| Title | 1: C vs I Coil NTV | ||
| Name: | Smith smithsp@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Study plasma rotation braking with C- and I- coils ramped up separately. Compare to calculations of NTV torque using M3D-C1 calculations of the plasma response. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Keeping other external torques constant, ramp up the current in the C- and I- coils in small steps. | ||
| Background: | Recent calculations of the NTV torque in QH-mode show a very different plasma response to the I- and C- coils | ||
| Resource Requirements: | All torque/heating possibilities. | ||
| Diagnostic Requirements: | CER, Thomson, MSE. | ||
| Analysis Requirements: | M3D-C1 calculations of the plasma response. | ||
| Other Requirements: | -- | ||
| Title | 2: Hysteresis and Turbulence Spreading in the H-L Back Transition | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G.R Tynan, G.R. McKee, Z. Yan, P.H. Diamond, L. Zeng, J.A. Boedo, T.L. Rhodes, E.J. Doyle | ITPA Joint Experiment : | No |
| Description: | For burning plasma operating near the LH power threshold, understanding the hysteresis and spatio-temporal evolution of the H-L back transition is crucial. In addition, if back transitions can be controlled or the transition sequence slowed down, safe ramp-down of the current and beta-pol without undue stress on the poloidal field system/plasma control may be achieved. The goal of this experiment is to understand the observed hysteresis in the L-H and H-L transition power, and the feedback cycle that potentially controls the back transition. The proposed experiment will investigate the evolution of the pedestal radial profiles (via fast profile reflectometry) and will allow allow mapping of ExB flow velocity and density fluctuation level profile (via DBS/BES), across the LCFS, with high time/spatial resolution during H-L back transitions. This will elucidate the back transition mechanism and allow an assessment of turbulence spreading. The understanding gained may suggest a possible approach for inducing/controlling the back transition sequence in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Two scenarios will be explored; an ITER-similar shape, and a low triangularity, low P_th shape (#149725) where limit cycle back transitions have been reliably obtained previously and probe insertion is feasible. We will use a) a combination of several de-rated beams to obtain a gradual reduction in P_NBI; b) a shape change - (either USN to LSN, or varying the x-point height) to induce a back transition during ELM-free H-mode.
It is important to avoid the customary type-I ELM trigger observed in many H-L back transitions, to decouple ELM/MHD and turbulence dynamics. Turbulence/flow coupling will also be investigated via the reciprocating probe (inserted up to ~ 1 cm inside the LCFS once repetable back transitions are obtained). The experiment will be carried out starting with two different L-mode densities, n. 2x 10^13 cm^-3, and 4-5x10^13 cm^-3. A dedicated experiment is needed to achieve H-L transition not triggered by ELMs and to allow a sufficient number of repeat shots to accommodate the midplane probe and DBS/BES diagnostic mapping. | ||
| Background: | The detailed H-L back transition process is not sufficiently understood. The spatial/temporal evolution of the pedestal radial profiles and turbulence characteristics during the back transition can now be addressed on DIII-D with superior diagnostic capabilities, and will likely add substantially to our understanding of forward L-H transitions. Knowing the L-H and H-L transition power hysteresis and its dependence on density is likewise important to ensure reliable H-mode operation. This experiment builds on previous L-H transition experiments focused on forward transitions, in a low triangularity shape favoring extended limit cycle transitions, and in an ITER-similar shape. | ||
| Resource Requirements: | Resource Requirements:30,330,150 Beams, possibly ECH | ||
| Diagnostic Requirements: | BES large array centered on pedestal region, DBS-5,DBS-8, midplane probe in the low triangularity scenario | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 3: The effect of LSN Shape on the q95 windows in RMP ELM Suppression | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The purpose of this experiment is to understand what selects the q95 windows that work for RMP ELM suppression (n=3) by using systematic shape changes in LSN discharges. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start with a standard LSN (ITER-like) shape with standard RMP suppression (q95 ~ 3.5, double I-coil row n=3) and vary the shape while measuring the q95 window of suppression. Vary triangularities, and squareness values. Use Ip ramps. | ||
| Background: | We know that shape modifies the q95 windows in LSN discharges ( see Schmitz et al, NF 52, 043005 (2012) ), from these experiments done on two distinct LSN shapes. But no systematic experiments have been done to map out the q95 regions with continuous shape variations to understand the trade-offs between the local resonance (outboard field line pitch) and the global resonance (q95 value). | ||
| Resource Requirements: | 1 day experiment 2 days if compelling results obtained. Standard DIII-D RMP experiment, operational hardware and diagnostics. Consider the trade-offs between using lower BT (more I-coil current headroom), diagnostic coverage, and possible use of ECH. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 4: Turbulence and Intrinsic Rotation | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | George McKee | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to connect intrinsic rotation in DIII-D with turbulence. We hope to test if the turbulent spectral features such as frequency, magnitude and correlation lengths are coupled with the magnitude and direction of the intrinsic rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use shape variations to modify the measured turbulence spectra in ECH-dominated H-modes and measure the intrinsic rotation profiles. | ||
| Background: | It has been shown that the BES turbulent spectra depend upon the plasma squareness, done using a DND shape in DIII-D. (See Holcomb PoP 16, 056116 2009). Theories of the generation mechanism for intrinsic rotation in the interior of a tokamak invoke plasma turbulence, and indicate that the direction of the rotation drive could depend upon the flavor of turbulence, e.g. TEM vs ITG. We want to measure whether there is any signature change in the intrinsic rotation profiles with changes in the turbulence spectra induced by shape change | ||
| Resource Requirements: | 1 day experiment. Standard Intrinsic Rotation experiment with ECH H-modes; need all gyrotrons and modest NBI for diagnostics. Beyond diagnostic NBI blips, the only NBI drive required will be the minimum needed for BES, and thus possibly a balancing counter beam. All possible compatible turbulence diagnostics should be used. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 5: Measure Intrinsic Rotation Size scaling in DIII-D alone -II | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | Wayne Solomon, Brian Grierson, 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 Alfvén 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. Nice to have main ion CER, where the shape allows coverage. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 6: Prompt torque and zonal flow damping | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | 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 an 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. 6-8 NBI sources. All ECH gyrotrons | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 7: Investigate angular momentum diffusion and pinch using off-axis torque | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | 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. This experiment is best done in reverse Ip plasmas because the orbit shift of the 150 beam particles is outward, leading to a prompt torque input that is further off axis. | ||
| 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 for rotation measurements. Complete profile diagnostics for background plasma profiles | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 8: Compare Mach probe and CER measurements of intrinsic rotation profile of main ions in helium plasma | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | J. Boedo, J.S. DeGrassie, W.M. Solomon, B.A. Grierson | 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: | Helium plasmas with ECH | ||
| Diagnostic Requirements: | Mach probe. Standard profile diagnostics including CER. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 9: Test of Neoclassical Toroidal Viscosity theory using modulated I-coil currents | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, W.M. Solomon, S. Smith | 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 create maximum nonresonant n=3 magnetic field. C-coil configured for error field correcton. 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. Use both main ion and carbon CER measurements. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 10: Maintain low-rotation QH-mode with ECH only | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | 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: | The target plasma for this work will be the NRMF sustained QH-mode with net zero NBI torque which as been developed over the past several years. We will established the QH-mode 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 remains in QH-mode with the NRMF assist. | ||
| 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 2009-2012 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque or small co-Ip 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 or slightly co-Ip for these shots, there were still 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. 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. 6-7 gyrotrons. C-coil configured for maximum n=3 field, 7 kA current. I-coil configured for error field correction and as much n=3 field as possible. | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 11: Investigate in-out density asymmetry at large toroidal rotation | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | 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 was 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: | Reverse Ip for QH-mode. Proper beam modulation to get best CER measurements inside magnetic axis. | ||
| Diagnostic Requirements: | All CER chords including those viewing points inside magnetic axis. All profile diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 12: Diagnostic spatial cross calibration using edge sweeps in QH-mode | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | 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 now that we have the benefits of improved spatial calibrations using the component measurement arm. Getting the spatial cross calibration of the MSE views of the 30 and 210 beams is an essential first step in using MSE to determine the edge current density profile. | ||
| Resource Requirements: | Reverse Ip. 8 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 | 13: 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: | Inductive Scenarios | 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 fluorine into these plasmas using either LiF pellets or tetrafluoromethane gas so that we can study impurity transport with a non-recycling impurity. Use injection 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. 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 fluorine is that it does not recycle, thus allowing direct measurement of the edge loss rate from the decrease of core density of fluorine. If we want a heavier, non recycling impurity, we could use chlorotrifluoromethane. | ||
| Resource Requirements: | Either lithium pellet injector or standard D III-D gas injector with fluorocarbon gases. | ||
| Diagnostic Requirements: | All standard profile and fluctation diagnostics, especially edge BES and ECE-I for EHO studies.CER tuned to F lines for impurity transport study | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 14: Control of toroidal mode number of EHO | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, M. Lanctot | ITPA Joint Experiment : | No |
| Description: | Investigate how the toroidal mode number n of the edge harmonic oscillation(EHO) changes with plasma shape, q, density and n-number and parity of the externally imposed nonresonant magnetic field (NRMF) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Run QH-mode plasmas and make systematic single parameter scans to see the effect of each parameter on the n-number of the EHO. For example, in shots with the same shape as 141439, use odd parity, n=3 NRMF from the I-coil and
systematically vary I-coil current and find the threshold where the EHO switches from n=1 to n=3. For the investigation of the q dependence, we will need both current and toroidal field scans. Experiments in 2011 have shown that toroidal field scans can be done dynamically during one shot. Another variable to investigate is the n-number of the NRMF to see if n=2 NRMF has the same n-number changing effect on the EHO that the n=3 NRMF does. If there is an effect of the n=2 NRMF, then we need to investigate whether this can be further optimized by changing the phasing between the upper and lower set of I-coils. | ||
| Background: | The edge harmonic oscillation (EHO) is a key feature of the QH-mode which provides the extra particle transport to allow the plasma to reach a transport steady state at edge parameters below the explosive ELM limit. Experiments in
2011 and 2012 revealed that control of the toroidal mode number of the EHO is an important part of running QH-modes with low or co-Ip torque. If the EHO has a toroidal mode number n=1, it can lock to the wall at low rotation. EHOs with n=2 or greater or the broadband MHD do not have this problem. Over the years, we have empirically developed techniques to change the n-number. These include increasing the plasma density, changing the plasma shape, altering the edge q and using odd parity n=3 NRMF. One of the most fascinating observations is the switch of the EHO from n=1 to n=3 when odd parity NRMF is applied to the plasma. However, we have never systematically investigated these to establish the range over which they work and, more importantly, how to optimize them. The goal ofthe present experiment is to perform that systematic investigation. | ||
| Resource Requirements: | Reverse Ip. 8 NBI sources. | ||
| Diagnostic Requirements: | All profile diagnostics. Fluctuation diagnostics for edge measurements of EHO, especially BES and ECE-I | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 15: Edge rotation shear threshold for QH-mode access | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | A.M. Garofalo | ITPA Joint Experiment : | No |
| Description: | QH-mode requires sufficient shear in Er/RBtheta. A key question is what is the proper frequency to compare this to in order to produce a nondimensional quantity. Basic MHD theory suggests the Alfven frequency is the correct one for comparison. This experiment seeks to test that. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will be done in reverse Ip QH-mode plasmas with shape similar to that in shot 141439. The threshold in Er/RB_theta for the QH-mode will be found at each condition by varying the rotation and, hence, Er by changing the input neutral beam torque. The density and toroidal field will be varied and the threshold found at various conditions. Edge sweeps will be used in order to improve the spatial resolution of the Er measurement. Data in QH-mode and ELMing H-mode will be obtained. Data both with and without NRMF fields will also be obtained. | ||
| Background: | Shear in the edge toroidal rotation frequency associated with the E x B drift Er/RB_theta has been shown to exceed a threshold when the plasma is in QH-mode. This shear correlates much better with the existence of QH-mode than the shear
in the edge carbon rotation speed. In plotting the shear in Er/RB_theta, previous work chose to normalize to the Alfven frequency, since that frequency plays a key role in many MHD modes. Using this normalization, data from shots with different Alfven frequencies all showed the same threshold. However, we do not know if the Alfven frequency normalization is really correct. This experiment is designed to check that normalization by studying the threshold as a function of density and toroidal field, which are the physics variable that enter into the Alfven frequency. | ||
| Resource Requirements: | Reverse Ip. 6-8 neutral beam sources. | ||
| Diagnostic Requirements: | All profile diagnostics, especially edge CER. Fluctuation diagnostics optimized for edge views of the EHO, especially BES and ECE-I. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 16: Further development of QH-mode with strong co-Ip NBI torque | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | T.H. Osborne, P.B. Snyder, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | Use systematic, theory-guided parameter scans to broaden operating range for QH-mode with strong co-Ip torque discovered in 2008 and produced serendipitously in 2011 | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The set of experiments listed here are designed to 1) optimize QH-mode operation under the conditions used in the 2011 experiments and to 2) broaden the QH-mode operating space. The discharges will start from the conditions of shots like 147293 and 147354
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,3) determine minimum NBI torque which can sustain co-injected QH-mode without using NRMF 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. | ||
| Background: | QH-mode with all co-injection was discovered during serendipitously during the 2008 campaign and a dedicated experiment was performed for one day. In 2011, we again serendipitously produced co-injected QH-mode without NRMF at NBI torques as low as 2 Nm. We have just barely begun the investigation of the QH-mode with strong co-Ip torque. The goal of the present proposal is to use our knowledge of QH-mode with counter-Ip NBI to find ways to broaden the QH-mode operating space with strong co-Ip NBI so that this 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 Bt operation. 8 NBI sources. Recent boronization with deuterium carrier gas. | ||
| Diagnostic Requirements: | All profile and edge fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 17: 3-D Fields and ECH Density Pumpout | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & 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, varying beam mix, at varying levels of applied 3-D field, or error field if there is an effect for small levels of perturbation. | 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, I-coils,C-coils. | ||
| Diagnostic Requirements: | detailed kinetic profiles, reflectometry profiles, rotation profiles, main ion rotation valuable | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 18: High Collisionless NBI Torque Drive for GAMs, aka the VH-mode path? | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | George McKee, Terry Rhodes | ITPA Joint Experiment : | No |
| Description: | *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 | 19: Effect of n=1 applied and error fields on n=3 RMP ELM suppression | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Carlos Paz-Soldan, Matt Lanctot | ITPA Joint Experiment : | No |
| Description: | This experiment will investigate the effect of error fields on the q95 windows in standard shape LSN n=3 RMP ELM suppression. Controlled scans and variations of the added n=1 field will be tested by varying the level of C-coil error correction, and adding n=1 field perturbations to the C-coils with varying amplitude and toroidal phase. The goal is to determine to what extent, if any, the n=1 field (background) is determining the standard n=3 ELM suppression q95 windows. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A reference n=3 RMP ELM suppression condition will be established. First, an added n=1 rotating perturbation will be added to the C-coils and any effect on an otherwise clean window of ELM suppression will be measured. The perturbation amplitude will be raised until locking, or the return of ELMs takes place. The results will guide the amplitude and phase of the n=1 parameter search to be conducted. Ip ramps will be used to look for variations in q95 suppression windows. | ||
| Background: | * As yet, no predictive model tells us where the windows of RMP ELM suppression will take place in DIII-D. Some hypotheses are that a) a resonant surface must be located at the ??correct? place relative to the electron pressure pedestal, b) that perhaps n=1 error fields are in some way giving preference to q95 windows approaching q95 ~ 3/1, c) that the kink response of the plasma is favoring the low q95 region for suppression windows.
* This experiment seeks to investigate the n=1 error hypothesis. | ||
| Resource Requirements: | One day. Standard for RMP n=3 ELM suppression. ECH available | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 20: L-H transition Trigger Physics/ Main Ion Momentum Balance | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G. Tynan, G.R. McKee, Z.Yan, L. Zeng, J.A. Boedo, T.L. Rhodes, E.J. Doyle | ITPA Joint Experiment : | No |
| Description: | This experiments attempts to simultaneously measure all terms in the radial main ion momentum balance across the LH-transition with high time/spatial resolution, including the Reynolds stress, main ion poloidal and toroidal velocity, and ion diamagnetic flow component as well as total ExB flow. Helium plasma will be used to enable main ion edge CER. The goal is to establish simultaneously and with high time/spatial resolution the turbulence fields (,E_pol~), turbulent eddy topology changes (via BES), the Reynolds stress radial gradient, and the response of poloidal/toroidal ion velocity, in order to identify/characterize the L-H transition trigger. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Low triangularity, LSN He plasmas near power threshold will be used to allow probe measurements up to ~ 1 cm inside the LCFS (reference shots #140439 and #149725). The outer gap/ upper triangularity will be modified to achieve optimum spatial CER coverage of the edge/pedestal region.
The main diagnostic challenge is to optimize the CER SNR so that the expected initial poloidal/toroidal ion velocity excursion can be observed as the L-H transition sequence is triggered. Both "regular" and limit cycle transitions will be used. Extended (500 ms) limit-cycle transitions (ref. #149725 in D_2 plasma) allow phase-lock analysis of spectroscopic ion flow velocities, turbulence level, and ExB velocity (via DBS/BES) to substantially improve SNR and time resolution. This should allow us to obtain turbulence, flow, stress gradient, and nonlinear energy transfer data (from probe data) with sufficient accuracy to test against recent L-H transition models. | ||
| Background: | Recent experiments in DIII-D, EAST, and HL2-A have measured the Reynolds stress radial gradient and the rate of nonlinear energy transfer from the turbulence spectrum. These experiments were carried out in limit-cycle L-H transitions, where the transition dynamics can be investigated on an expanded time scale. However, DIII-D has the unique advantage of high spatial resolution diagnostic coverage via BES, DBS, and CER. The time evolution of the ExB velocity (and velocity shear) and, separately, the diamagnetic component of the ExB velocity (reflecting ion pressure profile evolution) have been successfully measured across the transition. However, the vXB term in the momentum balance has not been directly measured, but constitutes the essential link to establish poloidal/toroidal momentum drive during the transition. | ||
| Resource Requirements: | 30, 330, and 150 Beams, He plasma | ||
| Diagnostic Requirements: | BES large array, DBS5/DBS-8, V/Q-band profile reflectometry, midplane reciprocating probe,
CER tuned to He+ (4686A) | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 21: Magnitude of cross-field drifts in the Private Flux Region, divertor asymmetries and detachment | ||
| Name: | Groth groth@fusion.gat.com | Affiliation: | Aalto U |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | Cedric Tsui, Adam McLean, Steve Allen, Tony Leonard, Max Fenstermacher, Dmitry Rudakov, Jose Boedo, Rick Moyer, Jon Watkins, Rich Groebner, Dave Hill | ITPA Joint Experiment : | No |
| Description: | (1) Directly measure plasma conditions, potentials, and flows of deuterons and low charge state carbon ions in the divertor regions in L-mode plasmas. Connect these measurements to previous studies in 2004 and UEDGE simulations carried out since. Measure same plasma parameters with different diagnostics and assess/determine root cause for discrepancies (e.g., DTS versus X-point RCP versus target LPs versus IRTV). Connect carbon flows to deuteron flows: what physics is the dominant driver?
(2) Repeat in low-power H-mode plasmas (for RCP measurements): how do stronger (radial) gradients and ELMs affect the flow pattern and divertor asymmetries? |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Upstream density scan in low-power L-mode (10% beam modulation) and low-power H-mode (power within RCP limit)
- Density limit to determine three upstream densities for more detailed characterization - Three density points: low, high-recycling close to jsat rollover, detached - Constant density with strike point sweeps over DTS channels - 1-2 repeat discharges for spectroscopy and imaging, and data redundancy; diagnostic tweaks | ||
| Background: | Cross-field drifts in the private flux region play an instrumental role in creating in/out asymmetric divertor conditions. These drifts are predominately driven by gradients in Te in the radial direction. UEDGE simulations predict asymmetric divertor conditions, however, the database on direct measurements of the plasma potentials and flows is sparse. To exploit new capabilities in DIII-D, i.e., the swing-arm probe system in the inner divertor and the flow imaging diagnostics, repetition of parts of the 2004 divertor characterization experiment is required to fill in critical gaps. Furthermore, upgrade of the DTS and IRTV systems will elucidate physics and data discrepancy observed in the 2004 divertor characterization experiment. | ||
| Resource Requirements: | LSN with strike points on the 45-deg tiles (inner) and shelf (outer) for DTS and X-point probe measurements; large clearance (6cm) to upper outer limiter, 9 cm to outer midplane, high-density TS chords across (density) pedestal and SOL; L-mode with one NBI source, modulated, for dedicate edge measurements, co-current source; same configuration in H-mode with 1-2 sources, co-current | ||
| Diagnostic Requirements: | Swing-arm probes, flow imaging system, DTS, target LPs, midplane and X-point probes, divertor spectroscopy and tangential cameras | ||
| Analysis Requirements: | flow analysis for probes and imaging, plasma conditions from DTS and probes, 2-D reconstructions of D and C emission profiles | ||
| Other Requirements: | |||
| Title | 22: Assessment of main chamber carbon sources | ||
| Name: | Groth groth@fusion.gat.com | Affiliation: | Aalto U |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | Rich Groebner, Tony Leonard, Adam McLean, Max Fenstermacher, Al Hyatt | ITPA Joint Experiment : | No |
| Description: | Measure core carbon content (n4+ and n6+ carbon densities) as function of inner and outer gap in L-mode | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform three-point density scan in L-mode (low-recycling, high-recycling close to rollover, detached) with varying gap sizes to inner and outer midplane limiters (while maintaining sufficiently large clearance to upper outer limiter); suggested gap sizes: inner: 12, 6, 3 cm (OMP equivalent), outer 9, 6, 4 cm; ideally vary only one gap while maintaining the others; measure n4+ and n6+ densities in pedestal/SOL regions with CER (= 1 repeat shot per condition); measure low charge state carbon distributions with MDS and tangential cameras | ||
| Background: | Previous studies (2004 divertor characterization in L-mode) showed that the core carbon content is insensitive to (a) detachment of the outer divertor and (b) to the gap of the separatrix to the upper outer limiter. Since the inner divertor was detached even in low-recycling conditions, our hypothesis was/is that the main source for carbon in the core is from the inner divertor. To test this hypothesis against other main chamber sources from the inner and outer midplane regions, additional measurements are required. | ||
| Resource Requirements: | LSN, fwd BT, L-mode with 1 beam source for edge CER measurements, modulated, co-current; strike points on 45-deg tiles (inner) and shelf (outer) for DTS measurements; variation of inner and outer gaps while keeping other gaps (to top, upper outer limiter) constant | ||
| Diagnostic Requirements: | CER for n6+ and n4+ (7270 A) density profiles, divertor and main chamber spectroscopy, tangential cameras, DTS, target LPs, LPs at upper outer limiter | ||
| Analysis Requirements: | CER analysis, low charge state carbon profiles from MDS and 2-D reconstructions from tangential cameras | ||
| Other Requirements: | -- | ||
| Title | 23: RMP ELM Suppression at q95 above 4 | ||
| Name: | deGrassie degrassie@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use shaping to get out of the rut of n=3 RMP ELM suppression in DIII-D only for q95 < 4. Start with the odd parity high q95 (> 7) suppression (mitigation?) LSN condition and lower q95 by reducing the plasma cross section whilst maintaining field line alignment with the two row I-coils on the outside. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The starting condition is n=3 odd parity I-coil suppression (mitigation?) above q95 ~ 7 (e.g. 128464). Vary the plasma q95 value while maintaining alignment of the outboard field lines with the two row I-coil set, by reducing the plasma cross section. Obtain RMP ELM suppression in the range 4 < q95 < 5 in this fashion. | ||
| Background: | *With fixed RMP coils and essentially a fixed shape (ITER-like) the suppression, or non-suppression conditions have been established. Ultimately suppression coils should be designed for a specified equilibrium. As yet we do not know how to do the best such design.
*Understanding how to vary the q95 windows is necessary to gaining this design capability. *This technique has been used, albeit by luck, in shot 122445, with a skinny LSN and odd parity I-coils that resulted in suppression (mitigation?) in the range 4 < q95 <6. *At higher q95 the mitigation phenomenon may be more at work than that of suppression, using Chapman's explanation (APS Invited 2012), and that is why MAST results are described by mitigation. *Mitigation may be the best tool for ELM control if the resultant energy dumps are at the tolerable level. | ||
| Resource Requirements: | One day. Standard RMP ELM suppression experiment. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 24: Plasma response to 3D magnetic perturbations in the unfavorable grad-B drift direction | ||
| Name: | Battaglia dbattagl@pppl.gov | Affiliation: | PPPL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Todd Evans | ITPA Joint Experiment : | No |
| Description: | The experiment will aim to characterize the plasma response to n=3 magnetic perturbations in a LSN plasma with the toroidal field reversed (unfavorable grad-B drift). A LSN shape is desired in order to take advantage of the lower divertor diagnostics and connect to large database of LSN favorable grad-B drift direction discharges. Reversing the toroidal field will have several consequences that should provide a valuable test for ELM suppression and 3D transport theory: <br><br>The equilibrium will be right-handed, thus the plasma resonance will be on the positive side of the poloidal mode spectrum. The even parity poloidal mode spectrum is symmetric, thus any difference between right and left-handed resonances will be due to error fields. However, the odd parity poloidal mode spectrum is not symmetric, so reversing the toroidal field will have a large impact on the resonant components. Comparing the plasma response to odd and even parity in right and left-handed equilibriums will provide more information on the nature of the plasma response to the vacuum poloidal mode spectrum and error fields.<br><br>Neoclassical transport will change, altering the toroidal and poloidal flows near the separatrix. This most likely would change the plasma response by altering the penetration of magnetic perturbations.<br><br>The larger L-H power threshold will allow a test of plasma response in high-powered L-modes where a weak Te and ne pedestal exist. This may lead to a plasma response more like H-mode (pump-out with minimal temperature response) than L-mode (edge collapse of density and temperature).<br><br>Diagnostics such as the 3D magnetics and the soft x-ray camera would benefit from testing the detection and analysis of n=3 perturbations in LSN with reversed helicity and co-Ip torque. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The primary objective is to characterize the plasma response to an n=3 magnetic perturbation in L- and H-mode with the toroidal field reversed. This could include phase flips, RMP amplitude scan, even/odd parity and beta scan. The plasma shape would be chosen to match a low-collisionallity LSN ELM suppressed discharge in favorable grad-B direction, preferably with large heating power. If ELM suppression is achieved with reversed toroidal field, the experiment will focus on documenting the plasma response in this particular condition. | ||
| Background: | Previous experiments with RMP ELM suppression in the unfavorable grad-B drift direction used USN. As the I-coil current was increased, the discharge would back transition (H-L) before ELM suppression was observed. This experiment would be tried again using a LSN shape (more experience on density control in this configuration) and larger heating power (if available) to maintain H-mode. | ||
| Resource Requirements: | One run day to fully develop target discharge and document plasma response. | ||
| Diagnostic Requirements: | All available profile and divertor diagnostics. | ||
| Analysis Requirements: | XGC0, Kinetic EFIT, Varyped | ||
| Other Requirements: | -- | ||
| Title | 25: Transport mechanisms induced by RMP | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov, Z Unterberg, M. Schaffer | ITPA Joint Experiment : | No |
| Description: | RMP causes changes in the pedestal pressure gradient, keeping pedestal P-B stable. The detailed mechanism is not known. We propose to study the details of the plasma response to the RMP using probes. We also propose to create a particular plasma to enhance detectability of the RMP-induced response. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use ECH- heated discharges. Use elliptical, inner post limited discharges and n=1 (or n=2), m=3 perturbations to make islands visible and large in radius (1-2 cm) so the effects are detectable by all diagnostics. | ||
| Background: | Preliminary experiments using this configuration yielded intriguing but incomplete results. Completion of this work is important to understand underlying mechanisms | ||
| Resource Requirements: | ECH at 1-2 MW. COils | ||
| Diagnostic Requirements: | scanning probes, Xpoint SXR camera. BES, fluctuation diagnostics | ||
| Analysis Requirements: | TRIP3D, NIMROD, data analysis | ||
| Other Requirements: | |||
| Title | 26: Investigate the role of RS on inward momentum transport | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | G. Tynan, D. Rudakov | ITPA Joint Experiment : | No |
| Description: | Recent results to study intrinsic momentum sources have shown momentum transported inwards, against the velocity gradient. We propose to pursue this work and in particular to study the role of the Reynolds Stress on this. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Low power ECH-heated discharges for probe penetration. Beam blips for CER | ||
| Background: | Previous experiments using probes at the edge to elucidate the source of the intrinsic rotation yielded data that indicates momentum transport inward against the velocity gradient. | ||
| Resource Requirements: | DIII-D. ECH up to 1-2 MW. NBI diagnostic blips | ||
| Diagnostic Requirements: | CER, probes, BES | ||
| Analysis Requirements: | CER data anlysis, probe data analysis | ||
| Other Requirements: | |||
| Title | 27: COupling between core and SOL parallel flow | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov, C. Tsui, T. Ronglien, G. Porter, E. Belli, J. Candy. | ITPA Joint Experiment : | No |
| Description: | Recent results have shown that the intrinsic rotation source, co-Ip is large and concentrated right inside the LCFS. This source MUST diffuse out but there is no quantification of this process so we propose to quantify it. The intrinsic torque scales as GradP. Recent work also has calculated the poloidal variation of velocity, using NEO, , base on classical.neoclassical physics. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose to: 1) vary the intrinsic torque by varying GradP and inspect the SOL velocity for response and 2) measure the SOL/edge velocity at various poloidal locations to verify the (mostly) diamagnetic physics determines the poloidal variations | ||
| Background: | Previous work to investigate intrinsic rotation and poloidal variation of D+ parallel velocity | ||
| Resource Requirements: | DIII-D ECH and NBI blips and various diagnostics | ||
| Diagnostic Requirements: | scanning probes (mid and XPT ) and swing probes, CER | ||
| Analysis Requirements: | analyze above diagnostic data and Neo computations | ||
| Other Requirements: | |||
| Title | 28: Role of profile evolution and energy transfer in L-H transition | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G. Tynan, D. Rudakov, L. Schmitz, T. Rhodes, G. McKee, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Previous experiments using the i-phase regime as a tool to explore the detailed physics of the L-H transition revealed that slow profile evolution played an important role on the final transition dynamics. Energy transfer from high frequency to low frequency also plays a key role. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Propose to run I-phase discharges and study the evolution of the Grad-P terms and their role on the final transition. We will also take dedicated data to be able to calculate bi-coherency and make the flow of energy from high to low frequencies clear. | ||
| Background: | Recent work done in 2012 by Tynan, Schmitz and Boedo to clarify the role of Reynolds Stress in the L-H transition also found a significant effect from slowly evolving profiles. Also energy transfer to the zonal flows was identified. | ||
| Resource Requirements: | ECH and DIII-D some NBI and diagnostics | ||
| Diagnostic Requirements: | Scanning probes, reflectometry, BES, DBS, CER. | ||
| Analysis Requirements: | Of the diagnostics above | ||
| Other Requirements: | |||
| Title | 29: Modulation of Bootstrap Current | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | D. Thomas, H. Stoschus | 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 | 30: New Optimal Plasma Shape for AT Scenario? | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | For the high q_min, steady-state AT scenario, use the "ITER Similar Shape" (e.g., lower SND shape in shot 129323) rather the plasma shape from the standard unbalanced DND shape . In the low qmin hybrid scenario, the ISS is proved to have high beta limits (ideal with-wall limit greater than beta_N=5) and low electron heat transport. If these properties are present in the q_min>2 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 with qmin>2 but with the ISS 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. Stability analysis of kinetic EFITs with correct edge current density profiles using DCON found that the ideal n=1 with-wall limit was very high, more than beta_N=5. | ||
| 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: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 31: Test of Turbulence Spreading Using Turbulence Propagation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | 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. At least two ECH deposition positions should be studied, one near the axis to observe outward propagating turbulence and one near the edge to observe inward propagating turbulence. | ||
| 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: | GYRO simulations will be done after the experiment. | ||
| Other Requirements: | |||
| Title | 32: High Beta Hybrids and Pressure Profile Broadening | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | 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.
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) Begin with a 1 MA, high-beta hybrid case with 4 co-/on-axis beams that would serve as a fiducial. Lower BT until a hard beta limit is found (likely from a 2/1 mode); we expect to at least reach beta_N of 3.4-3.5 given previous results. (2) Repeat the fiducial case, but add the 150 beamline tilting fully downwards. 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. (3) 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 at least beta_N=3.5 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 (10 mV). TRANSP calculations show that these plasmas should be very close to fully noninductive. The ideal wall stability limit is calculated to be above beta_N=4 by DCON. During one half-day experiment, however, a lower stability limit of beta_N<3.2 was found. It was found that these discharges had a systematically more peaked pressure profile than the previous cases, which can explain the lower beta limit.
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. Based upon experimental experience that broader pressure profiles yield higher stability limits, some broadening of the pressure profile is 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 | 33: High Beta, Steady State Hybrids | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | 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. The optimization of the six gyrotrons and six co-beams will allow us to eliminate the residual 10 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, and possibly some broadening of the pressure profile using off-axis NBI, should allow us to increase beta closer to the ideal wall limit, which is above 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 (except possibly the 150 beam) 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 FNSF 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: | We expect to have to repeat shots several times to obtain full-length gyrotrons and beams simultaneously. (1) Start by repeating shot 133881. (2) Inject all 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. Hopefully we will have already done an experiment to determine if tilting the 150 beam increases the stability limit by broadening the pressure profile. (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. (5) If plasma current is overdriven (i.e. negative loop voltage), then increase plasma current to compensate. The density also can be adjusted. | ||
| 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 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 10 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. The evidence is that the lower 2/1 mode 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 are essential. 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 | 34: Electron Critical Gradient and Heat Pinch | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Luce, S. Smith, C. Holland | ITPA Joint Experiment : | No |
| Description: | Explore the relation between the critical gradient in the electron temperature and the electron heat pinch. Use heat pulse modulation to separate the "power balance" heat flux into its conductive and convective components. Determine if an inward electron heat pinch exists for the low gradient cases. Use plasmas previously found to give inward electron heat fluxes, i.e., high density and low plasma current L-mode plasmas. We want to document as many fluctuations as possible. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish L-mode plasma with Ip=500 kA and density=3e+19 m^-3. Use occasional short beam pulses to measure ion profiles. (2) Inject all 6 gyrotrons at rho=0.6. Modulate one gyrotron at 25 Hz. (3) Shot by shot, move one gyrotron from rho=0.6 to rho=0.4. (4) Document each case with fluctuation diagnostics such as CECE and DBS. (5) We can set up a "second" phase of the experiment to add NBI, as was done in 2011. (6) If time permits, repeat at a larger rho value (0.6-0.8). | ||
| Background: | This proposal builds upon the electron critical gradient experiment led by Jim DeBoo in 2011. By combining "power balance" heat flux with heat pulse propagation, the steady-state conductive and convective components of the electron heat flux could be determined. This allowed an electron critical gradient to be identified that is in agreement with fluctuation measurements and GYRO calculations. For both the rho=0.4 and rho=0.6 cases, there was evidence for a small heat pinch at the lowest electron temperature gradient. In this experiment we will change the plasma conditions to be more favorable for making electron heat pinches by increasing the density (which increases the ion-electron heat exchange) and lowering the plasma current (which reduces the ohmic heating). It is expected that this experiment will find clear evidence for an inward electron heat pinch, especially then the electron temperature is below the critical gradient. | ||
| Resource Requirements: | ECH: Desire all 6 gyrotrons.
NBI: Need 30LT, 330LT, 150LT sources. | ||
| Diagnostic Requirements: | All profile diagnostics, including density profile reflectometry. All fluctuation diagnostics. | ||
| Analysis Requirements: | Besides standard transport analysis, will need TGLF/GYRO calculations of the electron critical gradient, and TGLF simulations of the temperature profiles. | ||
| Other Requirements: | |||
| Title | 35: Non-Diffusive Heat Fluxes in the Edge "Shortfall" Region | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Luce, S. Smith, C. Holland | ITPA Joint Experiment : | No |
| Description: | Determine whether the "shortfall" by turbulent transport models in the edge heat flux for L-mode plasmas is due to an under prediction of the non-diffusive heat flux. The diffusive and non-diffusive heat fluxes will be measured in two locations: first, in the edge "shortfall" region; second, near the half-radius where transport codes do a better job of predicting the heat flux. The electron critical gradient, electron transport stiffness and fluxes will be determined from (1) by measuring the power balance and heat pulse transport, and (2) from the turbulence behavior (for the critical gradient). <br> <br>This proposal is essentially the same as #34 except the plasma conditions are more similar to the 2011 DeBoo experiment (higher current, lower density than for #34). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Use plasma condition for which a well established "shortfall" exists in the edge heat flux predicted by transport models. (2) Inject all gyrotrons at rho=0.8. with one gyrotron modulated at 25 Hz. (3) Shot to shot, move one gyrotron to rho=0.6 and measure change in electron temperature gradient and heat pulse propagation. (4) Document turbulence behavior during these discharges using CECE, DBS, etc. (5) Repeat all steps but with ECH moved to rho=0.6 and rho=0.4. (6) Can add a late beam-heated phase to all cases to measure BES and rotational effects. | ||
| Background: | It has been well established that in many L-mode plasmas, transport models such as TGLF and GYRO drastically under predict the ion and electron heat fluxes in the outer regions. This "shortfall" causes the transport simulations to essentially fail in the edge region. This experiment will determine if this theory/experiment disagreement is due to theory under predicting the effect of non-diffusive heat flux; the experiment will also measure the critical gradient and stiffness in the conductive electron heat flux, which may provide additional clues as to the origin of the disagreement. | ||
| Resource Requirements: | ECH: 6 gyrotrons desired.
NBI: 30LT and 330LT beams essential, 150LT desired. | ||
| Diagnostic Requirements: | All fluctuation diagnostics are required. | ||
| Analysis Requirements: | Transport analysis with ONETWO and poetAP. Turbulence and transport modeling with TGLF and GYRO. | ||
| Other Requirements: | -- | ||
| Title | 36: Natural ELM Pacing in High-Beta, Steady-State Plasmas | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | T. Luce | ITPA Joint Experiment : | No |
| Description: | High-beta, steady-state plasmas naturally have a very high ELM frequency (~300 Hz). We have not studied the size and footprint of the transient heat pulses in such plasmas, and therefore it is premature to assume that ELM suppression is required. Even if some type of ELM mitigation is needed in high-beta, steady-state plasmas, it will likely be easier to further increase the ELM frequency rather than completely suppress ELMs, and we need to determine what ELM mitigation (if any) is actually required. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using the established methods of characterizing the ELM size, as well as using the new capabilities of the IR periscope camera, we will study the ELM heat characteristics in AT plasmas with beta_N~3.5 and q_min~1.5. This should be done as a function of (1) beta_N, (2) collisionality, (3) q_95, and (4) triangularity. | ||
| Background: | ELM pacing experiments in DIII-D have demonstrated the ability to trigger ELMs with pellet injection, up to the 60 Hz capability of the hardware. However, another proven method to strongly vary the ELM frequency is to adjust the plasma beta (as well as the plasma shape). The ELM frequency is a strongly nonlinear function of the plasma beta, increasing from a few Hz at beta_N=1.8 to 300 Hz at beta_N=3.6. Therefore, it is not clear that, depending on details of the plasma shape, high-beta, steady-state plasmas cannot be naturally ELM pacing. | ||
| Resource Requirements: | NBI: 6 co-sources are required.
EC: 6 gyrotrons are desired. | ||
| Diagnostic Requirements: | Fast magnetics, MSE, LIBEAM, IR camera. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 37: Identification of I-mode through simultaneous core and edge turbulence measurements | ||
| Name: | White whitea@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Turbulence & Transport | Presentation time: | Requested |
| Co-Author(s): | D. G. Whyte, M. Fenstermacher, A. E. Hubbard, J. Hughes | ITPA Joint Experiment : | No |
| Description: | This experiment follows up on progress made in creating I-modes at DIII-D, MP #2012-93-04 <br> <br>I-mode on DIII-D will be created, e.g. based on target shots such as 149908, and changes in core and edge turbulence will be documented in order to identify I-mode. C-Mod measurements indicate that is it a combination of reductions in core and edge turbulence, rather that WCM appearance, that are clear indicators of I-mode [White TTF 2012] (in addition to increased in confinement with Pinj, and formation of Te-pedestals and not ne-pedestals). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Target shots are
The DIII-D I-mode experiment proposed here will consist of 1-day of run time. The first half day will consist of a wide parameter scan in current, density,and heating power to determine the conditions most favorable for I-mode (starting of course near the "best case" parameters from 2012). The second half day will consist of many dedicated repeat shots of "best I-mode targets" in order to document as clearly as possible fluctuations measured with BES, CECE, DBS, PCI, etc. It will be essential to target I-mode candidates with the proper density, B-field and NBI arrangement (e.g. on axis 150 L) to optimize fluctuation measurements. This was not done in the past exploratory experiments due to time constraints. | ||
| Background: | ELMs have the potential to destroy the first wall of ITER or a reactor. Many ELM control and mitigation techniques are being explored, but it is fruitful to consider an operating scenario that simply does not require ELMs occur.
QH-mode and I-mode are two such candidate regimes. DIII-D is well poised to explore and promote such alternative regimes for ITER and future reactors. I-mode, in contrast to QH-mode, has not been thoroughly explored on DIII-D and this proposed experiment will build on experience gained from MP #2012-93-04 in order to 1) create I-mode and 2) unambiguously identify it as such via detailed core and edge turbulence measurements. This experiment follows up on progress made in creating I-modes at DIII-D, MP #2012-93-04 "Exploration of I-mode operating space", that was led by Dennis Whyte last summer. It was observed at DIII-D that high confinement plasmas with a temperature pedestal but no density pedestal could be created and sustained. Additionally, as input power was increased, confinement did not degrade. However, there was no evidence of the edge localised Weakly-Coherent Mode (WCM) on edge midplane fluctuation diagnostics, which is taken to be a signature of I-mode at C-Mod [Whyte NF 2010]. Recently at C-Mod, there is mounting evidence collected during JRT 2012 experiments that I-modes can occur without clear WCM features observed at the outboard midplane, and in fact, other changes in edge and core turbulence simultaneously may in fact be clearer signatures of I-mode. Using the extensive set of core and edge fluctuation diagnostics at DIII-D we can probe the changes in both edge and core turbulence that occur at the L-I mode transition, thus providing better evidence that in fact I-mode has been produced at DIII-D and at C-Mod in cases when the WCM is not clearly visible in the measurements. For more details, please see slides presented at July 26th 2012 805 meeting. | ||
| Resource Requirements: | Experimental set-up as on 149908, but with ON AXIS 150L NBI for optimum BES measurements, and density scan to optimize DBS measurements. | ||
| Diagnostic Requirements: | Full profile diagnostics where possible (Thomson, ECE, CER, MSE...). Full turbulence suite (BES, CECE, DBS, PCI, edge probes, etc.) | ||
| Analysis Requirements: | Profile analysis, CER, MSE, turbulence data, TRANSP, ONETWO, TGLF, BOUT++, EPED...etc. | ||
| Other Requirements: | -- | ||
| Title | 38: Calibrated passive FIDA spectrum | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Nathan Bolte | ITPA Joint Experiment : | No |
| Description: | Repeat best case from the 2012 piggyback experiments with the new camera. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | One source injects at 50% duty cycle for one second as the plasma current is gradually varied. | ||
| Background: | Excellent data were obtained in 2012 for Bolte's thesis project but the CCD camera that acquired the spectra died before an intensity calibration was performed. | ||
| Resource Requirements: | Piggyback experiment: one beam source for approximately one second at end of shot | ||
| Diagnostic Requirements: | New CMOS FIDA camera | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 39: Measure and understand energetic particle profiles in steady-state plasmas | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | John Ferron, the Energetic Particle group | ITPA Joint Experiment : | No |
| Description: | Acquire FIDA profiles & measure coherent instabilities in a set of steady-state plasmas with different values of qmin | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce the 2009 betan=2.7 qmin scan. Modify the beam programming to acquire FIDA profiles from all three systems. Optimize fluctuation diagnostics for Alfven eigenmode measurements. Also reproduce the q95 scan. | ||
| Background: | The FIDA active beam diagnostics require beam modulation patterns that are rarely achieved during usual steady-state operation. By operating at betan=2.7, we can achieve a condition that is relevant to the highest performance plasmas while maintaining sufficient flexibility to obtain good data. | ||
| Resource Requirements: | All beams except 210LT. | ||
| Diagnostic Requirements: | FIDA, ECE & ECEI, BES, UCSD camera + .... | ||
| Analysis Requirements: | NUBEAM & FIDASIM calculations for all good shots. Alfven eigenmode stability calculations. | ||
| Other Requirements: | |||
| Title | 40: MGI with applied n=1 RMP fields | ||
| Name: | Izzo izzo@fusion.gat.com | Affiliation: | Fiat Lux LLC |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Investigate the role of the m=1/n=1 mode phase in determining radiation toroidal peaking by attempting to lock the phase of the 1/1 mode with externally applied fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform single valve (high-Z) massive gas injection experiments with standard diverted target plasma. Use I-coils to apply n=1 RMP fields beginning shortly before MGI pulse. Vary the phase of the n=1 RMP field to look for an effect, especially on radiated power measurements. | ||
| Background: | NIMROD simulations of MGI in DIII-D indicate that the toroidal location of the radiated power peak, as well as the degree of toroidal peaking, during the thermal quench is determined by the phase of the m=1/n=1 mode. A basic confirmation of this effect is needed experimentally, which can be accomplished by controlling the phase with external fields. If the phase can be successfully locked to the applied fields, the code predicts that a systematic variation in the radiated power measured at a single toroidal location should be observed as the phase is varied. The code also suggests that a variation in impurity assimilation may be seen. | ||
| Resource Requirements: | 1/2 run day, possibly combined into a full run day with another MGI experiment, such as multi-valve MGI. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 41: Investigation of the effects of the EHO toroidal mode number on the edge particle transport | ||
| Name: | Diallo adiallo@pppl.gov | Affiliation: | PPPL |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, R. Maingi, A. Garofalo(?), E. Doyle(?), S. Gerhardt | ITPA Joint Experiment : | No |
| Description: | The goal of this proposal is to investigate the impact of the edge harmonic oscillation (EHO) characteristics (e.g., toroidal mode number n, and relative amplitude of each n-components, finite n-spectrum vs broadband) on the edge pedestal particle transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental plan will utilize the established experimental parameters developed for the toroidal mode control of the EHO proposed by K. Burrell in ROF 14. Using similar experimental conditions to select the toroidal mode number dominant contribution of the EHOs, we will quantify the edge particle transport for various EHO characteristics. This quantification is to be performed using the following three approaches in order of priority.
