recent measurements with Doppler Reflectometry on ASDEX-U [5]. Therefore it would be very useful to document the density and q95 scaling of GAM/ZMF zonal flow. The newly upgraded and wider-field BES diagnostic also provides larger spatial coverage and better signal to noise ratio for these measurements.
[1] G. McKee, et al PPCF 45 A477 2003 [2] D. Gupta, et al PRL 97 125002 2006 [3] G. McKee, et al PPCF 48 S123-S136 2006 [4] Z. Lin, PRL 83, 3645 1999 [5] G. Conway, IAEA, 2010. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | BES, DBS, CECE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 539: High time-resolution profile and turbulence measurements across LH transition | ||
| Name: | Yan yanz@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Transport: Other | Presentation time: | Not requested |
| Co-Author(s): | G. McKee,George Tynan and P. Diamond | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to obtain a radial profile and timeā??resolved turbulence measurements across LH transition, which will provide investigations of the role that turbulence, zonal flow shear and mean ExB shear play during the LH transition, and the connection between core turbulence and transport and the rapidly evolving edge pedestal at/after the L-H transition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main idea of this experiment is to get full radial profile and high time ā?? resolved turbulence and plasma profile (ion, electron temperature and density) measurements across LH transition at co, counter and balanced beam momentum input. The BES diagnostic is essential and will be configured with two 2x16 2D array separated poloidally. This will provide both turbulence and zonal flow profile measurements during the LH transition. The zonal flow shear will be compared with mean ExB shear to investigate similar (or different) role they play before, at and after the LH transition. This could also provide opportunity of studying the core/edge correlations. The turbulence suppression dynamics from the core to edge will be compared with the simultaneous measurements, and also correlated with the local profile gradient variation at the transition. The CER and Thomson diagnostics will be deployed to obtain high time-resolution measurements across the transition. | ||
| Background: | It is well known that edge turbulence is rapidly suppressed at the L-H transition (in less than 100 microsec), and more recent measurements indicate that the core turbulence is also suppressed on a relatively fast time scale (< 10 ms), suggesting a link between the edge and core turbulence. It is not clear that profile changes are rapid enough to explain the observed changes. Obtaining high time-resolution measurements of core profiles across the transition will help elucidate whether local gradient changes can explain the fast turbulence evolution, or whether a turbulence spreading mechanism may be at play between the edge and core. In addition, the existence of GAM and the zero-mean-frequency (ZMF) zonal flow has been clearly identified experimentally in tokamak and stellarator plasmas [1,2]. Those flows are predicted to be generated by the plasma turbulence and may relate to the shear flow mechanism thought to drive the L- to H- mode transition [3]. However the interaction between spatial and time resolved turbulence, GAM/ZMF zonal flow and mean ExB flow across LH transition hasnā??t been addressed in detail.
[1] G.R.Mckee, et al., PoP, 10, 1712, (2003) [2] A.Fujisawa, et al., PRL 93, 165002 (2004) [3] K. H. Burrell, PoP 4, 1499 (1997) | ||
| Resource Requirements: | 7 neutral beams | ||
| Diagnostic Requirements: | BES, DBS, CECE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 540: Joint experiment on RWM stabilization physics | ||
| Name: | Sabbagh none | Affiliation: | Columbia U |
| Research Area: | RWM Physics including Rotation Dependence | Presentation time: | Not requested |
| Co-Author(s): | J.M. Hanson, J.W. Berkery, M.J. Lanctot, H. Reimerdes, I.T. Chapman, M.A. Van Zeeland, W.W. Heidbrink, M. Okabayashi, E.J. Strait, Y. In, R. La Haye | ITPA Joint Experiment : | Yes |