1.The first approach will focus on non-recycling transport analysis induced by EHO. This approach will rely gas injection using silane (SiH4) or carbon tetrafluoride (CF4) to study the transport of non-recycling particles as a function of EHO n-spectrum. 2.The second approach will target the electron transport. This approach will reconstruct the pedestal density and temperature as a function of the multiple phases of the dominant magnetic signature of the EHO. The goal is to elucidate the relationship between particle and heat transport by providing fast reconstruction of the edge pedestal structure through the full cycle of the dominant mode in the magnetic signature of various types of EHOs. 3.Finally, the last approach will ??consider? a modulated gas puff (at rate near particle confinement time) to determine the edge transport parameters (particle diffusion coefficient D and pinch velocity v) for the dialed EHOs n spectrum. Note that this approach will be used in tandem with the fast reflectometer density prole measurements sampling the pedestal region for a complete characterization of the density fluctuations induced by the gas modulation. In addition, measurements of the density profile pedestal structure can be obtained from the fast reflectometry at each phase of the modulation of the gas puff. For completeness, this modulation experiment could in principle be repeated for obtaining the transport coefficients of impurities such as Helium. | ||
| Background: | The viability of ITER as a test fusion reactor hinges on achieving steady state edge particle transport for ELM avoidance. A potential scenario candidate is the quiescent H-Mode (QH) discovered on DIIID. Experiments have shown that EHO is a key ingredient in achieving QH mode. The EHO has been extensively studied in ITER relevant scenarios (low-torque) and a range of toroidal mode numbers have been shown to be associated with EHOs. In addition, EHO was shown to produce electron, main ion, and impurity particle transport at the plasma edge at a rapid rate compared to that produced by ELMs under similar conditions. However, no systematic investigations of the particle transport dependence on the toroidal mode spectrum characteristics has been performed to establish to optimum mode number and extrapolate to ITER conditions. The goal of this experiment is two-fold. First, we document the particle transport explicitly as a function of EHO toroidal mode content. Second, we identify the key physics mechanism for enhanced particle transport. | ||
| Resource Requirements: | Fast modulation of gas feed (Frequency TBD). Fast-sweeping reflectometer sampling the pedestal region. Impurity injection capability. | ||
| Diagnostic Requirements: | All profiles diagnostics. DBS measurements and reflectometry, BES large array centered on pedestal region, and calibrated midplane Dalpha signals to study the particle flux in EHOs. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 42: Compatibility of High Performance Plasmas With a Puff-and-Pump Radiating Divertor | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | C. Holcomb | ITPA Joint Experiment : | No |
| Description: | This study will combine all essential elements for making the first real test of the puff-and-pump concept applied to high performance DIII-D plasmas. Neon is injected into the private flux region (PFR)of the lower divertor. Deuterium plasma flow toward the lower divertor target on the low field side is enhanced by a combination of deuterium gas injected upstream of the divertor target and active cryo-pumping at the target. Previous puff-and-pump experiments with standard ELMing H-mode plasmas have shown that biasing the plasma magnetic balance toward the lower divertor (e.g., dRsep = -0.5 cm)and directing the gradB drift toward the upper divertor yield the best chance of optimizing the stability and transport benefits of (near) DN operation while maintaining plasmas relatively clean of impurity accumulation with low peak heat flux. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The base plasmas are near-DN high performance H-mode plasmas that can be reliably maintained for at least two seconds flattop. The lower divertor cryo-pump is maintained at liquid helium temperature. The gradB ion drift direction is toward the upper divertor, and dRsep can range from -0.3 to -0.5 cm. "High performance" refers to betaN > 3, H89p ~ 2.5, and qmin >1.5. This experiment can best be done as follows:
* First establish the sensitivity of the high performance plasma to deuterium gas injection. Scan the deuterium gas puff rate to establish the operational limit as to how much D2 the plasma can accommodate before there is appreciable degradation in AT properties. * Scan of the neon injection rate at the "best" D2 injection rate established above. Understanding the sensitivity of high performance DIII-D plasmas embedded in a "radiating divertor" environment is paramount. Important measurables from this experiment are changes in energy confinement, stored energy, and qmin. Other important measurables include changes in the radiated power distribution and heat flux, changes in the density and temperature at the divertor targets, and the accumulation of neon in the core plasma. | ||
| Background: | Several years of study here at DIII-D and other tokamak facilities have demonstrated the viability of a puff-and-pump radiating divertor in reducing power loading at the divertor targets while at the same time preserving high energy confinement and betaN in the core. However, for high performance plasmas, the cooling of the divertor and SOL by radiative means can lead to excessive cooling of the pedestal plasma and marked degradation of favorable plasma pedestal and core properties.
Previous attempts using "radiaing mantle" approaches were successful in knocking down the peak heat flux by ~50% without serious degradation of core plasma properties. However, in this experiment, we believe that divertor heat flux can be reduced much further without seriously impinging on plasma performance. This is done using a combination of impurity (neon) injection plus D2 injected from a non-divertor (upstream) location. Experiments with puff-and-pump radiating divertor conducted here at DIII-D (2006-2010)on standard H-mode plasmas indicate that this approach can be extended to high performance discharges with a reasonably good chance of success. | ||
| Resource Requirements: | This experiment can use dedicated time (2-4 shots) at the end of the day in which SS Heating and Current Drive experiments were featured. We basically followed this prescription in the 2012 campaign. Cumulatively, ~0.5 day is needed for this experiment. | ||
| Diagnostic Requirements: | Standard core plasma diagnostics, plus these divertor diagnostics: Asdex gauge in all three baffles,lower divertor IR camera, bolometer, lower divertor and centerpost Langmuir probes, filterscopes, core SPRED, and lower divertor tangential visible TV. | ||
| Analysis Requirements: | SOLPS/UEDGE, and ONETWO | ||
| Other Requirements: | -- | ||
| Title | 43: Density Control and Active Impurity Removal from Double-null H-mode Plasmas | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | This experiment explores the possibility of using changes in the magnetic balance and gas puff program to both 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) has been shown to reduce pedestal (and line-averaged) density by up to 50%. 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 while maintaining core density by making small changes in dRsep. With simultaneous pumping on both outer divertor legs of a magnetically balanced high-triangularity DN, DIII-D IS UNIQUELY CONFIGURED TO MAKE A DEFINITIVE STATEMENT. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment is straightforward. Start with a DN shape and maintain a constant density throughout the discharge by putting the system in density feedback control starting at t = 2.0 s. The direction of the ion gradB drift is toward the lower divertor. Argon impurities are first injected into the private flux region of both divertors. Wait for steady conditions; this should take the discharge out to about t = 3.5 s. Between t = 3.5 s and t = 3.8 s, change dRsep from 0 to +0.5 cm. Hold dRsep = +0.3 cm from t = 3.8 s to 4.3 s, as argon is expelled from the plasma; density feedback will help maintain core plasma density. Then return dRsep to 0 at t = 4.3 s and finishing up at 4.6 s. Compare argon impurity density before dRsep is changed (t=3.45 s) with the impurity density after dRsep is restored to dRsep = 0 (t = 4.6 s). How long does it take the argon density to return to its original value? ---Repeat with neon. | ||
| 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 reduce the impurity content from the core plasma of near-DN plasmas while largely maintaining core density, 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 5 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, bolometer, core SPRED, and CER. | ||
| Analysis Requirements: | SOLPS/UEDGE, ONETWO | ||
| Other Requirements: | -- | ||
| Title | 44: Can Nitrogen impurity Seeding Improve Radiating Divertor Performance? | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | 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 examine<br> this hypothesis on DIII-D. | 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: 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 obtaining 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 improve performance with nitrogen. In fact, A. Kallenbach has suggested that DIII-D at least examine this possibility. This experiment does just that 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, bolometer, and CER. | ||
| Analysis Requirements: | SOLPS/UEDGE and ONETWO | ||
| Other Requirements: | -- | ||
| Title | 45: Differences in Impurity Accumulation Between Resonant and Non-resonant RMP Plasmas | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | -- | 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 previous 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 ELMing 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 responsible for the similarity in argon impurity accumulation in the main plasma between RMP-resonant and non-RMP discharges. 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: | Six beam sources. This is a 0.5-1.0 day experiment. | ||
| Diagnostic Requirements: | IR camera, CER, Thomson scattering, bolometers, CO2 interferometers, and SPRED. | ||
| Analysis Requirements: | MIST (or equivalent), ONETWO | ||
| Other Requirements: | -- | ||
| Title | 46: Does Added Baffling From Re-Configured Tiles on the Lower Divertor Shelf Facilitate Detachment? | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | M.A. Mahdavi | ITPA Joint Experiment : | No |
| Description: | A toroidally symmetric row of graphite tiles on the low field side of the lower divertor shelf will be installed for the 2013 campaign. The purpose of this reconfiguration is to aid in trapping neutral particles at high values of the outer separatrix strike point, thereby facilitating detachment and low peak heat flux. For a fixed value of the outer strike point, i.e., Rosp = 1.71 m, we will compare existing data from the 2012 campaign (open configuration) with data from the upcoming 2013 campaign (closed configuration). Three separate densities will be directly compared: ne = 3.2-, 4.2-, and 5.2s10^19 m^-3. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Shots 149609-611 serve as the model shots for comparison. These shots represent the open configuration. These are very high X-point (75 cm off the lower divertor floor), a = 51 cm, Pinj = 5 MW, and Rosp = 1.71 m. We will re-run these shots at the three densities listed above.
Based on our SOLPS modeling, we fully expect to see a partially-detached outer leg in the new (closed) configuration at lower density than we found in the model shots. | ||
| Background: | We expect the added baffling at the edge of the shelf to greatly facilitate detachment, compared with the present open configuration. We base this prediction on SOLPS modelling and the experimental results from open vs closed divertor plasma behavior studied in the 2012 campaign.
This experiment will help QUANTIFY the advantages of baffling in future divertor design relying on large Rosp values. | ||
| Resource Requirements: | Machine time 0.25 day (4-5 good shots), ion gradB drift direction is downward, 5 co-beams. | ||
| Diagnostic Requirements: | Core Thomson scattering, CER, divertor fixed Langmuir probes on the shelf, lower divertor IR camera, visible tangential TV in lower divertor, and bolometer. | ||
| Analysis Requirements: | SOLPS | ||
| Other Requirements: | -- | ||
| Title | 47: Quantifying the TBM torque, both in magnitude and radial location | ||
| Name: | Tala Tuomas.Tala@vtt.fi | Affiliation: | VTT Technical Research Centre |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, H. Reimerdes, A. Salmi, J. Snipes | ITPA Joint Experiment : | No |
| Description: | Understand the physics of TBM induced rotation changes and the origin (magnitude, radial location, beta dependence) of the TBM torque (NTV torque). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Based on the single shot where the TBM modulation at 5Hz was applied in the 2011 TBM experiments on DIII-D, this method turned out to be a very good one to study the rotation changes induced by the TBM. The analysis of the single shot shows that the effect of the TBM on rotation originates from the edge, as published in H. Reimerdes et al., IAEA FEC 2012 paper. In order to be able to calculate quantitatively the TBM torque, a few more pulses with TBM modulation would be needed combined with good measurements of momentum transport within the same shots. What is needed for this is a scan of TBM perturbation amplitude (either by coil current, plasma movement, I-coil error field correction) so that the amplitude of the rotation change varies, and simultaneously apply NBI modulation to measure the momentum transport coefficients within each shot. So, each of the shot in this scan should have a phase with TBM modulation and a phase with NBI modulation.
Another important parameter to be scanned is beta. At high beta, larger changes in rotation were observed. There, the effect of TBM on rotation could be different (possibly a combination of edge and core contributions), and using the TBM modulation technique, this difference could be studied, in particular where the TBM torque sources/sinks are radially located. A beta scan with TBM modulation and NBI modulation within a shot to nail down the reason why at higher beta the TBM influences more rotation is proposed. | ||
| Background: | 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. 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.
Further experiments to pin down the physical mechanism how the TBM affects rotation, by inducing a torque source/sink and/or changes in momentum transport and edge rotation. Using the TBM modulation at 5Hz frequency, the evidence point out towards the fact that TBM is inducing a counter edge torque and this decrease is then further propagated to the core. To quantify the TBM torque, a TBM perturbation scan with TBM modulation combined with a separate beta is needed for ITER extrapolation. | ||
| Resource Requirements: | TBM mock-up active, Co and counter NBI | ||
| Diagnostic Requirements: | Full profile diagnostics, especially CER, plus FIDA and main ion CER. | ||
| Analysis Requirements: | TRANSP, intrinsic torque + modulated transport analysis | ||
| Other Requirements: | |||
| Title | 48: Dependence of momentum transport and intrinsic torque on rho* and beta | ||
| Name: | Tala Tuomas.Tala@vtt.fi | Affiliation: | VTT Technical Research Centre |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, A. Salmi, C. Angioni | ITPA Joint Experiment : | Yes |
| Description: | This proposal consists of two ITPA experiments TC-15 and TC-17 experiment. TC-15 is about studying parametric dependences of momentum transport coefficients (pinch + diffusion) on several plasma parameters and TC-17 is about determining the intrinsic torque. The objective of this experiment is to study the dependence of the momentum pinch and Prandtl numbers on beta and rho* while a similar, strongly linked proposal by W. Solomon focuses on the intrinsic torque and its dependences on the same parameters rho* and beta. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The first quantity to be scanned is rho*. A similarity experiment with JET is proposed by scanning from small rho* plasmas (high B JET pulses) to large rho* plasmas low B DIII-D pulses). This experiment is already proposed in JET (accepted as a back-up experiment in C31 and planned as a main experiment in C32). The rho* scan is extremely important for the intrinsic torque part of the experiment (proposed in W. Solomonâ??s proposal id=?).
The session plan is such that each pulse has a fast NBI modulation phase (~10Hz) for momentum transport studies (this proposal) and slow modulation (~2Hz) for intrinsic torque studies (W. Solomonâ??s proposal). The synergy is also desirable from the scientific point of view as we can check the momentum balance and confirm each otherâ??s data by feeding both the momentum transport coefficients and intrinsic torque into the momentum transport equation. The second scan is the beta scan. Standard technique to carry out the beta scan by keeping other dimensionless parameters as constant as possible will be used. 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. The outcome of the experiment is to confirm (or not) the theory and gyro-kinetic predictions for the dependence of momentum pinch on beta. | ||
| Background: | In recent years, several tokamaks have shown that a significant inward momentum pinch exists. 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, DIII-D and AUG. On the other hand, no significant collisionality dependence has been found on neither on DIII-D nor on JET (the first part of the TC-15 experiment was carried out on DIII-D and published in PoP in 2010 by W. Solomon et al and IAEA FEC 2012 by T. Tala). While the parametric dependencies (collisionality, R/Ln, q) of the momentum pinch ITG dominated plasmas were included in TC-15 joint experiment, no studies of momentum pinch in high beta plasmas was included and no studies of rho* dependence were done. According to 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.
Similarly to momentum transport studies, or even more importantly, no rho* scaling experiment to study the dependence of intrinsic torque has been performed. This is definitely needed for any reliable intrinsic torque extrapolation to ITER, and in 2013, this ITPA TC-17 experiment will be pushed in JET to get experimental time in the C32 campaign. | ||
| Resource Requirements: | Co- and counter NBI | ||
| Diagnostic Requirements: | Full profile diagnostics, especially CER, plus FIDA and main ion CER. | ||
| Analysis Requirements: | Standard, well-developed techniques will be used to analyse the NBI modulation and intrinsic torque analysis. | ||
| Other Requirements: | |||
| Title | 49: Turbulent transport in detached divertor | ||
| Name: | Boedo boedo@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov, C. Tsui, J. Watkins | ITPA Joint Experiment : | No |
| Description: | Detached divertors are known to feature enhanced transport (not only atomic-based dissipation) that spreads particle and heat loads. But little is known about the details and we propose to systematically investigate the turbulence and transport properties of detached divertors | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Detach the lower and upper divertors (not simultaneously) and evaluate rms levels of density and potentials and if possible (geometry-limited) the turbulent transport. | ||
| Background: | divertor detachment is a solution for large heat and particle fluxes into the divertor components. There is still much unknown about the detached divertor dynamics and transport. In particular, enhanced transport helps to further spread heat and particle loads. So we propose to investigate that mechanism | ||
| Resource Requirements: | DIII-D, lowish power NBI (2 MW), ECH (1-2 MW) | ||
| Diagnostic Requirements: | Xpoint and midplane scanning probes, floor probes, turbulence diagnostics (BES, reflectometry, etc) | ||
| Analysis Requirements: | diagnostic analysis and if possible fluid and turbulence simulatiosn (UEDGE, SOLPS, BOUT) | ||
| Other Requirements: | -- | ||
| Title | 50: Improved Heat Load Reduction Using Long Outer Divertor Leg Geometry During Puff and Pump Operation | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | This experiment is the logical extension to our puff-and-pump studies and, separately, our divertor shaping studies. It is designed to provide a clear assessment of possible advantages provided by increasing the poloidal length (or parallel SOL connection length) of the outer divertor leg. Based on 1-D two-point modeling: ntar is proportional to [Lpar]^6/7 [nsep]^3 and Ttar is proportional to 1/{[Lpar]^4/7 [nsep]^2}, where nsep is the midplane separatrix density and Lpar is the parallel SOL connection length between Xpoint and target. These scalings suggest that conditions conducive to a radiative divertor solution can be achieved at low nsep by increasing Lpar. Our data from the 2011 and 2012 campaigns are consistent with the above Lpar scalings.<br>In this experiment, we test whether this line of thinking can be successfully coupled to puff-and-pump operation. We compare the performance of the core and divertor plasmas in these long-legged outer divertor configurations with much shorter-legged ("standard") outer divertor plasmas, under radiating divertor, puff-and-pump conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Our base case is a high X-point plasma shape (poloidal outer leg length ~75 cm), modelled on LSN shot 149603: Ip = 0.8 MA,Bt = 2.0 T, Pinj = 5 MW. Rtar = 1.32 m, if only the present lower divertor view by the IR is available; Rtar =1.35 m, if the new periscope IR camera system is available. For this experiment the gradB drift direction is toward the upper divertor. Deuterium gas puffing is done from either gasA (or GasD, if necessary), while neon is injected into the private flux region of the lower divertor. For steady trace neon injection, three steady values of D2 puffing are used: 0, 50, 100 tl/s. Then, for steady D2 injection during the shot (best D2 puffing rate from above), use two values neon injection at perturbing levels---these levels generate radiating divertor conditions. This process is repeated for the low-xpoint configuration (poloidal outer divertor length ~20 cm).
Measurables include heat flux, density and temperature at the outer divertor target via IR camera and Langmuir probes. Behavior of upstream separatrix and pedestal density and temperature, and impurity neon accumulation and energy confinement in the main plasma are measured by Thomson scatter, CER, SPRED, and magnetics. | ||
| Background: | From 2003-2006 the focus was on active particle exhaust (and the conditions needed to optimize pumping). From 2006-2010, the focus was on radiating divertor physics (and the conditions needed to optimize radiating divertor operation). And from 2011-2012, the focus was on how divertor shaping affects the divertor and upstream plasma conditions in the SOL (and how these divertor shaping changes might be set up to optimize heat load reduction). This experiment brings all three of these facets together. Based on our understanding of the past decade of divertor activity in these areas, one expects that the high X-point, "isolated" divertor to generate more radiated power outside the main plasma, lower peak heat flux at the outer divertor target, and a much "cleaner" core plasma with better performance measurables, as when compared with more standard, lower X-point configurations. | ||
| Resource Requirements: | 0.5-1.0 day of experimental time and minimum of 5 co-sources. | ||
| Diagnostic Requirements: | Divertor IR camera, floor Langmuir probes, bolometer, Thomson scattering, core SPRED, lower divertor tangential camera | ||
| Analysis Requirements: | SOLPS and ONETWO | ||
| Other Requirements: | -- | ||
| Title | 51: Study of chirping Alfvén eigenmodes in DIII-D | ||
| Name: | Podesta mpodesta@pppl.gov | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | DIII-D EP group | ITPA Joint Experiment : | Yes |
| Description: | Identify critical parameters regulating the dynamics of bursting/chirping TAE/RSAE modes in DIII-D, with focus on AI startups. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce baseline discharge 144889 with bursting/chirping AEs. Study the dependence of AE dynamics on fast ion and plasma parameters through scans of toroidal field and NB parameters (injection voltage, injection geometry at fixed total NB power). Map out conditions for the development of AEs from quasi-stationary to bursting/chirping. Time permitting, establish connection to NSTX scenarios, where bursting/chirping AEs are commonly observed, by lowering the magnetic field (~0.6 T). | ||
| Background: | AE modes showing amplitude bursts and/or frequency chirping can develop into the extreme case when so-called 'avalanches' occur, with greatly enhanced fast ion transport and losses. Bursting/chirping AE instabilities in the TAE/RSAE frequency range were recently observed in "advance inductive" startup scenarios on DIII-D. This study will investigate the conditions leading to the development of strong AE bursts/chirps and its impact on fast ion transport. | ||
| Resource Requirements: | Variable mix of NB sources (co- vs. counter, different NB injection energies)
Scenarios at low magnetic field, approaching the standard values of NSTX (~0.55 T), may be included in the experimental plan. | ||
| Diagnostic Requirements: | Fast ion diagnostics (neutrons, FIDA, FILD). Internal mode structure measurements (ECEI, BES). Profile diagnostics (Thomson, CER, MSE). | ||
| Analysis Requirements: | EFIT, TRANSP, NOVA. Possibly FIDASIM. | ||
| Other Requirements: | -- | ||
| Title | 52: Control of TAE/RSAE modes via external perturbations | ||
| Name: | Podesta mpodesta@pppl.gov | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | DIII-D EP group | ITPA Joint Experiment : | No |
| Description: | Explore the possibility of affecting the TAE/RSAE dynamics by means of externally applied, low-n perturbations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce plasmas with robust TAE/RSAE activity. Better if TAEs appear as 'cluster' of modes with comparable frequency. Aim at low rotation discharges so that the Doppler shift of TAE frequencies is small and can be matched by the Internal Coils. Perform amplitude and frequency scans of n=1 perturbations imposed through the Internal Coils and characterize modifications of AE dynamics. Focus on startup phase, possibly w/ balanced NBI to avoid fast plasma rotation. | ||
| Background: | Low-frequency MHD modes have been observed to couple with TAE modes through three-wave coupling (DIII-D, NSTX). The proposed experiment is aimed at inducing a similar coupling between pairs of TAE (and, possibly, RSAE) modes and externally applied perturbations. If entanglement of the TAEs is observed, a parametric study of the amplitude/frequency of the external perturbation will reveal the threshold for effective coupling and the feasibility of this method to regulate the dynamic of bursting/chirping Alfvénic modes which are responsible for fast ion loss and redistribution. | ||
| Resource Requirements: | Internal coils operating at relatively high frequency (~1 kHz), n=1 configuration. | ||
| Diagnostic Requirements: | All fast ion diagnostics, ECEI, possibly reflectometers. Profile diagnostics (TS, CER, MSE). | ||
| Analysis Requirements: | EFIT, TRANSP, NOVA | ||
| Other Requirements: | |||
| Title | 53: QH-mode with low torque start-up | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | W.M. Solomon, A.M. Garofalo, M. Lanctot | ITPA Joint Experiment : | No |
| Description: | We seek to develop techniques to allow QH-mode operation with low co-Ip NBI torque (< 1 Nm) throughout the discharge. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The basic approach is to start with an ECH H-mode, turn on the NTV torque using the nonaxisymmetric coils, and then determine the NBI torque limit
1. Use Ohmic startup followed by ECH H-mode to get torque-free H-mode a. Requires SND with grad-B drift towards divertor or DND b. Use reverse Ip, standard Bt, LSND or DND c. This current and field combination allows operation under standard QH-mode conditions in USND to assess wall conditions 2. Need best possible error field correction; accordingly, use I-coil for n=1 error field correction and some even parity n=3; use C-coil with D1 power supply for n=3 NRMF 3. Determine whether n=3 NRMF inhibits L to H transition a. If so, turn on after transition b. If not, turn on before 4. Add sufficient counter beams to create QH-mode a. Use beta feedback to keep beta_N at 2 5. Scan amount of counter torque and start time of beams to determine counter torque limit for QH-mode As part of this investigation, we should also see whether using n=2 NRMF rather than n=3 NRMF provides any advantage. | ||
| 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 2009-2012 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque or small co-Ip torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. These shots, however, used large (4 Nm), counter-Ip NBI torque to create the QH-mode; the torque was then changed during the QH-mode to the small, co-Ip values. To convince the fusion community that QH-mode is a feasible operating mode in future devices, we must extend our previous demonstration to discharges where the NBI torque is at small, co-Ip levels throughout the discharge. Initial attempts at creating these plasmas in the 2012 campaign were only partially successful; we did discover, however, that we need to use Ohmic startup rather than turning the beams on early as we have done for the counter-Ip NBI cases. The present plan builds on that observation, coupling it with creation of ECH H-mode with zero NBI torque. | ||
| Resource Requirements: | Reverse Ip. 6-7 gyrotrons. C-coil configured for maximum n=3 field, 7.1 kA current. I-coil configured for error field correction and as much n=3 field as possible. | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 55: RMP Elm Suppression in DN Plasmas | ||
| Name: | Lazarus lazarus@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Evans, Battaglia | ITPA Joint Experiment : | No |
| Description: | Achieve ELM suppression in DN plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The plan is in miniproposal 2012-01-06 from last year. It was unsuccessful because it was attempted shorty after a boronization. Needs to be tried again. | ||
| Background: | Many tools for the analysis of 3D effects are from the stellarator world and are constrained to stellarator-symmetric equilibria. In D3D this reduces to up-down symmetry and the even parity
I-coil configuration | ||
| Resource Requirements: | standard as detailed in the mp | ||
| Diagnostic Requirements: | standard as detailed in the mp | ||
| Analysis Requirements: | EFIT kinetic equilibrium
VMEC 3D equilibrium | ||
| Other Requirements: | |||
| Title | 56: Sawtooth Control | ||
| Name: | Lazarus lazarus@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Plasma Control | Presentation time: | Not requested |
| Co-Author(s): | Evans, Luce, Lanctot | ITPA Joint Experiment : | No |
| Description: | Investigate control of the sawtooth, in particular its collapse time, with the application of an m/n=1/1 boundary perturbation as detailed in the miniproposal 2012-99-50 | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Return to bean/oval sawtooth experiment and observe the effects of the C-coils/I-coils on sawtooth characteristics | ||
| Background: | Recent calculations with VMEC (Lazarus, 2012 IAEA TH/P3-02)show that an equilibrium state can exist with q0<<1 if there is a mn=1/1 boundary perturbation. This has properties similar to
an ideal saturated internal kink. There are enticing points of agreement with experimental observations (ISX-B, TEXTOR, JET, DIII-D). | ||
| Resource Requirements: | standard, as detailed in mp 2012-99-50 | ||
| Diagnostic Requirements: | standard, as detailed in mp 2012-99-50 | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 57: Develop a high performance L-mode plasma. | ||
| Name: | Lazarus lazarus@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | General Physics | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Develop an L-mode plasma with a high performance core that sustains q0~2/3 and ql<2 (thus disruption-free). For details see my science meeting presentation of Nov. 16, 2012. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | There are 4 separate threads to be pursued and then combined. Each is non-trivial. (1)Attain operation with a m/n=1/1 boundary perturbation that sustains q0 at near 2/3. (2) Develop the capability to stabilize the external kink, as done on RFX. This will allow ql<2. (3) Investigate core parameters in such an L-mode case. Interchange stability and its consequences on electron and ion transport is of particular interest. In general, characterize confinement within a/2 vs plasma density, Ti/Te, beta_p_1, shape. (4) Investigate pumped limiter operation. | ||
| Background: | In recent work RFX demonstrated stabilization of the external kink and operated at ql~1.8. Recent VMEC calculations offer a plausible understanding of the TEXTOR q0 results (q0~2/3) as a consequence of relatively small n=1 error fields,
suggesting this as a stable state of the tokamak (albeit with a helical core). Sawtooth experiments in oval shapes suggested dismal electron confinement, but excellent ion confinement. These might combine to produce a high performance core. It represents a plausible path towards an FNS. It should be considered as orthogonal to ITER, NOT as opposition to ITER. | ||
| Resource Requirements: | Varied. None are extraordinary. It may prove that power supply upgrades for the I-coils are needed. Without some initial efforts to stabilize the external kink, it is premature to speculate on this. | ||
| Diagnostic Requirements: | Nothing extraordinary. | ||
| Analysis Requirements: | To be developed. | ||
| Other Requirements: | A few weeks of run time per year for a few years.
There is limited overlap with other proposed work. | ||
| Title | 58: Progressing Towards a Staged Approach to Disruption Mitigation | ||
| Name: | Parks none | Affiliation: | GA |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | Valerie Izzo, Eric Hollmann | ITPA Joint Experiment : | No |
| Description: | We are trying to do three things in one experiment to save run time: (1) Validate the pellet ablation model for Boron pellets, a proxy for Be pellets considered for ITER; (2) Establish whether we can get a purely radiative pre-emptive thermal collapse with deep Boron pellet penetration (cooling from the â??inside outâ?? with no magnetic reconnections, instead of usual â??outside inâ?? cooling which contracts J profile, excites external kink modes or surface tearing modes with eventual m/n = 1/1 internal kink, reconnection, and core heat dump on divertor/first wall; (3) Determine whether an acceptable Current Quench (CQ) rate is accessible in DIII-D by pinning the CQ floor temperature at a stable point near the 20 eV local minimum in the Boron radiation emissivity curve. If the CQ floor temperature drops below ~ 20 eV, infer whether discharge was spoiled by native impurities or post TQ carbon influx from wall. NIMROD modeling is needed to predict whether a uniform radiatively induced temperature collapse is feasible, with minimal MHD activity. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | |||
| Background: | For many years, disruption mitigation experiments on existing tokamaks have focused on pre-emptive shutdown methods that induce a thermal quench (TQ) by injecting a radiative species (gas or pellets) into a normal, or pre-disruptive discharge. Unfortunately, no pre-emptive MGI experiment has come close to achieving the abnormally high â??criticalâ?? densities, needed for runaway electron (RE) suppression, largely on account of the observation that â??directâ?? penetration of a gas jet into a normal plasma discharge does not occur. A staged approach was suggested [Parks, NF 2011] where the TQ is triggered by injecting deep into plasma only a small quantity of impurity, e.g., using Be pellet at ~ 600 m/s. Then to reach the critical density, a separate high-density D2 gas jet is injected deeply into CQ discharge using Burst Disc Gas Injection (BDGI) [Parks, NF 2011]. The merit of this approach is the potential for RE avoidance with an acceptable CQ rate in ITER. We use Boron in place of Be pellets in DIII-D to study pellet ablation and plasma response to pellet. A rigorous, unified theory of pellet ablation was recently developed that includes single species pellet materials (icy or refractory) with arbitrary Z. The model can be validated on DIII-D with Boron pellet. Be has a local minimum emissivity L at Te = 8.5 eV where it becomes a â??dimâ?? helium-like radiator since its two strongly bound electrons are very difficult to excite. Boron has a local minimum emissivity L at Te = 20 eV where it is triply ionized and thus â??hydrogen likeâ?? with its singly bound electron very difficult to excite. This experiment is designed to investigate whether an acceptable CQ rate is possible by accessing the â??marginallyâ?? stable point near the local minimum at 20 eV. The CQ time is too short with argon pellet injection because its emission intensity is 3 orders of magnitude larger than boron. Since the â??properâ?? pellet mass M* scales as M* ~ 1/Sqrt[L], the very high L for argon and neon implies exceedingly small pellets must be used, which means penetration is impossible. | ||
| Resource Requirements: | Requires readiness of the GA shell pellet injector. One run day | ||
| Diagnostic Requirements: | UCSD Camera imaging: for pellet trajectory light. Utilize existing diagnostics for radiation attributes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 59: Effect of Changing the Grad-B Drift Direction on Snowflake Divertor Behavior, Including Detachment | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | S.L. Allen and V.A. Soukhanosvskii | ITPA Joint Experiment : | No |
| Description: | This experiment explores the role that particle drifts play in determining the plasma properties of the Snowflake divertor. Here, the focus is the direction of the ion grad-B drift direction, which previous studies on standard DIII-D divertors have shown to play a major part in determining important divertor properties, such as detachment, heat flux distribution, particle exhaust, and injected impurity behavior. In this experiment, we compare Snowflakes having identical shape but with their ion grad-B drift direction opposite each other. We examine the Snowflake (minus) in ion both grad-B drift directions at 3 densities. In each case, neon is injected into the private flux region at trace levels. Beam power is fixed at 6 MW. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with the ion grad-B drift direction toward the main divertor. On the first shot, start with standard lower SN divertor with trace neon injection into the PFR but without D2 injection; after equilibration of neon in core, then switch to SF(-)later in the shot. Next, program standard SN for the entire shot; density to increase during the shot so that density reaches 80% of the Greenwald limit by the end of the current flattop; no neon is injected on this shot. Choose three density levels that characterize (or span) this density range; at each density (maintained with feedback control), start with standard SN and inject a steady stream of neon into the private flux region at a trace level; as neon equilibrates in the main plasma, change to Snowflake(-). At higher density, equilibration time for the standard SN divertor might be too long, so a separate shot focusing on the SF(-) phase may be necessary. This part of the experiment should require only 7-8 good shots.
On the second day of this experiment when the ion grad-B drift direction is reversed, repeat the process established on Day #1. If possible, match the fixed density levels established on day #1. Measurables include density and temperature at the divertor targets, as well as heat flux profiles (with IR camera) and imaging of the lower divertor in CIII/Dalpha light using the lower divertor tangential camera. In addition to upstream density and temperature (TS), we will determine the amount of neon in the core plasma (CER and SPRED). | ||
| Background: | The Snowflake divertor configuration is based on the creation of a second-order null point via bringing together two first order null-points of a standard divertor. The result of this is an extended region of reduced magnetic field, with favorable consequences for plasma stability, transport, and heat flux reduction. The latter was clearly demonstrated on DIII-D during the last campaign. However, virtually nothing is known about the divertor properties of a Snowflake plasma having its ion grad-B drift directed away from the main divertor. Detailed studies with standard divertor configurations on DIII-D have shown that some divertor and core plasma properties improve when the ion grad-B drift is directed away from the main divertor, among them control over particle exhaust, reduced impurity accumulation in the main plasma, and a wider density operating range. In this experiment, we expect to find out if these favorable divertor and core properties in standard divertors with ion grad-B away from the divertor carry over to Snowflake. We also expect to have a detailed characterization of the divertor plasma of Snowflakes with opposite ion grad-B drift directions. | ||
| Resource Requirements: | 0.3 day in "forward" BT and 0.3 day in "reverse" Bt, for a total of 0.6 day. Six co-beam sources, lower divertor cryo-pump at liquid helium temperature. | ||
| Diagnostic Requirements: | Core Thomson scattering, CER, Langmuir probes, IR camera, bolometer, Asdex gauge in the lower divertor pumping plenum, core
SPRED, tangential lower divertor visible-light camera. | ||
| Analysis Requirements: | SOLPS/UEDGE, ONETWO | ||
| Other Requirements: | -- | ||
| Title | 60: Effective NTM control in low torque plasmas | ||
| Name: | Prater prater@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, E. Kolemen, R. La Haye, F. Turco, A. Welander | ITPA Joint Experiment : | No |
| Description: | Determine whether narrow, accurately placed ECCD is more effective in avoiding NTMs in low torque discharges than broad ECCD. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Restore a low torque discharge that had 2/1 NTMs without ECH but no NTMs with broad ECH on. Then narrow the ECCD profile and determine the best location through scanning. Then see how little EC power can be used to fully avoid the NTM. | ||
| Background: | In past work, experimenters working on low torque high performance discharges have found that use of the ECH system is effective at avoiding NTMs, but it has not been clear what the physical effect is, and sometimes ECH works as well as ECCD, and in either case broadly spread profiles work sufficiently well. Because that addresses the problem, work hasn't been done to determine whether narrow, well placed ECCD might be more effective with lower power. The point of this proposal is to see whether properly located ECCD can be optimally effective. | ||
| Resource Requirements: | At least 5 gyrotrons. | ||
| Diagnostic Requirements: | The usual, plus ECEI. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 61: Energy Transport During Electron-Dominated Heating of ITER-Relevant H-Mode Discharges | ||
| Name: | Taylor gtaylor@pppl.gov | Affiliation: | PPPL |
| Research Area: | Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | N. Bertelli, J.C. Hosea, R.J. Perkins, C.K. Phillips, P.M. Ryan, D.R. Smith, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | This experiment will study electron transport and plasma turbulence in DIII-D Advanced Inductive (AI) and ITER Baseline Scenario (IBS) H-mode discharges that are predominantly heated by electron cyclotron (EC) power and it aims to identify the dominant mechanism(s) responsible for enhanced electron transport when EC power is applied to these ITER-relevant scenarios, especially as produced on DIII-D with NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | All discharges should be run with balanced NBI to minimize the applied torque, and they should be run with no beta feedback on NBI power so that the NBI power remains constant.
The run plan is as follows: 1.Begin with an AI discharge similar to shot 146571 (the outer gap can be larger since there will be no fast-wave (FW) heating), with sufficient NBI power to transition to and sustain an H-mode. 2.Apply an EC heating pulse that is considerably shorter than the NBI pulse (~1 s). Vary the EC pulse timing and duration. Measure the change in stored energy at the turn-on and turn-off of the EC pulse. 3.Repeat 2 with increasing EC power (eg. 2, 4 and 6 gyrotrons). 4.Repeat 3 with one or possibly two gyrotrons modulated to study electron transport with ECE etc. 5.Repeat 1-4 for an IBS discharge similar to shot 150840 (once again, the outer gap can be larger since there will be no FW heating). Typically the IBS discharges in 2012 had to be run at much higher densities than the AI discharges (~5.5x1019 m-3 for IBS compared to ~ 3.5x1019 m-3 for AI) to avoid NTMs. The higher density in the IBS discharges caused the plasma to go overdense for second harmonic ECE, so the density should be lowered to get core ECE data. If NTM??s appear some of the gyrotron launchers should be configured for ECCD in order to stabilize NTM??s. [NOTE: For the 2013 campaign it is hoped to upgrade at least one, possibly more, of the EC launchers so that they can be changed from ECCD to ECH orientation in 0.5-1 s, compared to ~ 5 s at present.] 6.Compare heating efficiencies for ECCD and ECH for best heating case of 5. Couple EC power from mirrors configured for ECCD followed by coupling power from the mirrors configured for ECH and vice versa. Perform in successive shots, or if possible, using two ECH pulses in the same shot with fast mirror movement or with power configured for ECH and ECCD. 7.Run AI and IBS H-mode discharges with EC heating only (no NBI) at the highest gyrotron power available to assess how well EC heats an ECH-only H-mode discharge. | ||
| Background: | ITER will utilize virtually torque-free, fuelling-free, dominant electron heating to generate and sustain plasmas in the H-mode regime. EC heating will play a major role in generating H-mode discharges in ITER. However there is growing evidence for significantly increased electron transport when EC heating is applied to DIII-D ITER-relevant H-mode discharges produced with NBI. Comparison of EC and FW heating of AI H-mode discharges in 2011 showed similar core electron heating and heating efficiency based on the time evolution of stored energy for the first ECH pulse and the FW heating pulse applied on top of the first ECH pulse, whereas very little heating and increase in stored energy was obtained with a second ECH pulse on top of the first (eg. shots 146571 and 146574). In 2012 a saturation of stored energy was observed in DIII-D IBS discharges when increasing levels of EC heating were applied (eg. shots 150840 and 150821). The radiated power observed when the EC power was applied to either the AI or the IBS discharge scenarios was essentially proportional to the total power, suggesting that the observed behavior is due to enhanced electron transport in these scenarios. It is imperative that the dominant mechanism(s) causing the enhanced electron transport are identified so that ITER-relevant scenarios with reduced electron transport during EC heating can be developed. | ||
| Resource Requirements: | Machine Time: 1-1.5 days (Steps 1-4 of the run plan for the AI target discharges can be completed in about 0.5 days, similarly Step 5 for the IBS target discharges can be completed in about 0.5 days, and the remaining steps of the run plan may take another 0.5 days)
Number of gyrotrons: 6 (7 if available) Number of neutral beam sources: 4, plus beam blips for BES and MSE | ||
| Diagnostic Requirements: | ECE, BES, CHERS, , MSE, UCLA reflectometry for oblique angles, and other diagnostics for measuring turbulence | ||
| Analysis Requirements: | TORAY, GENRAY, TORBEAM, TRANSP | ||
| Other Requirements: | Also submitted to Transport & Turbulence and Steady State Heating and Current Drive - please discuss placement with that group | ||
| Title | 62: RMP assisted snowflake divertor | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | T.E.Evans et al. | ITPA Joint Experiment : | No |
| Description: | Application of 3D field - for instance n=3 fields for ELM control - have to be shown to lead to striation of the divertor particle and heat fluxes. These striations are caused by the decomposition of the stable and unstable manifolds of the separatrix. Recently direct imaging confirmed this structure formed due to the divertor X-point as hyperbolic fixed point. In the snowflake configuration a secondary X-point is generated which in the snowflake + configuration is located inside of the wall. This configuration is prone to a strong interaction of the manifolds of both hyperbolic fixed points. This is likely to facilitate opening of the magnetic field region at the separatrix and distribution of the heat and particle outflux into many lobes born from both X-points. In addition, the high magnetic shear in the plasma edge facilitates island overlap and stochastization which can be beneficial for pedestal stability control. Investigation of the impact of snawflake configuration with RMP fields on edestal stability and the transient and stationary divertor fluxes is proposed in this experiment. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | - Setup optimized snowflake + configuration (e.g. 149736 at 3750 ms)
- Apply RMP field from I/C-coils - scan mode number and RMP strength - measure heat and particle fluxes in divertor - measure pedestal profiles - investigate divertor conditions AND pedestal stability | ||
| Background: | This experiment makes use of the unique 3D coil capabilities of DIII-D and the flexibility of PCF and field coils for generating snwoflake. The results can enhance the snowflake divertor concept and may show an improved method to control edge stability and transients in snwoflake configuration | ||
| Resource Requirements: | One day at DIII-d, snowfalke patch panel, Icoils on SPAs, C-coils for EFC | ||
| Diagnostic Requirements: | divertor IR
divertor visible cameras (DiMeS TV and TTV) with D_a, and CII, CIII - soft X-ray at Xpoint with low and high energy cut off filters - edge Thosmson scattering | ||
| Analysis Requirements: | - pre experiment: explore with TRIP-3D field structure (ongoing),explore impact on edge transport with EMC3-Eirene (grid modification required for snowflake)
- post experiment: TRIP3D and EMC3-Eirene, EPED/ELITE for pedestal stability | ||
| Other Requirements: | -- | ||
| Title | 66: 13C-methane injection into USN re C migration to walls and ITER decision about PFC target material | ||
| Name: | Stangeby peter.stangeby@utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | Plasma-material Interface | Presentation time: | Requested |
| Co-Author(s): | Tony Leonard, Richard Pitts (ITER), Dmitry Rudakov, Clement Wong | ITPA Joint Experiment : | Yes |
| Description: | Motivation: the proposed experiment will provide important information for the decision which will be made in late 2013 whether ITER should start with CFC or W at the divertor targets.<br>One of the arguments that continues to be strongly maintained by some, e.g. at 2012 ITPA DIVSOL, against starting with C at ITER targets is that the C will migrate out of the divertor and so the C contamination of ITER will not be removed by replacing the divertor. That C will then be problematic subsequently by causing increased tritium retention.<br>Based on our 13CH4 injection expt in DIII-D into the secondary divertor (unbalanced DN), however, such a migration pattern seems unlikely.<br>DIII-D is the only tokamak that can inject 13CH4 simultaneously into both outer and inner primary divertors (using USN), toroidally symmetrically. Thus DIII-D can do a definitive experiment here.<br>Inject 13CH4 into USN using both inner and outer pumping plenums, to establish if C from primary targets migrates out of divertor. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | On the last day of the campaign, inject 13CH4 into USN using both inner and outer pumping plenums, to establish if C from primary targets migrates out of divertor. One day (20 shots), repeat condition, high power, ELMing H-mode discharges, with partial detachment of the outer divertor - as ITER plans to use. A comprehensive set of tiles removed for measurements of 13C surface density by Bill Wampler, Sandia. | ||
| Background: | This will be the 4th 13C-methane injection experiment done in DIII-D but the first to measure the migration out of the divertor of carbon released into the divertor plasma, simulating carbon sputtered from the targets. DIII-D's ability to inject trace gases toroidally symmetrically and using large entrance orifices, puts it in a unique position to do interpretable trace impurity expts. In other tokamaks the injection is toroidally non-symmetrical and/or involves gas injection through small orifices. Non-toroidally-symmetric injection results in non-toroidally-symmetric deposition of the 13C, which makes it difficult/impossible to find most of the 13C and makes the interpretation of the expt difficult/impossible. Typically only a few percent of the injected 13C is found in 13C-methane injection experiments in other tokamaks, while in the three DIII-D expts, ~ 50% of the 13C was found. Injection of gas through small orifices almost certainly disturbs the local plasma - although that is not possible to confirm by direct measurements since the disturbance would be quite local. Even with the far more spatially distributed gas injections used in DIII-D - where diagnosis of the plasma in the injection region is possible - we were able to measure some plasma changes to the local plasma, changes which we then incorporated in the interpretive modeling. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 70: DIMES test of misaligned CFC tile to establish size of misalignment needed to disrupt ITER | ||
| Name: | Stangeby peter.stangeby@utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | Plasma-material Interface | Presentation time: | Requested |
| Co-Author(s): | Tony Leonard, Richard Pitts (ITER), Dmitry Rudakov, | ITPA Joint Experiment : | No |
| Description: | Motivation: the proposed experiment will provide important information for the decision which will be made in late 2013 whether ITER should start with CFC or W at the divertor targets.<br>There will be 200,000 tiles in the ITER divertor. Off-normal events, e.g. disruptions, could cause tile misalignment.<br>AUG and C-mod have shown that a small, 0.3 mm, misalignment of a single W-tile causes disruptions.<br>It is assumed, but has never been demonstrated, that comparable - or even larger - misalignments of CFC tiles will not cause disruptions when exposed to the same power loading.<br>It is important to confirm experimentally that this expectation is valid and to establish how large a misalignment of a CFC tile will likely cause a disruption in ITER.<br>It is proposed to carry out DiMES tests with an intentionally misaligned CFC-tile to establish how large a misalignment is required to cause a disruption. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | High power, ELMing H-mode, USN discharge. The bottom separatrix slowly lowered onto DiMES with a CFC cap having a 1 mm high edge, X 1 cm wide X 1 cm long, simulating a mm misaligned tile. If the plasma does not disrupt, replace the DiMES sample with a higher exposed edge. Use several discharges with DiMES sample removed for measurements after each shot, to establish how the surface profile evolves through the exposures, i.e. does it "fire polish" as is known to happen with isotropic graphite, or do jagged CFC surfaces persist? The latter might not cause disruptions but could cause unacceptably high rates of carbon release into the plasma. | ||
| Background: | The carbon blooms in JET, TFTR were due to tile misalignments and sometimes caused disruptions:
"If a carbon bloom occurs near the density limit during belt limiter operation on JET, a MARFE usually appears on the small major radius side of the plasma column. The MARFE does not usually result in a disruption even though the radiated power is nearly equal to the input power." Mike Ulrickson, Journal of Nuclear Materials 176 & 177 (1990) 44. CFC may be more problematic than isotropic carbon: extreme thermal stresses may cause the CFC to disintegrate, with fragments of significant size then penetrating the confined plasma, potentially causing a disruption. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 71: 4. ELM Characterization to Validate BOUT++ ELM Simulations | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | S.L. Allen, C.J. Lasnier, I. Joseph, X.Q. Xu | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to obtain a complete set of high temporal resolution data during moderate size Type-I ELMs to be used for validation of BOUT++ non-linear simulations of DIII-D ELMs at experimental conditions. The development of BOUT++ models that can simulate Type-I ELMs at realistic low experimental H-mode collisionalities is part of an internall funded project at LLNL. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use standard ELMing H-mode in LSN plasma shape optimized for fast diagnostic characterization of ELM transient behavior. Likely will need to use high triangularity so that OSP is within view of vertical viewing fast IRTV. Adjust power to achieve moderate frequency (~20 Hz) and moderate size (delta_W_ped/W_ped <~ 10%) Type-I ELMs that are optimized of BOUT++ simulation. Obtain simultaneous fast measurements of all possible pedestal, SOL and divertor transients during ELM events, eg. fast heat flux at both strike-points, fast radiated power (DISSRAD), fast magnetics, fast edge CER, fast target tile currents (if possible) etc. | ||
| Background: | Initial detailed fast ELM characterization experiments done previously (summarized in PPCF, 45, (2003), pp.1597-1626) showed many features of the perturbation of the pedestal, SOL and divertor during Type-I ELM events. many diagnostic measurements that can contribute to detailed non-linear modeling of ELM evolution have been improved since then, eg. new high resolution Thomson, high temporal resolution CER, edge current profile measurements with LiBeam and edge MSE (co plus counter NBI) etc. Also there has been substantial development of the BOUT++ non-linear ELM simulation capability for cases at realistic, low collisionality experimental pedestal conditions. A new complete ELM characterization dataset would contribute significantly to validation of these new BOUT++ simulation capabilities. | ||
| Resource Requirements: | Standard ELMing H-mode, LSN (probably moderately high triangularity, shape optimized for fast pedestal, SOLand divertor measurements, co and counter NBI to optimized edge MSE bootstrap current measurement, all fast edge, SOL and divertor diagnostics. This could be done in a half day experiment. | ||
| Diagnostic Requirements: | All fast edge, SOL and divertor diagnostics including both fast vertically viewing IRTV and periscope IR | ||
| Analysis Requirements: | Profile fitting, kinetic EFITS, pedestal Er and bootstrap current analysis all fed into BOUT++ simulations of individual ELM events. Detailed comparison of BOUT++ results (synthetic diagnostic development) with fast measurements. | ||
| Other Requirements: | -- | ||
| Title | 72: Energy Transport During Electron-Dominated Heating of ITER-Relevant H-Mode Discharges | ||
| Name: | Taylor gtaylor@pppl.gov | Affiliation: | PPPL |
| Research Area: | Turbulence & Transport | Presentation time: | Requested |
| Co-Author(s): | N. Bertelli, J.C. Hosea, R.J. Perkins, C.K. Phillips, P.M. Ryan, D.R. Smith, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | This experiment will study electron transport and plasma turbulence in DIII-D Advanced Inductive (AI) and ITER Baseline Scenario (IBS) H-mode discharges that are predominantly heated by electron cyclotron (EC) power and it aims to identify the dominant mechanism(s) responsible for enhanced electron transport when EC power is applied to these ITER-relevant scenarios, especially as produced on DIII-D with NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | All discharges should be run with balanced NBI to minimize the applied torque, and they should be run with no beta feedback on NBI power so that the NBI power remains constant.
The run plan is as follows: 1.Begin with an AI discharge similar to shot 146571 (the outer gap can be larger since there will be no fast-wave (FW) heating), with sufficient NBI power to transition to and sustain an H-mode. 2.Apply an EC heating pulse that is considerably shorter than the NBI pulse (~1 s). Vary the EC pulse timing and duration. Measure the change in stored energy at the turn-on and turn-off of the EC pulse. Substitute ECH power with similar level of NBI power. 3.Repeat 2 with increasing EC power (eg. 2, 4 and 6 gyrotrons). 4.Repeat 3 with one or possibly two gyrotrons modulated to study electron transport with ECE etc. 5.Repeat 1-4 for an IBS discharge similar to shot 150840 (once again, the outer gap can be larger since there will be no FW heating). Typically the IBS discharges in 2012 had to be run at much higher densities than the AI discharges (~5.5x1019 m-3 for IBS compared to ~ 3.5x1019 m-3 for AI) to avoid NTMs. The higher density in the IBS discharges caused the plasma to go overdense for second harmonic ECE, so the density should be lowered to get core ECE data. If NTM??s appear some of the gyrotron launchers should be configured for ECCD in order to stabilize NTM??s. [NOTE: For the 2013 campaign it is hoped to upgrade at least one, possibly more, of the EC launchers so that they can be changed from ECCD to ECH orientation in 0.5-1 s, compared to ~ 5 s at present.] 6.Compare heating efficiencies for ECCD and ECH for best heating case of 5. Couple EC power from mirrors configured for ECCD followed by coupling power from the mirrors configured for ECH and vice versa. Perform in successive shots, or if possible, using two ECH pulses in the same shot with fast mirror movement or with power configured for ECH and ECCD. 7.Run AI and IBS H-mode discharges with EC heating only (no NBI) at the highest gyrotron power available to assess how well EC heats an ECH-only H-mode discharge. | ||
| Background: | ITER will utilize virtually torque-free, fuelling-free, dominant electron heating to generate and sustain plasmas in the H-mode regime. EC heating will play a major role in generating H-mode discharges in ITER. However there is growing evidence for significantly increased electron transport when EC heating is applied to DIII-D ITER-relevant H-mode discharges produced with NBI. Comparison of EC and FW heating of AI H-mode discharges in 2011 showed similar core electron heating and heating efficiency based on the time evolution of stored energy for the first ECH pulse and the FW heating pulse applied on top of the first ECH pulse, whereas very little heating and increase in stored energy was obtained with a second ECH pulse on top of the first (eg. shots 146571 and 146574). In 2012 a saturation of stored energy was observed in DIII-D IBS discharges when increasing levels of EC heating were applied (eg. shots 150840 and 150821). The radiated power observed when the EC power was applied to either the AI or the IBS discharge scenarios was essentially proportional to the total power, suggesting that the observed behavior is due to enhanced electron transport in these scenarios. It is imperative that the dominant mechanism(s) causing the enhanced electron transport are identified so that ITER-relevant scenarios with reduced electron transport during EC heating can be developed. | ||
| Resource Requirements: | Machine Time: 1-1.5 days (Steps 1-4 of the run plan for the AI target discharges can be completed in about 0.5 days, similarly Step 5 for the IBS target discharges can be completed in about 0.5 days, and the remaining steps of the run plan may take another 0.5 days)
Number of gyrotrons: 6 (7 if available) Number of neutral beam sources: 4, plus beam blips for BES and MSE | ||
| Diagnostic Requirements: | ECE, BES, CHERS, , MSE, UCLA reflectometry for oblique angles, and other diagnostics for measuring turbulence | ||
| Analysis Requirements: | TORAY, GENRAY, TORBEAM, TRANSP | ||
| Other Requirements: | Also submitted to Inductive Scenarios and Steady State Heating and Current Drive - please discuss placement with that group | ||
| Title | 73: Test optimized TBM error field correction at higher values of beta | ||
| Name: | Reimerdes reimerdes@fusion.gat.com | Affiliation: | CRPP-EPFL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Hanson, R. La Haye, M. Lanctot, N. Oyama, C. Paz-Soldan, J. Snipes | ITPA Joint Experiment : | No |
| Description: | This experiment will test the hypothesis that n=1 TBM error field correction (EFC) can suppress the deterioration of confinement with increasing beta once beta normalized exceeds values of approximately 2.0 [H. Reimerdes, et al., IAEA 2012]. A confirmation of this hypothesis would be a favorable result for ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The empirically optimized n=1 I-coil EFC currents found in 2011 will be applied in the corresponding TBM reference scenario (ELMy H-mode, q95~4.1, beta normalized~1.8, co-NBI) and beta increased. Any increase in the torque due to the TBM field and energy and particle confinement degradation will be compared against a case with optimized EFC of the intrinsic error alone. In order to check that the optimized currents do not change with beta small scans around the previously found currents will be applied. Such changes can for example be caused by the changing PF coil currents. The capability of the EFC currents to suppress the n=1 magnetic plasma response to the TBM field can also guide an empiric optimization (possibly by using magnetic feedback, i.e. dynamic error field correction). | ||
| Background: | In the 2009 TBM experiments it was found that the TBM mainly affects plasma rotation and to a smaller degree energy and particle confinement. While the rotation decrease is modest at beta normalized of 1.8, the confinement degradation rapidly becomes worse with beta as beta normalized exceeds a value of 2.0 [J.A. Snipes, EPS 2010, M.J. Schaffer, et al., Nucl. Fusion 51 (2011) 103028]. The beta dependence of the confinement degradation suggests that it is caused by an increasingly amplified n=1 â??kink mode resonantâ?? component of the TBM field.
In 2011 the n=1 I-coil error field correction was optimized in the presence of the TBM using the plasma angular momentum as the optimization criterion. It was found that at beta normalized of 1.8, n=1 I-coil currents with the I-coils configured with 240 Deg phasing can only recover a quarter of the TBM induced rotation decrease. At the same time the currents almost completely suppress the magnetic n=1 plasma response to the TBM field measured at the outboard midplane. It is therefore speculated that the n=1 EFC will suppress the increase of the magnetic plasma response with beta and hence the confinement degradation that has been seen for values of beta normalized above 2.0 [H. Reimerdes, et al., IAEA 2012]. | ||
| Resource Requirements: | - TBM mock-up coil
- 5 co-NBI | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 74: Optimize TBM error field correction against field penetration in low NBI torque H-modes | ||
| Name: | Reimerdes reimerdes@fusion.gat.com | Affiliation: | CRPP-EPFL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Hanson, R. La Haye, M. Lanctot, N. Oyama, C. Paz-Soldan, J. Snipes | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to evaluate the potential of low n external fields to ameliorate TBM effects in ITER relevant slowly rotating H-modes, where error fields can limit operation either by causing field penetration or triggering an NTM. This is achieved by optimization EFC currents by maximizing the n=1 error field tolerance and/or minimizing the NBI torque required for stable operation. Low density locked mode experiments suggest that n=1 correction alone could be sufficient to restore the resilience against field penetration, which would be a favourable result for ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | One possible approach is a measurement of the tolerance to an n=1 proxy error in compass scans. Comparing origin and radii of the circles of maximum tolerable proxy field in discharges with and without TBM field yield correction currents and quantifies the detrimental effect of the TBM using optimized EFC currents. Another approach is based on measuring the minimum tolerable NBI torque for various correction currents. This approach requires more discharge since each discharge only yields a measurement for one n=1 I-coil current. The approaches are complementary. | ||
| Background: | The limited effectiveness of n=1 EFC in ameliorating the TBM induced rotation decrease in rotating H-modes [H. Reimerdes, et al., IAEA 2012] contrasts its ability to fully recover a low locking density in L-modes [M.J. Schaffer, et al., Nucl. Fusion 51 (2011) 103028]. This suggests that the components of the external field that cause braking at higher beta and higher rotation (presumably through non-resonant braking) differ from the components that cause field penetration (presumably through resonant braking). If rotation is key to the successful recovery of the locking density, EFC should also be more effective in ITER relevant slowly rotating H-modes. | ||
| Resource Requirements: | - TBM mock-up coil
- 5 co-NBI, 2 ctr-NBI | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 75: Evaluate the potential of correcting more than one mode component of the TBM error field | ||
| Name: | Reimerdes reimerdes@fusion.gat.com | Affiliation: | CRPP-EPFL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Hanson, R. La Haye, M. Lanctot, N. Oyama, C. Paz-Soldan, J. Snipes | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to improve the structure of the TBM correction field. Possible improvements are based on improving the m spectrum of an n=1 correction field, adding n>1 components to the correction field or a combination of both. Here the optimization of m and n components are addressed separately in order to facilitate the link to theory. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment requires a target that clearly shows the effects of the TBM and is reproducible at the same time. The target could be the fast rotating H-mode developed in 2011, but any other scenario that meets the above criteria would work, too.
In order to address the poloidal spectrum of an n=1 field, an optimization procedure using external fields that substantially differ from the 240 Deg phasing I-coil field is carried out. This includes an n=1 field applied with the C-coil alone as well as model based (e.g. vacuum, IPEC) C- and I-coil combinations that are predicted to improve the correction. The performance is compared to the standard n=1 I-coil field with 240 Deg phasing. If the poloidal structure of the n=1 field is important, the recoverable rotation should change. In order to address the importance of higher n components the optimization procedure used for n=1 is repeated using an n=2 field. This requires the control of each individual I-coil with an independent power supply. Optimum n=1 and n=2 EFC currents are then applied alone and simultaneously. | ||
| Background: | The 2011 TBM experiments showed that in rotating ELMY H-modes at beta normalized of 1.8, n=1 EFC using 240 Deg I-coil phasing can only correct 25% of the TBM induced rotation decrease. The remaining 75% of the rotation decrease must be caused by either secondary n=1 components of the TBM field or by n>1 components of the TBM field [H. Reimerdes, et al., IAEA 2012]. | ||
| Resource Requirements: | - TBM mock-up coil
- 5 co-NBI, 2 ctr-NBI (depending on the target). - Independent control of 12 I-coils | ||
| Diagnostic Requirements: | - 3D magnetics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 76: Connection between particle and momentum transport | ||
| Name: | mordijck smordijck@wm.edu | Affiliation: | The college of William and Mary |
| Research Area: | Turbulence & Transport | Presentation time: | Requested |
| Co-Author(s): | Ed Doyle | ITPA Joint Experiment : | Yes |
| Description: | Theoretical work [P.H. Diamond et al. 2009 Nucl. Fus. 2009] shows that there is a strong connection between particle transport and momentum transport. Recent experimental results on ASDEX (see IAEA McDermott) shows that density peaking depends on rotation profile. However, in these plasmas they change from ITG to TEM regimes (through linear analysis), which was shown in Cmod to strongly affect rotation profiles (bifrucation) and is strongly correlated with collisionality and density. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | DIII-D is in a unique position to investigate the connection between momentum and particle transport and its relation with Te/Ti, TEM vs. ITG, density limits and collisionality (these seem to be the recurring parameters). The experiment should not only look at the changes in micro turbulence but also at the changes in macroscopic transport through perturbative techniques (gas puffs as well as beam blips). | ||
| Background: | |||
| Resource Requirements: | Counter Bean, electron and ion heating | ||
| Diagnostic Requirements: | Reflectometer and every fluctuation measurement + good CER | ||
| Analysis Requirements: | SOLPS5, TGLF, GYRO, profile analysis | ||
| Other Requirements: | |||
| Title | 77: Influence of rotation and rotation shear on ELM suppression | ||
| Name: | mordijck smordijck@wm.edu | Affiliation: | The college of William and Mary |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Rick Moyer, Jim Callen | ITPA Joint Experiment : | No |
| Description: | -- | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform a typical ELM suppression experiment, full co-beams, in the resonant window, at the lowest possible I-coil current for suppression. Than replace 1 of the co beams with a counter beam, to change the toroidal rotation (and therefore Er), but same power input and other relevant parameters. Observe whether ELM suppression is still obtained at same I-coil current (or less/more current is needed) in correlation with the Er profile. If no suppression is observed, perform q-scan to see whether a new suppression window can be found. Document how changes in Er affect ELM suppression (other techniques on how to change Er close to the top of the pedestal are also welcome). | ||
| Background: | Recent results presented at the APS meeting [Moyer et al.] indicate that when ELMs are suppressed there is a change in Er close to the top of the pedestal. A similar trend has been observed for q95 experiments on DIII-D. This flattening of the Er close to the pedestal can change the stability of the plasma and might connect the DIII-D experiments with experiments on other machines, where even with strong density pump-out, ELMs are not suppressed. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 78: 3D magnetics plasma response measurements of n=1 perturbations for RWM analysis & model validation | ||
| Name: | King kingjd@fusion.gat.com | Affiliation: | Department of Energy |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | Edward (Ted) Strait | ITPA Joint Experiment : | No |
| Description: | Using the first plasma data obtained from the 3D magnetics upgrade, reconstruct the full 3D structure of the plasma response to an applied n=1 perturbation. The goals are to quantitatively validate code predications of MARS-F (and others) in full 3D, and determine mechanisms for code over prediction of the plasma response above 80% of the no-wall βN limit. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The emphasis of this experiment is obtaining the largest applied kink mode perturbation possible to maximize signal strength for initial measurements. This will be achieved by applying a full current, slowly rotating, I-coil n=1 perturbation to a neutral beam heated H-mode plasma. Because of the small predicted signals (δB3D/B0 ~ 10-4), synchronous detection is required.
(1) Generate a high βN discharge with q95 and I-coil phasing set to maximize the n=1 perturbation (240 degree upper-lower phase difference) (2) Scan βN between 1.2 and 2.4 at fixed q95 and plasma shape. The maximum βN will need to exceed the no-wall limit. For each shot, measure the toroidal variation of response (Bp and Br) at 7 poloidal locations (4 new). Generate a full 3D reconstruction of the plasma response at the vessel wall on both the HFS and LFS. Models predict the finest structure on the HFS, so the two vertical arrays of the 3D magnetics upgrade will be absolutely necessary to this experiment. | ||
| Background: | It was shown in M.J. Lanctot, et.al, Phys. Plasmas (2011) that the magnetic plasma response amplitude of an n=1 applied perturbation quantitatively agrees with linear ideal MHD predictions up to 80% the no-wall βN limit. Physics not included in this model results in significant over predictions of the response measurement above this point. The full structure of the mode has never been measured. This experiment will test the hypothesis that the poloidal spectrum varies with βN as the no-wall limit is approached. | ||
| Resource Requirements: | I-coil quartets are needed, requiring 3 bi-polar SPAâ??s to drive a rotating n=1 field. | ||
| Diagnostic Requirements: | The complete installation, testing and calibration of the 3D magnetics upgrade is required. Also, soft x-ray measurements are desired for determining the internal toroidal amplitude and phase of the perturbation. | ||
| Analysis Requirements: | Existing magnetics plasma response analysis tools are needed. 3D reconstruction tools, used for 3D magnetics design work, will need to be further developed to use new sensor data and synchronous detection. Also, predictive modeling with MARS-F & IPEC will be a benefit. | ||
| Other Requirements: | For reconstructions at the vessel wall to be related to the full 3D equilibrium at the surface of the plasma both the shape and q-profile need to be nearly identical for each value of βN scanned. | ||
| Title | 79: 3D magnetics equilibrium measurements to applied n=3 perturbations | ||
| Name: | King kingjd@fusion.gat.com | Affiliation: | Department of Energy |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | E.J. Strait and M.J. Lanctot | ITPA Joint Experiment : | No |
| Description: | Make first 3D magnetics measurements of the poloidal spectral changes on the inner wall due to an applied n=3 external I-coil perturbation. Specifically variations resulting from changes in q-profile, I-coil parity and βN are to be explored. This will provide a first test of the 3D magnetics sensitivity, and allow for detailed quantitative model validation (MARS-F, ect.) of the full eigenstructures resulting from applied n=3 perturbations. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The initial focus of this experiment is obtaining sufficient 3D signal for a physically relevant measurement, and will be followed by parameter scans. A lower single null ELMing H-mode discharge with the highest possible βN/li will be used. The plasma shape will be chosen to maximize kink mode coupling (upper triangularity ~ 0.1). The I-coils need to run at full current and polarity flips will be used to allow synchronous detection.
Odd Parity: (1) Run at q95 ~ 5.0. (2) At fixed q95 ~ 5.0, scan βN as low as possible before 3D signal is lost. (3) At fixed βN/li, scan q95 as low as possible before disruptions and/or 3D signals are no longer observable. Even Parity: (4) Switch to even parity and repeat (1) through (3) Using the new 3D magnetics pair of vertical inner wall arrays, measure the poloidal spectrum. These spectra can be combined with the LFS measurements to obtain a full 3D reconstruction at the vessel wall. Note, for reconstructions at the wall to be related to the full 3D equilibrium at the surface of the plasma the shape will need to be fixed while q95 is scanned. | ||
| Background: | It has been predicted using MARS-F and IPEC models that fine 3D eigenstructures are apparent along the inner wall during external n=3 I-coil perturbations. Measuring changes in these inner wall poloidal spectra may elucidate a variety of physics topics associated with applied 3D fields (e.g. RMP and NTV). Testing this capability is the purpose of this experiment. | ||
| Resource Requirements: | I-coils require 2 bi-polar SPAâ??s to drive an oscillating n=3 field. | ||
| Diagnostic Requirements: | The complete installation, testing and calibration of the 3D magnetics upgrade is required. | ||
| Analysis Requirements: | Existing magnetics plasma response analysis tools are needed. 3D reconstruction tools, used for 3D magnetics design work, will need to be further developed to use new sensor data and synchronous detection. Also, predictive modeling with MARS-F & IPEC will be used. | ||
| Other Requirements: | |||
| Title | 80: Poloidal spectral measurements of n=3 RMP ELM suppression using the 3D magnetics upgrade | ||
| Name: | King kingjd@fusion.gat.com | Affiliation: | Department of Energy |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | E.J. Strait and M.E. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | Use new 3D magnetics upgrade to measure suspected changes in the inner wall poloidal spectrum associated with n=3 RMP ELM suppression. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | By modifying collsionality, generate identical discharges with and without ELM suppression to study differences in mode structures. Apply an n=3 RMP. C-coil EFC may be useful to maximize the ELM suppression window. Maintain high βN (~2.2) in initial discharges to maximize 3D signal. Fixed parameters needed for successful 3D magnetics measurement: βN, shape, I and C-coil amplitudes, and q95. Also, variations in the overall q-profile should be minimized. In all cases the I-coils will need to oscillate in phase by 600 for synchronous detection.
High Collisionality: (1) ELM suppression shot: Apply an oscillating, odd parity n=3 RMP and maintain suppression through I-coil 600 phase shifts (2) Loss of ELM suppression shot: Reduce ν < 1.0 while maintaining βN constant by reducing density (stronger pumping), and increasing temperature (stronger NBI). Low Collisionality: (1) ELM suppression shot: Apply an oscillating, even parity n=3 RMP and maintain suppression through 600 phase shifts (2) Loss of ELM suppression shot: Increase ν > 0.4 while maintaining βN constant by increasing density (weaker pumping), and decreasing temperature (weaker NBI). After achieving both high and low collisionality cases at high βN, scan to lower βN until suppression is completely lost. | ||
| Background: | It is hypothesized that there is a change in the poloidal spectrum associated with the transition of an applied RMP to an ELM suppressed state. For example, does the perturbation become more or less ballooning-like during ELM suppression? This experiment aims to answer such questions. | ||
| Resource Requirements: | 2 bi-polar SPAs to drive an oscillating n=3. 3 SPAs to drive the C-coil. | ||
| Diagnostic Requirements: | The complete installation, testing and calibration of the 3D magnetics upgrade is required. | ||
| Analysis Requirements: | Existing magnetics plasma response analysis tools are needed. 3D reconstruction tools, used for 3D magnetics design work, will need to be further developed to use new sensor data and synchronous detection. Also, predictive modeling with MARS-F & IPEC will be needed. | ||
| Other Requirements: | |||
| Title | 81: Active Particle Exhaust of H-mode Plasmas in the Snowflake-minus Divertor Configuration | ||
| Name: | Petrie petrie@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | S.L. Allen, Vlad Soukhanovskii | ITPA Joint Experiment : | No |
| Description: | This experiment is the first attempt at evaluating how efficiently particles in an H-mode discharge can be actively removed in a Snowflake-minus configuration via cryo-pumping. Particle exhaust by the lower divertor cryo-pump is done at several radial locations relative to the entrance to the lower divertor pumping plenum. The pumping exhaust rate of the Snowflake plasma is plotted as a function of distance from the plenum entrance. These values are compared with the pumping exhaust rate of the corresponding standard configuration, for reference. This is done by pumping on the "standard" divertor case during the first part of the discharge until the pumping rate and core density are steady, then pulling the configuration into a Snowflake-minus configuration for the remainder of the shot. Care must be taken to insure that the floor tiles are properly conditioned, since putting the outer strike point(s) on unconditioned tiles will result in outgassing that will significantly complicate the analysis.<br><br>Measurables: Plot of particle exhaust rate and core line-averaged and pedestal density as a function of Rosp. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For this experiment, the ion grad-B drift is directed toward the divertor. To condition the divertor tiles, we first do a radial scan of the outer strike point of a standard divertor at high beam power: Rosp = 1.2 m at 1.5s, sweep radially to Rosp = 1.40 m by 3.5 s, then sweep back to 1.2 m by 5.5 s. Then repeat the shot.
The "data shots" include the following radial strike point locations: Rosp = 1.20 m, 1.25 m, 1.30 m, 1.335 m, and 1.37 m. Each shot includes both a standard divertor phase and a Snowflake-minus phase. Two shots at each Rosp location are taken, in order to establish that the tiles under the OSP are indeed properly conditioned and not evolving significant deuterium gas during the shot. | ||
| Background: | NSTX-U plans to make the Snowflake divertor an important part of their upcoming program. Their planning for operating with SF includes the capability to actively exhaust particles. Modeling has been done with SOLPS to try to simulate a cryo-pump capability similar to the DIII-D approach, but their is no hard data anywhere in the world for benchmarking the modeling. According to V. Soukhanovskii, this issue of adequately being able to pump Snowflakes is very high on the priority list for NSTX-U.
DIII-D has the capability (a) to run a Snowflake configuration compatible with pumping and (b) to obtain the relevant pumping data that would be very helpful in NSTX-U planning. | ||
| Resource Requirements: | 0.5 day experiment, 6 co-beam sources, some shaping preparation for maintaining SF(-) at slightly non-standard Rosp near the plenum entrance. | ||
| Diagnostic Requirements: | Asdex gauge in lower divertor plenum, IR camera (preferably the "periscope" system), filterscopes, bolometer, core Thomson scattering, and Langmuir probes (floor). | ||
| Analysis Requirements: | SOLPS | ||
| Other Requirements: | -- | ||
| Title | 82: IR Imaging of I-coil Induced Fast Ion Losses | ||
| Name: | Van Zeeland vanzeeland@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | W. Heidbrink, C. Lasnier | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to use the LLNL IR/Visible periscope to investigate fast ion losses from I-coil imposed field perturbations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will reproduce shot 150417 from 8/16/12, which utilized a 25Hz rotating n=2 I-coil field applied to a small circular plasma and was observed to induce measurable fast ion losses. The fast ion heat load on the wall will be monitored with the LLNL IR periscope for no applied field, applied rotating field, and successive shots with the field phasing stepped toroidally but held constant during a given shot. Fast ion losses will also be monitored with both FILDs, and the impact of the fields on the plasma will be measured with BES imaging, lithium beam spectroscopy, and reflectometry.
Note, this proposal is for a half-day experiment or less. | ||
| Background: | Energetic particle populations in tokamaks exhibit increased transport and possible loss due to the presence of non-axisymmetric (3D) fields such as those arising from test blanket modules, internal MHD instabilities, toroidal field ripple, general error fields, and edge localized mode/resistive wall mode (ELM/RWM) control coils. Losses of energetic particles cause localized heating with the potential to damage first wall components, making this an important practical issue for future burning plasma experiments, particularly for those considering coil sets to specifically impose non-axisymmetric fields.
Fast ion losses due to TBMs were studied in detail in recent DIII-D experiments and in August 2012, an experiment to investigate I-coil induced losses due to a rotating n=2 perturbation was also carried out. During planning, it was hoped the new LLNL IR periscope would be available to monitor the heat footprint of lost fast ions on the wall but the experiment was carried out before the full camera installation was complete. | ||
| Resource Requirements: | 210R, 30L, n=2 even parity icoil | ||
| Diagnostic Requirements: | IR Periscope, FILD | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 83: Establishing stationary I-mode with high performance | ||
| Name: | Whyte whyte@psfc.mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | A. Hubbard, M. Fenstermacher, A. White, J. Hughes, S. Gerhardt, R. Maingi | ITPA Joint Experiment : | Yes |
| Description: | The goal of this proposal is to establish I-mode in a stationary mode (constant heating power, density, etc.) with sufficient energy confinement performance that it can be regarded as a viable burning plasma operating scenario that does not require ELMs for particle and impurity control. The results will be combined with those from Alcator C-Mod (and ASDEX-Upgrade) to help extrapolate I-mode to ITER and to contribute significantly to the 2012 Joint Research Target of OFES. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The proposal builds on the successes of the 2012 I-mode exploration on DIII-D. Those experiments explored a wide operating space in Ip, q95, density and heating power with ion grad-B drift pointed away from the X-point (the configuration in which I-mode is typically found). An I-mode was tentatively identified (#149908) during the exploration, showing a substantial edge T pedestal, with an L-mode density profile, and significant improvements in confinement (H98 up to 0.8) compared to standard L-mode. There were qualitative indicators of reduced fluctuations in the pedestal during the I-mode phase, however this diagnosis was not optimized.
The initial experiment indicated that either shape (triangularity) and/or divertor topology were critical in obtaining I-mode in LSN, with a preference for smaller lower triangularity and the outer strikepoint positioned on the top of the lower divertor baffle. Due to the limited time a full exploration of the effect of shaping was not possible. Also, the I-mode was not maintained in a stationary condition due to the power scan that was present in each discharge. The proposed experiment would use the 2012 results as a starting point for a) establishing the I-mode in a stationary condition with constant heating. b) fully diagnose evolution of edge fluctuation characteristics c) explore the effect of shape/divertor changes on the I-mode operating window (X-point height, total triangularity) | ||
| Background: | I-mode is a relatively new operating regime that has been recently explored on C-Mod and ASDEX-Upgrade. At its simplest, I-mode features an edge energy barrier in the form of a temperature pedestal, yet retains L-mode particle confinement with no evident edge density barrier (making it distinct from H-mode). The results on C-Mod and AUG indicate establishment of H-mode energy confinement without the need for ELMs, making it attractive as an alternate operating regime in ITER to avoid ELM-induced damage. Obviously it will be important to compare and contrast I-mode on several devices to determine effects of heating schemes, wall/divertor/plasma geometry, size, etc. | ||
| Resource Requirements: | One day of experimental time
Reverse-B Beam and ECH heating Divertor cryopumping | ||
| Diagnostic Requirements: | Full set of pedestal / boundary fluctuation diagnostics to examine changes in broadband fluctuations and any weakly coherent, high frequency modes often associated with I-mode in other devices. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 84: Influence of resonant/non-resonant magnetic perturbation on particle confinement with n=2 | ||
| Name: | Jakubowski marcin.jakubowski@ipp.mpg.de | Affiliation: | Max-Planck Institute for Plasma Physics |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Todd Evans, Andrew Kirk, YoungMu Jeon, Wolfgang Suttrop | ITPA Joint Experiment : | Yes |
| Description: | Aim of the experiment is to understand what type of interaction between magnetic perturbation and the plasma equilibrium leads to the enhanced particle transport and thus so-called pump-out. This is part of the broader inter-machine comparison with ASDEX-Upgrade, KSTAR and MAST and supports high priority ITPA activities defined as the goals for 2013 within PEP-23 and TC-24 joint experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a standard LSN (ITER-like) shape with n= 2 RMP mitigation (q95 ~ 3.35, double I-coil row) and vary the pitch of the magnetic perturbation by changing the phase of top I-coils while keeping bottom coils fixed (6 steps). At each step perform an additional discharge with n=2 and q95 at the edge of the resonant window (q95 ~ 3.2). | ||
| Background: | According to the present understanding RMPs reduce pressure gradients in the pedestal region by introducing a stochastic boundary allowing suppressing or mitigating ELMs while keeping the outward impurity transport enhanced. On DIII-D (in contrast to ASDEX-Upgrade) the pump-out is almost immanently associated with the ELM suppression or mitigation by external field. Also on DIII-D the pump-out window is much wider as compared to the ELM suppression window, which indicates that pump-out could be an effect of non-resonant field interacting with plasma. However, the results from MAST suggest that only pitch resonant coupling of magnetic perturbation with plasma results in enhanced particle transport. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 85: Current profile control for stable ITER baseline scenario plasmas | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys, G. Jackson, T. Luce, W. Solomon, M. Walker | ITPA Joint Experiment : | No |
| Description: | Analysis of ~100 ITER b.s. discharges has indicates that detailed features of the current profile in the outer half of the plasma are closely linked to the triggering of tearing modes that limit the performance and duration of this scenario. Global changes in the current profile are likely responsible for changes in the classical tearing stability term (Î?â??), that destabilize the 2/1 tearing mode. While direct stabilization of an island has already been demonstrated, this analysis and experiment is aimed at modifying the global current profile to avoid having to directly stabilize any island and without pre-emptive stabilization.
This experiment would provide â?¢ a direct test of the hypothesis postulated from a 2-yr analysis (see background section) using a controlled environment where the current profile changes are systematic and externally controlled â?¢ a method to control the current profile in the region where it matters for tearing stability, to steer the plasma away from tearing stability boundaries, without sacrificing betaN, performance or rotation |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Run an ITER discharge placed on the stability boundary (possibly low torque) and monitor the MSE pitch angle time history of the stable and unstable branch in the region of interest. Correlate with RT-efit reconstructions of J. Place single gyrotrons at the locations between MSE channels 11-44, roughly between rho=0.5 and rho=0.9.
b. Checkout the new MSE-ECCD feedback algorithm (take data in the PCS and check the response, without activating the algorithm â?? described under analysis requirements), while using feed-forward, shot-to-shot control of J: for the unstable case, repeat adding ECCD at the location that shows narrowing of the MSE channels (supposed to lead to instability). Create cases that become unstable or remain stable â??at willâ??, by using co- and counter-ECCD at different locations. c. Use feedback to control MSE channels, and hence J: link the difference between 2 contiguous MSE channels signals to the gyrotron(s) aimed at the location between those two MSE channels - when the difference falls below a threshold (TBD), turn on the corresponding gyrotron(s). Check the threshold values, the timing of the feedback, the necessary gyrotron power (modulated ECCD power ok?). d. Run discharges with the feedback on, in all the locations of the stability boundary that would not be stable otherwise (low/zero torque, low density, lower li, etc). The algorithm will measure and control the MSE pitch angles (channels 11-44), by modifying the current profile by means of ECCD, to reduce or increase the MSE pitch angles differences locally (which represent the amount of plasma current present at that location). | ||
| Background: | The ITER baseline scenario is know for being prone to the triggering of m=2,n=1 tearing modes, that spoil the performance, perturb the current profile in a way that is not recoverable, and often lock to the wall forcing an early termination of the discharge. It was recently demonstrated that the evolution of the current profile is the main effect responsible for the triggering of this instability[1]. Analysis of >100 ITER baseline scenario shots, and new experiments in the 2012 DIII-D campaign have explored the tearing stability boundaries in betaN, li, torque, and density. The statistical analysis shows that all the discharges that have â??j95nâ??>0.5 (average normalized current at 95% of the poloidal flux) are stable, while with lowed j95n only high density shots are observed to be stable. Independent analysis of raw MSE pitch angles time histories for a small sample of the discharges reveals that the statistical study is consistent with the trends observed in the raw data. In the unstable cases, the MSE pitch angles evolve in such a way that less current is detected at rho~0.8 and more current builds up at rho~0.6 (the total Ip is kept constant). The q=2 surface is located between these two rho values. This points to a rather simple explanation for the triggering of the limiting instability in these discharges: when less current is present outside the q=2 surface (whatever the physical mechanisms that caused this), more current will build up on the inner side because the total current is kept constant. This causes a higher current gradient to be created at the rational surface, which is the main destabilizing effect for tearing modes.
Since the raw MSE time histories provide direct information about the local current profile, independently from any equilibrium reconstruction, this study gives an independent confirmation of the statistical observations, which are based on reconstructed efit equilibria. We propose to directly check if the physics hypothesis is correct, by controlling, off-line, in feedforward and in feedback, the local features of the current profile, that were observed to be related to the tearing stability of the discharges. Note that this method does not rely on direct stabilization of an existing island, nor on the continuous ECCD power to pre-emptively stabilize tearing mode, but instead it constitutes a physics basis for tailoring the current profile to avoid having to use direct stabilization at all. This will also directly provides a simple way to steer the discharges away from instability, without sacrificing performance (i.e. not lowering betaN), and potentially with minimal extra injected power during the discharge. [1] F. Turco and T.C. Luce, Nucl. Fusion 50 (2010) 095010 | ||
| Resource Requirements: | 30 and 330 NBI sources, both 210 NBI sources. 6 gyrotrons at max power (stand-by power capability needed). | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer, reflectometer | ||
| Analysis Requirements: | Create and implement MSE-ECCD feedback control algorithm in the PCS before the experiment. Basic logic needed:
- measure RT-MSE pitch angles and have at least channels 11-44 available in the PCS algorithm. Compute differences between contiguous channels (40-11, 41-40, etc). Connect this category to a dud trip that switches phases in the gyrotron power category. - Assign one (or 2, depending on the configuration, TBD) gyrotron power signal to one MSE difference-signal (gyrotrons have fixed positions, the assignments are done before the experiments, off-line) - Set a threshold â??Tâ?? for the MSE differences (TBD): when 1 or more (TBD) of the signals go below T, turn on the corresponding gyrotron(s) - When the MSE-difference signal returns below T, turn off the corresponding gyrotron. | ||
| Other Requirements: | |||
| Title | 86: Safety Factor Scaling of Turbulence and Transport in Hybrid Scenario Plasmas | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, C. Holland, J. Kinsey, T. Rhodes, L. Schmitz, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Energy confinement in H-mode discharges varies strongly with plasma current (with q as the relevant dimensionless parameter) as approximately tau_E ~ Ip^(0.9). This basic dependence is well known empirically but lacks a clear fundamental explanation in terms of basic turbulent transport that is consistent with theory and simulation. Connecting this empirical dependence with the underlying turbulence behavior is crucial to understanding confinement scaling as well as to validating transport simulations over a wide range of parameters. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will perform this experiment in quasi-steady hybrid H-mode plasmas and vary the safety factor systematically. The toroidal field will be held constant (2.0 T) and plasma current will be adjusted to vary q95 in the range of 3.0 | ||
| Background: | Many experiments have shown a strong dependence of confinement and transport on plasma current. Previous DIII-D experiments have shown that turbulence and zonal flow/GAM characteristics vary with q95. This experiment would seek to obtain comprehensive turbulence and full transport measurements with multiple fields, a wide wavenumber range, and over a wide radial extent in well-characterized discharges. Calculations of turbulence and transport by gyrokinetic simulations have generally agreed reasonably well with measured turbulence parameters and inferred heat fluxes at mid-radii in L-mode plasmas [Holland-PoP-2009], while at larger radii, simulations routinely underestimate both fluctuation amplitudes and transport. This experiment will seek to determine wether this "L-mode shortfall" is present in H-mode plasmas; work by Kinsey suggests that there is a larger shortfall at higher q95, which is consistent with preliminary analysis of a 2012 L-mode experiment [T. Rhodes et al.] These experimental results will form a foundation for future investigations aimed at optimizing transport in advanced scenarios with non-monotonic q-profiles. | ||
| Resource Requirements: | Neutral beams (all) | ||
| Diagnostic Requirements: | Fluctuation diagnostics: BES, DBS, CECE, ECEI, FIR, PCI, UF-CHERS
Full Profile diagnostics | ||
| Analysis Requirements: | Transport Analysis; TGLF; GYRO | ||
| Other Requirements: | |||
| Title | 87: farSOL expts at high n/nGW to establish if ITER farSOL will be high-recycling/detached | ||
| Name: | Stangeby peter.stangeby@utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | Tony Leonard, Richard Pitts (ITER), Dmitry Rudakov, Clement Wong, | ITPA Joint Experiment : | No |
| Description: | Motivation: the proposed experiment will provide important information for the ITER decision about PFC choice for targets at the 2ndary divertor at the top (ITER will operate with unbalanced double-null). <br>ITER specs for region outside 2nd separatrix, "farSOL", are based on very little experimental information - particularly for high n/nGW, where most needed, since ITER will use high n/nGW. These plasma conditions apply to the 2nd strike point at the top, where ITER is now considering replacing the Be with W, because of the expected, I.e. ASSUMED, intensity of the plasma-solid interaction there.<br>The present ITER design/planning assumes the farSOL is Sheath-Limited, S-L. For S-L flux tubes, there is ~no drop in plasma temperature along the length of the flux tubes and so the sputtering rate can be high at the solid surfaces which terminate the flux tubes. In ITER there will be two types of farSOL flux tubes involved in the plasma contact with the Be surfaces: divertor-type and limiter-type. The divertor-type is due to the fact that ITER will use unbalanced DN with the 2ndary divertor at the top; this is therefore where the most intense plasma-Be interaction will occur. In addition, limiter-type contact will occur over a much larger fraction of the total Be wall area of 700 m2, albeit at a less intense level, but with potentially major consequences for the erosion-wear rate of the thin Be armour and tritium retention by Be-codeposition . Both types of plasma-Be contact are therefore important in ITER and for both types it is important to know if the farSOL flux tubes involved are in the S-L regime, or in the much more attractive H-R (high-recycling) or DET (detached) regimes, where there is substantial drop of plasma temperature along the length of each flux tube, resulting in reduced sputtering at the Be surfaces at the ends of the flux tubes. <br>There is evidence in DIII-D that even for limiter-type wall contact, the farSOL flux tubes may be in H-R: in DIII-D window-frame expts [Leonard JNM (2007]: "At high density, the baffle probe saturation current is higher than the midplane probe and may indicate .. local recycling." There is a still stronger tendency for divertor-type flux tubes to transition into H-R/DET than limiter-type flux tubes, so DIII-D should study both at high n/nGW.<br>The farSOL at high n/nGW may be H-R/DET (high-recycling/detached) with major practical - positive! - implications for ITER.<br>It is proposed to carry out farSOL expts at high n/nGW in DIII-D using (1) unbalanced DN, and also (2) SN with window-frame, to establish if ITER farSOL is likely to be H-R/DET.<br>In 2008 we carried out unbalanced DN expts (13C-methane injection) to simulate the ITER farSOL, using n/nGW - 0.63. The proposed expts will use the same configuration but raise n/nGW to 0.8 or higher. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | High power, H-mode. Two densities at similar power, one at highest n/nGW, one 20% lower n/nGW. Two values of dsrsep. IR measurements of the 2ndary target area. (1) Use unbalanced DN with 2ndary divertor at the bottom, where edge diagnosis is better, to study the divertor-type farSOL. (2) Also use LSN with small wall gap to the 'nose' of the upper outer pumping plenum, which creates toroidally symmetrical wall-contact, thus facilitating farSOL analysis using the 'window-frame' method for limiter-type contact farSOL studies. | ||
| Background: | Until ~10 years ago it was thought that there was very little plasma contact with the main walls of tokamaks. It is now known that at high n/nGW - where ITER will operate - the total ionic flux to the main walls becomes comparable to the total ionic flux to the primary divertor. This "farSOL" plasma has major implications, direct and indirect, for (i) power loading of the walls, (ii) fuel recycling, (iii) impurity generation by wall sputtering. Direct: due to the ion impact on the walls. Indirect: due to the charge exchange neutral particles that result from the ion-wall contact. Re sputtering, the latter can be more important than the former because the charge exchange neutrals arising from main wall contact can be very energetic, originating from deep in the confined plasma.