| Description: | Leverage differences between DIII-D, NSTX, MAST, and other devices (e.g. data from JT-60U) to test the theory of kinetic RWM stabilization (B.Hu, et al. PRL 93 (2004) 105002) which has shown to reproduce experimental results on both NSTX (Berkery PRL 104 (2010) 035003; Sabbagh NF 50 (2010) 025020; Berkery PoP 17 (2010) 082504) and DIII-D (Reimerdes IAEA 2010 paper EXS/5-4). This is in contrast to past models of RWM stabilization that may have explained RWM stability physics in individual devices, but that have failed to unify results, as has been recently shown (Sabbagh IAEA 2010 paper EXS/5-5). A companion experiment will be proposed on NSTX, following similar experiments conducting in NSTX in 2008 and 2009. The key variable examined in each device, and across devices, will be the fast ion distribution. This can be varied in DIII-D several ways, including the new capability of off-axis NBI. In addition to minimizing RWM stability in DIII-D as conducted by Reimerdes, et al. in MP20101402. The approach to marginal stability by using n = 3 non-resonant NTV braking (Zhu, PRL 96 (2006) 225002; Sabbagh, PRL 97 (2006) 045004) as has been conducting in NSTX for several years, will also be used to best compare results between devices. Active MHD spectroscopy will be used to experimentally evaluate RWM stability in general, and the experiment will attempt to generate a linearly unstable RWM similar to results published from NSTX. The MISK code, used for analysis of both NSTX and DIII-D in the publications above, will continue to be a common tool for analysis of this joint experiment. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Discharges similar to those used in MP20101402 will be used to best approach the point of RWM linear instability, as was diagnosed in that experiment by active MHD spectroscopy. Variations to fast ion profile will made from this point by changing NBI mix, especially using the new off-axis NBI capability, plasma density, and plasma current at fixed q. An equivalent, clean dataset varying Ip at fixed q was obtained in NSTX XP1020 in 2010, and serves as an initial comparison. Non-resonant fields with dominant toroidal mode number = 3 will be used to vary the plasma rotation profile as well, in more similar ways to NSTX in an attempt to access less stable rotation profiles as indicated by the MISK code. | ||
| Background: | Further testing of the Hu-Betti RWM stabilization, which is being tested quantitatively in present analysis, is critical for extrapolation to future burning plasma experiments. References to existing publications in testing this model are given in the Description section above. One manifestation of the importance of this study is that the ITPA joint experiment MDC-2 is now focused on the validation of RWM stability models and cross-comparison of analysis codes. This experiment would add directly to this task The Hu-Betti model presently does well in describing detail of the RWM marginal stability point in NSTX, especially the observation of RWM instability at relatively high plasma rotation (Berkery PRL 104 (2010) 035003). In NSTX, we are now at the level of examining omissions of terms in the model that might account for ~ 10 - 20% variations in gamma*Tau_wall. In DIII-D, The ability to generate the linearly unstable RWM in DIII-D without off-axis fishbone instability is still a matter of discussion, and a separate goal of this experiment. Note that the generation of unstable RWMs in DIII-D after the onset of off-axis fishbones may be due to a change in the stabilizing fast ion population, which is also consistent with the Hu_Betti model and embodied in the MISK code. | ||
| Resource Requirements: | Co-directed plasma current. 8 NB Sources required. I-coil and C-coil use (n = 1 error field correction, n = 1 active MHD spectroscopy, n = 3 NTV magnetic braking. High performance plasmas to ensure high beta operation. | ||
| Diagnostic Requirements: | Equilibrium diagnostics including kinetic profiles (Thomson, CER, MSE) and fast ion diagnostics. | ||
| Analysis Requirements: | TRANSP/ONETWO calculations of fast ion pressure, Kinetic EFIT, MISK, MARS-K analysis. HAGIS code analysis will be requested (I. Chapman). | ||
| Other Requirements: | |||
| Title | 541: Effect of the fast-ion pressure gradient on Alfven Eigenmode stability | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Energetic Particle | Presentation time: | Not requested |
| Co-Author(s): | EP working group | ITPA Joint Experiment : | Yes |
| Description: | Use the off-axis neutral beam to alter the fast-ion pressure gradient grad beta_f. Measure the stability threshold for TAEs and RSAEs | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use a standard AE current-ramp discharge such as 142111. Do beam power scans to find the stability threshold. Use the 150 beams (at maximum off-axis angle) to flatten beta_f. | ||
| Background: | In standard TAE theory, the fast-ion drive term is proportional to grad beta_f. This would be a direct experimental test of the validity of this term. | ||
| Resource Requirements: | All NBI sources desirable. | ||
| Diagnostic Requirements: | Only need the fluctuation and neutron diagnostics for this experiment but, since it will likely be combined with other studies, all fast-ion diagnostics are desirable. MSE essential. | ||