The vast majority of SOL research to date has focussed on the nearSOL, specifically the narrow power-channel that goes to the primary targets. By contrast the farSOL is still largely unexplored, particularly at high n/nGW. ITER will operate with unbalanced double-null with the 2ndary divertor at the top. Thus the most intense farSOL wall contact in ITER will be divertor-like rather than limiter-like. This increases the probability that the farSOL in ITER will not be in the sheath-limited regime - the usual assumption - but could be in the high-recycling state or even detached. This would have major - largely positive - implications for ITER, certainly for power-loading on the 2ndary targets (the cx fluxes, on the other hand, would increase, although not necessarily the cx sputtering rate). In the present ITER design the 2ndary targets are to be clad with the low-melting metal Be and it is being considered to replace the Be there with W. The latter, however, would have significant potential to cause high W concentrations in the confined plasma. It is therefore important to know if Be targets will be acceptable - which they may be if the 2ndary outer divertor is high-recycling or detached. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 88: inner column limiter expts to establish if ITER wall-limiter design should include �??funnel effect' | ||
| Name: | Stangeby peter.stangeby@utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | Tony Leonard, Richard Pitts (ITER), Dmitry Rudakov, Clement Wong, | ITPA Joint Experiment : | No |
| Description: | Motivation: The proposed experiment will provide important information for the ITER decision in 2013 about the Be wall-limiter design.<br>ITER will startup/rampdown on inner wall as limiter.<br>Present ITER wall design assumes q-parallel loading only and does not include the "funnel effect" (direct q-perpendicular loading).<br>On TFTR, JET there was/is power deposition at inside midplane that can not be explained by q_par loading and has been ascribed to a funnel effect (q_perp loading).<br>There is also some other experimental evidence for the funnel effect; however, it remains a poorly explored, understood effect; accordingly the decision was made to not include it in ITER Be wall design shaping; this is now being questioned; final design decision will be next year.<br>It is proposed to run limiter plasmas on DIII-D central column, with ir viewing, to get data on funnel effect for ITER wall design decision. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | It is proposed to run L-mode limiter plasmas on DIII-D central column, with ir viewing, to get data on funnel effect for ITER wall design decision. Use the same range of densities and plasma currents as in the previous DIII-D experiments in support of establishing a startup/rampdown database for ITER, Dmitry Rudakov [2011 APS]. | ||
| Background: | The "funnel effect" is described in Sec. 25.2 of Stangeby "The Plasma Boundary of Magnetic Fusion Devices", and references there. See also Baelmans, Reiter et al, JNM 290-293 (2001) 537, which report B2-EIRENE code analysis of the funnel effect, including application to the TEXTOR ALT-II limiter. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 89: Mo, W, Al, Be DiMES net/gross erosion experiments | ||
| Name: | Stangeby peter.stangeby@utoronto.ca | Affiliation: | U of Toronto |
| Research Area: | Plasma-material Interface | Presentation time: | Requested |
| Co-Author(s): | Chris Chrobak, Bill Wampler (Sandia), Tony Leonard, Dmitry Rudakov, Clement Wong, | ITPA Joint Experiment : | Yes |
| Description: | In theory, prompt local deposition of sputtered particles should reduce net relative to gross erosion. Definitive experimental evidence is needed.<br><br>Al: proxy for Be (PhD project of Chris Chrobak). The DIVIMP code is being used by ITER to design the Be wall and needs to be bench-marked. Chris made a start in 2012. First Al layer melted, 2nd did not. With new solutions, try again.<br><br>Mo: while the net/gross erosion of Mo was in good accord with expectations, transport of the Mo which did not deposit on the sample (~50%) seems to migrate further than one would expect. Evidently caused by a strong mixed-materials effect due to the substrate material that surrounded the Mo sample, namely C, which erodes very quickly cf Mo, thus rapidly "pulling the rug out from under" deposited Mo. Repeat the Mo expt but with W as the surrounding substrate material. The sputtering rate for W is significantly less than for Mo and so should make for a more interptretable expt.<br><br>W: the most important high-Z. We didn't start with W because of uncertainties in the S/XB, but now we have demonstrated a non-spectroscopic method of measuring gross erosion so we can usefully do W studies with DiMES. This will provide a measurement of S/XB which will be valuable for applications generally, specifically to W expts in other tokamaks where only a spectroscopic method can be used, requiring reliable a value for photon efficiency (S/XB).<br><br>For safety reasons, DIII-D cannot use significant amounts of Be and the permitted level is now 10X lower than it was the last time Be DiMES experiments were done, in the 1990s. This may preclude experiments on net/gross erosion, which require a 1 cm diameter Be layer; however, it will still be possible to perform an extremely important experiment to measure the absolute erosion (gross erosion) rate, i.e. sputtering yield for D-ion impact, Y_Be_D. This will be done using very small, 0.01 cm2 area (vs ~ 1 cm2 before) and very thin, 20 nm (vs 100 nm before) coatings, using our new discovery of a non-spectroscopic method for measuring gross erosion, namely by measuring the net erosion of a very small sample using ex situ ion beam analysis. For very small samples, gross erosion ~ net erosion. The present database for Y_Be_D has a major problem since the values from careful PISCES experiments are an order of magnitude less than measured with ion accelerators and spectroscopically in JET. This has serious consequences for ITER since it causes an uncertainty of an order of magnitude in the predictions for the lifetime of the Be wall as well as for the rate of tritium retention by Be-codeposition. DIII-D is in a unique position to make high quality measurements of Y_Be_D in the actual tokamak environment, under well-controlled, well-diagnosed conditions - and without the uncertainties of the spectroscopic method for measuring atomic influx rates in a tokamak. The uncertainties in ion beam analysis are much smaller than the spectroscopic one | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 90: Carbon-13 in He Plasma erosion/deposition experiment to benchmark DIVIMP code being used by ITER | ||
| Name: | Chrobak cchrobak@cfs.energy | Affiliation: | Commonwealth Fusion Systems |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | P. Stangeby, A. Leonard, G. Tynan | ITPA Joint Experiment : | No |
| Description: | Provide experimental data for low-Z (carbon) physical erosion and deposition under simple as possible helium plasmas in LSN configuration for benchmarking plasma material interaction codes DIVIMP and WBC-Redep. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The ideal material to expose for the purposes of these experiments is Be, but due to its extreme safety hazards, a suitable substitute may need to be used. Al has been identified as a suitable proxy for Be, but after testing in 2012 was found unable to withstand high power loads before melting. Thus, carbon is left, requiring the use of He plasmas to remove the chemical erosion factor and a Carbon-13 enriched erosion sample is required to detect its presence over the carbon background. It is proposed here to expose a specially-designed sample with a carbon-13 enriched surface layer to dedicated, repeat, well-characterized plasma shots. Measurements of the sputtering yield and material influx plume would be done spectroscopically using the MDS spectrometer view chords and narrow band filtered visible light cameras, as well as post-mortem measurements of the exposed surface by ion beam nuclear reaction analysis. | ||
| Background: | Theory predicts that for erosion of high-Z materials, prompt-local deposition of eroded material is dominant, and for low-Z materials, long-range transport and deposition is dominant. However, initial measurements of Mo erosion in 2011 found that only 20% of the net eroded material was found immediately surrounding the sample [4,5]. By contrast, for the low-Z Be DiMES sample in the 1996 experiment [3], only about half the Be that was eroded from the Be sample was found on the graphite surface of the DiMES head. These discrepancies indicate gaps in the current theory, and further stress the need for accurate measurements of gross and net erosion from PFCs.
From our 13C-methane injection experiments in DIII-D [6] we know that much of the low-Z launched from the main wall is transported long range, e.g. from the top of DIII-D to the bottom. About half the total 13C that was injected at the top of the LSN discharges was found in the bottom divertor. We have less of a handle on the other half of the injected 13C, but it appears to have been deposited short-range, on the main wall (short as distinguished from local) The idea that the wall is a source of sputtered impurity which all ends up in the divertor sink is too simplistic: parts of the wall are in a state of net erosion while other parts are in a state of net deposition from wall sources elsewhere. Additionally, within the region immediately surrounding the strike points there are net erosion and net deposition zones that change with varying strike point location and plasma confinement mode [7]. ITER urgently needs proper benchmarking of edge impurity transport codes DIVIMP for low-Z impurities to estimate the net erosion of the Be wall armor and the tritium retention by Be co-deposition [1]. Experiments prepared on EAST and JET to benchmark low-Z erosion/deposition (C in EAST, Be in JET) examine only limiter-type plasma contact. Although most of the area of the ITER Be wall will experience limiter-type contact, most of the actual erosion will occur at the upper, second divertor, and will therefore involve divertor-type plasma contact. Hence, it is essential to properly benchmark the DIVIMP code for low-Z materials in divertor-type plasma contact. DIII-D is in a unique position to provide these results due to the high quality divertor and edge diagnostic suite and DiMES material exposure system. | ||
| Resource Requirements: | 13C-coated Sample Fabrication
PIGE Nuclear Reaction Analysis with Ion Beam | ||
| Diagnostic Requirements: | DiMES TV camera view with spectral filter for C emission
MDS spectrometer views 1) on DiMES and 2) just off DiMES Tangential TV camera views of He and C emission Langmuir probes Divertor-Thomson DiMES Thomson CER | ||
| Analysis Requirements: | TBD | ||
| Other Requirements: | |||
| Title | 91: Complete the Development of PCS n=2 Error Field Control Algorithm | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | C. Paz-Soldan, J.M. Hanson, E.J. Strait, J. King, R.J. LaHaye, A.M. Garofalo, N. Ferraro, J.-K. Park, N. Logan | ITPA Joint Experiment : | No |
| Description: | Complete compass sweep experiments at two values of q95 in order to obtain the fit coefficients that will allow a PCS algorithm to be implemented. A PCS algorithm has been implemented and will be tested as part of this ½-day experiment. This will establish sufficient n=2 EFC for low beta operation. The experiment naturally incorporates a stringent test of plasma response and NTV theory. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Follows jist of MP 2012-83-09, approved but not given time in 2012)
1. Repeat compass scan at q95=3.4 with reference 149315, but at ne~0.75e13 cm-3 to increase the sensitivity of the plasma to locking, i.e. less n=2 current needed. 2. Perform compass scan also at q95=4.6 3. Compare low-density locked mode limit for both cases to no correction case At least 10 good ohmic shots | ||
| Background: | Results from 2012 error field control experiments using the compass sweep technique indicate there is a significant n=2 error field in DIII-D, consistent with in situ vacuum measurements from 2001 and 2003. The inferred control currents are ~50% of the standard n=1 control currents. Using the expected intrinsic error field from SURFMN, plasma response calculations with the IPEC code find that the n=2 pitch-resonant fields are ~5-10x smaller compared to n=1, which suggests the n=2 error field limits performance through non-resonant braking. This is consistent with the experimental observation that the application of a large n=2 I-coil field leads eventually to the onset of an n=1 locked mode. These conditions are ideal for understanding the relative importance of NTV torque in error field compensation metrics compared to kink-resonant and pitch-resonant harmonics. | ||
| Resource Requirements: | This low power ohmic experiment does not require auxiliary heating of any kind. We also operate at reduced toroidal field and plasma current. However, it requires all available 3D coil power supplies (4 SPAs and 2 C Supplies) in order to power 6 independent I-coil circuits. | ||
| Diagnostic Requirements: | Essential diagnostics are magnetics & CO2 and 288 GHz interferometers | ||
| Analysis Requirements: | Standard analysis of compass sweep results. Results Comparisons with plasma response codes: IPEC and M3D-C1 | ||
| Other Requirements: | |||
| Title | 92: H-mode Optimization of n=1&2 Error Field Control Currents | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | C. Paz-Soldan, J.M. Hanson, E.J. Strait, J. King, R.J. LaHaye, A.M. Garofalo, N. Ferraro, J.-K. Park, N. Logan | ITPA Joint Experiment : | No |
| Description: | Maximize the ratio of the plasma rotation to the injected neutral beam torque by optimizing n=1&2 error field control currents in H-mode | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Determine the optimal n=2 I-coil error field control currents in H-mode by measuring changes in the plasma rotation as a function of I-coil current. Identify the amplitude and phase of the n=2 coil currents by maximizing the plasma rotation. Use plasma beta ramps and simultaneous n=1&2 RWM feedback (n=2 algorithm to be developed) with I & C coils to minimize magnetic field asymmetries. Compare H-mode results with other 2013 experiments aimed at developing a standard n=2 error field control algorithm in ohmic discharges. | ||
| Background: | Previous experiments by Garofalo and others have shown that the optimal error field control currents in ohmic and high-beta experiments are different. Typically, the toroidal phase of the n=1 control currents are within 10-20 degrees, but the amplitudes can differ by up to 2x. This may be due to differences in the intrinsic error field (say from changes in the F & B coil currents), or in the plasma response. In ohmic discharges, the minimization of the low-density locked mode threshold identifies the optimal currents, while in H-mode, the optimal currents are found by maximizing the plasma rotation or minimizing magnetic field asymmetries using RWM feedback.
See proposal #91 for further background on n=2 error field control. Following the completion of the 3D magnetics upgrade in 2013, DIII-D will have a dramatically improved capability to detect n=even magnetic perturbations at multiple poloidal locations. The development of a PCS n=2 RWM control algorithm is planned for 2013, which will allow simultaneous n=1&2 dynamic error field control. | ||
| Resource Requirements: | Requires 3 SPAs on C-coil and Audio Amplifiers on I-coil. | ||
| Diagnostic Requirements: | 3D magnetic arrays, CER rotation measurements, and other diagnostics for kinetic equilibrium reconstructions | ||
| Analysis Requirements: | Use established techniques to analyze coil-current-dependent rotation changes. Compare results with first principles error field control metrics. Compare results with standard n=2 EFC algorithm. | ||
| Other Requirements: | PCS development of n=2 RWM feedback algorithm | ||
| Title | 93: Active Control of the EHO in Quiescent H-modes | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Plasma Control | Presentation time: | Requested |
| Co-Author(s): | H. Reimerdes, K.H. Burrell, A.M. Garofalo, J.M. Hanson, P.B. Snyder, W.M. Solomon, B. Tobias, L. Yu | ITPA Joint Experiment : | No |
| Description: | This experiment will test the feasibility of actively driving the edge harmonic oscillation, which is the essential feature of the Quiescent High confinement mode regime. The EHO is a naturally occurring edge-localized mode in the 5-10 kHz range that enhances particle transport without degrading the thermal transport barrier. This continuous mode provides density control with H-mode confinement without the deleterious effects of edge-localized modes. The prevailing theory of the EHO posits that it is a saturated kink-peeling mode driven by shear in the edge rotation. This rotation shear requirement limits the QH-mode operating space, which might be extended if external control of the EHO is possible. In this initial experiment, we will attempt to couple to a pre-existing saturated EHO using externally applied fields. We will also attempt to drive the mode at the boundary of the operating regime. Following such a demonstration, future experiments will focus on expanding the QH-mode operating regime and probing the stability of the EHO. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This plan for a Thursday evening experiment requires access to QH-mode. Therefore, it should be conducted following a full-day experiment that will operate in the QH-mode regime.
Shot 1. Reproduce a counter-rotating QH-mode. Use magnetics, BES, and ECEI to measure natural EHO amplitude and frequency. In the second half of the QH-mode phase, slowly decrease the neutral beam power to find the threshold where the ELMs return (expect ~ 3MW). Note changes in EHO structure and frequency during power scan, if any. Shot 2. This shot has two phases: one with sufficient beam power to access QH-mode, and the other with the NBI power epsilon-smaller than the threshold found in shot 1. Throughout the shot, apply a magnetic perturbation with the I-coil at the EHO frequency. Measure I-coil induced changes in the EHO amplitude during QH-mode phase. Look for any I-coil enhanced D-alpha emission. Observe if EHO is driven to finite amplitude and ELMs avoided when beam power is below the threshold for QH-mode. Shot 2a. If there is evidence for active control of the EHO, repeat shot 2 with half the I-coil current. Note change in driven EHO amplitude, if any. Shot 2b. If there is evidence for active control of the EHO, repeat shot 2 with a smaller outer gap during the second phase to vary the plasma-coil distance. Shot 2c. If there is evidence for active control of the EHO, repeat shot 2 with a programmed triangle waveform in the I-coil frequency during the second phase. Shot 3. Reproduce a counter-rotating QH-mode. In the second half of the QH-mode phase, puff gas to raise the density until ELMs return. Shot 4. Repeat shot 3 with I-coil field. Observe if EHO is driven to finite amplitude and ELMs avoided when the density exceeds the threshold identified in shot 3. Shot 4b. If there is evidence for active control of the EHO, repeat shot 4 with even higher density. | ||
| Background: | The EHO is a key feature of the QH-mode. It provides the sufficient particle transport to allow the plasma to reach a transport steady state at edge parameters below the explosive ELM limit. Previous experiments have demonstrated that external fields can influence the toroidal mode number of the EHO using odd parity n=3 NRMF. One of the most fascinating observations is the switch of the EHO from n=1 to n=3 when odd parity NRMF is applied. This is evidence that external fields can interact with the EHO, although this interaction may be indirect, i.e. via changes in the kinetic profiles. Examples of direct interaction between magnetic fields and other MHD modes are well documented. For example, internal saddle coils have been used on JET to drive global Alfven waves in the range from 30-70 kHz, and on DIII-D to drive marginally stable RWMs in the range from -40 to +60 Hz (the sign denotes the direction w.r.t. the plasma rotation). If the EHO is indeed a kink-peeling mode, then, in principle, it should be possible to drive this MHD mode to finite amplitude when it is close to marginal stability provided there is sufficient coupling between the I-coil field and the stable EHO. | ||
| Resource Requirements: | I-coils with audio-amplifiers, preferably with two parallel amplifiers powering pairs of I-coils. Because of this special I-coil patch panel, this experiment is unlikely to be done during a dedicated QH-mode experiment. | ||
| Diagnostic Requirements: | 3D magnetics, BES, and other diagnostic for kinetic equilibrium reconstructions | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 94: Turbulence Structure and Amplitude Variation with ExB Shear in H-mode Plasmas | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | Burrell, Holland, Petty, Rhodes, Schmitz, Yan | ITPA Joint Experiment : | No |
| Description: | Determine how the 2D eddy structure of low-wavenumber turbulence as well as its magnitude and spectra vary as a function of rotationally controlled ExB shear in long-duration hybrid H-mode plasmas. Quantify the effects of shear flow on core turbulent and transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Determine turbulent eddy structure as a function of Mach number in hybrid discharges by directly measuring eddy correlation function, magnitude, decorrelation rates, and radial & poloidal correlation lengths, in these hybrid discharges with the expanded 2D BES system, as well as the multichannel Doppler reflectometer system. Long-duration, steady hybrid discharges will be employed. The low-amplitude of fluctuations in the core of hybrid plasmas makes their measurement more challenging, but the steady qualities (several seconds) allow for ensemble-averaging of the fluctuation characteristics with good resulting signal-to-noise. This will allow us to examine the improved transport in hybrid discharges, and specifically the Mach number dependence, as well as to more broadly and generally examine the ExB shear effects on turbulence and transport. Discharges similar to those already developed by C. Petty et al. will be used, with the exception that the neutral beams used for beta feedback will be changed to allow for the BES measurements (which require a steady 150 left beam). We will run relatively high q95 (~5.5) to increase turbulence magnitude. Several repeated discharges would be performed for full radial measurements. | ||
| Background: | The control and suppression of turbulence by ExB shear is fundamental to confinement physics in toroidally-rotating plasmas. Its explanation in terms of comparing turbulence growth rates and measured ExB shearing rates, and general agreement with simulations, is a significant success in fusion research. Despite this triumph, there remains a lack of fundamental quantitative experimental measurements that link the measured changes in localized core turbulence characteristics with the applied rotation and resulting shear. This experiment aims to resolve this gap. Transport in Hybrid scenario discharges has been shown to depend strongly on the toroidal Mach number (M = v_tor /c_s). 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 (Polizer/Petty-Nuclear Fusion, 2008). This has been shown to be consistent with the a reduction in ExB shearing at lower Mach number from GLF23 modeling. Furthermore, previous measurements of turbulence characteristics in hybrid discharges (McKee, APS-2005) with BES, showed that turbulent eddies exhibit a strongly tilted structure in co-injected hybrid discharges. This is in sharp contrast to the more radially-poloidally symmetric eddy structure typically observed in the core of (rotating or non-rotating) L-mode discharges. The direction of this tilted eddy structure is consistent with the ExB shear flow in these plasmas, although it the shear magnitude didn't appear to be large enough to bring about the measured eddy tilt. | ||
| Resource Requirements: | All co and counter-neutral beams | ||
| Diagnostic Requirements: | Fluctuation diagnostics: BES, DBS, CECE, ECEI, FIR, PCI, UF-CHERS
Full Profile diagnostics | ||
| Analysis Requirements: | Transport Analysis; TGLF; GYRO | ||
| Other Requirements: | |||
| Title | 95: Experimental Validation of a Revised PCS Error Field Category | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Plasma Control | Presentation time: | Not requested |
| Co-Author(s): | D. Humphreys, J.M. Hanson, J. King, E.J. Strait, G. Jackson, A.M. Garofalo | ITPA Joint Experiment : | No |
| Description: | In a Thursday evening session, perform a thorough validation of a revised version of the PCS Error Field Category against the existing one | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | During this session, we will test the basic features of the revised error field category as well as vet the RWM feedback algorithm. | ||
| Background: | The capability to simulate the 3D coil and magnetic feedback system on DIII-D would help users to test, understand, and optimize the system performance. Although various RWM feedback models exist, a model of the full system on DIII-D does not. Beyond increasing confidence in the existing algorithms, a complete and well-understood model would also enable new advanced controllers to be developed, simulated, and implemented into the PCS.
The GA Control group has recently developed software tools that allow the PCS to integrate C source code generated by the Simulink Embedded Coder. These tools enable Matlab/Simulink-based control models to be directly incorporated into the PCS. A Matlab/Simulink model of the Error Field Category and PID-based RWM feedback algorithm will be developed in 2013 and the Embedded Coder will be used to generate the new PCS code. During this process, the new 3D magnetic diagnostic signals will be incorporated into the feedback algorithm. The performance of the new category and the existing one will be compared, providing a valuable test of the Embedded Coder workflow. | ||
| Resource Requirements: | PCS, 3D coil systems | ||
| Diagnostic Requirements: | New 3D magnetic diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 96: Effect of Error Field Control Currents on Strike Point Splitting | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | A. McLean | ITPA Joint Experiment : | No |
| Description: | Establish if a connection exists between the structure of the outer strike point (i.e. strike point splitting) and optimal error field correction. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Compare strike point splitting with Dimes TV with and without optimal error field correction in LSN discharges. Also consider a case with a 3D field that adds to the intrinsic n=1 error field.
This scoping experiment can be done in a single LSN discharge. During flattop, vary the amplitude and phase of the n=1 EFC currents. Monitor strike point with Dimes TV. | ||
| Background: | 3D magnetic fields are known to strongly modify the magnetic topology near the X-point. 3D effects on the X-point are also observed in vacuum and plasma response calculations. The degree of strike point splitting and non-axisymmetry may allow a new way to evaluate the quality of error field correction. Since most EFC experiments are done using biased up double-null discharges (to avoid ohmic H-mode), it is not possible to obtain Dimes TV data during these experiments. A shot is needed to vary EFC current in a LSN discharge. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Dimes TV. Other divertor diagnostics are desirable. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 97: Time-resolved measurement of plasma response spectrum | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | M. J. Lanctot | ITPA Joint Experiment : | No |
| Description: | The proposed experiment would test an innovative technique to identify the frequency dependence of the n=1 plasma response, allowing for a time-dependent resolution of physical parameters in a single-mode plasma response model. In previous work, the spectrum has been identified by combining data from multiple discharges probed with single-frequency perturbations. The proposed method would obtain the same information using a perturbation with several superposed traveling waves. Since this technique would necessarily involve higher amplitude perturbing currents than are normally used in single-frequency active MHD spectroscopy, experimental time is required to assess the possible deleterious side effects for the plasma, such as mode-locking. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Apply an n=1 spectroscopic waveform containing 5 frequency harmonics. Evaluate the beta dependence of the growth rate and coupling parameter from the 5-frequency response, and compare with the results from traditional, single-frequency spectroscopy. The measured dependencies will be compared with the predictions of plasma response codes such as VALEN and MARS. The measurements will be extended to n=2 if time allows. | ||
| Background: | Measurements of the plasma response to applied low-n magnetic perturbations can be used to assess the proximity to marginal RWM stability. The amplitude and toroidal phase of the plasma response can be related to the damping rate and mode rotation frequency of the stable RWM via a single-mode model [Reimerdes, et al, Phys. Rev. Lett. 93 (2004) 135002]. The link between the plasma response and stability can be understood in terms of the energy and torque required to perturb the plasma. As the plasma approaches marginal stability, less external energy is required to drive a fixed amplitude perturbation at the plasma surface. The plasma response is therefore a direct measurement of the proximity to marginal stability.
A fit to multiple frequency components is needed to simultaneously determine both a complex coil-mode coupling parameter, wall eddy-current decay time, and the RWM growth rate in the single-mode model. The RWM growth rate can be calculated from single-frequency plasma response data by assuming the other parameters are fixed. However, the coupling parameter may vary with plasma equilibrium parameters such as shape and outer-gap. In addition, analysis of 2-frequency plasma response measurements in NSTX (assuming a fixed coupling parameter) revealed a strong beta-dependence of the wall eddy-current decay time [J.-K. Park, et al., Phys Plasmas 16 (2009) 082512]. The multi-frequency technique may provide a more reliable and direct comparison with RWM stability modes, compared to single-frequency plasma response measurements. Initial tests of this technique in startup plasmas show that the multi-frequency response is consistent with the single-mode plasma response model over a limited beta range well below the no-wall limit. However, additional investigations are needed to demonstrate the reliability of this technique above the no-wall limit and to address whether physics model parameters (other than the RWM growth rate) stay fixed as plasma parameters change. | ||
| Resource Requirements: | H-mode plasma, sufficient NBI power to vary normalized beta above the no-wall limit. | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 98: Develop feedback-controlled MHD spectroscopy | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | M. Lanctot, G. Navratil, P. Martin, P. Piovesan, D. Shiraki, E. Strait, F. Turco, F. Volpe | ITPA Joint Experiment : | No |
| Description: | Develop and test a new active MHD spectroscopy technique in which the plasma response is held at a fixed amplitude using feedback. The required feedback currents should depend on the stability properties of the closed-loop system. Thus, this technique may provide a real-time stability assessment while simultaneously maintaining control of the stability. In addition, it provides a means of evaluating new feedback algorithms in piggyback. A similar technique was used to demonstrate sawtooth control in RFX-MOD tokamak experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use an established intermediate-beta H-mode plasma target. Program a rotating, fixed amplitude sensor offset in the PCS error field category. Enable fast RWM feedback with standard settings. Using the feedback error as a performance metric, evaluate the following dependencies: (a) rotation frequency of sensor offsets, (b) feedback gain and phase-shift, (c) plasma normalized beta. Compare the beta-dependence of the applied feedback field with the beta dependence of the plasma response to â??traditionalâ?? MHD spectroscopy in similar discharges. | ||
| Background: | In more traditional active MHD spectroscopy techniques, the plasma response to a fixed amplitude applied perturbation is measured, and can be linked to plasma stability. Less stable plasma equilibria typically have a larger amplitude plasma response, making the driven plasma response a useful measurement for disruption avoidance. However, the plasma response begins to increase more quickly as marginal stability is approached, and can lead to detrimental effects, such as plasma rotation braking. This new technique would automatically vary the perturbation level needed to maintain a fixed, low-amplitude plasma response, independent of plasma stability. Thus, measurements can be made in close proximity to the closed-loop marginal stability point.
There are several other possible advantages conferred by this technique: (a) feedback algorithms can be evaluated without the requirement of unstable plasma modes, (b) the closed-loop marginal stability point likely has higher plasma performance (eg beta) than the open-loop marginal point, (c) plasma modes driven using this technique have lead to tokamak sawtooth suppression in RFX-MOD, and (d) the feedback currents may be sensitive to residual uncorrected error fields. This experiment can be done using the present capabilities of the PCS Error Field category. That is, no PCS algorithm modifications are needed. | ||
| Resource Requirements: | H-mode plasma, SPA or AA power supplies on I-coils | ||
| Diagnostic Requirements: | Magnetics, MSE, Thomson scattering, CER, ECE radiometer, density interferometer. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 99: Understand and control resistive wall mode stability in high-qmin plasmas | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Berkery, M. Lanctot, G. Navratil, S. Sabbagh, E. Strait, F. Turco | ITPA Joint Experiment : | Yes |
| Description: | The goals of this experiment are to understand and control the stability of the resistive wall modes encountered in 2012 experiments with betan ~ 3 and qmin ~ 3. The following questions will be addressed: <br> <br>(a)Is dependence of the RWM growth rate on plasma parameters, such as rotation, consistent with previously measured dependencies of the driven plasma response and the stability theory incorporating kinetic modifications to ideal MHD? <br> <br>(b)How do the driven plasma response and plasma rotation behave as the marginal stability boundary is crossed? <br> <br>(c)Can the transition to instability and associated beta collapses be avoided by controlling the plasma response using NBI feedback? | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce DIII-D shot 150301, which suffers an unstable n=1 mode and beta collapse starting at t=2265 ms. Re-optimize error field correction using slow RWM feedback. Vary the injected NBI torque to determine the sensitivity of the stability to plasma rotation. Document the change in plasma response as marginal stability is approached using active MHD spectroscopy. Attempt to avoid beta collapse by maintaining a safe level of plasma response using NBI feedback. | ||
| Background: | Previous experiments have shown that the rotation-dependence of the driven plasma response is consistent with the predictions of a theory that includes kinetic modifications to ideal MHD [H. Reimerdes, et al, Phys. Rev. Lett. 106 (2011) 215002], and it is expected that unstable RWMs will exhibit a similar sensitivity. This experiment will provide crucial stability threshold data for comparison with predictions of kinetic stability codes such as MISK and MARS-K.
Control of the driven plasma response has been demonstrated using NBI feedback below the no-wall beta limit [J. M. Hanson, et al, Nucl. Fusion 52 (2012) 013003]. However, control above the no-wall limit is expected to be more challenging due to the increased importance of kinetic effects, and the non-linear dependence of the plasma response on plasma stored energy. An assessment of this control technique in a regime near the RWM??s marginal stability point will establish its usefulness for disruption avoidance. | ||
| Resource Requirements: | At least 7/8 NBI sources, including both 210 sources | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | Kinetic equilibrium reconstructions, ideal MHD and kinetic stability calculations | ||
| Other Requirements: | -- | ||
| Title | 100: Understand resistive wall mode control physics in low-q95 plasmas | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Bialek, A. Garofalo, G. Jackson, M. J. Lanctot, E. Lazarus, J. P. Levesque, P. Martin, M. E. Mauel, G. A. Navratil, M. Okabayashi, C. Paz-Soldan, P. Piovesan, E. Strait, F. Turco, A. Turnbull | ITPA Joint Experiment : | No |
| Description: | Unstable resistive wall modes (RWMs) and the loss of active RWM control are problems that lead to disruptions in plasmas with q95 < 2. This experiment would address the control issues, and if successful, lead to a feedback-stabilized q95 < 2 discharge or a determination of the feedback limit (eg, the maximum open-loop growth rate or minimum q95 that can be stabilized using feedback). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (a) Measure latency and transfer function of the RWM feedback system (sensors, PCS, amplifiers, coils) during maintenance or plasma start-up time (no plasma required). These measurements will serve as the basis for RWM control modeling and controller design, to be performed prior to (c) â?? (e).
(b) Perform companion experiments on the HBT-EP device to address aspects of control at low q. For example, investigate the impacts of latency and saturation, test advanced control algorithms. The goal is to leverage the flexibility of HBT-EP to develop control techniques and experience that help inform the DIII-D experiment. (c) Reproduce shot 150593 without feedback to establish a baseline reference (d) Test for a possible additional uncorrected error field by applying slow RWM feedback. Follow the established strategy of iterating over several shots, using the feedback currents from the previous shot as feedforward waveforms for the next shot. (e) Instability onset is still expected as q95 is decreased. If no unstable modes are encountered, attempt a further reduction in q95 and steady operation with q95 = const < 2. In the likely event that a discharge-terminating instability is encountered, apply RWM feedback using the algorithm and settings motivated by (a) and (b). | ||
| Background: | The low-q95 RWM could become a â??standard candleâ?? for developing RWM control knowledge, similar to the low-density locked mode used for error field control studies. Recent attempts to access q95 < 2 in diverted DIII-D plasmas were met with apparent unstable resistive wall modes (RWMs) when q95 was ramped to ~2. Applying RWM feedback control facilitated temporary access to q95 < 2, for a duration of about 400 ms in the most successful case. However, feedback control was ultimately lost when control coil power supplies reached their voltage limits, allowing an unstable RWM to grow and cause a disruption. Preliminary analyses indicate that a slowly-evolving error field and pickup from higher frequency plasma modes are two issues that affected feedback performance in these discharges. Additional experimental time is needed to confirm or rule out the presence of an uncorrected error field, and to demonstrate robust stability control in this regime.
Two innovative RWM control algorithms have been developed and implemented in the DIII-D PCS. The first compensates feedback sensor measurements for known ac vacuum coil couplings. An improved feedback transfer function should result, because many of the ac couplings have non-trivial frequency dependent phase-shifts. (At present dc-only compensation is used.) The second algorithm is based on a 3d VALEN model including the DIII-D wall, coils, sensors, and plasma mode [J. Bialek, et al, Phys. Plasmas 8 (2001) 2170]. The model serves as a basis for a state-space linear quadratic Gaussian (LQG) control algorithm. The LQG algorithm exploits the knowledge contained in the model in two ways: (1) sensor measurements are reconciled with model predictions to reduce the impact of noisy signals, and (2) the optimal gain for minimizing feedback effort and mode amplitude is used. The HBT-EP experiment is a high aspect ratio, circular cross-section tokamak specifically designed for MHD control studies [D. A. Maurer, et al. Plasma Phys. Control. Fusion 53 (2011) 074016]. Low-q regimes with current-driven instabilities can be accessed using plasma current ramps. The mode control system consists of a powerful GPU-based controller and arrays of in-vessel poloidal sensors and radial coils, similar to sensors and I-coil actuators used for RWM feedback on DIII-D. HBT-EPâ??s high availability and flexible control system make it an ideal device for developing fusion-relevant mode control techniques. | ||
| Resource Requirements: | 1 â?? 2 co-Ip NBI sources, audio amplifiers on I-coils, SPA supplies on C-coils. | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | Determination of RWM feedback system transfer function and feedback modeling, prior to experiment. | ||
| Other Requirements: | |||
| Title | 101: Investigation of n=1 RMP effects on ELMs with pitch-aligned configuration of I-coils | ||
| Name: | Park jpark@pppl.gov | Affiliation: | PPPL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | T. E. Evans, R. M. Nazikian, Y. M. Jeon | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to test if n=1 RMPs can alter ELMs in DIII-D, when the n=1 configuration of I-coils is adjusted for better pitch-alignment from the nominal configuration that is optimized for error field correction and for the coupling to the kink mode. For this goal, 120-180 poloidal phasing in the n=1 I-coil configuration will be used rather than the typical 240-300 poloidal phasing, and will be applied to ELMing plasmas. Comparison with KSTAR results is also an important goal, and thus target plasmas will be designed to match as closely as possible the operating parameters used in KSTAR RMP experiments. The q95 variation will be particularly important since resonant surfaces are fewer for n=1 and so the location of each surface or the number of resonant surfaces in the pedestal will be critical. Experimental results, whether or not it is successful for ELM alteration, will be very useful to guide RMP physics study and future RMP experiments in other devices including KSTAR. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Two different methods of the n=1 I-coil phasing vs. q95 will be used to see if n=1 RMPs can modify ELMs. First, two-step I-coil phasing, either by a pairs (120 phasing to 180 phasing) or (180 phasing to 240 phasing), will be applied depending on q95 of target plasmas (q95=6.5, 5.0, 3.5). Second, the fixed I-coil phasing (120 and 180 phasing) will be applied to q95 varying targets, by changing IP=0.9MA to 1.5MA during the shots. To compare results with KSTAR, NBI power will be also decreased down to PNBI=2.5MW if possible, but the higher heating up to PNBI=7-9MW will be also tested to see if n=1 RMPs can modify ELMs in the low collisionality regime. | ||
| Background: | In KSTAR, n=1 RMPs have been successfully used for ELM mitigation and suppression when the coil configuration is optimized for pitch-alignment, 90 phasing. However, another configuration optimized for the kink mode, 180 phasing, locked plasma or caused H-L back transition as typically expected by n=1 fields. In general the KSTAR results imply the importance of the field optimization for edge, regardless of specific toroidal mode n, which can be put to the test in DIII-D using I-coils. Although KSTAR has three rows of coils and DIII-D I-coils are two rows without the midplane array, the coupling to the pitch-alignment by 90 phasing and to the kink by 180 phasing in KSTAR can be similarly produced by 120-180 phasing and 240-300 phasing, respectively, in DIII-D. Nominally DIII-D n=1 I-coils are configured to 240-300 phasing for the kink mode, which may be the reason why n=1 applications were not successful for ELM alteration in DIII-D, as found with 180 phasing in KSTAR. Therefore, it will be interesting to check if the 120-180 phasing of I-coils can mitigate or suppress ELMs. Another important difference between KSTAR and DIII-D experiments is the q-profile. KSTAR had higher q95 when ELMs were suppressed, and the high q95 may be essential in n=1 applications in order to have sufficient number of rational surfaces in the pedestal. So q95 scans in this experiment are also desired. Results of this experiment, whether or not successful, can be compared with KSTAR and will be very useful to understand 3D field effects on ELMs. | ||
| Resource Requirements: | Standard RMP ELM control hardware and heating systems. | ||
| Diagnostic Requirements: | Standard RMP ELM control diagnostics. | ||
| Analysis Requirements: | Kinetic equilibrium reconstructions. | ||
| Other Requirements: | |||
| Title | 102: TBM mock-up effects on confinement at high β | ||
| Name: | Snipes Joseph.Snipes@iter.org | Affiliation: | ITER Organization |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Hanson, Y. In, R. La Haye, N. Oyama, J-K. Park, C. Paz-Soldan, H. Reimerdes, W. Solomon, T. Strait, T. Tala | ITPA Joint Experiment : | No |
| Description: | This proposal seeks to operate the TBM mock-up in high β H-mode plasmas to clearly determine how much the change in confinement due to the TBM mock-up fields can be affected by optimizing the I-coil error field correction. The 2009 TBM mock-up experiments clearly showed that the effects of the TBM mock-up fields on energy and particle confinement increased with increasing βN [1,2]. The 2011 TBM mock-up experiments [3] were operated at relatively low βN < 2 where effects on rotation were observed, but there were no clear changes in energy and particle confinement (except for a reduction in the confinement of energetic neutral beam ions [4]). Several methods to optimize the n=1 error field correction were attempted and a 25% increase in rotation was observed under conditions believed to optimize the n=1 error field correction in the presence of the TBM mock-up field. However, it is not clear that the optimum error field correction was actually found and the optimum is likely to depend on βN. So, this proposal aims to revisit the error field correction in the presence of the TBM mock-up fields to further optimize the correction with the I-coils at high βN. In particular, we will minimize the n=1 resonant field amplification magnetic response of the plasma to the TBM mock-up perturbation for conditions with βN > 2.5 in an ITER similar shape. In addition, since the attempts to use Dynamic Error Field Correction (DEFC) were not optimized in the 2011 experiments, additional run time should be devoted to DEFC to compare these two techniques at high βN. The new 3D magnetics array will allow up to n=2 DEFC. Another approach that should be attempted to correct the TBM error field is to obtain a more localized correction around the TBM by optimizing the two C-coil currents nearest the TBM in addition to the n=1 error field correction. The main purpose of these experiments is to quantify how much optimum error field correction can reduce the impact of the TBM mock-up fields at high βN. This is best carried out in highly rotating plasmas to avoid locked modes and disruptions. If time permits, the error field correction optimization could also be carried out at high βN with balanced NBI and ECH to operate in low rotation conditions. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiments should be carried out in an ELMy H-mode with ITER similar shape with βN > 2.5. EFC will be applied with the I-coils in the presence of the TBM mock-up fields. The currents will first be optimized by minimizing the n=1 magnetic response of the plasma to resonant field amplification. Since the TBM mock-up fields can increase the likelihood of locked modes under low rotation conditions, the experiments will be carried out with co-NBI in highly rotating plasmas. After optimizing the error field correction by minimizing the n=1 magnetic response of the plasma, the optimum I-coil currents will be maintained during the TBM mock-up pulse throughout the plasma flattop for several discharges to check reproducibility. ECH may be applied to reduce NTMs, if necessary. DEFC should also be applied to re-optimize up to n=2 error field correction, now enabled with the new 3D magnetics array, in the presence of the TBM mock-up fields at the same βN. Then, additional corrections should be attempted by optimizing the two C-coil currents nearest the TBM mock-up. These three methods to optimize error field correction should be attempted and several discharges may be repeated with the optimum correction to check reproducibility and ECH may again be applied to reduce NTMs, if required. If time permits, low NB torque plasmas could also be investigated at the same value of βN, possibly with additional ECH to reduce NTMs or reach sufficient βN. The error field correction techniques will likely need to be re-optimized under low rotation conditions. The optimum error field correction will be compared for each of these conditions and techniques and the effect on particle and energy confinement of the TBM mock-up fields will be quantified comparing optimum error field correction in the absence of the TBM mock-up fields with that in the presence of the TBM mock-up fields. | ||
| Background: | The 2009 TBM mock-up experiments clearly showed that the effects of the TBM mock-up fields increase with increasing βN. The 2011 TBM experiments provide a good starting point to determine the optimum error field correction based on minimizing the n=1 magnetic response, but they will need to be re-optimized at βN = 2.5. The 2011 TBM experiments also attempted DEFC, but were unsuccessful so more experimental time would be required to optimize this technique. The higher n error field correction with the 2 nearest C-coils is a more recent idea that should also be attempted. The new 3D magnetics array will allow up to n=2 DEFC. | ||
| Resource Requirements: | TBM mock-up coil, co- and possibly also counter-NBI. ECH may also be required to reduce NTMs and to operate at high βN with low rotation if time permits. The I-coils and 2 nearest C-coils to the TBM mock-up and associated power supplies are also required. | ||
| Diagnostic Requirements: | Locked mode (RWM) sensors. Particle and energy confinement and rotation measurements. Interferometer. Thomson scattering and ECE measurements. | ||
| Analysis Requirements: | Analysis of error fields and their correction. Standard energy and particle confinement. | ||
| Other Requirements: | -- | ||
| Title | 103: Probing and controlling the L->H and H->L transitions by small pellet injection | ||
| Name: | Hahn hahn76@kfe.re.kr | Affiliation: | Korea Institute of Fusion Energy |
| Research Area: | L-H Transition | Presentation time: | Requested |
| Co-Author(s): | P. Gohil, G. Tynan, L. Baylor, P. H. Diamond, D. Battaglia, Kazuhiro Miki, Weiwen Xiao, L. Schmitz, Z. Yan, G. McKee, M. Xu | ITPA Joint Experiment : | No |
| Description: | We propose studies of stimulated L-->H and H -->L transitions by small/shallow pellet injections. The motivations for this study are both pragmatic and scientific. <br><br>For pragmatic motivations, we have<br> i) ITER will operate at marginal Pth. Control of back transitions is of great interest as well as L-->H in a practical view. Previous work done at D3D indicated possibilities of reduction of the transition heating power by 30% in the pellet-stimulated transition. Since the pellets are very fast actuators, the work may address an effective, minimum invasive method for controlling enhanced/controlled hysteresis.<br> ii) From the previous work done at various machines/apparatus, the key of the phenomena seems to be edge radial electric field (Er) development. That would evoke questions on the critical deposition depth & size of the particle injections which may develop stiff density profiles to produce enough Er. <br> iii) The pellet-stimulated transition has a lifetime per injection, hence it would provide a way to study the back transitions without changing major parameters (current / heating / shape �¢?�¦).<br><br>For physics/scientific motivations, we note:<br>i)Pellet-induced profile perturbations are techniques for explorations of the mean flow-zonal flow-turbulence system which controls the L-->H transition. The response of the system to localized gradient perturbations and cooling fronts can be probed with modern fluctuation diagnostics. <br>ii)Explore the physics of the �¢??stimulated transition�¢?, as opposed to usual spontaneous transitions �¢?? Are they at different transition pathways? What role does the ZF, seemingly critical to the spontaneous transition, play in this stimulated transition?<br>iii)Such perturbative experiments are ideal for critical tests for any L-->H models. A 1D transport model has been used to predict the existence of novel states, such as driven sustained H-mode with subcritical heat power. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | H-mode D2 discharge with EC power is desired for usage of LFS probe arrays, in order to identify MF shear
1) Do a D2 LSN discharge with a EC power scan, from (0.5, 0.6, 0.7, 0.8, 0.9,1.0, 1.1, 1.2) x Pthr in a single discharge. Use EC modulations for the power change knob. 2) Inject a pellet per each power step. Observe if there's L-H-L transition under Pthr. 3) Choose the power level with L-H-L, make a scan of period. If period of L-H-L is tau, period is (0.3 0.6 0.9 1.5)x tau. 4) If possible, change pellet length to bigger/smaller size and repeat 2). Find L-H-L period. repeat 3) with the power level found. | ||
| Background: | Experimental observations of invoking L-H transitions by strong gas injections have a long history: A time-limited transition triggered by rapid gas puff / LiD pellet at Tuman III showed limited evidences the mode was triggered by edge radial E changes (Askinazi et al, 1993). After a decade, the DIII-D experiments by LFS pellets found and measured that the pellet triggered immediate edge Er change, which reduced the Pthr by ~30 % (P. Gohil et al, PPCF, 2003). No fluctuation data was available at that time to relate the observation with the mechanism of turbulence suppressions.
Those previous works exclusively focused on L-> H dynamics and did not try to deal with H-> L control issue. Recent experiments done at HL-2A (Duan, NF 2010) and KSTAR, using a supersonic molecule beam injection (SMBI), gave chances to revisit the dynamics including the back transition. The SMBI-stimulated transition has shallower depositions of the particles on the edge, and has a lifetime depending on the amount of dosage and base pressure, which can be used as a very handy technique to control the L-H-L cycle. The SMBI has very similar characteristics as LFS small pellets used in DIII-D for ELM mitigation studies (L. Baylor, IAEA 2012). | ||
| Resource Requirements: | 6-12 shots of total run might be desired, can be divided by 2 sessions if pellet length change is allowed:
1.3 mm x (0.9 - 1.8mm) D2 pellet injection system at LFS (midplane is ideal) EC heating & modulations enough to get an spontaneous H-mode (3-6 Gyrotrons) | ||
| Diagnostic Requirements: | Bolometers & Zeff
Filterscopes (D-alpha) BES large array focused on the edge DBS Microwave reflectometry CER at edge : Er, Ti and rotations Edge Thomson for pedestal Te / ne profile LFS probes for MF shear measurement | ||
| Analysis Requirements: | N/A | ||
| Other Requirements: | N/A | ||
| Title | 104: Comparision of small-ELM regimes in DIII-D and NSTX | ||
| Name: | Sontag sontag@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | R. Maingi, S. Gerhardt | ITPA Joint Experiment : | No |
| Description: | Compare edge regime characteristics in DIII-D during RMP mitigated ELMs to small-ELM regimes in NSTX. Filament structure, heat flux, collisionality and shape dependence will be examined. This is a JRT13 proposal | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This proposal can be completed in piggy-back mode during RMP experiments that achieve ELM mitigation. Using shots with up/down (stellarator) symmetry allows analysis with stellarator equilibrium and stability tools, so data from ROF proposal #55 by Lazarus on ??RMP ELM suppression in DN plasmas? would be significantly helpful. | ||
| Background: | Understanding the factors that lead to small or suppressed ELMs allows development of ITER operational scenarios that have steady plasma edge conditions and predictable heat loading on PFCs. This proposal is to examine plasma edge characteristics in DIII-D during RMP mitigated ELMs and compare to small-ELM regimes in NSTX. Filament structure will be compared since some n=3 mitigated ELMs on DIII-D exhibit overlapping ELM structures, which is indicative of possible mixed modes, as is also observed in NSTX when n=3 fields are applied to ELMing discharges. Examine collisionality dependence of small-ELM regimes (can use existing data). Also look at dr-sep dependence of ELM characteristics. IR camera data will show the peak divertor heat flux variation with ELM size and frequency. | ||
| Resource Requirements: | Normal operations with RMP ELM suppression. | ||
| Diagnostic Requirements: | All profile diagnostics, BES and DBS to look at edge turbulence, IR camera for peak heat flux. SXR and upgraded 3D magnetics for filament mode structure. | ||
| Analysis Requirements: | Kinetic EFIT, VMEC | ||
| Other Requirements: | -- | ||
| Title | 105: Optimization and exploration of the q95<2 scenario in DIII-D | ||
| Name: | Piovesan paolo.piovesan@igi.cnr.it | Affiliation: | Consorzio RFX |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Bialek, A. Garofalo, G. Jackson, J. Hanson, M. J. Lanctot, E. Lazarus, L. Marrelli, P. Martin, G. A. Navratil, M. Okabayashi, C. Paz-Soldan, E. Strait, F. Turco, A. Turnbull, P. Zanca | ITPA Joint Experiment : | No |
| Description: | In the 2012 campaign, discharges with q95<2 have been realized in DIII-D in the context of the TJA, following an idea originally developed at RFX-mod. Careful optimization of the plasma start-up and shape and magnetic feedback control of MHD stability were crucial to the success of this proposal. <br>Building on these results, we propose an experimental plan to fully explore and optimize this new scenario. The proposal would not only develop an interesting fusion scenario, but it would also give precious information on the physics of MHD stability control. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The following options are listed in order of priority, with each step requiring the previous one to be completed.
(a) Error field correction. Starting from shot 150593, further reduce the impact of error fields through the standard dynamic error field correction algorithm, i.e. apply feedforward coil currents based on what feedback ''suggests'' in previous shots and iterate for a few shots. This will allow to discriminate among error field correction and direct stabilization of the 2/1 RWM. (b) RWM FEEDBACK. If the 2/1 mode is still unstable, optimize fast feedback using the I-coils. To this end, we plan to use AC compensation and complex gains, varying the phase between the n=1 applied field and the n=1 plasma response. If these first two steps are successful, we plan to try and sustain the q95<2 phase as much as possible in L-mode. Fast feedback may be switched-off in short windows to test the 2/1 mode stability. (c) H-mode. Then we propose to try a transition to H-mode in the q95<2 phase by increasing the NBI power. The MHD stability of such a plasma is unknown and may require to re-optimize feedback control, and in particular DEFC and fast RWM feedback as in (b). (d) LOWER q95. As a last point, we propose to explore even lower q95 values by lowering the toroidal field since the beginning of the discharge. Even this step may require further optimization of magnetic feedback. | ||
| Background: | The 2012 TJA experiment on the q95<2 scenarios showed that overcoming the q95=2 limit is possible by magnetic feedback control of MHD stability. Given the limited experimental time, a full optimization of the feedback control was not possible. In fact, basically due to feedback limits the 2/1 RWM was not completely suppressed and eventually leaded to a disruption. Analysis of these experiments suggests two possible causes and solutions.
First, a significant uncorrected error field is still present and tends to excite the 2/1 RWM at a finite amplitude. Dynamic error field correction should allow to further reduce it to lower level. If the mode is marginally stable, this may even be sufficient to avoid its growth, as observed with beta-driven RWMs. Second, modeling of this discharges with a code developed and tested on RFX-mod tokamak plasmas [P Zanca et al 2012 Plasma Phys. Control. Fusion 54 094004], and recently adapted to DIII-D, suggests that a small error in the 2/1 RWM phase used for feedback forces the mode into rotation. This is indeed observed in the experiment and causes the voltage saturation eventually responsible for the disruption. Such an error can be compensated using a complex gain, which can be determined in the experiment by varying the phase between the n=1 applied field and the n=1 plasma response. With this correction, the mode rotation should be avoided and the model predicts full suppression of the mode. A better estimation of the mode amplitude and phase would be obtained using the AC compensation technique [L Piron et al 2011 PPCF 53 084004] instead of the DC one, already developed and tested in DIII-D, but not used in the 2012 TJA experiment. The work described above is preliminary to any further optimization and exploration of the q95<2 scenario. Once these steps will be successfully completed, the experiment will continue by exploring H-mode and/or even lower q95 values, where new MHD stability windows may open. | ||
| Resource Requirements: | 1,2 co-Ip NBI sources, audio amplifiers on I-coils, SPA supplies on C-coils. | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 106: Magnetic turbulence measurements in high beta H-mode discharges | ||
| Name: | Guttenfelder wgutten@pppl.gov | Affiliation: | PPPL |
| Research Area: | Turbulence & Transport | Presentation time: | Requested |
| Co-Author(s): | J. Zhang, N.A. Crocker, E.J. Doyle, C.H. Holland, G.R. McKee, W.A. Peebles, C.C. Petty, T.L. Rhodes, C. Rost, G. Wang, Z. Yan | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to document fluctuation characteristics in high beta H-mode discharges where magnetic (microtearing) turbulence is theoretically most likely to be present. Microtearing turbulence is predicted to have spatial characteristics that are distinctly different from ITG/TEM (outlined below). Data will be obtained with turbulence diagnostics in an attempt to cross-correlate between fluctuations in density (BES, DBS, PCI, mm backscattering), temperature (CECE), and magnetic field (recently commissioned UCLA polarimeter) to identify features that are, or are not, consistent with microtearing predictions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment will focus on high beta H-mode discharges, while limiting density below n<6E19 m-3 to allow optimal access for the UCLA 288 GHz polarimeter (minimizing refraction). Upper single null (Zmag~+7.5cm) is preferred for alignment with the polarimeter line-of-sight to maximize sensitivity to magnetic fluctuations. The experiment will focus around two discharge conditions (low BT and high BT) that offer different advantages. When possible, attempts will be made to flatten the density profile, which is expected to enhance microtearing turbulence, through variations in the strike point location, plasma shape, or beam power.
(A) The high BT scenario will be based on the reference target 128413, part of beta scaling confinement experiments, where low-k microtearing modes have previously been predicted to be unstable around r/a=0.5-0.7, R=200-215 cm. The high-BT scenario should allow for CECE simultaneous with DBS, BES, PCI, polarimetry measurements. A scan in ECH power will be used to increase Te (beta_e, a/LTe). A limited scan in density may be performed to optimize trade-off between high beta & collisionality, and optimal polarimetry access. (B) The low BT scenario will be based on the â??NSTX-matchedâ?? discharges following the joint NSTX/DIII-D poloidal rotation experiment (20100111) by K.H. Burrell (e.g. 140989, BT=0.56T, Ip=0.63MA, betaT~6.5%). These types of plasmas in NSTX are often (theoretically) unstable to microtearing modes and provide a logical scenario for comparison to DIII-D for similar magnetic turbulence validation studies. While the low field removes CECE (and MSE, ECE) capabilities, all other fluctuation diagnostics should remain available. A limited density scan will be attempted, depending on difficulties with locked-modes, as well as a narrow range BT scan to minimize MHD activity. | ||
| Background: | Theoretical evidence suggests that microtearing modes (generally unstable at relatively high beta_e, nu_e, and a/LTe; and weak a/Ln~0) may sometimes be an important component of electron thermal transport in the core and near the pedestal top of of spherical tokamaks (NSTX, MAST), conventional aspect ratio tokamaks (AUG, DIII-D, JET), and RFPs (RFX). There is clear motivation to measure turbulence characteristics associated with microtearing modes.
Non-linear simulations for both NSTX and AUG illustrate that microtearing turbulence is expected to be distinctly different from all other core-related micro-turbulence mechanisms (ITG, TEM, KBM, ETG). Most notably: (1) magnetic perturbations (k_theta*rho_s<0.2) are radially broad (compared to radially narrow for ITG/TEM), and (2) density and electron temperature perturbations exhibit narrow corrugations around rational surfaces (with spacing Drat=1/k_theta*s_hat). It is also expected that ne-Te cross phases should be different from other turbulence mechanisms. Based on the distinct spatial structures, cross-correlating turbulence measurements (ne, Te, Br) from multiple diagnostics may help identify features that are (or are not) consistent with microtearing turbulence. Previous calculations (based on NSTX simulations) using a synthetic diagnostic approach illustrate the broad magnetic perturbations of microtearing turbulence may be detectable by line-integrated polarimetry measurements [Zhang, APS 2011]. The UCLA polarimeter has now been successfully commissioned on DIII-D during the 2012 campaign, and both the hardware and synthetic diagnostic code have been validated. Chris Holland has previously identified a DIII-D discharge (128413) that exhibits unstable microtearing modes (k_theta*rho_s<0.2) between r/a=0.5-0.8. This provides one of the initial target discharges of the experiment. The second target would complement a similar experiment originally proposed for NSTX using polarimetry, BES, and high-k scattering. The UCLA 288 GHz polarimeter has recently been installed on DIII-D in this configuration, so DIII-D is in a unique position to obtain data on microtearing turbulence. The results would likely provide valuable information for, and comparison with, measurements and analysis on NSTX-U (beginning 2014 and beyond). These measurements will also contribute to Jie Zhang thesis data. | ||
| Resource Requirements: | Beams (heating, BES, MSE), ECH | ||
| Diagnostic Requirements: | BES, DBS, polarimetry, CECE, mm backscattering, PCI
CER, TS, ECE, MSE | ||
| Analysis Requirements: | EFIT, ONETWO/TRANSP, TGLF, GYRO | ||
| Other Requirements: | |||
| Title | 107: Error field detection by application of a "spiraling" error field | ||
| Name: | Shiraki shirakid@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | N. Logan, E. Olofsson, F. Volpe | ITPA Joint Experiment : | No |
| Description: | The n=1 intrinsic error field can be inferred from a single discharge by the application of a slow "spiraling" error field (uniform rotation with growing amplitude) in the presence of a 2/1 locked mode. The mode comes into torque balance such that it locks to the total n=1 error field, which is the sum of the static intrinsic error field and the spiraling applied error field. The intrinsic error field is deduced from analysis of the locked mode phase relative to the applied phase of the spiraling field, as well as an analysis of the torque experienced by the mode from the total error field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A non-disruptive 2/1 locked mode will be created by a beta ramp in a low-rotation (balanced injection) plasma. Once the mode is locked, a slow (~1Hz) n=1 spiraling error field will be applied by the I-coils to vary the total EF and therefore the phase of the locked mode. The amplitude of the spiraling field will begin at less than the expected intrinsic EF (~1kA in the I-coils) and increase to greater than this value over a few cycles, efficiently sampling the amplitude-phase EFC operating space. The inferred value of the EF from this method gives an empirically optimized set of currents for n=1 EFC.
Upon optimization of the EFC currents, a rotating EF can be super-imposed to the optimized EFC currents in an identical discharge, and the quality of the EFC can be assessed by the uniformity of the locked mode rotation. The same technique will be used with the C-coils to optimize the currents for C-coil EFC. If time permits, a spiraling n=2 field applied to a 3/2 locked mode may be used to study the intrinsic n=2 EF. | ||
| Background: | Sampling a large range of amplitude and phase with continuous torque measurements due to a spiraling EF ensures a wide range of rotation behavior while providing a detailed mapping of the EFC contours in a single discharge. The torque experienced by the locked mode from the spiraling EF can be calculated based on a model for the perturbed currents in the island, or computed directly through measurement of the Maxwell stress tensor. The new 3D magnetics should allow improved measurements of the Maxwell stress.
This technique provides an alternative to (and can be benchmarked against) the current method of error field detection based on the low-density locked mode threshold. The spiraling EF technique has the advantage of only requiring a single non-disruptive discharge, and not being limited to low-density ohmic plasmas. | ||
| Resource Requirements: | I240- and C-coils with SPAs | ||
| Diagnostic Requirements: | 3D magnetics, ECE | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 108: Disruption avoidance by forced magnetic spin-up of NTMs | ||
| Name: | Shiraki shirakid@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | N. Eidietis, R. La Haye, M. Okabayashi, E. Olofsson, F. Volpe | ITPA Joint Experiment : | No |
| Description: | Disruptions caused by the locking of a 2/1 NTM are to be avoided by using a rotating n=1 field from the I-coils to entrain a mode which would otherwise lock, with possible recovery of H-mode. This is to be done with and without simultaneous application of modulated ECCD to help suppress the island entirely. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A 2/1 NTM will be created and allowed to approach locking, by a beta ramp in a low-rotation (balanced injection) plasma. When the mode is detected to be near locking (f<1kHz), a fast rotating n=1 field (~100Hz) will be applied to entrain the mode in the rotating frame of the applied field and prevent locking. The required currents for disruption avoidance by entrainment of the mode and/or H-mode recovery are to be studied. The required currents may be reduced if modulated ECCD is simultaneously applied (at the entrainment frequency and "phase") to assist with mode suppression, and the required EC power in combination with the magnetic rotation will also be studied. | ||
| Background: | Rotating n=1 fields applied by the I-coils have previously (2011) been used on DIII-D to steer and control 2/1 locked modes. This was done in conjunction with modulated ECCD to help stabilize the mode. A pre-emptively applied fast rotating field may prevent locking to the wall and/or static error field before it happens, and sufficient rotation may stabilize the mode through rotational stabilization and/or rotational shear.
If successful, forced magnetic rotation of 2/1 NTMs will help eliminate a leading cause of disruptions due to locked modes, as well as providing valuable data on current and power requirements for disruption avoidance on ITER. | ||
| Resource Requirements: | I-coils in 240 phasing with SPAs
Modulated ECCD | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 109: Electron Heat Transport in E-Dominated Heating of ITER-like H-Mode Discharges with Modulated ECH | ||
| Name: | Diem sjdiem@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | P.M. Ryan, M. Murakami, J.M. Park, J.C. Hosea, R.J. Perkins, G. Taylor | ITPA Joint Experiment : | No |
| Description: | This experiment will modulate the ECH power to study the heating deposition and electron transport in ITER relevant discharges on DIII-D. EC modulation at frequency ~5 Hz has been chosen to study the CD efficiency and q0 evolution using break-in-slope and Fourier transform analyses. This frequency was chosen to avoid characteristic ELM and sawtooth frequencies (7-10 Hz). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | All discharges should be run with no beta feedback on NBI power so that the NBI power remains constant.
Run plan: 1. Begin with Q-H mode target discharge 150840 (dominant electron heating discharge). Density should be lower than the baseline discharge so that 2nd harmonic ECE is not cut off. The beams should be balanced to minimize the applied torque. 2. Apply modulated radial launch EC heating at a rate of 5 Hz. 3. Increase density to that of original discharge, 150840 and then repeat the modulated radial launch EC heating at rate of 5 Hz. This will require 3rd harmonic ECE monitoring of Te. 4. Reduce counter and increase co-current NBI to increase torque/plasma rotation and run modulated EC heating at 5 Hz for low-density conditions of [1]. 5. Repeat 5 with high-density conditions of [4]. 6. Choose the conditions of [1]-[6] that demonstrate the best heating and increases in stored energy. Replace the radial ECH injection with equivalent amount of tangentially injected counter-ECCD and repeat modulation experiments to evaluate the change in q0. | ||
| Background: | The direct electron heating comparison made between FW and ECW in 2012 showed that both techniques had difficulty in consistently and efficiently heating the core plasma for the ITER Baseline Scenario. Understanding the EC power deposition and electron heat transport is important both to evaluate a proposed ITER operation scenario that uses net counter-ECCD to control q0, and to improve our ongoing analysis of FW heating of IBS plasmas.
The target discharge proposed in this experiment is close to the target discharges in ROF proposals 118 and 257. Combining these proposals would simply require repeating several discharges during the shot development and scans in proposals 118 and 257 while adding ECH and ECCD modulation. | ||
| Resource Requirements: | Machine time: 2 days of machine time
Number of gyrotrons: 6 Number of neutral beam sources: 2, plus beam blips for MSE | ||
| Diagnostic Requirements: | ECE, 3HECE, CHERS, MSE, UCLA reflectometor, ORNL reflectometor | ||
| Analysis Requirements: | TRANSP, CURRAY | ||
| Other Requirements: | Also submitted to Transport and Turbulence - please discuss placement with that group | ||
| Title | 110: Electron Heat Transport in E-Dominated Heating of ITER-like H-Mode Discharges with Modulated ECH | ||
| Name: | Diem sjdiem@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Turbulence & Transport | Presentation time: | Requested |
| Co-Author(s): | P.M. Ryan, M. Murakami, J.M. Park, J.C. Hosea, R.J. Perkins, G. Taylor | ITPA Joint Experiment : | No |
| Description: | This experiment will modulate the ECH power to study the heating deposition and electron transport in ITER relevant discharges on DIII-D. EC modulation at frequency ~5 Hz has been chosen to study the CD efficiency and q0 evolution using break-in-slope and Fourier transform analyses. This frequency was chosen to avoid characteristic ELM and sawtooth frequencies (7-10 Hz). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | All discharges should be run with no beta feedback on NBI power so that the NBI power remains constant.
Run plan: 1. Begin with Q-H mode target discharge 150840 (dominant electron heating discharge). Density should be lower than the baseline discharge so that 2nd harmonic ECE is not cut off. The beams should be balanced to minimize the applied torque. 2. Apply modulated radial launch EC heating at a rate of 5 Hz. 3. Increase density to that of original discharge, 150840 and then repeat the modulated radial launch EC heating at rate of 5 Hz. This will require 3rd harmonic ECE monitoring of Te. 4. Reduce counter and increase co-current NBI to increase torque/plasma rotation and run modulated EC heating at 5 Hz for low-density conditions of [1]. 5. Repeat 5 with high-density conditions of [4]. 6. Choose the conditions of [1]-[6] that demonstrate the best heating and increases in stored energy. Replace the radial ECH injection with equivalent amount of tangentially injected counter-ECCD and repeat modulation experiments to evaluate the change in q0. | ||
| Background: | The direct electron heating comparison made between FW and ECW in 2012 showed that both techniques had difficulty in consistently and efficiently heating the core plasma for the ITER Baseline Scenario. Understanding the EC power deposition and electron heat transport is important both to evaluate a proposed ITER operation scenario that uses net counter-ECCD to control q0, and to improve our ongoing analysis of FW heating of IBS plasmas.
The target discharge proposed in this experiment is close to the target discharges in ROF proposals 118 and 257. Combining these proposals would simply require repeating several discharges during the shot development and scans in proposals 118 and 257 while adding ECH and ECCD modulation. | ||
| Resource Requirements: | Machine time: 2 days of machine time
Number of gyrotrons: 6 Number of neutral beam sources: 2, plus beam blips for MSE | ||
| Diagnostic Requirements: | ECE, 3HECE, CHERS, MSE, UCLA reflectometor, ORNL reflectometor | ||
| Analysis Requirements: | TRANSP, CURRAY | ||
| Other Requirements: | Also submitted to Inductive Scenarios - please discuss placement with that group | ||
| Title | 111: Dependence of I-mode on current direction and heating mix | ||
| Name: | Hubbard hubbard@psfc.mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Dennis Whyte, Anne White (MIT), Max Fenstermacher (LLNL), Alberto Loarte (ITER Organization). George McKee (U. Wisc). | ITPA Joint Experiment : | Yes |
| Description: | The I-mode regime is extremely attractive for fusion in that it combines a thermal barrier and high energy confinement with high particle transport, preventing accumulation of impurities and removing the need for ELMs. Multi-machine studies are being carried out through the ITPA (TC-18 and 19, PEP-31). It is under consideration for ITER. However, issues remain for the extrapolation to ITER. The key concern is that the regime is usually obtained in the unfavorable drift configuration, which for ITER means reversing field and current. This would mean that NBI is in the counter direction. The same would be true on JET, which for this reason has not yet explored I-mode.<br>At the recent ITPA meeting, Alberto Loarte suggested that D3D could play a critical role in investigating dependences on current direction, and of heating source and direction. It is unique in being able to independently vary Ip and Bt direction; C-Mod, AUG and JET all need to maintain the same helicity for various technical reasons. It also has a flexible heating mix including ECH and co and counter NBI. We therefore propose to compare I-mode access and properties with different Ip directions, and different heating sources. A positive outcome would motivate future experiments on JET, and support the development of the regime for ITER. <br>This experiment should follow the establishment of a robust I-mode regime (see proposals from Dennis Whyte and Anne White). It will support the FY13 JRT. Experimental details will be added in the following sections. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment should follow successful completion of Dennis Whyte??s Idea 83 and/or Anne White??s idea 37, demonstrating robust I-mode with LSN, reversed BT, standard Ip (co-NBI).
We will use a fixed shape and parameter set from that experiment. The experiment will be done in two parts, on different days (order and priority TBD): Day 1: Reversed Ip, keeping everything else fixed. (ie counter NBI). This will also have the advantage that standard helicity optimizes the BES spatial resolution, important for GAM measurements (see idea #362) 1.Assess power range, performance, pedestal properties and compare with above. 2.Vary NBI direction. Co vs counter vs balanced/mix. Based on July expts we expect to need 2-3 MW to get I-mode. 3.Compare balanced NBI with predominantly ECH. Day 2: Normal Ip, but reversed Ip, as on XP37. Repeat steps 2 and 3 above. Taken together, these experiments will assess and separate effects of torque and of fast ion losses. | ||
| Background: | I-mode is a stationary, high performance regime without ELMs ?? attractive in many respects for fusion, including ITER. It has been robustly obtained over wide parameter ranges on
C-Mod, and also observed on AUG over several years. Initial assessments of extrapolation to ITER, by Dennis Whyte, look promising [See Marmar APS 2011 and Hubbard IAEA 2012 talks]. To obtain the unfavorable drift configuration, ITER would need to use Reversed BT and Ip, LSN. The ITER Team is interested in assessing I-mode. However, the biggest potential obstacle is that reversed Ip would mean counter-NBI. It is not clear how this would affect I-mode access and performance. D3D is UNIQUELY capable to assess this question. It can run - Different Ip directions with SAME BT, and divertor geometry, and - Different heating mixes (ECRH, co vs counter or balanced) with SAME Ip and BT direction. Other devices (C-Mod, AUG, JET) are restricted to a single helicity, meaning that Ip and Bt need to be changed together; changing Ip direction implies a major change in the configuration (USN vs LSN). Also, C-Mod has no NBI and AUG and JET only one direction. At the recent ITPA meeting, IO (Loarte) suggested and group agreed that this D3D experiment be a priority in 2013 (for TC-19, PEP-31). If results are positive, will motivate a JET experiment with counter-NBI in 2014. | ||
| Resource Requirements: | NBI, both co and counter. ECH. 3 MW each system. | ||
| Diagnostic Requirements: | Top priority on edge profile diagnostics, and edge and core fluctuation diagnostics. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 112: 2D imaging of small ELMs | ||
| Name: | Tobias tobias@lanl.gov | Affiliation: | Los Alamos National Laboratory |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Produce small ELMs appropriate for 2D ECE-Imaging and BOUT++ simulation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment can take any number of forms and still meet its stated goals, and so the specifics of the discharges depend more upon selecting a small ELM regime that is of interest to a broader range of experiments. The only obvious exception being that we do not want to do this during discharges where the ELMs have been artificially modified by one of the other mitigation techniques, i.e. pellet pacing, RMPs, etc. Certainly the experiment can be done as a piggy-back if there are other experiments with an interest in either Type-III or other small ELM regimes. | ||
| Background: | In the past we've put much effort into characterizing and modeling large Type-I ELMs for the purpose of validating BOUT++ simulations [Fenstermacher, M., et al., Fast Pedestal, SOL and Divertor Measurements from DIII-D to Validate BOUT++ Nonlinear ELM Simulations. 2012.] . The large ELMs selected from previous campaigns present a variety of challenges to modeling, and so there is interest in attempting to model small ELMs in the hopes that the code will produce more consistent results.
From the perspective of the experimentalist, there are a number of reasons why large, Type-I ELMs become the most challenging to draw detailed measurements from. Of course, the ELM evolution is rapid, and there is often little or no hint of any precursor or evolving linear mode. Aside from this, ECE diagnostics have special difficulties imaging these modes on DIII-D. Strong mm-wave bursting of unknown origin obscures the ELM structure in these discharges [Yu, L., et al., Characterization of Intense Bursts of mm-wave Emission Using New RF Spectrometer on the DIII-D Tokamak. Bulletin of the American Physical Society, 2012. 57, B. Tobias, et. al., Intense millimeter wave radiation from the H-mode pedestal on DIII-D at ITER relevant collisionality, 39th EPS]. The fact that this bursting is not observed under the higher collisionality conditions at AUG and KSTAR suggests that it will not be a complication for imaging small ELM regimes. If this is successful, and we are able to get good imaging data for an ELM not obscured by mm-wave bursting, there will be a wide range of opportunities to compare this to modeling and also to results obtained on other machines. In particular, we are motivated to touch base with the results reported last year on ASDEX, where small ELM regimes were accompanied by both coherent and broadband fluctuations with much shorter poloidal wavelength than edge fluctuations routinely observed on DIII-D by our ECE-Imaging system [Boom, J., et al., Characterization of broadband MHD fluctuations during type-II edge localized modes as measured in 2D with ECE-imaging at ASDEX Upgrade. Nuclear fusion, 2012. 52(11): p. 114004.]. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | ECE-Imaging
MIR desirable (expected avail. nlt June 2013) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 113: Heat Load during slow current quench phase of disruptions/VDEs and its relation with poloidal halo c | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | M. Sugihara, E. Hollmann, R. Pitts, A. Loarte, V. Izzo, N. Eidietis, D.
Humphreys |
ITPA Joint Experiment : | No |
| Description: | Measure heat deposition due to particles together with halo current. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Set up fast diagnostics, especially IR cameras and wall probes, for accurate (total) heat load measurements. Set up also fast diagnostics for only radiation power deposition to derive the net heat load only due to particles. Simultaneously measure the halo current. It is ideal if plasma temperature can be directly measured, but if not, temperature should be estimated from the current quench rate. From these measurements, heat load associated with the halo current is to be derived.
Create intentional downward hot VDEs and repeat to get shot-shot repeatibility. Create several different current quench (CQ) speeds, e.g., fast quench, slow quench and intermediate quench speed. In order to create different CQ speeds, a scan of the initial plasma thermal energy and a series of hot VDEs (no mitigation) and a series of mitigated VDEs by using different species of impurity and amount (from H2/D2 to Ne or Ar) for triggering thermal quench during vertical movement should be performed. It is expected that the halo current magnitude as well as the dissipated energy fraction by radiation and particles is very different for these different CQ speed discharges, which should make the derivation of the relation between the heat load and the halo current clearer and more reliable. | ||
| Background: | During the current quench phase of VDEs (center disruption case also), ITER plasma will have always strong contact with the wall/divertor. So far, ITER has assumed that radiation energy dissipation will dominate during this phase, so that no significant heat load has been specified. However, recent experiments in various machines, e.g., JET ILW, indicate that significant fraction of magnetic energy seems to be dissipated convectively and/or conductively. This indication is supported by the observed small radiation power during this phase, especially for slow current quench discharges. In ITER, melting of beryllium wall is a large concern, if heat load is localized to the upper and lower first wall region. Thus, following information is particularly important for the assessment of the impact of the heat load during CQ phase;
(1) Width of the convective/conductive heat flux during CQ phase and their parameter dependence, (2) Relation of these heat flux width to the halo current width. | ||
| Resource Requirements: | 1 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, Tile current monitor and Rogowski loops for halo current measurement. | ||
| Analysis Requirements: | some analysis will be required to estimate plasma temperature and Zeff. | ||
| Other Requirements: | None | ||
| Title | 114: Toroidal peaking of radiation, response time for triggering TQ and mitigation efficiency during mass | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | M. Sugihara, R. Pitts, A. Loarte, E. Hollmann, V. Izzo, N. Eidietis,D. Humphreys | ITPA Joint Experiment : | No |
| Description: | Measure toroidal peaking of radiation, pre-thermal quench time duration and heat flux reduction on target plate. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Choose reference discharge, e.g., typical ELMy H-mode, and perform experiments with changing the injection rate/amount of impurity, e.g., neon. Radiation asymmetry, or toroidal peaking if it can be assessed, should be measured by AXUV during pre-TQ and TQ phases. Simultaneously time duration of pre-TQ should be derived from the time sequence data for edge and core Te, Prad and others. Set up fast diagnostics, especially IR cameras and wall probes, for accurate heat load measurements on the divertor target region during TQ phase and derive the reduction of the heat flux by the impurity injection with varied injection rate/amount. These experiments should be repeated for similar reference discharges to derive the database of the inter-relations between toroidal peaking, pre-TQ time duration and heat flux reduction with respect to the injection rate/amount. | ||
| Background: | During the massive gas injection, possible large radiation peaking is a large concern in ITER. In particular, recent NIMROD calculation shows that large peaking is generated by m/n=1/1 MHD activity during thermal quench phase. In this case, increasing the number of injection port location would be less effective to reduce the peaking factor and other control scheme needs to be developed. It is conjectured that one potential control scheme is to optimize the injection rate and amount. One could expect that with decreasing the injection rate and amount, the injected impurity will distribute around the torus more uniformly, which may reduce the radiation peaking factor both during pre-TQ and TQ phases. On the other hand, time duration of pre-TQ will be prolonged (prolonged latency time), by which we will lose fast response for triggering the TQ, and in addition, mitigation performance (reduction factor) of the heat flux on the target plate could be degraded. Key point is whether we can find balanced operation window between lower peaking factor and longer response time and degraded mitigation performance in ITER. For this purpose, systematic database of these inter-relations is very important for DMS design of ITER and they have not been studied yet. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | AXUV, 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. | ||
| Analysis Requirements: | detailed analysis will be required to evaluate the radiation peaking factor, heat flux reduction factor. | ||
| Other Requirements: | None | ||
| Title | 115: QH-mode at low co-Ip torque using n=2 NRMF | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | M.J. Lanctot | ITPA Joint Experiment : | No |
| Description: | The goal of the experiment is to determine whether using nonaxisymmetric fields from the C-coil wired for n=2 gives improved QH-mode operation at low co-Ip torque. These data will also be used as a test of NTV torque theory. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create QH-mode shots similar to 149220 but with the C-coil connected in the n=2 configuration. The I-coil will be used for error field correction. Scan neutral beam torque in the stationary phase of the shot to determine the maximum co-Ip NBI torque which can be used. Compare this to results from 149217-221 to see the change in peak NTV torque. Vary the toroidal phase of the n=2 field to examine the importance of the intrinsic n=2 error field. To see how different n-numbers combine, add as much even parity n=3 field as possible from I-coil while maintaining good n=1 error field correction | ||
| 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 2009-2012 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque or small co-Ip torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. The most recent results showed that NRMF from the C-coil alone allow operation at co-Ip torque up to about 1 Nm; scaled to ITER, this is 3 times the NBI torque that it will have. The goal of the present experiment is to investigate the effect of wiring the C-coil with an n=2 configuration. Due the coil geometry, the n=2 configuration has toroidal sidebands, namely a significant n=4 magnetic field at roughly 50% of the n=2 field. The resulting multi-harmonic field may lead to increased NTV torque over the n=3 configuration, which creates a relatively pure n=3 field. | ||
| Resource Requirements: | Reverse Ip. C-coil configured for maximum n=2 field, 7.1 kA current. I-coil configured for error field correction and possible even parity n=3 NRMF. | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 116: Influence of non-axisymmetric fields (TBMs / RMPs) on fast ion confinement and loss | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Energetic Particles | Presentation time: | Requested |
| Co-Author(s): | S Pinches, WW Heidbrink, GJ Kramer, MA Van Zeeland, M GarcÃa-Muñoz, X Chen | ITPA Joint Experiment : | Yes |
| Description: | The aim is to establish the expected fast ion losses in the ITER baseline scenario arising from the non-axisymmetric field configuration present. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use a low-current (0.6 MA), low-field standard H-mode configuration and slowly ramp the currents in the TBM mock-up coil and in the I-coils and C-coils to mimic those appropriate for ELM control in ITER. Use all available diagnostics to measure the changes in the fast ion distribution and losses.
By using a plasma current of 0.6 MA, the orbit width of 75 keV beam ions in DIII-D is approximately 70% of that of alpha particles in ITER and the beam ions become a reasonable proxy for the behaviour of fusion born alpha particles. For fields lower than 1.4 T the beam ions may also be sufficiently fast compared to the Alfvén velocity to destabilise TAE modes and the combined influence of all toroidal asymmetries may be studied. | ||
| Background: | The plasma configuration in the ITER baseline scenario will not be axisymmetric: One of the ELM control strategies foreseen for ITER is the application of non-axisymmetric magnetic fields using so-called ELM control coils, whilst the presence of ferritic material in the TBMs will further destroy the axisymmetry of the plasma configuration.
A consequence of breaking the toroidal symmetry, is that canonical angular momentum will not be conserved and particle drift orbits move radially leading to a degradation of fast particle confinement and increased losses. The exact details are sensitive to the configuration (plasma current and field) and so it is proposed to use an ITER-like baseline H-mode scenario to undertake measurements of the fast ion losses arising from the use of RMP coils for ELM suppression and the mocked-up TBM. If it is additionally possible to destabilise Alfvén Eigenmodes in this scenario then further information can be gained regarding all the non-axisymmetric fields considered responsible for the transport and loss of alpha particles in ITER. | ||
| Resource Requirements: | NBI system, TBM mock-up, I-coils and C-coils | ||
| Diagnostic Requirements: | Fast ion diagnostics (FIDA, FILD, NPA, neutrons), instability diagnostics (magnetics, SXR, ECEI), equilibrium diagnostics (core profiles including MSE) | ||
| Analysis Requirements: | Data provision to ITPA EP TG for modelling by fast ion codes that can treat plasma asymmetries. | ||
| Other Requirements: | |||
| Title | 117: Conditions for nonlinear Alfvén Eigenmode evolution and impact on fast ion confinement and loss | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Energetic Particles | Presentation time: | Requested |
| Co-Author(s): | S Pinches, WW Heidbrink, SE Sharapov | ITPA Joint Experiment : | Yes |
| Description: | The aim of this experiment is to: (1) Experimentally establish the criteria for the transition from steady-state, constant frequency, Alfvén Eigenmode activity to nonlinear behaviour in which the frequency and amplitude rapidly change (e.g. as seen in chirping events); and (2) Quantify the impact on fast ion confinement and/or losses that results from this change in behaviour. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Select reference case in which NBI driven AE are observed to behave nonlinearly. (This will probably be a low-field case (Bt < 1.4 T) such that the beam ions are super-Alfvénic.) Scan in toroidal magnetic field and/or plasma density (and possibly electron temperature using ECRH) to change collisionality at the resonant velocity and transition between types of nonlinear behaviour, and observe effects upon fast ion population. | ||
| Background: | The effect of dynamical friction at resonance has been shown to have a profound effect upon the nonlinear evolution of fast ion driven modes. Rather than saturating at a steady amplitude there is a complex interplay between the processes trying to establish the equilibrium fast ion distribution and the extraction of energy to drive the mode.
Waves excited by energetic particles exhibit several types of temporal behaviour experimentally: (1) a steady-state regime; (2) a regime with periodic amplitude modulation (the â??pitchfork splittingâ?? [A.Fasoli et al., PRL 81 (1998) 5564]; (3) a chaotic regime [R.F. Heeter et al., PRL 85 (2000) 3177]; and (4) an â??explosiveâ?? regime [S.D.Pinches et al., PPCF 46 (2004) S47]. A comparison between ICRH-driven TAEs and NBI-driven TAEs on different machines [M.Lilley et al., PRL 102 (2009) 195003] shows that there is a tendency for ICRH-driven TAEs to exhibit the first three regimes above whilst NBI-driven TAEs are prone to the explosive regimes with a bursting temporal evolution. It is important to understand the reason for such differences in order to correctly predict the behaviour of alpha-particle driven modes in ITER. It was recently suggested that the difference in mode evolution is caused by differences in the relaxation processes which tend to restore the unstable distribution function of energetic particles: quasi-linear RF diffusion for ICRH cases and Coulomb dynamical friction (electron drag) in the majority of NBI cases. | ||
| Resource Requirements: | NBI heating system. | ||
| Diagnostic Requirements: | Fast ion diagnostics (FIDA, FILD, NPA, neutrons), instability diagnostics (magnetics, SXR, ECEI), equilibrium diagnostics (core profiles including MSE). | ||
| Analysis Requirements: | Equilibrium reconstruction; Magnetics analysis (toroidal mode number and frequency evolution); Nonlinear fast ion codes that can be validated against these results and used to predict behaviour in ITER. | ||
| Other Requirements: | |||
| Title | 118: Dominant electron heated ITER baseline scenario studies | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | J. Hanson, G. Jackson, T. Luce, J. Navratil, W. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to study the performance and the MHD stability of the dominant electron heated ITER baseline scenario. These discharges would have all the ITER-relevant characteristics reproduced together for the first time: q95=3.1, betaN~1.8-2.2, low li, low/zero rotation, and Te>Ti (expected Te/Ti~>1.15). It is necessary to assess the performance and the MHD stability of this scenario with few or zero fast particles, and transport properties characteristic of electron heating. We propose to explore the betaN-li-torque-Te/Ti stability map with the methods that have proven effective in the past 2 DIII-D ITER campaigns, as well as to assess the effects of EC heating and current drive in different depositions on the plasma cross-sections. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a.Reproduce shot #150840, with all 6 gyrotrons (3.5 MW of ECH?), q95=3.2 (instead of q95=4.2), no FW, no pellets, possibly with no NBI power (the target shot had 1 MW of NBI). If necessary (i.e betaN too low?), add progressively enough beam power to sustain betaN>~1.8 without H-to-L back transitions.
b.Scan the initial li (varying the Ip ramp rate) to assess if there is a ??preferred? state less prone to developing tearing instabilities. Scan Te/Ti adding NBI power with balanced injection (keep rotation as constant as possible). Scan Te/Ti with co-injection, obtain rotation scans. How fast can the current be ramped (i.e. how low an li can be obtained), before an RWM sets in? Can this boundary be changes adjusting the beam power or the ECH power location? c.Move part of the ECH power to ECCD off-axis to tailor the current profile and make it more tearing stable. | ||
| Background: | A significant amount of work has been done to characterize the low-torque ITER b.s. scenario, which uses counter-NBI sources to lower the plasma rotation. The current profile evolution was found to be the main actuator to affect the tearing stability of the discharges. betaN, li and torque scans were performed to map the stability space, and at the end of the campaign a series of discharges were attempted, which used full power ECH (central radial injection), and various amounts of beam power (1-2.5 MW total). Since MHD stability was a problem at q95=3.1 and the focus of the experiment was not on obtaining precise ITER-like conditions, the discharges were moved to a safer q95=4.2 value. It is now time to move to the more challenging, but also more relevant, q95=3.2 value, and assess if and how discharges can be run under Te>Ti conditions, with dominant electron heating. This experiments is needed to seek answers to questions such as
-Is the li-betaN-torque stability map different from the one obtained with dominant beam power? -How fast can the current be ramped (i.e. how low an li can be obtained), before an RWM sets in? -Can this boundary be changes adjusting the beam power or the ECH power location? | ||
| Resource Requirements: | 30 and 330 NBI sources, both 210 NBI sources. 6 gyrotrons at max power and max duration. | ||
| Diagnostic Requirements: | Magnetics, MSE and CER when NBI usage allows, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 119: Impact of fast-ions on the RWM stability boundary | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Berkery, J. Bialek, J. Hanson, M. J. Lanctot, G. A. Navratil, M. Okabayashi, C. Paz-Soldan, S. Sabbagh, T. Strait, A. Turnbull | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to study the physics of the RWM in the absence of fast-ions, in order to determine if the apparent stability of DIII-D discharges to the destabilization of the RWM is due to the presence of stabilizing fast particles. We propose to eliminate the presence of fast-ions completely, and asses the RWM stability under those conditions, by producing ECH-only plasmas (similar discharge obtained in the 2012 campaign, plus an extra 0.9 MW of ECH available in 2013) and then ramping the plasma current up quickly to produce an unstable current-driven RWM. The goal is to map the stability boundary in the complete absence of fast-ions, then progressively add NBI power (and hence adding fast-ions to the scenario) to determine whether the stability boundary moves, and ultimately if the RWM is stabilized in the presence of fast-ions, and by what amount and distribution of particles. When adding NBI power, care must be taken to use balanced co-counter injection to obtain the same level of rotation with and without beams.
Even though the proposed scenario has intrinsically rather low betaN (~1.8), the physics of the destabilization of the RWM with and without fast-ions is universal and this experiment would provide crucial information on whether RWM stability poses a threat to the future machines that will not rely on fast-particle-producing heating systems. Moreover, if a way is found to run the scenario with qmin>1 for long enough, the experiment would provide the perfect (and only) platform to benchmark the stability codes (MARS-K, MISK, etc) that are needed to predict the MHD characteristics of future machines. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Produce an ECH-only, diverted H-mode discharge (a starting point is 150840), with all the available ECH power (3.6 MW?). Insert a fast enough Ip ramp to destabilize a current-driven RWM.
b. Obtain some data on the Ip ramp-rate necessary to destabilize the mode, and attempt to modify the scenario to obtain qmin>1 (early heating, co-ECCD off-axis, counter-ECCD on axis?). c. When the RWM is systematically reproducible, add balanced NBI power, as much as it can be used while maintaining a fixed betaN level (estimated 1-3 MW NBI), with on-axis or off-axis sources. Observe how (and if) the RWM stability boundary changes: is the original discharge stabilized? Is a faster/slower Ip ramp necessary to destabilize the mode? How does the stability change with on-axis vs off-axis fast-ion distributions? | ||
| Background: | Theoretical models suggest that the RWM stability is strongly affected by the presence of fast particles. In particular, is has been postulated that the RWM is stabilized in the DIII-D scenarios because of the fast ions produced by the NBI power, and machines that do not rely on NBI for heating and current drive may have issues with unstable RWMs when operating above the no-wall limit. Studies to assess the effects of fast-ions on the RWM stability have been preformed in DIII-D. However, while it was possible to modify the localization of the fast-ions and the fast-ion distribution function by means of the OA-NBI system in the previous campaign, this approach has proven difficult and the interpretation of the results hard to determine. This experiment proposes to tackle the problem in a different way, focusing on nailing down the physics in a scenario that allows for significant changes in the amount of fast-ions, and for a simple way to assess the RWM stability. The ECH-only shot would have no fast particles, and the discharges with different levels of NBI power and different injection angles will provide the direct knob to determine the effect of the amount and distribution of fast-ions. This test will allow us to create a map of the RWM stability under controlled conditions for fast-particles and rotation profiles. | ||
| Resource Requirements: | 30 and 330 NBI sources, both 210 NBI sources. 6 gyrotrons at max power and max duration. | ||
| Diagnostic Requirements: | Magnetics, MSE and CER when NBI usage allows, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 120: Optimizing non-resonant, nonaxisymmetric fields in QH-mode | ||
| Name: | Burrell Burrell@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | M.E. Fenstermacher, M.J. Lanctot | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to determine whether the poloidal mode structure of the nonaxisymmetric fields affects the NBI torque that can be used to sustain QH-mode plasmas with small or co-Ip NBI torque. The experiment allows the poloidal spectra to be ordered according to the net torque, thereby providing a test of plasma response and NTV theory. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce QH-mode shot similar to 141436, which used I-coil for NTV and C-coil for error field correction. Use torque ramp during stationary phase of discharge to determine NBI torque limit. Vary phase between upper and lower rows of I-coil to determined phase dependence of this limit. Include single row cases to evaluate the benefits of multi-row coil systems for rotation control in future devices. | ||
| 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 2009-2012 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque or small co-Ip torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. Unfortunately, with the present D III-D set of nonaxisymmetric coils, there is no way to alter the poloidal mode spectrum. If we use an n=2 configuration for the I-coil, we can continuously adjust the phase between the upper and lower rows of the I-coil, thus altering the poloidal mode spectrum. For example, the spectrum can be adjusted so it is neither pitch resonant nor kink resonant. | ||
| Resource Requirements: | Reverse Ip. C-coil configured for n=1 error field correction. I-coil configured for n=2 with the upper and lower rows independently powered. | ||
| Diagnostic Requirements: | Standard profile and all fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 121: Testing Emerging Two-Fluid Model for RMP Plasma Response in Straightforward IWL Plasmas | ||
| Name: | Shafer shafer@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Z. Unterberg, A. Wingen, T. Evans, N. Ferraro, D. Battaglia, J. Boedo | ITPA Joint Experiment : | No |
| Description: | Simpler conditions are needed to provide an ideal test bed to benchmark plasma response models (e.g. extended MHD vs. ideal vs. vacuum) through more direct measurements, which are necessary for predictive capability for future machines. This experiment is designed to simplify plasma conditions to allow more straightforward measurements of plasma response to RMPs and compare to modeling (e.g. M3D-C1, TRIP3D). The elongated inner wall limited (IWL) shape provides several advantages to achieve these simpler conditions: 1. No homoclinic tangle structure without an X-point. 2. Control of lower edge magnetic shear to avoid a â??pile-upâ?? of RMP-resonant surfaces. 3. Potentially large islands in the field of view of the tangential SXR Imaging system and Thomson (both core and divertor). A lower power phase can be used to allow detailed probe measurements. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. IWL L-mode plasma with high density for SXR measurement and 2-3 neutral beams (similar to 146519, but shifted vertically down for better Thomson measurements and lower q_edge).