| Analysis Requirements: | NOVA-K | ||
| Other Requirements: | |||
| Title | 542: ITER ELM Pacing using NB | ||
| Name: | Leuer leuer@fusion.gat.com | Affiliation: | GA |
| Research Area: | Atternative Techniqes for ELM Control | Presentation time: | Not requested |
| Co-Author(s): | J. Ferron, M. Fenstermacher, T. Strait | ITPA Joint Experiment : | No |
| Description: | Perhaps the most overlooked ELM pacing actuator for the DIII-D ITER Demo discharges is the Neutral Beams. In many ITER discharges the beta level, beta/NB feedback and timing of the diagnostic beam team-up to regulate the ELM frequency to almost exactly 10Hz. We propose varying the NB Beta/feedback parameters, NB characteristics and NB timing to regulate (pace) the ELM frequency. We will determine the ELM frequency and magnitude relative to beam average power level, oscillatory frequency and magnitude and beam characteristics (like left/right, co/counter, voltage). We will scale the results to concepts for implementation and utilization in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish type I ELMā??ing H-mode discharge, preferably using the ITER Scenario 2 shape and conditions. Establish constant power output beam power necessary to maintain Beta (use typical beta feedback to establish average power). First order experiment is oscillating the magnitude and frequency of injected power input about the mean; 2nd order is change mixture of NB actuators (i.e. left, right, co, counter, voltage). Overall we will characterize ELM frequency and magnitude with 1) average beam output, 2) beam oscillatory frequency, 3) oscillatory magnitude, 4) mix of different source type, and 5) other NB parameters.
Some preliminary results could be obtained in piggyback by changing the PCS beam feedback characteristics and diagnostic beam timing. | ||
| Background: | In establishing a ā??disturbanceā?? model for ITER from our DIII-D ITER Demo discharges it was observed that for almost all ELMing ITER discharges the ELM frequency is highly regulated by beam timing. For our standard reference ELMing ITER discharge (131498 Doyle IAEA08) the ELM frequency is seen to lock into exactly 10Hz with phase exactly every on the tenths time (i.e. .1, .2, .3 s) during a good fraction of the pulse (See Fig. 1 below). It seems associated the beam/beta feed back frequency and the natural ELM frequency being of order 10Hz and locks into the pulsing of the 30L diagnostic beam every 100ms (i.e. .1, .2, .3 s ā?¦). In this discharge the diagnostic beam is pulsed every 100ms (i.e. 0.1, 0.2, 0.3s => 10Hz) and the system locks into this pulse. The ELM almost always occurs at the minimum of the beam power input. This phenomenon is observed in many ITER discharges and is seen in some plasmas with higher frequency in the diagnostic beam (see 131722). There are a number of examples where ELM frequency is increased when overall beam frequency characteristics are increased; and some where this does not happen. The bottom line is the Beams are the best actuators for ELM pacing and control in the DIII-D ITER demo discharges. We propose to establish the relationship between beam parameters, especially frequency to determine ELM level and response.
ITER Relevance: It is also observed that the elm occurs at approximately the minimum of the beam power input and this would indicate that this method could quantitatively be applied to ITER. The overshoot/undershoot nature of the beam feedback would indicate that a similar oscillation in the ITER NB (or ECH) power input could regulate the ELM frequency. Since ITER is close to the H-L threshold power at in its nominal operation this would mean we would be just sliding slightly up and down the H-L transition to get the ELM frequency what we want. The question is how much power oscillation is needed and what control is needed to regulate the ELM frequency. Main question is to determine what degradation in performance would be expected at different levels and what beam power oscillation is acceptable to ITER relative the main power being input by Alpha heating. | ||
| Resource Requirements: | ELMing H-mode with Type I elms, preferably in the ITER configuration. Could piggy back some of this on ITER Sc2 discharge days by changing Beam/Beta feed back. Dedicated experiments of 2 x half days for oscillatory beam experiment. | ||
| Diagnostic Requirements: | Fast magnetics, MSE, Thompson Scattering, Spread, Visible camera, Bolometers, IR camera, CO2 Interferometer, ECE, SXR, Edge diagnostics | ||
| Analysis Requirements: | efit, transport | ||
| Other Requirements: | |||