2. Low-power phase (1 NB vs 2-3) for probe measurements near middle to end of the shot. 3. Break-up day into n=1 and n=3 segments. Use rotating n=1 at several RMP amplitudes and 2-3 different qâ??s. Use n=3 phase-flipping in a similar fashion. 4. Potential rotation scan with co vs. counter beams. | ||
| Background: | M3D-C1 modeling has been compared to both n=3 and n=1 RMP plasmas showing good agreement of the boundary displacements. This modeling has yet to compared directly to an RMP-induced island inside the plasma, which is an underlying component to the emerging model at DIII-D for ELM suppression (RMP-induced island forms near the top of the pedestal given two-fluid response and limits inward expansion of pedestal avoiding ELM crash).
SXR Imaging has already provided two interesting cases of plasma response measurements: 1.) a helical displacement in the steep-gradient region that compares well to the M3D-C1 modeling and 2.) Using a low-energy filter to measure USXR emission, lobes created by the split manifolds were measured that compare well to vacuum predictions. This diagnostic could provide a unique view in the IWL test bed plasma. Development of this scenario was based on analysis of a previous IWL L-mode plasma (146519). Modeling is underway for this shot with M3D-C1, but initial analysis indicates a relatively low screening response allowing the growth of islands. SXR Imaging shows reasonably strong signal given a sufficient density. This plasma had a relatively low Te, which effectively filtered out the region of psin>0.85 in the SXR Images given the metallic filter. However, Thomson, BES etc, can easily cover this region. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | SXR Imaging, Thomson, BES, BES Imaging, ECEI, MSE. | ||
| Analysis Requirements: | M3D-C1, IPEC, TRIP3D-MAFOT. | ||
| Other Requirements: | |||
| Title | 122: Bid farewell to locked modes - NTM-locking avoidance by feedback synchronization assisted with ECCD | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | Daisuke Shiraki, Francesco Volpe, Andrea Garofalo, Rob Lahaye, Ted Strait | ITPA Joint Experiment : | No |
| Description: | Comprehensive utilization of Internal non-axisymmetric coils, ECCD system and a PCS with an appropriate choice of feedback algorithm would reduce or even eliminate the risk of NTM-locking disruptions in a tokamak reactor. <br> <br>The feedback experiments in piggyback a few years ago had revealed very interesting results on the tearing mode behavior. The feedback operated with toroidal phase shift can synchronize the I-coil currents to the tearing mode, and not only prevent the mode from locking, but also modify the mode characteristics. <br> <br>The successful maneuverability of tearing mode with Internal non-axisymmetric coils is very crucial for the Error Field Correction, but also carry very significant impact on determining the use of internal coils in the ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Observations in piggyback experiment are:
(1) The Feedback can shift freely the tearing mode direction and frequency without losing the stored energy. The feedback fields accelerate/decelerate toroidally the mode and, interestingly, the mode frequency is lowered by ten-hundred times within short period with minimum stored energy loss. (2) The synchronization-transition from 0.5-1 kHz frequency to the feedback-controlled frequency (20-60 Hz) is smooth when the phasing is properly adjusted. The choice of phase shift seems less sensitive to plasma condition. These results suggest that the tearing mode is quite maneuverable so that the mode can be slowed down to low frequency range, and then can be controlled. Although it is to be examined further whether the tearing mode can be quenched in time, it seems, at least, we can sustain as a feedback-forced slowly-rotating-tearing mode. However, caution is to be made that this can occur only when the error field correction is extremely well compensated and of course only with high bandwidth amplifier-coil system. However, the feedback itself should perform simultaneously as a dynamic error field correction process, if resonant field amplification occurs during NTM-locking period. Experimental Approach We apply feedback by Dud detector onset: (1) To survey systematically the impact of feedback toroidal phase on tearing modes (2) To explore possibilities of quenching by varying the phase/gain in time (3) To explore the implication to ITER internal coil option. (4) this approach can be combined with ECCD, - no ECCD, or minim ECCD power - increase overall Q in ITER, - slower frequency with o- or x- points sychronization easier (5) while rotating, the electromagnetic force can be added (6) same time, dynamic error field correction takes place, if resonant field amplification occurs due to equilibrium shift during NTM This technique should be tested in any plasma conditions to observe the versatility of this approach. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 123: Bid farewell to locked modes - NTM-locking avoidance by feedback synchronization assisted with ECCD | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Daisuke Shiraki, Francesco Volpe, Andrea Garofalo, Rob Lahaye, Ted Strait | ITPA Joint Experiment : | No |
| Description: | Comprehensive utilization of Internal non-axisymmetric coils, ECCD system and a PCS with an appropriate choice of feedback algorithm would reduce or even eliminate the risk of NTM-locking disruptions in a tokamak reactor.
The feedback experiments in piggyback a few years ago had revealed very interesting results on the tearing mode behavior. The feedback operated with toroidal phase shift can synchronize the I-coil currents to the tearing mode, and not only prevent the mode from locking, but also modify the mode characteristics. The successful maneuverability of tearing mode with Internal non-axisymmetric coils is very crucial for the Error Field Correction, but also carry very significant impact on determining the use of internal coils in the ITER. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Observations in piggyback experiment are:
(1) The Feedback can shift freely the tearing mode direction and frequency without losing the stored energy. The feedback fields accelerate/decelerate toroidally the mode and, interestingly, the mode frequency is lowered by ten-hundred times within short period with minimum stored energy loss. (2) The synchronization-transition from 0.5-1 kHz frequency to the feedback-controlled frequency (20-60 Hz) is smooth when the phasing is properly adjusted. The choice of phase shift seems less sensitive to plasma condition. These results suggest that the tearing mode is quite maneuverable so that the mode can be slowed down to low frequency range, and then can be controlled. Although it is to be examined further whether the tearing mode can be quenched in time, it seems, at least, we can sustain as a feedback-forced slowly-rotating-tearing mode. However, caution is to be made that this can occur only when the error field correction is extremely well compensated and of course only with high bandwidth amplifier-coil system. However, the feedback itself should perform simultaneously as a dynamic error field correction process, if resonant field amplification occurs during NTM-locking period. Experimental Approach We apply feedback by Dud detector onset: (1) To survey systematically the impact of feedback toroidal phase on tearing modes (2) To explore possibilities of quenching by varying the phase/gain in time (3) To explore the implication to ITER internal coil option. (4) this approach can be combined with ECCD, - no ECCD, or minim ECCD power - increase overall Q in ITER, - slower frequency with o- or x- points sychronization easier (5) while rotating, the electromagnetic force can be added (6) same time, dynamic error field correction takes place, if resonant field amplification occurs due to equilibrium shift during NTM This technique should be tested in any plasma conditions to observe the versatility of this approach. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 124: Electron Energy Transport during q0 control using modulated EC power | ||
| Name: | Ryan ryanpm@ornl.gov | Affiliation: | ORNL |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | S. Diem, M. Murakami, J.M. Park, J.C. Hosea, R.J. Perkins, G. Taylor | ITPA Joint Experiment : | No |
| Description: | This experiment will modulate the ECH power to study the heating deposition and electron transport in ITER relevant discharges on DIII-D. EC modulation at both low frequency (~5 Hz, to study the CD efficiency and q0 evolution) and high frequency (~30 Hz, to study heating efficiency and deposition profile) will be employed with break-in-slope and Fourier transform analyses. These frequencies were chosen to avoid characteristic ELM and sawtooth frequencies (7-10 Hz). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | All discharges should be run with no beta feedback on NBI power so that the NBI power remains constant.
Run plan: 1. Begin with ITER baseline scenario discharge 150828. Restriction on outer gap is not required as FW will not be used, although we will start with the same gap to maintain similar confinement times. Density should be lower than the baseline discharge so that 2nd harmonic ECE is not cut off. The beams should be balanced to minimize the applied torque. 2. Apply modulated radial launch EC heating at a rate of 5 Hz. 3. Apply modulated radial launch EC heating at a rate of 30 Hz. 4. Increase density to that of original discharge 150828 then repeat the modulated radial launch EC heating at rates of 5 Hz and 30 Hz. This will require 3rd harmonic ECE monitoring of Te. 5. Reduce counter and increase co-current NBI to increase torque/plasma rotation and run modulated EC heating at 5 Hz and 30 Hz for low density conditions of [1]. 6. Repeat 5 with high density conditions of [4]. 7. Choose the conditions of [1]-[6] that demonstrate the best heating and increases in stored energy. Replace the radial ECH injection with equivalent amount of tangentially injected counter-ECCD and repeat modulation experiments to evaluate the change in q0. | ||
| Background: | he direct electron heating comparison made between FW and ECW in 2012 showed that both techniques had difficulty in consistently and efficiently heating the core plasma for the ITER Baseline Scenario. Understanding the EC power deposition and electron heat transport is important both to evaluate a proposed ITER operation scenario that uses net counter- ECCD to control q0, and to improve our ongoing analysis of FW heating of IBS plasmas. | ||
| Resource Requirements: | Machine time: 1 day of machine time
Number of gyrotrons: 6 Number of neutral beam sources: 2, plus beam blips for MSE | ||
| Diagnostic Requirements: | ECE, 3HECE, CHERS, MSE, UCLA reflectometor, ORNL reflectometor | ||
| Analysis Requirements: | TRANSP, CURRAY | ||
| Other Requirements: | |||
| Title | 125: Observation of non-linear interactions in fast-ion loss spectra | ||
| Name: | Kramer gkramer@pppl.gov | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Xi Chen and the fast-particle group | ITPA Joint Experiment : | No |
| Description: | Alfven Eigenmode (AE) induced beam-ion losses after one poloidal transit have been observed during the current ramp phase in standard DIII-D discharges for studying AEs. Those modes are a good tool to study details of the wave-particle interaction. In those experiments a non-linear coupling between reversed shear AEs (RSAE) and toroidicity induced AEs (TAE) has been observed in the loss spectra as measured with FILD. We would like to study the behaviour of the non-linear interaction when the q-profile relaxes from reversed shear to normal shear. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create a standard DIII-D AE discharge (pulse 146096) and stop the current ramp at 0.6 MA when the AE-induced "prompt" losses from the 30L beam reach the FILD-2 detector and observe the evolution of the losses. Repeat the experiment and stop the current ramp at 0.73 MA when the AE-induced "prompt" losses from the 330L beam reach FILD-2 detector. | ||
| Background: | -- | ||
| Resource Requirements: | 30L and 330L beams | ||
| Diagnostic Requirements: | FILD-2, ECEI, MSE, FIDA, all fast-ion diagnostics | ||
| Analysis Requirements: | Main analysis codes: EFIT, NOVA, SPIRAL | ||
| Other Requirements: | -- | ||
| Title | 126: Triggering ELMs at high frequency with PPPL Lithium granule injector | ||
| Name: | Mansfield dmansfie@pppl.gov | Affiliation: | Retired from PPPL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | D. K. Mansfield, A. L. Roquemore, R. Maingi, G. Jackson, W. Wu, P. B. Parks, L. Baylor | ITPA Joint Experiment : | No |
| Description: | Use the PPPL Li granule injector to trigger and pace ELMs at high frequency (0-1000 Hz). Compare the results against D2 pellet pacing. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform a typical ELM-pacing experiment using plasma conditions and parameters similar to those that obtained during previous D2 pellet pacing discharges. Inject 1 mm Li granules at different velocities and at low frequency (say 45 m/s, 65 m/s and 85 m/s @ 30 Hz) and compare results. If needed, repeat the velocity scan for smaller granule size (1mm, 0.8 mm, 0.6 mm or 0.4 mm granules available). After establishing the optimum size and speed, inject at increasingly higher frequencies up to 1000 Hz or whatever frequency the plasma can tolerate. Some of this scanning work can be done during the same discharge because both the pacing frequency and the injection velocity can be swept independently in real time because of the nature of the injector design. | ||
| Background: | Recently the injection of Li granules was observed to trigger/pace ELMs on the EAST device in China. This work was accomplished with the PPPL Li granule injector. This observation allows one to contemplate pacing ELMs on ITER with Be or perhaps BeD2 granules. A better understanding of how metallic pellets trigger ELMs is needed before this could be undertaken. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | Similar / same as previous diagnostic coverage for D2 pellet pacing XPs. PPPL fast video camera required as part of injector hardware. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 127: Characterization of Intense Bursts of Millimeter Wave Emission in QH-mode Plasma | ||
| Name: | Yu lbyu@ucdavis.edu | Affiliation: | UC, Davis |
| Research Area: | Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | B.J. Tobias, M.E. Austin, C.W. Domier, N.C. Luhmann, Jr. | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to characterize the intense bursts of the millimeter wave emission in QH-mode using the ECE radiometer system, ECE imaging system and the newly installed RF spectrometer, in order to lead to a plausible explanation for these bursts. This work will be the most important part of my dissertation. In this experiment, only several simple QH-mode will be made to restore the bursts. This experiment will use QH-mode, low collionality discharges with BT optimized for ECE and ECEI view. Several key parameters will be varied including magnetic field, neutron beam power. Highly spatially and temperally measurements of the bursts location, frequency, bandwidth and intensity will be obtained by ECE, ECEI, RF spectrometer. In addition, obtain the best data from CER, BES, FIDA, SXR are very important to interpret the data and explain the phenomenon. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For the diagnostic, the new RF spetrometer which has been installed on DIII-D since Aug will help characterize the bursts. It can take the signal from both the low frequency antena or one vertial channel of ECEI system, then divided into two part, one goes to regular ECEI module, the other goes to the modified module. In last campaign, it was connected to the low frequency lenses and antenna, which had successfully demonstrated that the bursts from both systems are millimeter wave radiation instead of the low RF frequency interference. For this campaign, it will be connected to the center ECEI channel, which will provide 7 contiguous mimic ECE channels on different toroidal angles. This will help determine the toroidal mode number and structure of the bursts.
For the plasma conditions, two key parameters should be varied after bursts appear. One is toroidal magnetic field which might affect the upshifted frequency; the other is neutron beam power, which will affect the collion frequency νei (which is trusted to be important in some static experiences in Ref 1) without big affect on the electron. Reproduce shot 149135, then increase the Bt from 2.0 to 2.1 T during 2500-3500 ms; then turn on the neutron beam at time 2500 ms; then try to reproduce the shot 149135 with Bt=1.9 T and 2.1 T | ||
| Background: | The ECE radiometer system and ECE imaging system are both hetedyne systems view at the 2nd harmonic X-mode ECE radiation. Super intense bursts, which go up to 10000 times the blackbody radiation temperature, with duration of 5-10 μs have been observed by both system mainly on three plasma events: disruptions, ELMs in H-mode and in QH-mode. The bursts in EHO in QH-mode have specific phase realtionship with edge magentic fluctuation data (Ref 1) and propagate the same poloidal veloctiy with n=1 EHO (Ref 2); also they are proved to be narrowbanded (1 GHz compared to 3-5 GHz in ELMy cases); all these feastures make them easier to be charactized.
There is two kind of bursts in QH-mode which varied in many parameters. One has strong narrowbanded EHO with BT=2 T. It has a MHD threshold for the bursts to initiate and it is highly related to the odd-mode amplitude of δB[Ref. 1]. The bursts are very strong that the actually ECE signal will get saturated most of the time during the bursts as shot 149135. And these bursts has a very narrowband which is smaller than 1 GHz and their upshifted frequency from the EHO locations is 3-5 GHz. The other has broadband EHO with BT=1.9 T. These bursts initiate and disappear without clear connection with δB. The bursts on ECE channel 1 in shot 146473 look like some intense spikes. And their upshifted frequency from the EHO locations is 1-2 GHz. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | ECE, ECEI, RF Spectrometer, MSE, CER, SXR, FIDA | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 128: Carbon-13 in He Plasma erosion/deposition experiment to benchmark DIVIMP code being used by ITER | ||
| Name: | Chrobak cchrobak@cfs.energy | Affiliation: | Commonwealth Fusion Systems |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | P. Stangeby, A. Leonard, G. Tynan | ITPA Joint Experiment : | No |
| Description: | Provide experimental data for low-Z (carbon) physical erosion and deposition under simple as possible helium plasmas in LSN configuration for benchmarking plasma material interaction codes DIVIMP and WBC-Redep. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The ideal material to expose for the purposes of these experiments is Be, but due to its extreme safety hazards, a suitable substitute may need to be used. Al has been identified as a suitable proxy for Be, but after testing in 2012 was found unable to withstand high power loads before melting. Thus, carbon is left, requiring the use of He plasmas to remove the chemical erosion factor and a Carbon-13 enriched erosion sample is required to detect its presence over the carbon background. It is proposed here to expose a specially-designed sample with a carbon-13 enriched surface layer to dedicated, repeat, well-characterized plasma shots. Measurements of the sputtering yield and material influx plume would be done spectroscopically using the MDS spectrometer view chords and narrow band filtered visible light cameras, as well as post-mortem measurements of the exposed surface by ion beam nuclear reaction analysis. | ||
| Background: | Theory predicts that for erosion of high-Z materials, prompt-local deposition of eroded material is dominant, and for low-Z materials, long-range transport and deposition is dominant. However, initial measurements of Mo erosion in 2011 found that only 20% of the net eroded material was found immediately surrounding the sample [4,5]. By contrast, for the low-Z Be DiMES sample in the 1996 experiment [3], only about half the Be that was eroded from the Be sample was found on the graphite surface of the DiMES head. These discrepancies indicate gaps in the current theory, and further stress the need for accurate measurements of gross and net erosion from PFCs.
From our 13C-methane injection experiments in DIII-D [6] we know that much of the low-Z launched from the main wall is transported long range, e.g. from the top of DIII-D to the bottom. About half the total 13C that was injected at the top of the LSN discharges was found in the bottom divertor. We have less of a handle on the other half of the injected 13C, but it appears to have been deposited short-range, on the main wall (short as distinguished from local) The idea that the wall is a source of sputtered impurity which all ends up in the divertor sink is too simplistic: parts of the wall are in a state of net erosion while other parts are in a state of net deposition from wall sources elsewhere. Additionally, within the region immediately surrounding the strike points there are net erosion and net deposition zones that change with varying strike point location and plasma confinement mode [7]. ITER urgently needs proper benchmarking of edge impurity transport codes DIVIMP for low-Z impurities to estimate the net erosion of the Be wall armor and the tritium retention by Be co-deposition [1]. Experiments prepared on EAST and JET to benchmark low-Z erosion/deposition (C in EAST, Be in JET) examine only limiter-type plasma contact. Although most of the area of the ITER Be wall will experience limiter-type contact, most of the actual erosion will occur at the upper, second divertor, and will therefore involve divertor-type plasma contact. Hence, it is essential to properly benchmark the DIVIMP code for low-Z materials in divertor-type plasma contact. DIII-D is in a unique position to provide these results due to the high quality divertor and edge diagnostic suite and DiMES material exposure system. | ||
| Resource Requirements: | 13C-coated Sample Fabrication
PIGE Nuclear Reaction Analysis with Ion Beam | ||
| Diagnostic Requirements: | DiMES TV camera view with spectral filter for C emission
MDS spectrometer views 1) on DiMES and 2) just off DiMES Tangential TV camera views of He and C emission Langmuir probes Divertor-Thomson DiMES Thomson CER | ||
| Analysis Requirements: | TBD | ||
| Other Requirements: | |||
| Title | 129: Optimization and Control of ITER-like MGI | ||
| Name: | Wesley wesley@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Evaluate optimization and control of low-quantity high-Z MGI applied in the context of meeting ITER DMS requirements for thermal radiation duration and symmetry followed by control of the current decay rate. Assess with single and two-location fast valves, with pure high-Z (neon or argon) and mixed D2 + high-Z gases. Goals include assessing feasibility of controlling TE radiation duration and symmetry and ensuing current decay, providing data on/for scaling of same with target plasma configuration and plasma size/configuration/energy (basis for comparison of DIII-D attributes with larger/smaller plasma size, etc.) Establish insight for ITER DMS gas delivery and control requirements. Assess variance and degree of repeat controllability. Assess RE seed generation and/or progress to RE plateau. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | 1-valve and 2-valve MGI with neon, argon and mixed gases ~ 200 Torr-liter neon input, LSN and vertical stable IWL targets with q ~3, ohmic -> max NBI input. | ||
| Background: | ITER proposes low-Q neon MGI as candidate disruption mitigation technique. Thermal energy radiation duration and symmetry must satisfy no-wall-melt requirements. Subsequent rate of current decay must fall within 50-150 ms limits. Feasibility of fulfilling these sequential requirements needs to be assessed in an integrated experiment. Critical issue/requirement for 2016 Final Design Review | ||
| Resource Requirements: | MEDUSA and CEREBUS MGI systems, standard NBI and ECH, IWL and LSN equilibrium | ||
| Diagnostic Requirements: | Usual disruption diagnostics including multi-azimuth bolometry, 3-D magnetics, periscope and divertor IRTV, fast camera imaging, RE seed diagnostics, during injection TS profiles, | ||
| Analysis Requirements: | 3_D plume models, fast EFIT or JFIT reconstructions, NIMROD simulations, .... | ||
| Other Requirements: | |||
| Title | 130: Synergy between core-localized Alfven eigenmodes and edge localized TBM fields on fast ion losses. | ||
| Name: | Kramer gkramer@pppl.gov | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Fast-particle group | ITPA Joint Experiment : | Yes |
| Description: | Fast-ion loss experiments with the test blanket module (TBM) mock-up in DIII-D have focused so far on beam-ion losses in MHD quiescent steady-state plasmas where it was found that the losses originate from near the plasma edge. Core-localized Alfven eigenmodes can efficiently transport fast ions from the plasma center to the edge where the TBM fields are strong and enhance the TBM-induced fast ion losses which are concentrated on the TBM surface as found in previous experiments. We want quantify those losses and use the data as a bench-mark for ITER simulations to quantify the synergetic effects of TBM fields and Alfen eigenmodes on fast-ion losses. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Run standard DIII-D AE experiments (similar to shot 146096) with and without TBM fields present. Measure the heat loads on the TBM surface with an IR camera. Measure the TAE-induced fast-ion losses with the FILD detectors. Measure also changes in the fast-ion population with the FIDA systems. Compare the data with fast-ion loss simulations. Because the experiments are performed in the current-ramp up, it might be beneficial to build-in constant current phases at around 0.6 and 0.7 MA where AE-induced first orbit losses are well observable with FILD-2. | ||
| Background: | |||
| Resource Requirements: | TBM mock-up, Co-, counter, and off-axis NBI | ||
| Diagnostic Requirements: | FILD, MSE, FIDA, IR-cameras, all fast-ion diagnostics | ||
| Analysis Requirements: | Main analysis codes: EFIT, NOVA, SPIRAL, etc. | ||
| Other Requirements: | TBM mock-up, NSTX IR camera for measuring heat loads on the TBM surface. | ||
| Title | 131: Understanding runaway equilibrium to minimize damage hazard | ||
| Name: | James jamesan@fusion.gat.com | Affiliation: | |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | The goal of this proposal is to investigate runaway equilibrium to improve the understanding of radial position control of runaway electrons after rapid-shutdowns. Radial control of runaway electrons is important for ITER and other large tokamaks, where position control and rampdown of runaway current may be an important component of the disruption mitigation system.<br><br>A theory of the plasma/runaway beam system was previously developed [Z. Yoshida, J. Phys. B, 1989], but has not been experimentally investigated before at parameters close to DIII-D or ITER. The proposed experiments will study the equilibrium by measuring the current profile, and observing the effect of radial position actuators. Data from these experiments will also supplement a study of the critical density and electric field for avalanche suppression proposed by R. Granetz as an ITPA joint experiment. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Impurity-free runaway discharges will be established by initializing to just above the critical density for avalanche, then either:
a) holding at constant loop voltage while the density is gradually dropped, or b) holding at constant density while the loop voltage is gradually increased until the discharge runs away. By avoiding shutdown gas injection and atomic line emission interfering with MSE, the current profile evolution will be measured throughout the runaway portion of the discharge. Visible synchrotron emission will be compared with projected q-contours to study runaway confinement. If necessary after the runaway discharge is established, thermal plasma current can be further suppressed by injecting deuterium or a small amount of impurity gas such as helium or nitrogen, as long as the MSE system continues to operate normally. After a discharge is established with low thermal current, actuators such as current in positioning coils and loop voltage can be scanned to study their effects on the equilibrium shape, position, and current and density profiles. The observed effects will be compared to results from the ORNL VMEC code to solidify understanding of experimental observations, and to revise theory if necessary. | ||
| Background: | Runaway discharge position control has been demonstrated in the vertical direction, but not in the radial direction. Prior efforts at radial position control met difficulties separating the LCFS from the inner wall, which was achieved only by ramping up the current.
The plasma-beam equilibrium was investigated at least in the 1989 article by Zensho Yoshida, and at the small TBR-1 tokamak in Brazil. | ||
| Resource Requirements: | cryo-pumping | ||
| Diagnostic Requirements: | MSE, UCSD fast camera
Thompson, density reflectometer | ||
| Analysis Requirements: | ORNL VMEC for studying runaway equilibrium response to actuators
EFIT modified to include the beam pressure | ||
| Other Requirements: | -- | ||
| Title | 132: Dissipation & Control of RE from ITER-like LSN targets | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | Humphreys | ITPA Joint Experiment : | No |
| Description: | This experiment aims to produce runaway electron beams from vertically unstable ITER-like LSN targets and assess 2 scenarios:
(1) The catch-all ITER case: Uncontrolled VUD or VDE - What is optimal level of (primary/secondary) impurity injection to minimize RE current before final RE loss? This is a race between CQ and vertical motion. - What is the condition for final RE loss in a VDE, particularly the location of the hot RE core? (2) Optimal ITER case: Time for setup - Assess if D3D coils set to ITER controllability limits can catch and hold RE from ITER-like LSN target. - Examine how that controllability improves as target moved towards the vessel neutral point |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start with ITER-like LSN target. 2012 data indicates very low densities aid RE production, so utilize optimal error field correction to enable ultra-low density operation. Kill the plasma using small argon pellet.
For scenario 1, fire secondary MGI (neon, argon) into RE plateau during vertical motion. Vary quantity from shot-shot to alter RE current quench and observe point (if any) at which RE final loss current is mimimized. For scenario 2, set D3D vertical system to mimic controllability limits of ITER in-vessel coils. From shot-to-shot, move target plasma toward neutral point to assess if Z control can be established. | ||
| Background: | RE suppression in ITER is uncertain at best, and even if viable methods are developed, there will be times when they fail or do not have sufficient warning to react. Therefore, we must develop methods for minimizing the impact of a fully-formed RE channel, and have *validated* models of those scenarios. The first, and worst assumption is that the RE channel cannot be controlled by the ITER in-vessel coils (highly likely). In that case, the RE dissipation must be optimized to minimize the RE final loss. Note that the VDE speeds up if the CQ RE dissipation speeds up, so "as fast as possible" is not necesarily the best answer. On the other hand, we also want to experimentally verify the controllability of ta real RE beam using ITER-like Z control system, in order to validate models. In addition, given some warning, ITER may be able to reposition the target plasma closer to thte neutral point. we want to verify that the RE vertical controllability improves as the taret plasma is moved closer to the neutral point. | ||
| Resource Requirements: | - small argon pellet injector
- MGI | ||
| Diagnostic Requirements: | - SXR (disruption mode)
- Fast visible carmeras - full vessel IR camera - neutron counters | ||
| Analysis Requirements: | Use these RE VUD results to benchmark RE generation codes. | ||
| Other Requirements: | |||
| Title | 133: Fast Runaway Shutdowns by MGI | ||
| Name: | Wesley wesley@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Investigate use of pre-emptive (before CQ and post-emptive high-Z MGI to effect a fast (< 100 ms) runaway electron current shutdown in vertically-stable IWL and elongated/vertically-unstable LSN/'ITER-like target plasmas. Clarify target configuration, impurity assimilation and RE current dissipation attributes identified in 2010 and 2012 campaigns. LSN aspect requires development of more-reliable RE initiation method(s) | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Make REs with Ar pellet and/or Ar MGI; kill REs with remnant or added gas, low-Z and high-Z. Assess passive and active control effects and resulting to-wall and in-situ energy depositions. Compare continuous rampdown versus partial rampdown an crash scenarios. Compare D2, helium, carbon (pellet), neon, argon, xenon, krypton injection. | ||
| Background: | ITER needs a post-emptive RE dissipator. High-Z MGI before and after CQ gives promising results | ||
| Resource Requirements: | Ar pellet injector, MGI systems (at least two different gases in separate systems), IWL and LSN low-density EC heated targets. | ||
| Diagnostic Requirements: | , standard and new RE diagnostics, fast camera and IR views of inner wall and top/bottom divertor for vertically-unstable RE channels. | ||
| Analysis Requirements: | RTEFIT and JFIT, dynamic RE gain/loss models, .... | ||
| Other Requirements: | |||
| Title | 134: Pedestal stability and divertor transport studies with feedback-controlled snowflake configuration. | ||
| Name: | Soukhanovskii vlad@llnl.gov | Affiliation: | LLNL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | Plasma Control Group (Kolemen? Ferron? Hyatt?), S. L. Allen, T. Petrie, D. D. Ryutov | ITPA Joint Experiment : | No |
| Description: | Initial snowflake divertor configuration studies have been highly successful, however, they clearly demonstrated the need for magnetic feedback control. We propose to implement the X-point tracking algorithm in PCS, and experiment with control actuators and limits. The ultimate goal is to be able to control distance between null points and their relative orientation. This would enable utilization of snowflake divertor configuration as laboratory for pedestal and SOL transport studies. Precise feedback control would enable studies of ideal and snowlfake-plus configurations which are now hardly possible. Physics studies will include 1) Pedestal profiles and magnetic shear will be measured as functions of sigma (distance between null points). 2) Divertor heat flux, T_e and radiation profiles will be measured as functions of sigma (and effectively, as functions of connection length, flux expansion, and additional strike points). Feedback control would also enable snowflake configurations of medium and lower triangularities, effectively placing the snowflake divertor over the region diagnosed with divertor Thomson scattering system, as well as combining ITER shape with snowflake configurations. Feedback control would also enable studies of snowflake divertor compatibility with cryopumping, important for NSTX-U program and ST-FNSF concept development. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Need to elevate priority of the snowflake feedback control with the Plasma Control Group.
2) Implement and test null point tracking algorithm for snowflake configuration identifications in PCS 3) Model and develop control scenarios, e.g., control hight and radius of X-points, X-pt vs OSP, etc 4) Execute experiment to test scenarios and demonstrate controllability of sigma and nulls orientation 5) Execute physics study to measure pedestal and divertor profiles as functions of sigma for snowflake-minus configuration 6) Execute experiment to compare with snowflake-plus and ideal snowflake (which may have more impact on pedestal stability) | ||
| Background: | -- | ||
| Resource Requirements: | Need to elevate priority of the snowflake feedback control with the Plasma Control Group. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 135: Measurement of RE diffusion to center post | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux | ITPA Joint Experiment : | No |
| Description: | Try to measure RE radial diffusion coefficient. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create RE plateau with small Ar pellet injection. Hold beam and move into center post, then scan vertically along center post to get best chance of observing SXR emission spot. If successful, repeat with D2-diluted RE beam to try to reduce radial transport coefficient by pumping out argon content. | ||
| Background: | Anomalous loss of REs during RE plateau is suspected to be due to radial transport. Attempt to observe this transport directly from SXR emission spot on center post. | ||
| Resource Requirements: | 1 day run time. Small Ar pellet injector. D2 injector (SPI or MGI). | ||
| Diagnostic Requirements: | 165 CdTe array with raised view chords | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 136: Disruption main chamber heat loads | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | C. Lasnier, R. Pitts, M. Sugihara | ITPA Joint Experiment : | Yes |
| Description: | Measure main chamber heat loads (conducted and radiated) during different types of disruptions (mitigated and unmitigated). | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create disruptions (density limit, current limit, and beta limit) of reasonably reliable timing. In each case, optimize main chamber IR timing in view to get information both on overall heat load distribution as well as separation of TQ and CQ heat loads at location of max heat loads. | ||
| Background: | Predicting disruption survivability in future tokamaks requires characterization of disruption heat loads. At the moment, very little data is available on main chamber heat loads (conducted and radiated). | ||
| Resource Requirements: | 1 run day. | ||
| Diagnostic Requirements: | Main chamber and divertor IR cameras. Fast bolometry | ||
| Analysis Requirements: | IR thermography analysis. | ||
| Other Requirements: | |||
| Title | 137: 3D material migration during RMP ELM suppression | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | A. McLean, T.E.Evans, R. Laengner | ITPA Joint Experiment : | Yes |
| Description: | Measure erosion yields in 3D helical lobe structure induced by RMP fields. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use the DiMeS Porous Plug Injector (PPI) to directly measure the chemical erosion yield and the local C sputtering from MDS spectroscopy.
Apply PPI to L-mode and H-mode with RMP ELM suppression. Outer leg has to be placed on outer shelf and swept accross PPI location maintaining ELM cuppression. This has to be developed. | ||
| Background: | Formation of 3D striated heat and particle flux pattern have been observed as a feature accompanying ELM suppression in low collissionality plasmas in ITER similar shape. This is likely to change local erosion characteristics, material migration and eventually the re-deposition. Suitable numerical tools to address the impact of this observation on the divertor life time and integrety at ITER are under way. We are coupling EMC3-Eirene to ERO to address this. However, experimental data to validate this code package are required and the D3D setup is an ideal test bed for this important exploration of a basic physics aspects accompanying RMP ELM suppression. The actual experiemnt connects to a first of its time measurement in L-mode sicharges which have shown a reduction of the chemical ersion yield with C surface material. Similar measurements are obtained at TEXTOR and this experiment adresses ITPA task PEP19 issues. | ||
| Resource Requirements: | ITER04 patch panel, Icoils on C-supplies, n=1 EFC, PPI with 13CH4 loaded - the PPI head has to be enhanced for H-mode application. FZJ can contribute with design and manufacture effort to make this happen. A PhD thesis is launched in this field and hence appropriate technical and scientific support is provided. | ||
| Diagnostic Requirements: | MDS spectrometer, DiMeS camera, PPI, TTV cameras, IR TVs, Xpt soft X-ray with low energy cut off | ||
| Analysis Requirements: | TRIP3D, EMC3-Eirene/ ERO package, M3D-C1 for plasma response being included into EMC3-Eirene modeling (IO task) | ||
| Other Requirements: | -- | ||
| Title | 138: SSI target improvement by minimizing global low-n MHD activity in 1.5 | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Francesca Turco, Chris Holcomb, John Ferron, Tim Luce | ITPA Joint Experiment : | No |
| Description: | One of SSI promising configurations is with 2 > q_min >1.5, betan~4, ECCD-NTM suppression ( Bt> 1.6T) with currently available hardware limit. But, the achievable beta seems limited since the condition is approaching transport limited. However, there may be a room for optimization with present hardware. In SSI 2011 discharges of betan=3.5-4.0, we observed various slow low-n modes, such as ELM driven n=1, OFM driven RWM, now we see mysterious seedless RWM-like low-n mode â??magnetic bubbleâ??. This suggests that the plasma status could be close to the â??physics-based operational limitâ??. If so, we may not be able to achieve high plasma pressure simply by increasing the available axilliary-heating power. The improvement on the MHD stability is the key for stable steady high beta operation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Here, assuming that the plasma condition is close to the physics operational limit.
- Homework: systematic numerical study of optimizing internal kink limit condition with fixed boundary condition (Francesca Turco has started). The main parameters are geometrical characteristics, flatness of q_min, and off-NBI energy/angle. This process is similar to that as done for classical fishbone. - Experiments: Tight coupling to the wall (forcing modes onset to internal mode) and increasing the triangularity within the limit of proper divertor functions. Fast low-n feedback should be seriously pursued. These low n modes are likely to be just marginally stable, but, the magnitude was amplified by the resonant field amplification through uncorrected error field. If we take this hypothesis, fast DEFC should be applied. The fast time constant should be useful for mode above marginal condition. The feedback on one particular mode can not bring ultimate universal solution, but, careful categorization of level of achievement on feedback parameters will lead to determine the next step toward better SSI operation even q_min>2 configurations. | ||
| Background: | Although, at present, it is not clear about the couplings between low n MHD activity and transport processes, the fact that best shots were achieved with feedback implies that some reduction of low n mode may have been useful in an indirect way for minimizing mode couplings which are causes of enhanced transport.
One hypothesis of success / limit of 2012 q_min~1.5 is that plasma configuration is close to â??true marginal stability of n=1 kink with resistive wall including stabilization rotation and kinetic effect and thus, residual low n MHD modes were visibly excited. Thus, stability marginality against ideal internal modes may have been the key factor, namely, â??physic operational limitâ??. A hypothesis is that the increase of this marginality should reduce the seeds of turbulence onset, hopefully, improving transport properties. The essence of important mode can be estimated by the internal modes with fixed boundary regardless of final appearance of mode itself like RWM or OFM. Optimization of geometrical characteristics, curvature drift and flatness of q_min can lead to higher stabilizing conditions, as was in classical fishbone minimization. Serious attempt of fast feedback utilization should be attempted to improve the marginality of achievable beta, since here we are looking for 20% range improvement. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 139: TQ radiation distribution from R-2 MGI | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux, R. Moyer, C. Lasnier | ITPA Joint Experiment : | No |
| Description: | Evaluate if R-2 MGI gives more uniform radiated heat loads than R+1 MGI. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Shut down LSN plasmas with R-2 MGI. Then shut down same target plasma with R+1 MGI. Optimize diagnostics to get good visible and IR camera imaging of injection location. | ||
| Background: | ITER is planning on shutting down disrupting shots with R+1 MGI. However, poloidal flows seem to move the injected impurities rapidly to the center post, so main chamber melting at the injection port and at the center post are a concern. It is possible that using R-2 injection will spread Prad loads more uniformly. Also, injecting at several locations simultaneously may help. | ||
| Resource Requirements: | 135R-2 MGI injector | ||
| Diagnostic Requirements: | Fast bolometer, main chamber IR camera | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 140: q0 control with Counter-ECCD in ITER Baseline Discharges | ||
| Name: | Murakami masanori_murakami@att.com | Affiliation: | Retired |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | P.M. Ryan, S.J. Diem, J.M. Park, R. Prater, G. Jackson, et al. | ITPA Joint Experiment : | No |
| Description: | Performance of counter EC for H and CD is compared with radial-launch EC and counter-FWCD in ITER Baseline discharges. <br>This could be a basis of ITER Baseline discharge simulations with ECCD using the theory-based transport models and ITER H&CD systems <br>Detailed pedestal documentation | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using Counter-EC and radial-launch EC
Modulation fast (~30Hz) and slow(5Hz) EC power for FFT analysis J.M. Park, | ||
| Background: | ITER still plans to use counter-ECCD to make configuration control in particular, q(0) control to supplement ICRF (FWCD) or off-axis NBCD.
H&CD performance of the counter-EC need to be studied for control and research plans. This also gives us opportunity of comparing with FW experiments in DIII-D last year. | ||
| Resource Requirements: | NBI: >5 co-beams + 210RT
EC: >=5 gyrotrons | ||
| Diagnostic Requirements: | ECE, MSE, CER, TS | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 141: Studies of RE final loss | ||
| Name: | Hollmann ehollmann@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | A. Loarte, R. Martin-Solis, N. Commaux, C. Lasnier, R.Moyer, D. Rudakov | ITPA Joint Experiment : | No |
| Description: | Try to vary characteristics of RE final loss to understand Wmag - Wkin conversion, as well as heat load footprint. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create RE plateau with small Ar pellet injection. Move RE beam into upper, lower, or inner walls at varying speeds to change final loss speed. Also, try to image footprint of final loss on main chamber with IR camera. | ||
| Background: | Conversion of magnetic to kinetic energy in RE final loss is predicted to depend dominantly on the loss time of the RE beam to the wall. By varying the drift velocity of RE beams into the wall, it can be seen if this conversion is affected. Also, the footprint size of RE final loss heat loads is unknown, so it will be attempted to try to image this with the IR camera. | ||
| Resource Requirements: | 1 run day. Small Ar pellet injector. | ||
| Diagnostic Requirements: | BGO scintillator array. BGO energy counter. Main chamber and divertor IR cameras. Visible camera | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 142: ELM heat transport in the snowflake divertor. | ||
| Name: | Soukhanovskii vlad@llnl.gov | Affiliation: | LLNL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | S. L. Allen, T.Petrie, D. D. Ryutov | ITPA Joint Experiment : | No |
| Description: | Working together with LLNL Theory group, we would like to make a<br>quantitative assessment of the SF null-point heat convection theory on ELM<br>heat transport. <br>Fast IR thermography measurements of the inner and outer divertor strike points are needed for this assessment.<br>Other fast diagnostics will be needed as well, e.g., DISRAD for rad. power measurements, fast filterscopes, Langmuir Probes.<br>Upstream and downstream poloidal beta will be estimated from measurements.<br>Relative timing of ELM conductive and convective pulses arrival to the inner and outer strike points will be measured in the standard and in the snowflake (-plus, -minus) divertor configurations.<br>From these measurements, heat flux partition and relative loss could be assessed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | Initial snowflake-minus divertor studies in DIII-D indicated that ELM energy was universally reduced in the snowflake configuration, and ELM peak heat flux in the divertor outer strike point region was significantly reduced in the snowflake configuration with deuterium seeding.
Inner divertor measurements were not performed. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 143: RMP ELM control at low torque | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | The "worst case" scenario for ITER is that, due to limited torque input, the H-modes will be low torque and low rotation. Recent experimental and 2-fluid modeling work suggests that rotation - specifically, the electron perpendicular rotation = sum of the E x B and electron diamagnetic velocities-plays an important role in determining the plasma response to RMPs. The goal of this experiment is to explore the physics of plasma response to RMPs and the viability of ELM control for ITER-relevant H-modes with balanced torque and low toroidal rotation (limited to the intrinsic rotation level). Demonstration of RMP ELM control or suppression at low tourque would radially enhance the reliability of RMP ELM suppression in ITER. Even if ELM suppression isn't achieved, we will gain substantial information on plasma response and rotational screening in these low torque ELMing H-modes. A side benefit of this experiment will be the development of a high performance, low torque ELMing H-mode target plasma for a variety of ITER-relevant physics studies. If rotational screening models are correct, we should be able to access an ELM suppressed or mitigated regime in ELMing H-modes with low net toroidal rotation (auxiliary heating from balanced NBI plus ECH) using very little I-coil current, possibly only a few hundred amps. This would open a new field of active pedestal control by allowing the use of higher frequency (modulated and/or feedback controlled) RMPs, a first step toward integration of RMP ELM control with other modes of operation, including radiation-enhanced divertor operation, pellet fueling or a hybrid RMP-assisted pellet pacing. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a low net torque/near balanced NBI ELMing H-mode in ISS shape. This will establish a target ELMing H-mode with a zero crossing in the perpendicular velocity (unlike last year's counter NBI cases) but deeper in the core than desired for ELM suppression based on emerging understanding. Apply an n = 3 even parity RMP; this should induce stochasticity in the boundary (foot of the pedestal), and the subsequent radial ion current to maintain ambipolarity with increase the toroidal rotation and move the zero-crossing in the electron perpendicular velocity out toward the top of the pedestal. We will need to use the best error field correction and ECCD to control locking of MHD modes in the core to widen the operating window away from mode locking for this experiment (both techniques were routinely applied in the CY12 campaign. Scan the I-coil current for both phases (0° and 60°) and for both parities (even and odd) to vary the level of resonant vs. non-resonant perturbation applied and it's alignment to the intrinsic error fields. Document plasma response using new magnetics, transport and turbulence changes (especially intermittent transport) and stability. | ||
| Background: | Previous atempts have always approached RMP ELM suppression at low torque/low toroidal rotation by starting first with a highly co-rotating ELMing H-mode. Once even a small aount of counter-NBI is applied, the rotation drops until it bifurcates to a locked state starting with at the q = 2 surface. In essence, we keep trying to run these discharges in a controlled manner through the "prohibited" part of the rotation/locking bifurcation curve. It is ill-posed to attempt to "catch and hold" the discharge at the reduced rotation state due to the hard nature of the bifurcaton to locked rotation at the q = 2 surface. In this experiment, we propose to avoid passing through the bifurcation point by starting from a low rotation ELMing H-mode and applying the RMP. We already have a proof-of-principle of ELM control at near zero net toroidal rotation at moderate collisionality in the ITER shape using odd parity (for a lower level of resonant perturbation). This scenario provided an ITER-relevant level of ELM mitigation (replacing ELMs with increased intermittent transport) with little or no pedestal height/width reduction. | ||
| Resource Requirements: | Both 210 NBs for counter injection; 30L and 330R beam for balanced torque input. ECCD for control of core MHD modes (NTMs). Additional ECH heating to raise betan. | ||
| Diagnostic Requirements: | Ion (CER) and electron profile (ECE, Thomson, reflectometer) diagnostics; Fluctuation diagnostics: swap in 150R for 330R for BES; DBS; FIR, PCI, reciprocating probes.
New magnetics SPAs for error field control and n = 3 RMP | ||
| Analysis Requirements: | Significant profile analysis, kinetic EFITs, and stability analysis (ELITE); plasma response analysis and comparison to models (M3D-C1; NIMROD,M3D, MARS-F, XGC0)
Turbulence and transport analysis | ||
| Other Requirements: | -- | ||
| Title | 144: Validation of kink response models with rotating n = 1 and n = 2 and phase flip n = 3 RMPs | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | D.M. Orlov, B. Tobias, L. Zeng, E. Unterberg, M. Shafer, A. Wingen, A. Turnbull | ITPA Joint Experiment : | No |
| Description: | Emerging understanding of plasma response to RMPs indicates that the response includes both rotational screening and kink response. The former reduces some resonant harmonics in the vacuum field by factors of 1.5â??3.5, much less than predictions from slab 2-fluid models, while the second amplifies the corresponding harmonics in the vacuum field. DIII-D has demonstrated the ability to measure the displacments of the boundary due to the kink response at the crown (top) using Thomson scattering, outboard midplane (fast camera, ECEI, profile reflectometry), and X-point (EUV/SXR imager). We have in addition new magnetics to measure the plasma response at the wall. Dedicated run time is needed to use these new diagnostics (new since CY11) to validate models of plasma response to RMPs using rotating n = 1 and n = 2 RMPs, and phase-flip n = 3 RMPs. Understanding the plasma response to magnetic perturbations is critical for understanding how these magnetic field affect ELM behavior, and will lead to a predictive model for ELM suppression in ITER. Multiple models are now available, including the vacuum TRIP3D-MAFOT model, and the MARS-F, M3D-C1, and XGC0 models. In addition, these perturbative models generate a very different kink response on the HFS along the centerpost from the global VMEC model. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish "conventional" low collisionality ELMing H-mode. Apply rotating n =1 and n = 2 RMPs at 10 Hz, and phase flip n = 3 RMPs. Use existing diagnostics to resolve the boundary displacements versus poloidal angle, and to document transport changes and the external kink n is varied. Scan collisionality, RMP amplitude (I-coil current), non-resonant vs. resonant componets (parity) and phasing with respect to intrinsic error fields to document the variation in kink response. Reverse Bt to change the direction of the poloidal rotation of the displacement (change helicity of field-aligned structure). Vary q95 which impacts the magnitude of the displacement. Vary target plasma rotation which should vary the balance between rotational screening and kink amplification. | ||
| Background: | Understanding the plasma response to externally applied magnetic perturbations is critical for understanding how these magnetic fields effect ELM behavior, and will contribute to a predictive model for ELM suppression in ITER. The DIII-D program has demonstrated an extensive diagnostic set (high resolution Thomson scattering, fast BES imaging spectroscopy, EUV/SXR imaging, ECEI, profile reflectometry, fast CER, and improved magnetics) to measure the plasma response versus poloidal angle which allows us to separate the rotational screening, kink amplification, and vacuum manifold displacements. Despite substantial efforts in CY11 and CY12 to develop these diagnostic measurements, very little run time has been allocated to advance the preliminary measurements to their fullest capability of testing the plasma response models. This experiment will capitalize on these diagnostic investments. | ||
| Resource Requirements: | I-coil with n = 1, n = 2 and n = 3 rotation and phase flips
Bt = - 2T to provide ECE and ECEI data missing from CY11 data 150R beam unmodulated for BES imaging 210 beams to vary torque input and rotation Full pedestal and profile diagnostics UCSD 90R0 passive imaging of high field side UCSD 225R0 active BES of 150R | ||
| Diagnostic Requirements: | Primary diagnostics are the UCSD fast camera, requiring 150R unmodulated, and ECE measurements (ECE and ECEI), requiring increased toroidal field to at least 2 T.
Full turbulence diagnostics to monitor fluctuation response as the relative strength of kink versus rotational screening is varied to identify origin of the turbulence changes. | ||
| Analysis Requirements: | Profile and kinetic equillibrium analysis.
Plasma response modeling of selecting "best" cases: MARS-F, M3D-C1, VMEC, etc. TRIP3D-MAFOT determination of the manifold displacements for the vacuum and plasma response models. | ||
| Other Requirements: | -- | ||
| Title | 145: Explore RMP ELM suppression mechanism at moderate collisionality | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Use new diagnostics and emerging understanding of plasma response and RMP ELM suppression at low collisionality to explore the mechanisms for ELM suppression at moderate collisionality in DIII-D. Initial RMP ELM control results were obtained in ITER shape at moderate pedestal collisionalities of ~ 1 and above due to the lack of lower divertor pumping at high collisionality; representative "trophy" discharges include: 115467 and 119690. These discharges used an odd parity RMP which didn't produce density pump-out, a major advantage to ITER. In addition, to within experimental uncertainty, the pedestal profiles remained Peeling-Ballooning unstable,suggesting that an alternate path to ELM suppression may exist. Moyer et al. {PoP 2005] have shown that there is a significant increase in broadband magnetic and density fluctuations which are correlated with an increase in intermittent "blobby" transport. Because this path to RMP ELM control avoids the loss in core performance associated with existing RMP ELM suppression experiments, the mechanism(s) for this approach should be determined and the suitability for ELM control in ITER addressed. Another aspect of these ELM suppression shots is breaking of the plasma rotation across the plasma, making this one route to ITER-relevant low rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Re-establish RMP ELM suppression at moderate pedestal collisionality (~1) in the ITER shape. May be difficult due to lower divertor changes. Vary RMP current, phasing, and parity. Vary input momentum. Document plasma response at wall with new magnetics. Document plasma rotation, Er, pedestal profile, turbulence and transport changes. Scan collisionality (0.5 -> 4). | ||
| Background: | Previous RMP ELM suppression results at moderate pedestal colllisionality has some key aspects that make it appealing for ITER: the pedestal profiles are unchanged to within measurement uncertainty, and there is no core density/beta pump-out to lower core performance. Instead, ELMs are replaced by increased intermittent transport correlated with increased broadband magnetic and density fluctuations [Moyer PoP 2005]. Since upgrading diagnostics, and forming a picture of RMP ELM suppression at low pedestal collisionality, we haven't returned to these conditions to test the model.We should establish that this ELM suppression is either equivalent to or different from that at low collisionality. If different, we will need to determine the compatibility of this approach with low collisionality (that it doesn't require higher collisionalilty) to establish whether or not it can be used for ITER. | ||
| Resource Requirements: | I-coils for n= 3 even and odd parity; SPAs
well-conditioned vessel; glow to 7 min; no cryo-pumping co and counter NBI to vary torque input and power | ||
| Diagnostic Requirements: | CER (1 ms integration time) for ion channel profiles
Thomson scattering, ECE, and profile reflectometry for electron channel profiles Divertor and periscope IRTV; DiMES TV to monitor heat and particle flux splitting SXR imaging and X point tangential TV fast cameras with spectrometer and image intensifier ECEI (higher density makes this more feasible) turbulence diagnostics (DBS, BES, PCI, FIR) lithium beam to measure edge pitch angle and current profile | ||
| Analysis Requirements: | 'Standard' profile analysis and kinetic corrections to equilibrium
profile and fluctuation analysis | ||
| Other Requirements: | -- | ||
| Title | 146: ELM suppression with Saturated walls | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | |||
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| Other Requirements: | |||
| Title | 147: Natural and Pellet-induced ELM behavior at low rotation/torque | ||
| Name: | Moyer moyer@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | |||
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| Title | 148: Radiative divertor control development | ||
| Name: | Soukhanovskii vlad@llnl.gov | Affiliation: | LLNL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | S. L. Allen, T. Petrie | ITPA Joint Experiment : | No |
| Description: | Radiative divertor feedback control is being designed for ITER and has been demonstrated at several tokamaks. It is a high priority item for NSTX-U divertor Program. We propose to start development of the radiative divertor feedback control for DIII-D. In FY2013, we propose to perform radiative divertor experiments with D2 and CD4 seeding. In these experiments, several diagnostics will be used to document plasma characteristics and extent of detachment. Functional relation between divertor heat flux, radiation and gas injection rate will be measured. <br>The diagnostics include: infrared thermography for divertor surface temperature measurements, DISRAD for fast rad. power measurements, filterscopes, neutral pressure gages, and Balmer line spectroscopy for recombination monitoring. <br>Once the database is analyzed, we will propose one or two dedicated diagnsotics tto be implemented in FY2014 for further integration in PCS with Plasma Control Group. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use a simple lower single null plasma with 2-6 MW NBI heating and D2 and CD4 injections at several puffing rates.
On the basis of these experiments, identify diagnostics: (1) divertor plasma and PFC diagnostics and (2) the diagnostics characterizing the pedestal or core plasma that can be used as ??security? measures to insure the radiative divertor compatibility with H-mode confinement. The control diagnostic signals can include divertor radiated power, neutral pressure, spectroscopic deuterium recombination signatures, infrared thermography of PFC surfaces, as well as spectroscopic ??security? monitoring of possible confinement or pedestal degradation. The characteristic detachment onset time is slow. The control signal spatial resolution should be better than 1 cm, and the ability to distinguish between inner and outer divertor leg parameters is important. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 149: Measurement of ion temperatures in the divertor region of DIII-D using the DiMES platform | ||
| Name: | Donovan ddonovan@utk.edu | Affiliation: | U of Tennessee, Knoxville |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | David Donovan, Dean Buchenauer, Jon Watkins, Brian LaBombard, Dan Brunner, Regina Sullivan, Dmitry Rudakov, Jose Boedo, Clement Wong, Josh Whaley | ITPA Joint Experiment : | No |
| Description: | Ion temperature measurements are of great importance in determining sputtering and the total heat flux on the first wall of DIII-D, as well as in the analysis of various plasma properties that are determined by other diagnostics, particularly the Langmuir probe array. The Langmuir probe array is capable of providing electron temperature and plasma density at the probes. The properties determined by the Langmuir probes can then be used to infer the total energy of ions striking the wall (sheath drop + ion energy) and the heat flux at the probes by calculating the sheath power transmission factor (SPTF). While the Langmuir probe analysis and the determination of the SPTF both depend on the ratio of ion to electron temperature, the Langmuir probes alone are incapable of providing ion temperatures. Ion temperature measurements require diagnostics such as the retarding field analyzer (RFA) or the ion sensitive prove (ISP). <br> <br>The RFA uses biased grids to deflect electrons and ions below a specific temperature, thereby determining the energy distribution of the ion current by varying the bias on the discriminator grids. The ISP uses a recessed current collecting plate to measure ion current. The magnitude of the ion current at a given energy can be determined by matching the recessed plate depth to the Larmor radius of the ions as they spiral along the magnetic field lines towards the wall. Both of these probe designs have recently been tested with success on the Alcator C-MOD fusion experiment at MIT. The probe was inserted near the midplane using a retractable probe. The DiMES platform offers the ideal opportunity for testing a similar design in the divertor of DIII-D. These probes were designed to withstand heat fluxes at the mid-plane of C-Mod up to 700 W/m2. The ISP is more thermally robust than the RFA, but has been found to suffer from space charge accumulation on the electrically floating surfaces. This will require some modifications to the C-Mod design to compensate for the space charge buildup in order to obtain accurate Ti measurements. <br> <br>Because DiMES can be inserted and removed for a single shot, the probes can be inserted and exposed to a sweep of the strike point and then retracted. Total heat flux on the probe could be limited reducing the risk of damaging the probe during operation. Initial testing of the newly designed DiMES ion temperature probes can be done without requiring a strike point sweep by measuring scrape off layer currents. Later testing could be accomplished on piggyback shots with a short sweep over DiMES. The implementation of such a probe design will be done in collaboration with the C-MOD research group. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A DiMES head will be used that is equipped with a retarding field analyzer (RFA). The RFA contains two biasing grids, the first of which is used to repel ions below a certain energy and the second to repel electrons. A current collecting plate gathers the remaining ion current, allowing the energy distribution of the ion current to be determined. These grids will be susceptible to melting in excessive heat flux and therefore will require sweeps of the strike point. Preliminary testing can be done in the scrape off layer and will not require any interaction with the strike point. Once preliminary testing has been completed, piggyback shots will be requested to perform OSP sweeps over DiMES.
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. Another DiMES head will be designed containing an ion sensitive probe (ISP), which uses a recessed current collecting plate. The ISP is more robust than the RFA at withstanding high heat fluxes, but suffers from space charge accumulation on the electrically floating surfaces, which will require further design development to optimize the geometry of the head. Like the RFA, this probe will initially be tested in the SOL and will later be requesting piggyback shots to obtain OSP sweeps over DiMES. Both of these probe designs have been tested successfully on the midplane of C-MOD. The Sandia research group will be collaborating with the C-Mod diagnostic group to adapt the designs for the DIII-D divertor. Procedure Initial testing of the RFA and ISP probes will not require interaction with the strike point. The procedure below is for the eventual piggyback shots that will be requested. 1. Start with ohmic plasma with 250 msec inward sweep. Setup line scan for the IR camera to provide fast time resolution of the heat flux (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, Langmuir probe data and ion temperature for the DIMES Langmuir probe and surface temperature using IR camera in line scan (3 shots). 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, while watching 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 determination of the divertor plasma ion temperature. 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 requires further study. A study to better understand the role of probe geometry and magnetic sheath effects was carried out in the 2012 campaign. Measurements of the ion temperature will add to our understanding of the sheath power transmission and provide critical data on the ion energy striking the divertor surface. Since sputtering depends sensitively on the sheath potential and ion energy, these measurements are also critical to modeling of erosion and redeposition of divertor material. | ||
| Resource Requirements: | The hardware for the DiMES probe is available, along with instrumentation provided by the divertor Langmuir probe array. The experiment would require the IR camera and fast thermocouple array. Run time of approximately 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 the fast line scan mode of the IR camera (to improve time resolution during the x-point sweeps). Divertor Langmuir probes Ion temperture probes mounted on the DiMES system, + instrumentation IR camera (preferable in line scan mode) Fast thermocouple array Divertor spectroscopy Magnetics for EFIT determination of field angles Zeff C02 interferometer Thomson scattering Fast filterscope channels viewing the lower divertor Other useful diagnostics Tile current array Bolometers Edge CER for ion temperatures | ||
| 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. | ||
| Other Requirements: | -- | ||
| Title | 150: Sensitivity study of toroidally-asymmetric halo current to error field during VDE | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | It has been reported that toroidally-asymmetric halo currents associated with VDE are comparable to n=0 component. Configuration of VDE approaches toward helical equilibrium and locking is initiated at some final stage. Halo current characteristics are not fully consistent in various devices.
Objective in this proposal is to assess whether toroidally-asymmetric halo current component is related to the intrinsic and/or induced error field during the event of VDE. This assessment will help to understand the nature of halo current characteristics and clarify the possibility of n=1 error field involvement. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This proposal has passive / active documentation after VDE is induced by forced vertical motion, inducing VDE event. Plasma condition is reasonably high beta before VDE so that the plasma configuration becomes ideal kink unstable for helical distortion
Step one: passive documentation. To apply the constant n=1 non-axi-symmetric field by C-coil, Upper-I coil and Lower-I-coil and correlate with the halo current pattern. The toroidal direction of applied field will be scanned in standard compass approach. Step two : active documentation. If the step one implies the correlation between the halo-asymmetry and error field, it is worthwhile to apply the feedback using most promising coil set. According to the experience of NTM-locking avoidance by I-coil feedback, the feedback may be able to rotate the pattern. The feedback field should interact with magnetic islands. There are some cautions: - The available coil current by C- and I-coil are limited. Some homework is to be made for adequacy of asymmetric field availability. Thus, the measurement may face difficulty in analysis. Some homework is to be made for adequacy of asymmetric field availability. - we need to pay attention to the F7 / F6 combination as intrinsic error source. The large error field component may come from these coil current. | ||
| Background: | |||
| Resource Requirements: | |||
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| Other Requirements: | |||
| Title | 151: ITER SS Demo Discharge Documentations | ||
| Name: | Murakami masanori_murakami@att.com | Affiliation: | Retired |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | JM Park, T. Luce, J. Ferron, C. Holcomb, t. Osborne, G. Jackson, M. Okabayashi, P. Ryan, S. Diem, SSI Group | ITPA Joint Experiment : | No |
| Description: | ITER demonstration discharges are important data for ITER hardware heating and CD design and development of research plans. Theory-based integrated modeling for ITER SS scenario has been extensively used in ITPA-IOS group activities, such as H&CD mixes/upgrades studies. However, it needs new data based on improved hardware, diagnostic and analyses capabilities now available in DIII-D. This is a resubmission of the old ideas, but is significantly updated. In particular, it includes better more optimization, edge pedestal measurements, ITER-like off-axis NBCD, and higher ECCD capabilities. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Start with old 134372
2) Make closer to optimal ITER shape and dRsep scan 3) Error field minimization (with C and i-coils) 4) Document and evaluate f_NI, f_BS and G=betaN*H/q^2 5) Adjust timing of the high power phase at q_min=2 crossing 6) Add broadly distributed ECCD around rho=0.3=0.6 and move out 7) Higher Ip (equiv. to 11 MA) 8) Document and evaluate f_NI, f_BS and G=betaN*H/q^2 again | ||
| Background: | â?¢ ITER SS Goal must succeed as the first step toward for the Fusion Reactor Mission
â?¢ ITER SS Demo shot should aim at conditions within â??close to ITER hardware (or foreseeable upgradable) capabilities â?¢ 2 successful ITER SS Demo shots in 2008: equiv. Ip=8.5 MA and Ip=13 MA were carried out â?¢ ITER steady state scenario modeling using TGLF carried out using â??scaled experimentalâ?? edge from DIII-D ITER demo discharge (#134372) barely achieves fNI=100% and QDT=5, but performance with EPED1-edge is far too short. This is at least in part due to the EPED database not including steady-state type (in particular ITER Demo) discharges. This situation has to be alleviated â?¢ More credible ITER scenario development and addressing the H&CD mixes / Upgrade questions require good DIII-D ITER SS Demo shots followed by other devices in the future â?¢ We need ITER shape (dRsep) optimization and, good edge profiles (Ï? = 0.8 â?? 1.0) measurements that were not emphasized previously. â?¢ ITER-type off-axis NBCD and ECCD deposition are unique in DIII-D | ||
| Resource Requirements: | >5 NB sources (including (off-axis NBCD (â??5-deg steering)) in normal BT direction
>=5 ECCD | ||
| Diagnostic Requirements: | Full core diagnostics, MSE, Fast ion diagnostic (UC, Irvine), Edge TS, Edge reflectometer (UCLA, ORNL) | ||
| Analysis Requirements: | Scenario modeling with FASTRAN/ONETWO (TGLF), TRANSP, ONETWO time-dependent analysis; CURRAY/ONETWO, AORSA | ||
| Other Requirements: | 1 day min., 2 days desirable | ||
| Title | 152: Transport shortfall at high gyrobohm flux | ||
| Name: | Smith smithsp@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | Chris Holland, Nathan Howard, Craig Petty | ITPA Joint Experiment : | No |
| Description: | Chris Holland recently showed that for a database of shots, the shortfall of predicted transport is worst when experimental fluxes in gyrobohm units are high. This was most recently seen in the DeBoo 2011 electron critical gradient and stiffness experiment at the highest Te gradients and heat fluxes: GYRO predicted fluxes agreed well at lower fluxes and gradients, but not at the highest. The idea here is to perform a systematic radial scan of localized ECH heating to obtain the high fluxes at multiple radii. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use 5-6 gyrotrons aimed at the same location to create a high flux/gradient region just outside (in minor radius) of the deposition region. Possibly use the 6th for heat pulse measurements. Obtain steady state conditions over a scan of the radial deposition location. Possibly incorporate additional neutral beam heated phases, although it has been shown in the past that these are not so useful for getting to the highest Te gradients. There should be multiple opportunities to sample the ion kinetic quantities (an occasional beam blip.) | ||
| Background: | |||
| Resource Requirements: | 5 gyrotrons for power balance interpretive alaysis; 6 for additional heat pulse and therefore non-diffusive analysis. | ||
| Diagnostic Requirements: | ECE. Michelson ECE. CECE. Thomson. Profile Reflectometry. CER. DBS. Bolometer. | ||
| Analysis Requirements: | Power balance analysis with ONETWO or TRANSP. Fluctuation analysis by fluctuation experts. GYRO and TGLF runs to get fluxes. | ||
| Other Requirements: | |||
| Title | 153: Controlling VH-mode with RMPs and comparison to NSTX EPH-mode | ||
| Name: | Canik canikjm@ornl.gov | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | R. Maingi, S.P. Gerhardt | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to extend the duration of the VH-mode by applying RMPs, and to perform a comparison of the edge characteristics to those of the EPH-mode observed in NSTX, which has some features in common with VH, as part of the FY13 JRT. This is partially motivated by the recent progress in understanding ELM suppression via RMP application, and if successful will guide later experiments on NSTX aiming to control and extend the duration of the EPH-mode. The comparison to EPH-modes includes kinetic profiles, the structure of the ExB profiles will be documented, and the edge fluctuation characteristics. The measurements will be supplemented by analysis of the macro- and micro-stability properties of the edge plasma. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The first step in this experiment is to establish a robust, high quality VH-mode. To facilitate comparison to NSTX, this will be done in a DN shape with high triangularity. Once established, a sufficient number of shots will be taken to thoroughly measure the edge kinetic profiles and turbulence characteristics. Finally, n=3 RMP will be applied following the transition to VH-mode in order to attempt to extend the ELM-free VH phase. This will likely require tuning based on recent DIII-D analysis, including q-ramps to search for resonant windows and to place appropriate rational surfaces near the pedestal top in order to halt its inward growth. | ||
| Background: | The FY13 JRT is targeted at developing and understanding stationary enhanced confinement regimes without large ELMs. To date the VH-mode is not stationary, ending due to a large ELM and exhibiting a ramping density. However, the application of RMPs may allow stationary conditions to be achieved in VH, by suppressing the first large ELM as well as providing particle control via the routinely observed density pump-out. While this has been tried in DIII-D in the past, the understanding of RMP ELM-suppression has grown significantly since then, and the divertor geometry has been altered to improve pumping at high triangularity, so that new experiments are more likely to be successful. Further, the Enhance-Pedestal (EP) H-mode observed in NSTX experiments has several features that are similar to VH-mode (although there are also some differences). In particular, the EPH-mode exhibits a substantially wider edge region with significant ExB shear, along with a widening of the pedestal and higher pedestal-top temperatures; this results in higher confinement times (up to ~4 times L-mode scaling). Access to EPH-mode is also facilitated by achieving low-recycling conditions (at NSTX accomplished through lithium coatings applied to the PFCs), and shows reduced particle confinement compared to ELM-free H-modes. Part of the goal of the proposed experiment is to directly compare edge profiles and turbulence in EPH and VH-modes to explore these commonalities further. This effort will have a broad impact in developing these improved confinement regimes (both VH and EPH) towards steady-state. | ||
| Resource Requirements: | ~1 day, RMPs, cryopump | ||
| Diagnostic Requirements: | TS, CER (esp. edge), BES, reflectometry | ||
| Analysis Requirements: | Profile analysis, edge stability (ELITE), edge microstability (GS2/GENE), transport (ONETWO/TRANSP) | ||
| Other Requirements: | |||
| Title | 154: High frequency pellet ELM pacing | ||
| Name: | Baylor baylorlr@ornl.gov | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux, A. Loarte, M. Fenstermacher | ITPA Joint Experiment : | Yes |
| Description: | Extend the pellet induced ELM frequency up to 90Hz with the 0.9mm pellet size to investigate DelW vs 1/fELM beyond 60 Hz. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | |||
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| Title | 155: Investigation of pellet induced ELM heat flux footprint variation with q95 | ||
| Name: | Baylor baylorlr@ornl.gov | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux, A. Loarte, C. Lasnier, M. Fenstermacher, S. Allen | ITPA Joint Experiment : | No |
| Description: | Investigation of non-axisymetry in the heat flux footprint from pellet induced ELM filaments. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Vary q95 while observing the IR camera heat flux footprint in the divertor. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | Requires tangential periscope IR camera if available. | ||
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| Title | 156: Controlling the timing and evolution of L-H and H-L with 3D fields | ||
| Name: | Battaglia dbattagl@pppl.gov | Affiliation: | PPPL |
| Research Area: | Plasma Control | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Superconducting tokamaks, such as ITER, will benefit from a controlled and slow transition in and out of H mode. Maintaining good plasma control through the H-L transition is of particular concern for ITER. This experiment would examine the effectiveness of using 3D fields to trigger L-H and H-L transitions and slowing the pedestal pressure evolution to minimize dV/dt requirements on PF coils ("soft landing"). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Establish a discharge where L-H is triggered by a small increase in NBI and H-L is triggered by a decrease in NBI. This establishes baseline pedestal evolution associated with L-H and H-L.
2a. Program the n=3 to turn on right after L-H transition and then ramp to zero. Does this slow the pedestal pressure evolution? How does the evolution depend on n=3 amplitude and ramp rate to zero? 2b. Remove step down in NBI and try to trigger H-L with n=3. How does the timing and pedestal evolution following H-L change with n=3 amplitude? 3a. Try triggering L-H by applying n=3 during L-mode, increasing NBI, then notching out n=3. How long does the notch out need to be? 3b. If H-L timing is reproducible, trying reducing or removing n=3 soon after H-L transition to slow pedestal collapse. | ||
| Background: | H-L back transitions are observed on DIII-D with n=3. May be due to collapse of pedestal gradients and/or slowing the rotation. ITER will have internal non-axisymmetric coils that have faster time response than the PF coils. | ||
| Resource Requirements: | SPAs on I-coils | ||
| Diagnostic Requirements: | Profile diagnostics. L-H edge diagnostics. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | The middle of the pulse (H-mode) is available for other experiments (synergistic with ELM control). Also gets good data for 3D effect on L-H, H-L. | ||
| Title | 157: Imaging of pellet induced ELM events from mulitple simultaneous views. | ||
| Name: | Baylor baylorlr@ornl.gov | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux, R. Moyer, Z. Unterberg, S. Allen | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | Use up to 3 simultaneous views of LFS pellets triggering ELMs at 135 R-2 and 135 Midplane locations. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | New fast camera views from 90 Periscope and 135 Dimes upper port. | ||
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| Title | 158: 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: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Follow-up on the two high li experiment days in 2008 and 2012. Focus on eliminating the large perturbation of the first ELM that prevents consistent long pulse, high betaN, performance in the high li scenario and prevents operation at lower q95. This may involve testing a new method to form the high li discharge. Take advantage of heating and current drive upgrades since 2008: a neutral beam power level above 11 MW (providing that power is actually available this year) and a sixth gyrotron. 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 reliably and at lower values of q95, it is necessary to avoid the early, fast-growing n = 1 mode that has the appearance of the first ELM. Since a number of methods were tried in 2012, with apparent backwards progress, the starting point again would be the scenario used in 2008. Further analysis of the 2012 data is required in order to determine the next steps for the experiment. | ||
| 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. Attempts to improve this high li scenario were made during an experiment day during 2012. Those discharges still need to be analyzed carefully in order to determine the best next step. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | Would make use of FIDA. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 159: Collisionality dependence of pedestal height with RMP | ||
| Name: | Nazikian nazikian@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Requested |
| Co-Author(s): | Todd Evans, Tom Osborne, ... | ITPA Joint Experiment : | No |
| Description: | RMPs are known to suppress ELMs but at the cost of plasma performance at low collisionality. However on ASDEX-U ELM suppression is observed without significant impact on performance. This proposal aims to understand the role of collisionality and edge density on the edge pedestal pressure in ELMing discharges. They key question is whether the effect of density pump out on the pedestal height is consistent with P-B theory and experiment in plasmas where q-95 is outside of the ELM suppression window. The implication, if true, is that we can better predict the impact of the RMP on the pedestal height in ITER with the application of RMPs if the physics that determines the pedestal height with RMP in typical ELMing discharges is determined entirely by P-B theory and KBM physics. Basically, confirming the EPED1 model vs collisionality for RMP plasmas would help establish a predictive understanding of how RMPs may affect the pedestal height and width in ITER, leading to more reliable predictive understanding of RMP physics. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use the ITER baseline scenario. Repeat the experiment performed by Tom Osborne that showed the pedestal scaling with collisionality by adding RMP even parity and avoiding q-95 resonant window for ELM suppression. Use additional ECH and gas puff where needed to affect edge collisionality. Shift strike point or turn off cryo pump to further affect collisionality and edge density with RMP. Compile data that cam be compared with collisionality scans in the absence or RMP. Use this to extrapolate to pedestal performance in ITER with RMP. | ||
| Background: | |||
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| Title | 160: Dependence of confinement and stability on toroidal rotation in high li discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | 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: | |||
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| Title | 161: n=3 offset and ELM suppression with RMP | ||
| Name: | Nazikian nazikian@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Scan the n=3 offset with phase flips of the n=3 RMP in order to identify whether there are island structures at the top of the pedestal. Basically, we want to extend the offset level beyond previous offsets so that a minimum in the offset can be clearly identified by the minimum in the beta modulation. With the new 3-D coil set the offset can be better diagnosed. In addition the thicker filter planned for X-ray imaging will be more sensitive to core structures. The combination of the optimum offset with the new mgnetics and the thicker filter will be very effective in studying the existence of islands. In addition, performing this at several values of q-95 will help to identify the existence of an island response at the top of the pedestal. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Repeat phase flip experiments of the past campaign, add q-95 scan from shot to shot and optimize the offset with the new magnetics and measure the X-ray images and profile changes. | ||
| Background: | |||
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| Title | 162: 210RT MSE and low rotation, elevated q_min steady-state scenario discharges at high betaN | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | 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. Begin by adding relatively small amounts of counter-beam power from the 210 right source in order to determine whether MSE data viewing this beam can be relied upon to replace the 30 left channels when the 30 left beam line is tilted. 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. | ||
| 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. Finally, in the five-year plan proposal, tilting the 30 left beam is contemplated. That would disable the core MSE channels viewing that beam. We need to determine whether new MSE views are necessary or whether the view of the counter beam can be used in steady-state scenario discharges. | ||
| Resource Requirements: | |||
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| Title | 163: Exploration of DEFC in the feedback stabilized plasma in low-q discharges | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | Ted Strait, Yongkyoon In | ITPA Joint Experiment : | No |
| Description: | DEFC is an attractive tool to evaluate uncorrected error field in stable plasma condition. The necessary condition is simply that the resonant field amplification can take place to the applied non-axi-symmetric field. DEFC should function also even in feedback stabilized plasmas according to a simple model(YongKyoon In 2010 IAEA). However, the operation will be very sensitive to the gain setting. <br> <br>In addition, the magnitude of coil current of DEFC changes, confusingly, from marginally-stable condition to marginally-unstable since the coil current in unstable condition does overestimate the error field correction in contrast to underestimate in the stable condition. The improved approach near the marginal condition was proposed by Yongkyoon In andTed Strait. <br> <br>Recent achievement of q~2 condition approaches asymptotically toward marginal condition. This condition provides an excellent opportunity to verify the model prediction and improved approach. <br> <br>This exercise is useful not only for the DEFC in marginal condition of low-q, but also for direct feedback once we enter deeply in the unstable regimes. <br> <br>This study can be assisted by fast active MHD spectroscopy to determine the instantaneous stability condition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | When the plasma condition approaches towards marginal condition, we apply AMS of 250-500 Hz (Go Matsunaga measured reasonable response) to observe the plasma response. This ASM would be useful to assess EFC in low-q ~2 operation and high beta operation near ??real stability limit? including rotation and kinetic stabilization effect.
We need careful gain setting before and after the estimated marginal time period considering the model prediction. The resetting of error field correction based on the actual feedback current are also made based on the model prediction | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 164: Produce fNI = 1 discharges at q_min near 1.5 using off-axis injection and a model guided approach | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Focus on producing discharges which robustly have fNI = 1 with good current profile alignment. Run these discharges for as long as possible in order to allow good nvloop analysis of the electric field profile. Make use of a new empirical model that extrapolates from the existing database of steady-state scenario discharges in order to predict a self consistent set of parameters with fNI = 1. Use the long pulse discharge 147634 as the starting point, but operate at higher betaN and higher input power. Use the fNI = 1 discharge 133103 as a comparison point. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Most of the approach is described in the description paragraph. The starting point would be discharge 147634. Scans in beam power (to raise betaN) and Bt (at constant betaN, thus increasing beam power) would be used to find the fNI = 1 operating point that is most tearing mode stable. The results from the scan would provide additional data to contribute to our understanding of the scaling of fNI = 1 operating regimes. 147634 had q_min = 1.5 where confinement is good enough to perform this experiment with the available beam power and off-axis injection. | ||
| Background: | Recently, there has been a better understanding of the necessity to choose a self-consistent set of parameters in order to obtain fNI = fBS + fCD = 1. In particular, many of the steady-state scenario experiments that have been run in the past have been run under conditions that require too little beam power (e.g. good confinement or low betaN target). As a result, the beam-driven current fraction was not high enough to reach fully noninductive conditions. Discharges in fall 2011 were operated at betaN = 3.5 with relatively low fNI = 0.7 and relatively low beam power (in order to achieve a long pulse). These discharges should be scaled up, either in betaN or Bt in order to increase the necessary beam power and reach fNI = 1 (although at shorter pulse length). | ||
| Resource Requirements: | |||
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| Title | 165: D2 + few % Ne shattered pellets for disruption mitigation | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | P. Parks, L. Baylor, E. Hollmann, D. Humphreys, N. Eidietis, J. Wesley, V. Izzo | ITPA Joint Experiment : | No |
| Description: | Use the SPI to inject D2 pellets containing a few % (~5) of neon in an H mode plasma. Simulations carried out by P. Parks have determined that this approach should enable achieving higher densities than regular D2 shattered pellets without inducing too fast a current quench. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Inject these D2+Ne shattered pellets in H mode plasmas with various amount of injected power to vary the thermal energy content. Measure the final density using spectroscopic diagnostics (SPRED, CER,VB), interferometry (3rd colour)â?¦ | ||
| Background: | Achieving high current quench electron densities could prove key to suppress or at least to mitigate the runaway production in ITER. Present massive particle injection systems can achieve pretty high densities but they are still a factor of 5 too low assuming that reaching the Rosenbluth density is critical. The other issue is that injecting more particles could induce current quenches that are too fast in ITER. These could induce damages on the internal structures of the vessel because of the high level of Eddy currents. Simulations done by P Parks on shattered pellet injection showed that trace amounts of Neon in a deuterium pellet could potentially raise the final density significantly without increasing the current decay rate too much. This is what this experiment is proposing to test. | ||
| Resource Requirements: | The SPI. Regular tokamak systems. Good beams availability | ||
| Diagnostic Requirements: | Regular disruption diagnostics. The new IR periscope for heat load characterization. CER set for VB, SPRED | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 166: Effect of q95 on SPI shutdown | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, E. Hollmann, D. Humphreys, P. Parks, N. Eidietis, J. Wesley, V. Izzo | ITPA Joint Experiment : | No |
| Description: | Use the SPI to inject D2 pellets in an H mode plasma with different values of q95. Experiments carried out last year showed some interesting results concerning the onset of the current quench following an SPI injection. It seems quite different from MGI scenarios. But the data was not large enough for definitive conclusion. This year experiment would provide more data and also by varying Bt instead of Ip, determine the effect of ohmic heating and confinement on these shutdowns | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Inject these D2 shattered pellets in H mode plasmas with various Ip and Bt to vary q95. . Measure the final density using spectroscopic diagnostics (SPRED, CER,VB), interferometry (3rd colour)â?¦, current quench onset time, decay time, MHD activityâ?¦ | ||
| Background: | SPI has proven an efficient disruption mitigation system and is regarded as one of the prime candidates for the ITER DMS. But the lack of data on this new method for fast shutdown make it difficult to extrapolate to ITER. This experiment would help understand the basic mechanism of shutdown: what kind of MHD is triggered, do rational surfaces play a role, how does the current profile evolveâ?¦ | ||
| Resource Requirements: | The SPI. Regular tokamak systems. Good beams availability | ||
| Diagnostic Requirements: | Regular disruption diagnostics. The new IR periscope for heat load characterization. CER set for VB, SPREDâ?¦ | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 167: Control of radiation asymmetries by application of external field on SPI/MGI shutdown | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | V. Izzo, L. Baylor, E. Hollmann, D. Humphreys, P. Parks, N. Eidietis, J. Wesley | ITPA Joint Experiment : | No |
| Description: | Determine the role of the n=1 mode during the shutdown in the magnitude of the radiation assymetries. This is done by applying an external n=1 field during the shutdown process with MGI/SPI in order to change the behavior (lock ?) the n=1 mode responsible (according to NIMROD simulations) for the bulk of the radiation assymetries observed during a massive particle injection. Using the new IR periscope and scanning different phases of the n=1 could show that the IR footprint of the thermal quench evolves due to the n=1 seed. That would be a proof that the MHD activity is responsible for the asymmetry. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use the SPI and/ or MGI to inject impurities in an H mode plasma while applying a strong n=1 external phase using the I and C coils. Measure the IR footprint using the new IR periscope and the IR fast camera to determine the evolution of the footprint as a function of the phase and the intensity of the n=1 field. | ||
| Background: | Radiation assymetries induced by massive particle injection is a major concern for ITER. These asymmetries could induce local melting of the first wall if the toroidal radiation peaking factor reach a certain value. Injections at multiple toroidal locations are expected to mitigate this effect because it is assumed that the radiation asymmetry is due to the injection asymmetry. But recent results obatained by V. Izzo on NIMROD tend to show that these asymmetries could in fact be induced by a non rotating n=1 mode growing during the thermal quench and expelling heat from the plasma. The phase of that mode with respect to a massive particle injection would then determine the peaking factor. | ||
| Resource Requirements: | The SPI. Regular tokamak systems. Good beams availability. I and C coils | ||
| Diagnostic Requirements: | Regular disruption diagnostics. The new IR periscope for heat load characterization. CER set for VB, SPREDâ?¦ | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 168: Influence of the delay time between multiple massive gas injections at different toroidal locations | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, E. Hollmann, D. Humphreys, P. Parks, N. Eidietis, J. Wesley, V. Izzo | ITPA Joint Experiment : | No |
| Description: | Determine the effect of the delay between multiple injections using the new MGI array at 135R-2 and the existing MEDUSA array at 15R+1 by doing a delay scan between the 2 systems. This scan was previously done on Cmod but the system on DIII-D would have a significantly different disposition (1/1 vs 1/2 on Cmod). The bolometer arrays and IR cameras would provide valuable data on the magnitude of the radiation asymmetry as a function of the delay. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Inject 2 MGI pulses using the 2 MGI arrays with a set delay time between the 2 in a regular H mode discharge and scan that delay time between the 2 injections to determine what is the evolution of the IR footprint and radiation asymmetries. | ||
| Background: | Radiation asymmetries induced by massive particle injection is a major concern for ITER. These asymmetries could induce local melting of the first wall if the toroidal radiation peaking factor reach a certain value. Injections at multiple toroidal locations are expected to mitigate this effect because it is assumed that the radiation asymmetry is due to the injection asymmetry. But recent results obtained C-mod tend to show the opposite: simultaneous injections on C-mod at 2 opposite toroidal locations induce a significantly stronger asymmetry (during the thermal quench) than one single injection. Simulations on C-mod show that this could be due to a strong n=1 mode growing and locking during the thermal quench. The phase of that mode with respect to an injection site being critical. But these results on C-mod were obtained using a 1/2 injection symmetry. Which might force the n=1 mode to right in front of an injection site (worst case scenario). The system on DIII-D will be a 1/1 symmetry which might prevent that effect and mitigate the asymmetries. | ||
| Resource Requirements: | Both MGI arrays. Regular tokamak systems. Good beams availability. | ||
| Diagnostic Requirements: | Regular disruption diagnostics. The new IR periscope for heat load characterization. CER set for VB, SPREDâ?¦ | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 169: Runaway flattop mitigation by Neon shattered pellet injection | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, E. Hollmann, D. Humphreys, P. Parks, N. Eidietis, J. Wesley, V. Izzo | ITPA Joint Experiment : | No |
| Description: | Determine the efficiency of a neon shattered pellet at mitigating an existing runaway beam. For that use the usual runaway target and simply inject a neon shattered pellet in the runaway beam. The present SPI system is capable of doing that. Previous neon SPI experiment couldnâ??t be run because of neon blow by issues that were disrupting the plasma when firing the pellet but before the pellet could reach the plasma. In the case of a runaway plateau, this wouldnâ??t be so much of an issue since the runaway beam could sustain a small amount of neon with disrupting (already been tested with neon MGI on runaway plateaus). | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use the SPI to shoot a neon shattered pellet on the â??usualâ?? runaway target. For that produce a runwaway beam using a small (3mm) argon pellet and then inject a neon shattered pellet using the greased pellet scheme developed by ORNL. Measure any increase in the runaway losses using the BGO and plastic scintillators. Study the evolution of the equilibrium and Ipâ?¦ | ||
| Background: | Runaway formation in ITER might be impossible to avoid. There is thus an important need for a mitigation system working on an existing runaway beam. MGI was tested with some success in DIII-D previously using high Z impurities. SPI with its faster delivery and high ram pressure migh provide a more efficient impurity delivery for this scheme. | ||
| Resource Requirements: | The SPI. Regular tokamak systems. Good beams availability, Argon pellet injector | ||
| Diagnostic Requirements: | Regular disruption diagnostics. The new IR periscope for heat load characterization. CER set for VB, SPREDâ?¦ | ||
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| Title | 170: Nature of Energy Released During H-L Back Transition | ||
| Name: | Eldon eldond@fusion.gat.com | Affiliation: | GA |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | Boivin, Groebner, Osborne, Snyder, Tynan, Schmitz | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to determine the nature of energy released during the back transition out of H-mode and to test control of the transition speed by changing power ramp down rate. <br>The start of a transition from H-mode to I-phase is marked by an event characterized by a flash of D_alpha light and sometimes activity detectable by magnetic probes. This is followed by several smaller bounces of decreasing brightness (in D_alpha) which resemble limit cycle oscillations described by the predator prey model. Are these flashes the results of simple bursts of enhanced transport, or are they related to some unstable mode? How does energy loss to the divertor compare to ITER requirements? <br>The predator prey model predicts that adjusting the rate at which input power is decreased will control the length of an I-phase during the back transition. This will be tested. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Keep one beam on throughout the I-phase for diagnostics, power with de-rated beams so that power can be removed in small increments. Different power ramp down rates will be tested with multiple L-H and H-L transitions per shot, as in 148698 and similar. One hope is for bunch mode Thomson scattering to capture the start of the back transition, or very soon before it. All possible profile diagnostics are desired for stability analysis leading up to the H-L transition. One power ramp rate will be chosen for a target density scan. | ||
| Background: | -- | ||
| Resource Requirements: | 30, 330, 150 beams, ECH | ||
| Diagnostic Requirements: | BES array centered on pedestal, DBS, Thomson scattering in bunch mode, CER, profile reflectometry | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 171: Test of the argon pellet injector | ||
| Name: | Commaux commaux@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, S. Meitner, S. Combs, E. Hollmann, D. Humphreys, P. Parks, N. Eidietis, J. Wesley, V. Izzo | ITPA Joint Experiment : | No |
| Description: | The argon pellet injector is to be tested after upgrades and modifications to check if it can inject successfully 3mm argon pellets in the plasma without disrupting it too early. This test would also verify that these pellets produce efficiently significant amounts of runaway electrons. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use the upgraded argon pellet injector on the usual runaway generation target plasma (ECH heated circular limited discharge). Check if the pellets are observed in the plasma before the disruption (camera, ablation signal) and verify the production of runaways (scintillators, Ip)â?¦ | ||
| Background: | A new argon pellet injector was built, installed and commissioned on DIII-D in 2012 to enable the reliable production of runaways. It successfully injected pellets in the plasma but issues with the guide tube shattering the pellets and high vapor pressure of the argon ice made it difficult to operate and didnâ??t enable the production of runaways. The injector is being modified to address both issues (new guide tube, new stand and mechanical punch installed). The upgraded system is scheduled to be reinstalled on DIII-D before startup and has to be tested to be commissioned for experiments. | ||
| Resource Requirements: | The argon pellet injector. Regular tokamak systems. ECH | ||
| Diagnostic Requirements: | Visible cameras, pellet diagnosticsâ?¦. | ||
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| Title | 172: High betaN with off-axis injection at reduced Bt (if 2011-level beam power is available) | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Test the betaN limit with q_min >2 and broad pressure profiles produced by off-axis injection. Use as a basis one of the Bt = 1.4 T discharges from 2012 which were produced without the expected 14 MW neutral beam power. Do the experiment with both constant Bt and a negative Bt ramp as was studied during 2012. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | If neutral beam power about 14 MW is available during 2013 as it was during 2011, this experiment could be performed. During 2012 experiments we learned how to produce discharges with q_min >2 at Bt = 1.4 T. So that doesn't need to be redone.. We would start with one of those discharges and put in the higher neutral beam power in order to push to higher betaN. We would also test the betaN limit with the broader current profile anticipated with a negative Bt ramp, making use of what we learned about this type of discharge during 2012 experiments. | ||
| Background: | During 2012 experiments, the intention was to test the betaN limit at q_min >2 where betaN was limited to a maximum of 3.3 in 2011. The idea was to reduce Bt where less beam power would be required. However, the beam power actually available during 2012 experiments was less than the beam power available in 2011 so the maximum betaN achieved was about the same. | ||
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| Title | 173: Confinement and fast ion diffusion in high betaN, steady-state scenario discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Produce dedicated discharges for a study of confinement in steady-state scenario discharges at high betaN. Obtain discharges where an optimized comparison of predicted and measured fast ion density profiles can be made. 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. Also produce discharges optimized for obtaining fluctuation data for understanding of thermal transport. Do a scan of q_min. Look at confinement during the scan, and the scaling of the ratio of the transport code stored energy to the EFIT stored energy. Look for systematic discrepancies in measurements of the thermal stored energy. Correlate confinement to fluctuation levels for fast ion driven instabilities. Look for parameter regimes where fast ion losses can be minimized while still operating at high q_min >2. Coordinate this experiment with other experiments to measure the NBCD profile at high betaN. | 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. In addition, at q_min >2 in discharges with off-axis neutral beam injection, global confinement seems to be reduced as a result of enhanced fast ion transport while thermal confinement seems to be at expected H-mode levels. We need to understand the scaling of thermal and fast ion confinement in these discharges and where there are parameter regimes where good confinement is present so that high betaN can be achieved with the available neutral beam power. | ||
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| Title | 174: Exploration of DEFC in the feedback stabilized plasma in low-q discharges | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Ted Strait, Yongkyoon In | ITPA Joint Experiment : | No |
| Description: | DEFC is an attractive tool to evaluate uncorrected error field in stable plasma condition. The necessary condition is simply that the resonant field amplification can take place to the applied non-axi-symmetric field. DEFC should function also even in feedback stabilized plasmas according to a simple model(YongKyoon In 2010 IAEA). However, the operation will be very sensitive to the gain setting.
In addition, the magnitude of coil current of DEFC changes, confusingly, from marginally-stable condition to marginally-unstable since the coil current in unstable condition does overestimate the error field correction in contrast to underestimate in the stable condition. The improved approach near the marginal condition was proposed by Yongkyoon In andTed Strait. Recent achievement of q~2 condition approaches asymptotically toward marginal condition. This condition provides an excellent opportunity to verify the model prediction and improved approach. This exercise is useful not only for the DEFC in marginal condition of low-q, but also for direct feedback once we enter deeply in the unstable regimes. This study can be assisted by fast active MHD spectroscopy to determine the instantaneous stability condition. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | When the plasma condition approaches towards marginal condition, we apply AMS of 250-500 Hz (Go Matsunaga measured reasonable response) to observe the plasma response. This ASM would be useful to assess EFC in low-q ~2 operation and high beta operation near â??real stability limitâ?? including rotation and kinetic stabilization effect.
We need careful gain setting before and after the estimated marginal time period considering the model prediction. The resetting of error field correction based on the actual feedback current are also made based on the model prediction | ||
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| Title | 175: Beta collapse avoidance in SSI discharges by de-synchronizing toroidal harmonics by using feedback | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Go Matsunaga, Ted strait | ITPA Joint Experiment : | No |
| Description: | The OFM-ELM pacing study (2012IAEA) revealed how the coupling between toroidal harmonics in low n mode takes place.
(1) the poloidal wavelength of n=1,2,(3) is comparable at outboard side. This is why they couple well. (2) The toroidal phase of each harmonics are synchronized producing the total amplitude maximum at mid-plane and causes ELM crash, creating massive impurity inflow and betan collapse. Toroidal harmonics coupling has been said many years. However, until now, we have not visualized how the toroidal harmonics couple in two-dimensional mode structure. It is to be said that the fast growth of higher harmonic component is the major part leading to disruption, not the fundamental n=1. Although this was studied with OFM-ELM pacing, but, relation of fundamental /harmonics eigenmode patterns on surface (real part) should remain similar in any high-beta modes (because of ballooning effect), imaginary part varies depending upon the growth term. This behavior may be universal |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Now, we would like to utilize this knowledge in disruption / beta-collapse avoidance. It is better to control toroidal harmonics n=1, 2, 3 together. However, the present DIII-D RWM control system is only capable to inlcude n=1 and n=3. Thus, the initial experiment is a proof of principle trial with n=1 and n=3.
First, in order to test the phase relationsignificance, we need to slow down the mode frequency by reducing the plasma rotation and handling the system within reasonable time resolution. The SSI target plasma with additional 210 beams, we expect the mode frequency can be lowered considerably. We will compare the n=1 and 3 feedback in-phase and out-of-phase between n=1 and n=3 to observe the ratio evolution in toroidal harmonics n=1 and 3. Phase of n=3 harmonics should be desynchronized relative to n=1 toroidal phase When new magnetic sensors are available, it should be easier to look at n=1,2,3. | ||
| Background: | |||
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| Title | 176: Beta collapse avoidance in SSI discharges by de-synchronizing toroidal harmonics by using feedback | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Go Matsunaga, Ted strait | ITPA Joint Experiment : | No |
| Description: | The OFM-ELM pacing study (2012IAEA) revealed how the coupling between toroidal harmonics in low n mode takes place.
(1) the poloidal wavelength of n=1,2,(3) is comparable at outboard side. This is why they couple well. (2) The toroidal phase of each harmonics are synchronized producing the total amplitude maximum at mid-plane and causes ELM crash, creating massive impurity inflow and betan collapse. Toroidal harmonics coupling has been said many years. However, until now, we have not visualized how the toroidal harmonics couple in two-dimensional mode structure. It is to be said that the fast growth of higher harmonic component is the major part leading to disruption, not the fundamental n=1. Although this was studied with OFM-ELM pacing, but, relation of fundamental /harmonics eigenmode patterns on surface (real part) should remain similar in any high-beta modes (because of ballooning effect), imaginary part varies depending upon the growth term. This behavior may be universal |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Now, we would like to utilize this knowledge in disruption / beta-collapse avoidance. It is better to control toroidal harmonics n=1, 2, 3 together. However, the present DIII-D RWM control system is only capable to inlcude n=1 and n=3. Thus, the initial experiment is a proof of principle trial with n=1 and n=3.
First, in order to test the phase relationsignificance, we need to slow down the mode frequency by reducing the plasma rotation and handling the system within reasonable time resolution. The SSI target plasma with additional 210 beams, we expect the mode frequency can be lowered considerably. We will compare the n=1 and 3 feedback in-phase and out-of-phase between n=1 and n=3 to observe the ratio evolution in toroidal harmonics n=1 and 3. Phase of n=3 harmonics should be desynchronized relative to n=1 toroidal phase When new magnetic sensors are available, it should be easier to look at n=1,2,3. | ||
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| Title | 177: Possibility of â??Off-axis-Fishbone Mode and ELM Pacingâ?? due to Direct Energetic Particle Coupling | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Go Matsunaga, M. Okabayashi, J. Ferron, C. Holcomb, T. Luce, D. Pace, F. Turco | ITPA Joint Experiment : | No |
| Description: | One important question of â??Off-axis-Fishbone Mode(OFM)-ELM pacingâ?? is whether the ELM is triggered directly by Energetic Particle(EP) or indirectly through pressure or current profile change caused by enhanced EP transport. In this proposal, we plan to look into a possible hypothesis of direct EP coupling to ELM. This hypothesis is based on the EP loss observation near the edge by BES and onset of ELM within relatively-short time period after higher toroidal harmonic appearance in OFM waveform.
Firstly, as pre-requisite, fundamental study of enhanced EP loss mechanism by OFM will be documented by Full EP diagnostics and following series of EP orbit calculation. Secondly, the role of EP is examined by enhancing EP component near the edge as much as possible, simultaneously reducing OFM activity. EP amount near edge will be controlled by various combinations of Off-axis NBI parameters. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To observe the role of EP near edge, we will utilize configurations more stable to OFM. Here, as pre-requisite, we carry out systematic numerical study of MHD stability analogous to classical Fishbone activity reduction. Our initial trial is higher triangularity with various q_profile flatness. Then, off-axis-NBI parametric scan injection-angle / energy will be included. One uncertainty will remain how to estimate the divertor functioning compared with the present configuration.
In the experiment, first step of fundamental study of enhanced EP loss by OFM will be documented by full EP diagnostics and series of EP orbit calculation. This operation requires the normal Bt operation for full use of EP diagnostics. Second step of the role of EP component near the edge is investigated using a configuration more stable to internal modes based on numerical studies discussed above. the EP amount near edge will be controlled by various combinations of Off-axis NBI parameters. In addition, Go Matsunaga is considering to carry out the ICH edge heating in AUG as independent assessment of EP near the edge contribution. | ||
| Background: | Last several years, the understanding of global MHD behavior in relation to the overall plasma performance has been greatly advanced experimentally and theoretically.
However, recent SSI experiments suggest that EPs play a significant role in such way that OFM events induce global MHD modes, which impact critically on the plasma performance. More importantly, the relation of EP-driven OFM to the RWM / ELM is identical in DIII-D and JT60U devices even with completely-different NBI arrangement. Even waveform develops in time non-linearly in an identical manner in these two devices. This suggests important fundamental process takes place in the relation between EP, EP-driven OFM and global MHD modes. EP pressure likely remains as major plasma content even up to ignition (like 1 MeV in ITER NBI). Thus, it is important to investigate how the EP couples to global mode. This proposal is planned to look into possible mechanism of recent unique observed â??OFM-ELM pacingâ??. The details are presented at IAEA 2012. There are two ways to hypothesize the OFM coupled to ELM. One thought is that the OFM causes the PTh or JTh near the edge, which increases the local gradient and induce. In this proposal, we plan to look into a hypothesis of direct EP coupling to ELM, as we discussed above. | ||
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| Title | 178: Beta limit and bootstrap current fraction in ITER steady-state scenario discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| 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 and for optimization of steady-state scenario discharges in DIII-D. The portion of this work would be relevant to the IOS ITPA group. 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. Test both NCS and weak shear q profiles in order to determine how to achieve fNI = 1 at moderate betaN. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce the fNI = 1 discharges produced in 2008 and vary q_min, the central shear, 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. | ||
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| Title | 179: Snowflake in AT Plasma | ||
| Name: | Allen allens@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | C. Holcomb, V. Soukhanovskii | ITPA Joint Experiment : | No |
| Description: | As the snowflake divertor option is well suited to high triangularity plasmas, it seems like a good idea to try to develop a scenario that has AT plasma performance at high triangularity with the snowflake divertor. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Ideally, the snowflake at high triangularity would be developed in a single null configuration - or at least a drsep of several centimeters. Then, as confidence is obtained, an AT scenario closer to the current standard - i.e. almost double null with small drsep could be attempted. The snowflake divertor configuration should allow spreading of the heat flux by the geometric magnetic flux expansion. | ||
| Background: | Previous snowflake operation has shown effective reduction of the divertor heat flux - the snowflake is a compact divertor configuration. | ||
| Resource Requirements: | As the experiments move closer to the current "standard" AT configurations, i.e. with small drsep, Plasma Control of the separate x-points would be beneficial. | ||
| Diagnostic Requirements: | New IR periscope to view the divertor under the shelf. Possible new IR periscope viewing the divertor. | ||
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| Title | 180: Establish the incremental confinement of EC power in high betaN steady-state scenario discharges | ||
| Name: | Ferron ferron@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Produce a series of discharges dedicated to determining whether EC power can be used effectively to increase the stored energy. Do a several point scan of betaN and density and EC deposition location. Compare the incremental confinement with EC power and neutral beam power, both on and off-axis. Determine whether there are steady-state scenario relevant parameter regimes where density increases when EC power is applied in addition to regimes where density decreases when EC power is applied. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | See the description paragraph. | ||
| Background: | It is not clear that EC power, as it is presently used in steady-state scenario discharges, is effective at heating. In fact, there is evidence that when off-axis EC power is injected, additional neutral beam power is required in order to maintain betaN. The presently available neutral beam power at DIII-D is marginally low for reliably obtaining fNI = 1, so additional heating sources are required. A substantial upgrade in EC power is planned for DIII-D and this is a potential source of the power necessary to achieve high betaN, but it is necessary to understand how this power can best be applied in steady-state scenarios. We also need to know how to model H factors as a function of EC power. | ||
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| Title | 181: Heat pulse propagation in a perturbed magnetic topology | ||
| Name: | Evans evans@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | K. Ida (NIFS), Y. Suzuki (NIFS), S. Ohdachi (NIFS), S.Inagai (Kyushu University), E. Unterberg (ORNL), M. Shafer (ORNL), J. Harris (ORNL), O. Schmitz (FZ-Juelich), M. Austin (Univ. Texas), and ITPA PEP-19 Group Members | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to study changes in the transport and the structure of the equilibrium magnetic field when non-axisymmetric perturbation fields are applied to Ohmic, L-mode and H-mode plasmas. The ECH system will be modulated to increase Te near the center of the discharge in order to generate heat pulses that propagate radially outward to the boundary region. Data will be acquired on the propagation of these heat pulses from the core to the edge using the ECE system. Changes in the characteristics of the heat pulse propagation, as the magnetic topology and plasma conditions are varied, will allow us to understand how the plasma alters the applied vacuum magnetic perturbations. Data from this experiment will be compared with vacuum field calculations and results from similar experiments done in LHD and TEXTOR over the last few years. This techniques has the potential to be able to establish the extent to which the plasma response to externally applied RMP fields screens or amplifies resonant components and gives us a direct measurement of the energy transport in RMP H-modes and can be compared to ELMing H-modes. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The first step in this experiment is to use inner wall limited (IWL) Ohmic and NBI heated plasmas with an ECH pulse train modulated at 25 Hz between 0 and 2 MW and a large m/n = 3/1 island positioned close to the last closed flux surface. We need to develop discharge conditions that minimize the size and frequency of sawteeth since heat pulses generated by these instabilities contaminate the ECH pulses and make it difficult to analyze the ECE signals. The second step is to go to a diverted H-mode plasma with RMP ELM suppression and repeat the ECH pulses used in the previous step. This step will also require some discharge development to minimize sawteeth and to optimize the modulation frequency of the ECH since we want the the on-time of the ECH to be long enough to reach a saturated Te at the deposition radius and the off-time to be at least as long as the energy confinement time. It may be necessary to adjust the ECH on-off timing and modulation depth to match the discharge conditions. The toroidal field will be set to give the best possible ECE coverage of the pedestal and the ECH will aimed to heat near the rho = 0.2 surface while minimizing the current drive. | ||
| Background: | ECH pulses have been successfully used in LHD and TEXTOR to study changes in the magnetic topology, i.e., nested flux surfaces, small isolated magnetic islands, mixed islands and stochastic layers and regions of strong stochasticity, due to intrinsic resonant magnetic fields and applied RMP fields. These studies have been done primarily in helical (heliotron) and limiter (tokamak) plasmas under Ohmic and L-mode type conditions. During the 2011 DIII-D run period an initial set of data was obtained using IWL and diverted L-mode and H-mode plasmas. Two IWL discharges were obtained with and without RMP fields that had relatively small sawteeth (e.g., 146517). Several ISS ELM suppressed discharges were also obtained (e.g., 146797-146800) but these had significant sawtooth activity and the toroidal field was not well optimized for good ECE coverage of the pedestal. Nevertheless, with a careful analysis of the data several interesting and potentially important effects were observes. Based on what was learned from the 2011 data and the operational experience gained from these discharges we should be able to achieve better plasma conditions (i.e., with reduced sawteeth) and acquire better quality data that will answer several key physics questions about the plasma response to the RMP field in Ohmic, L-mode and RMP H-modes. This experiment is an important part of the ITPA PEP-19 work plan which is focused on understanding how 3D perturbation fields affect transport and confinement. | ||
| Resource Requirements: | Detailed resource, diagnostic, analysis and other requirements are listed in D3DMP No.: 2011-01-05.http://fusion.gat.com/pubs-int/MiniP/review/2011-01-05.pdf | ||
| Diagnostic Requirements: | ECE correlation-ECC BES reflectometer | ||
| Analysis Requirements: | Scheduling of this experiment needs to take into consideration the travel arrangements of international participants. | ||
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| Title | 182: Systematic test of paleoclassical transport | ||
| Name: | Smith smithsp@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | Jim Callen | ITPA Joint Experiment : | No |
| Description: | The goal is to run a plasma where paleoclassical heat transport dominates. Most likely this is an Ohmic discharge. Diagnose the plasma with DIII-D's modern set of profile and fluctuation diagnostics. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make ohmic plasmas at various toroidal fields and densities. If there is an adiabatic response to increasing density (p=nT => T=p/n) such that the temperature decreases, then the density should be pushed as high as possible. If the Greenwald limit is reached, then ECH power could be added to possibly overcome the Greenwald limit and reach for lower temperatures. | ||
| Background: | There is a predicted critical temperature, which depends on magnetic field, below which paleoclassical transport should dominate over gyrobohm transport. A previous study of various toroidal devices showed that paleoclassical transport may set the lower limit on transport [1]. A study of the pedestal database also indicates this [2].
[1] Callen, et al. 2007 [2] Smith, et al. 2012 | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | Standard. | ||
| Analysis Requirements: | Power balance analysis. | ||
| Other Requirements: | -- | ||
| Title | 183: Measurement of multiple mechanisms in the low-density limit via ellipticity in the compass scan | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This experiment proposes to perform a high-resolution low-beta compass scan, consisting of 6-8 points, in order to fully constrain an elliptical curve to the locked mode onset data. This elliptical curve will provide a unique test of theoretical models, requiring at least two physical mechanisms to fully define the locked mode onset. Fitting the elliptical curve will thus yield data to weight the relative importance of the two mechanisms. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment will be conducted on the standard low-q, low-beta target plasma used for DIII-D error field studies (for example 149499). An n=1 compass scan will be performed by ramping the applied field at a fixed toroidal phase until a locked mode is found. However, instead of the standard 3-4 phases, somewhere between 6-8 phases will be tried. This will fully constrain the elliptical model, which has 5 free parameters. The largest apparent ellipticity in existing scans occurred with the C-coil, thus it is proposed to use the C-coil for this experiment.
If ellipticity is verified, a modeling effort will be undertaken to understand this new result. A successful model of the dataset will yield an empirical weighting between competing physical mechanisms setting the low-density locked mode limit. This experiment could be accomplished in 6-8 good shots. | ||
| Background: | The standard technique for low-beta empirical error field correction on DIII-D is the 'compass scan', in which a correcting field is applied at various toroidal phases in order to induce a locked mode. The premise behind this approach is that locked mode onset is proportional to the magnitude of the applied field vector summed with the intrinsic machine error. Thus, the coil currents at which a mode appears can be fit (to zeroth order) to an offset circle in the compass scan phasor space (current amplitude and toroidal phase). DIII-D has performed 22 such 'compass scans' since 2004.
Despite the common interpretation of these scans, recent analysis indicates that nearly all compass scans systematically deviate from this picture. That is, that a measurable amount of -ellipticity- is present in fits to the locked mode onset. This ellipticity is not easily explained; as all known factors for locked mode onset are circular on the compass scan space, such as resonant and non-resonant (NTV) breaking. Thus, the ellipticity of the compass scan poses a unique challenge for modeling. | ||
| Resource Requirements: | No beams will be required for this experiment. A standard patch panel (3 SPAs on C-coil pairs) will be used. This experiment lends itself well to a short time slot (evening or startup, for example). | ||
| Diagnostic Requirements: | Thomson, magnetics, SXR, CO2 and 288 GHz interferometers. | ||
| Analysis Requirements: | All data analysis tools are already in existence. Modeling of the NTV for the equilibrium is already in progress in the context of the 2011 proxy field experiment (which shares the same target plasma) | ||
| Other Requirements: | |||
| Title | 184: Comparision DIII-D and AUG high collisionality ELM response to 3D magnetic perturbations | ||
| Name: | Evans evans@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | W. Suttrop, et al., and ITPA PEP-23+25 Members | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to reproduce ELM mitigation/suppression previously obtained in DIII-D with odd parity n=3 RMP fields and n=2 odd/even parity RMP fields in AUG at high density. A key question for understanding the interaction of RMP fileds with H-mode plasmas is whether the plasma response is dominated by collisionality or density when operating at high Greenwald fraction. This experiment will provide new information on the plasma response to n=3 field since we now have new diagnostic and operational capabilities compared to the original high Greenwald ELM suppression experiments done in DIII-D. In addition, comparisons with AUG results at high Greenwald fraction can provide important insights into the role of different RMP spetra on EML suppression physics and operational space. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | In this experiment we will start by reproducing the shape and operating parameter used in DIII-D high Greenwald fraction ELM suppression discharge 115467 with n=3 odd parity RMP fields. Once ELM suppression is obtained we will carry out a density scan to see if we observe a density threshold similar to that seen in AUG with n=2 RMP fields. Next we will increase the I-coil current from 4.0 kA (used in 115467) to 6.3 kA to see if higher I-coil currents produce effects similar to those seen in low collisionality/density ELM suppression discharges. The final step involved evolving the discharge shape over several shot to better match the shape used in AUG ELM mitigation discharges. | ||
| Background: | Previous experiments done on both DIII-D and AUG have produce ELM suppression/mitigation at high Greenwald fraction. While some of the characteristics observed in these experiments are similar (e.g., little of no change in the pedestal profiles) others are significantly different (e.g., effects on toroidla rotation and the absence of a q95 resonance window in AUG). Understanding the mechanisms involved in these similarities and differences is important for determining wheather ElM suppression at high Greenwald fraction (and low collisionality) is viable in ITER. This experiment will contribute to work being done in the ITPA PEP-23 working group and is an urgent ITER issue. | ||
| Resource Requirements: | Detailed resource, diagnostic, analysis and other requirements are listed in D3DMP No.: 2011-01-05.http://fusion.gat.com/pubs-int/MiniP/review/2011-01-05.pdf | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 185: Measure Plasma Response to RMPs | ||
| Name: | Chen chenxi@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Bill Heidbrink, Gerrit Kramer, Raffi Nazikian, Mike Van Zeeland, Jeremy Hanson and the EP group | ITPA Joint Experiment : | No |
| Description: | Use fast ion transport/loss to probe the plasma response to RMPs in three regimes: strong kink response, strong plasma screening and weak plasma response. Quantify fast ion loss (and transport) with the magnitude of kink response; Search for islands (location, size, strength); test/validate plasma response model/code,e.g. use the case where some islands are not screened when Vperp_e ~0 as screening model predicts. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using plasma current and beam settings favorable for fast ion first orbit loss detection:
* Apply n=1 rotating RMP: 1. In strong kink response regime, quantify loss/FIDA data with plasma kink response: - Beta-scan, collect data with different magnitude of kink response - scan the relative phase between upper and lower coils, collect data at all phase related to the mode - change from even parity to odd parity, collect data with different mode coupling (resonant vs. non-resonant) - vary I-coil currents 2. In strong plasma screening regime, search for islands, test screening model/code: - at a beta value when plasma kink is strong in co-rotating from #1, flip plasma rotation direction, i.e. ctr-rotating now, collect data (not screened vs. screened) - very slowly? rotating RMPs, collect data(the X- and O- point travel by) 3. Collect data in the regime where neither kink response nor screening is important for comparison with #1 and #2 4. If ELM suppression is achieved, slightly vary q95 near the marginal value, collect loss for case with magnetic island sitting on top of pedestal as predict by theory (ELM suppression) and without. * Apply n=3 RMP: - Scan plasma rotation, high co-rotating to high ctr-rotating, collect loss (without screening and with screening) - With ctr-rotating, vary plasma screening by + varying the toroidal phase of applied n=3 RMP + varying q95 - Flip polarity of RMP coils in one shot - Vary I-coils current * Use PCS, no RMPs: - Reproduce the plasma "movement"(variation in gapout, axis, x-point) caused by RMPs, collect loss/FIDA (only due to plasma "movement") | ||
| Background: | First orbit loss of fast ions is very sensitive to the plasma fluctuating field. Alfvn eigenmodes induced first orbit loss of beam ions have been observed where the mode amplitude B_tw/B, is on the order of 10E-3. The perturbations caused by RMPs have similar amplitude. Therefore, fast ion loss/transport measurement can contribute to the study of plasma response, field penetration/plasma screening.
* Quantify the fast ion loss/transport due to plasma kink response How much does plasma kink response affect fast ions during RMP? Will we find a connection between these two and use fast ion measurement to study the kink response? * Search for islands: We can use fast ions first orbit loss to search for magnetic islands due to RMPs. The particle orbits are different and the losses might be different: - with and without magnetic islands - when the ion travels through the x-point and the o-point of the island. * Test/Validate/Contribute to the development of plasma screening model --- "at certain plasma parameters and due to the non-linear evolution of the redial electric field produced by RMPs, the ExB rotation can be compensated by electron diamagnetic rotation locally. In this case, RMPs can penetrate and form magnetic islands." M. Becoulet et al. NF 52(2012) - Does the simulated fast ion loss with vacuum field or with plasma screening match the experimental fast loss signal better? - How does the fast ion loss vary with RMP strength/q95/rotation as the plasma screening changes with those parameters predicted by the modeling/theory? - To validate the case of island not screened (when the Vperp_e ~ 0) | ||
| Resource Requirements: | normal BT and Ip direction, n=1 rotating RMP, n=3 RMP, 30L, 150L&R, 330L, 210L&R | ||
| Diagnostic Requirements: | FILD1&2, CER, BES, ECEI, UCSD fast camera, FIDA, LIBEAM, MIR (if available), Thomson Scattering,... | ||
| Analysis Requirements: | Kinetic EFIT, TRANSP, NOVA-K, SPIRAL, M3D-c1, IPEC,... | ||
| Other Requirements: | -- | ||
| Title | 186: Optimal Mixing of I & C coils for n=1 EFC | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | E. Strait, M. Lanctot, R. La Haye, J. Hanson | ITPA Joint Experiment : | No |
| Description: | This experiment proposes to empirically obtain the optimal poloidal spectrum for n=1 error field control (EFC) via ??optimal mixing?? of I and C coil currents. Optimal mixing means proportionally trading-off currents in each coilset based on the optimum current of each coilset alone (see background). <br> <br>In the language of kink-resonant (IPEC single-dominant mode) vs. non-resonant (NTV producing) fields, optimal mixing means nulling the single-dominant mode field while systematically varying the other non-resonant harmonics. <br> <br>This proposal thus allows the role of second order effects (NTV, second-dominant mode) on the low-density limit to be directly measured (by holding the first-order effect constant), and also holds the promise of developing an improved n=1 error field control algorithms for general DIII-D use. The optimal mixing levels of I & C coils can also be compared to NTV theory from various codes. <br> <br>It is also likely that this experiment will produce measurable amounts of non-thermal electrons due to foreseen low-density operation. Piggyback experiments with the runaway electron group can be proposed to diagnose or control this population on the back-end of the discharge. Finally, this experiment provides an opportunity for the new 3-D magnetics diagnostic to measure the detailed structure of the n=1 locked mode found at low density. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Density rampdown discharges will be performed in the low-q, Ohmic plasma commonly used in EFC experiments. The currents in the I and C coil will be proportionally varied, such that a reduction in I-coil current is compensated by an increase in C-coil current, while the phase of both is held constant. The ??over-driven?? case will also be tried, where one coilset is compensated by the other coilset with a 180-degree phase difference from optimum levels. The minimum density prior to locked mode detection will be measured in each discharge and an optimum found along this scan. Additionally, a scan of increasing I & C coil currents together can further constrain the optimum field structure. | ||
| Background: | Previous compass scans have yielded an optimum current and phase for n=1 EFC for the I & C coils individually. Strikingly, it is found that the I & C coils are each correcting the -same- single-dominant mode of the intrinsic error field as calculated by IPEC, and thus both point in the same toroidal phase. For this reason, proportionally trading off the amplitude of I & C coil currents will maintain the same single-dominant mode field. Adding multiple coilsets thus allows the single dominant mode field to be held at zero while other parts of the spectrum can be optimized. | ||
| Resource Requirements: | Ohmic plasmas only. No beams, no ECH, etc. 4 SPAs required. | ||
| Diagnostic Requirements: | Thomson, 3-D magnetics, SXR, CO2 and 288 GHz interferometers | ||
| Analysis Requirements: | All presently in existence or underway. | ||
| Other Requirements: | -- | ||
| Title | 187: H-mode 3-D field optimization with purely kink-resonant and non-resonant n=1 fields | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | M.J. Lanctot, E.J. Strait, R.J. La Haye, J.M. Hanson, J. King, J-K. Park, N. Logan, N. Ferraro | ITPA Joint Experiment : | No |
| Description: | This experiment proposes to extend the new technique of â??H-mode compass scansâ?? to 3-D field configurations (n=1) that are as purely kink-resonant (kink) and non-resonant (anti-kink) as possible using DIII-Dâ??s I & C coils together. The kink and anti-kink fields will be designed to have an IPEC â??overlap integralâ?? to the dominant plasma mode of nearly 1 (full overlap) and 0 (no overlap), respectively. The H-mode compass scan would find the amplitude and toroidal phase of each field required to optimize various error field control metrics, such as the total angular momentum, plasma response, and plasma-wall Maxwell stress.
The motivation for these scans is several-fold. Firstly, the expected plasma response for the two fields should correspond to the first and second dominant modes, and thus provide a novel measurement for the new 3-D magnetics diagnostic. Second, each scan will add to a database which will well-constrain theoretical models of optimum error field correction in H-mode plasmas. Third, if the fields are truly â??orthogonalâ??, a combined kink-resonant and non-resonant correction algorithm can be devised simply by superimposing the optimal currents from each scan. Fourth, the amount of rotation braking from the kink and anti-kink fields can be compared to theoretical braking predictions for each field as well as Maxwell stress measurements at the wall. Fifth, this experiment would demonstrate that it is experimentally possible to optimize (reduce) NTV braking by adding an external field. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The 'H-mode compass scan' technique developed during the 2011 TBM campaign will be applied to these new 3-D field configurations. The plasma target will be identical to that of the 2011 TBM experiment so that the results from the two may be compared. The reference shot is 147135, with beta_N = 1.8, q95=4.1.
The technique consists of ramping the coil amplitude at 4-5 different toroidal phases. During the ramps the total angular momentum, plasma response, and plasma-wall Maxwell stress will be measured. A fit to these parameters will determine the optimum amplitude and phase for a given metric. To test the combined algorithm, the degree of rotation braking with each individual optimum current will be compared to the total optimum current, as long as the required currents are within the limits of existing power supplies. The experimentally derived optima will then be compared to calculations of both resonant and NTV braking from IPEC, M3D-C1, or other plasma response codes. The anti-kink field should produce dominantly NTV braking, while the kink field should be dominantly resonant. This experiment could be accomplished in 10-11 good shots. | ||
| Background: | Optimal error field correction of n=1 fields is now thought to be a competition between two potentially exclusive goals - the minimization of both resonant braking (which is dominantly induced by kink-resonant edge fields) and the global NTV braking due to non-resonant (and kink-resonant) fields. Experiments up to now have exclusively used 3D field spectra that contain elements of both, that is, their poloidal spectrum is both kink-resonant and non-resonant. This experiment will attempt to look directly at the relative importance of kink-resonant and non-resonant fields at braking the plasma in the high-beta scenario where this is most important, as well as and develop control algorithms which minimize both as much as possible. | ||
| Resource Requirements: | beta_N control and injected torque control are critical. All 3D power supplies (4 SPAs, 2Cs) will need to be available. A patch panel change will likely be necessary during the experiment. | ||
| Diagnostic Requirements: | SXR, CER, MSE, Thomson, 3D magnetics | ||
| Analysis Requirements: | TRANSP runs for all discharges, normal suite of 3D field analysis tools. | ||
| Other Requirements: | |||
| Title | 188: Extension of proxy error field experiments to H-mode plasmas | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R.J. Buttery, M.J. Lanctot, E.J. Strait, R.J. La Haye, J.M. Hanson, J. King, J-K. Park, N. Logan | ITPA Joint Experiment : | No |
| Description: | This experiment proposes to use the newly developed technique of the â??high-beta compass scanâ?? to find the optimal correction of a known proxy field under H-mode conditions. This experiment is a natural extension of the 2011 proxy experiment carried out in an Ohmic plasma [Buttery, Phys. Plasmas 2012].
This study is also complementary to 2011 TBM experiments (and 2013 proposals), with the important exception that the applied error field would be purely n=1, thus clarifying whether the cause of the incomplete rotation recovery in TBM experiments arose from n>1 fields or uncorrected n=1 components (see background). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The 'high-beta compass scan' technique performed twice in 2011 as part of the TBM campaign will be applied to the proxy field configuration. That is, the C-coils will apply a known (proxy) field while the I-coils will try to correct it. The discharge scenario will be identical to that of the 2011 TBM experiment so that the results from the two may be compared. The reference shot is 147135, with Beta_N = 1.8, q95=4.1.
The technique consists of applying the proxy field (C-coil), and then ramping the correcting field (I-coil) at various toroidal phases. It is further proposed to begin the ramp from the optimum levels found in the 2011 low-beta experiment. If the proxy error field induces tearing modes or other unwanted behavior, its amplitude will be reduced to manageable levels. During the ramp, metrics such as the total angular momentum, plasma response, and plasma-wall Maxwell stress will be measured. Fits to these parameters will be made and an optimum for each deduced. These measurements will be much enhanced by the new 3D magnetics capability. The optimum will then be applied in a dedicated discharge to observe the levels of rotation recovery and compare this to the values found with the TBM experiment. Comparison to theory for the various metrics will also be undertaken to benchmark our understanding of the relevant physics (NTV, plasma response, etc). This experiment could be accomplished in 5-6 good shots. | ||
| Background: | Previous work with the TBM mock-up in 2011 dramatically illustrated that optimum correction of the n=1 TBM error field only allowed a 25% recovery of the rotation from pre-TBM levels. The TBM spectrum, however, is complex and contains several different toroidal harmonics (n), leading to uncertainty as to whether the rotation degradation is caused by higher n modes or by incomplete n=1 correction. Knowing which path (n>1 or best n=1) leads to the most complete rotation recovery is a critical issue for devising error field control algorithms going forward. | ||
| Resource Requirements: | BetaN control and injected torque control are critical. Six 3D power supplies will need to be available (4 SPAs and 2 Câ??s). | ||
| Diagnostic Requirements: | ECE, CER, MSE, SXR, Thomson, 3D magnetics. | ||
| Analysis Requirements: | TRANSP runs for all discharges, normal suite of 3D field analysis tools. | ||
| Other Requirements: | |||
| Title | 189: Measurement of optimal n=2 phasing for RMP & ELM suppression without q95 windows | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | M. Lanctot | ITPA Joint Experiment : | No |
| Description: | This experiment will use n=2 I-coil phasing as the dynamic variable during the shot to measure the optimum phasing for a given set of plasma parameters. Recent results suggest that the optimum phasing may not be where the vacuum-field pitch-resonant harmonics (m = nq) are maximized, but instead at the kink-resonant maximum (m ~ 2nq). We propose a direct technique to test this hypothesis by dynamically scanning the phasing during the shot (at constant plasma parameters) and looking for ELM suppression windows --in phasing--. Due to the symmetry of the 3-D field spectrum with phasing, the optimum phasing should be in the center of the phasing suppression window.
Furthermore, it is possible and proposed to extend this technique to discover the optimum phasing as a function of relevant plasma parameters, likely the most important being q95 (but possibly also beta or rotation). Other key metrics, such as density pump-out, can be measured as a function of phasing in the same discharges. A demonstration of improved understanding is also proposed. By measuring the optimal phasing at a 2-3 values of q95, an algorithm can be developed which tracks q95 and maintains the optimal I-coil n=2 phasing. This algorithm can then be applied to a discharge with significant variations in q95, with suppression expected throughout - in contrast to the usual narrow q95 windows observed. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To develop an algorithm for q95 following, the required phasings at a few q95's must first be determined. To do this, the q95 will be held constant while the I-coil phasing is varied. The phasing could be either stepped or ramped. The sweep will center on the optimum value found in 2011 n=2 experiments (shot 145592). An ELM suppression window in phasing is expected, with the center point taken to be the optimum for that q95. This will be repeated at 1-2 more well-spaced values of q95 in order to build up an algorithm between q95 and phasing, which is expected to be linear. Greater diagnostic sensitivity can be achieved by modulating the outer gap during the experiment.
Once this mapping is determined, discharges for which q95 is ramped will be run. The q95 evolution will be known a-priori by using a well-known target discharge. As q95 is evolving, the n=2 I-coil phasing will be fed-forward based on the mapping. A successful outcome would be no ELMs found within the q95 range of the algorithm, and possibly beyond. This experiment will also give useful piggy-back opportunities to the new 3-D magnetics diagnostic. The optimal (and sub-optimal) n=2 field can also be rotated to enable synchronous detection and thus better noise rejection. This experiment could be accomplished in 9 good shots (4 per q95, 1 demonstration discharge). | ||
| Background: | n=2 RMP fields provide the capability (unlike n=3) of tailoring the poloidal structure of the 3-D field used for ELM suppression. This capability will be used to determine the optimum phasing at several q95s. In 2011 it was found that the q95 suppression window was different for different I-coil phasings, a favorable result for this technique to yield a successful measurement. | ||
| Resource Requirements: | This experiment will require 6 power supplies. Thus, C1, C2 and all SPAs must be operational. 1 SPA is dedicated to n=1 error field control with the C-coils. | ||
| Diagnostic Requirements: | Thomson, SXR, ECE, 3D magnetics, MSE, CER | ||
| Analysis Requirements: | M3D-C1 of n=2 plasma response and ELITE of the resultant pedestal profiles. | ||
| Other Requirements: | |||
| Title | 190: Measurement of n=3 error field effects in RMP phase-flip experiments | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | M. Lanctot | ITPA Joint Experiment : | No |
| Description: | This experiment proposes to apply the techniques of error field control (EFC) to understand and verify the offsets used in n=3 'phase flip' experiments commonly used to diagnose RMP ELM-suppression discharges. This will provide a physics basis for the continued use of this offset benchmarked against global EFC metrics, as well as providing useful data for the new 3-D magnetic sensors. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | First a high-beta 'compass scan 1-D analog' will be performed. Using the same target plasma as used for previous n=3 phase flip measurements, the n=3 field will be slowly ramped (or stepped) at +ve polarity, and in a subsequent shot at -ve polarity. Throughout, the NBI torque and beta will be held constant, and the global momentum of the plasma calculated. Where the momentum is maximized will be compared with the offset used in previous phase flip experiments. The n=3 plasma response minimum can also be compared, or in fact any number of parameters along this scan.
Second, a low-beta scan will be performed in which the key metric will be the locked mode onset. The plasma target would likely be the Ohmic test plasma commonly used in error field experiments. The n=3 field will be ramped until a mode appears, first at +ve polarity then at -ve polarity. The mean value will be compared with the offset used in RMP-ELM phase flips. (Standard C-coil n=1 EFC will be enabled throughout these discharges.) As the n=3 even parity field has only 1 free parameter (amplitude), the number of shots required for this study is very modest (~4). Considering the other n=3 parity would add another ~4 shots. | ||
| Background: | A very useful diagnostic technique for RMP-ELM experiments is currently the 'phase-flip' experiments, in which the n=3 field is quickly flipped from +ve to -ve polarity. However, it has been empirically found that a ~ 0.5kA offset is required to symmetrize the core Thomson response at each polarity, presumably due to an intrinsic machine n=3 error. As the value of this offset materially affects the measurements extracted during phase flips, it is important to understand its origin and confidently apply it in future experiments.
This experiment will seek to gain this understanding by applying the tools developed in n=1 error field correction to the special case of the n=3 phase flip experiment. This experiment is also complementary to proposed studies to further vary the offset for phase flip RMP experiments. | ||
| Resource Requirements: | Beta_N and injected torque feedback will be critical to the high-beta phase of the experiment. Bipolar current requirements necessitate the use of SPAs. | ||
| Diagnostic Requirements: | Thomson, CER, ECE, MSE, SXR, new 3-D magnetics | ||
| Analysis Requirements: | EFC Analysis tools may need slight modifications to treat n=3 fields. | ||
| Other Requirements: | |||
| Title | 191: Critical electric field for runaway electron growth and decay under quiescent conditions | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | N. Eidietis, R. Granetz, E. Hollmann, A. Tronchin-James, J. Wesley | ITPA Joint Experiment : | Yes |
| Description: | This experiment aims to study the critical toroidal electric field (Ecrit) required for runaway electron (RE) growth under quiescent plasma conditions. A plasma target developed for extremely low-density operation (125010, n_e < 5E12 cm^-3) will be used. Quiescent RE plasmas are of fundamental interest because they are free from the complicating transient and impurity effects associated with disruption-generated RE beams. <br> <br>To distinguish the observed RE dynamics from the Dreicer mechanism, the plasma current or density will be systematically increased and decreased during the discharge to assess hysteresis in the RE growth and decay (see background). These measurements will ultimately assess and clarify the anomalous RE loss mechanisms needed to match the data. <br> <br>Furthermore, the development of a robust and quiescent plasma target with a significant RE population could lead to significant advances in the science of RE control and measurement. Piggyback studies will use this experiment to improve equilibrium reconstructions of the RE beam. Depending on the robustness and magnitude of the RE beam, experiments suggested in other RoF proposals could also be applied to this target discharge in the future. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will begin by reproducing 125010, a discharge developed for error field control (EFC) measurements that was recently discovered to contain a measurable RE population that was quiescent for over 3 seconds without locked modes. Excellent EFC is critical to the success of this experiment, as the error field ultimately limits low-density operation via locked modes.
When low-density RE growth is observed, the plasma current (proportional to the toroidal electric field) will be slowly increased and decreased, in order to measure the levels of RE beam growth and decay. The density will then be decreased and increased slowly and the same measurements made. | ||
| Background: | The required Ecrit for RE growth is the dominant parameter for gauging whether collisional suppression of the RE beam is likely during MGI. However, observations on C-mod and DIII-D suggest that Ecrit is anomalously large when compared to existing collisional theories [R. Granetz, ITPA 2012], both in quiescent and post-disruption plasmas.
Furthermore, the Ecrit required for the suppression of an already-avalanching RE beam can be measured by decreasing the toroidal electric field (Ephi) until RE decay is observed. It is expected that the Ecrit for RE decay will be smaller than for RE formation. Equivalently, the required density for RE formation is expected to be lower than that for RE decay at constant Ephi. | ||
| Resource Requirements: | Mostly Ohmic only operation. Beam blips for MSE, 1 gyrotron for possible RE seed generation. 4 SPAs operational. | ||
| Diagnostic Requirements: | ZNS or FPLASTIC in low gain mode. Thomson, MSE, SXR, ECE, UCSD fast camera, CO2 and 288 GHz interferometers | ||
| Analysis Requirements: | May enable development of new tools for RE equilibrium measurement | ||
| Other Requirements: | -- | ||
| Title | 192: Measurement of heat flux on the divertor using the embedded thermocouple array | ||
| Name: | Donovan ddonovan@utk.edu | Affiliation: | U of Tennessee, Knoxville |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | David Donovan, Dean Buchenauer, Jon Watkins, Richard Nygren, Dmitry Rudakov, Charlie Lasnier, Josh Whaley | ITPA Joint Experiment : | No |
| Description: | Accurate heat flux measurements at the divertor are essential for understanding the material requirements of the first wall and benchmarking models for predictive behavior of future experiments. This understanding can be enhanced with better understanding and fuller use of data from the array of 16 fast thermocouples (FTCs) that are embedded 0.8 cm below the surface of the ATJ graphite tiles on the divertor floor and 1 cm below the surface of the divertor shelf tiles. The embedded FTC array has been used for several years now and was most recently repaired during the 2012-2013 vent. 1D heat conduction calculations have been made to estimate the surface heat flux during shots in which the location and power of the strike point was held constant. The heat flux measured during these shots was compared to the heat flux calculated by the divertor Langmuir probe (LP) array to obtain estimates of the sheath power transmission factor (SPTF). In order to obtain more accurate measurements of the surface heat flux, more sophisticated 2D and energy dependent modeling are planned. The measurements from the LP and the embedded FTCs will also be compared with the heat flux measurements taken by the IR camera.<br><br>Additional experimental time is requested to study the heat flux measured by the probes in relation to the FTC array and IR camera. The ideal condition to study the response from the FTC array is with a constant location for the strike point on the divertor and a fixed energy. The stationary shot would ideally be followed by a shot with a small sweep of the outer strike point over the divertor shelf LP array to provide better conditions for LP analysis. More sophisticated modeling may permit us to unfold, the time and energy dependent functionality that could then be included to allow for movement of the strike point and variation in energy deposited. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The embedded FTC array has already been in place for several years and has proven to be a useful diagnostic tool. The new capabilities added will be the heat conduction modeling capabilities provided in collaboration with Sandia. The main point for this proposal is that the modeling will provide (a) analyses that will help optimize the shot parameters and (b) data that will aid in the interpretation of the data from the IR, probes, etc. (See additional information in Background.)
We will require a magnetic configuration to be optimized to provide a stationary outer strike point on the divertor shelf at R = 150 cm with the input power held constant. This is the ideal condition for analysis of the heat conduction through the tiles to the embedded FTCs. A sample shot would be 145671. The following shot will contain an OSP sweep from 148 cm to 153 cm taking 500 msec and returning back in 500 msec. This sweep over the divertor shelf LP array will offer the ideal conditions for LP analysis. The sample shot for this is 145670. (The sweep rate may be revised based on the thermal analyses.) Procedure 1.Position OSP at 150 cm with constant NBI heating and no ELM suppression coils active. Collect embedded FTC data during these shots for analysis. (Sample shot 145671, shot duration may be revised based on thermal analysis). Ensure operation of Langmuir probe array. Perform at 50 MW/m^2 parallel heat flux to limit the effect of the ELM transient heat flux. 2.Repeat stationary OSP with 75 MW/m^2 parallel heat flux. 3.Repeat stationary OSP with 100 MW/m^2 parallel heat flux. 4.Position strike point at 148 cm at 2.5 sec. Move strike point to 153 cm by 3.0 sec. Return to 148 cm by 3.5 sec. Ensure operation of Langmuir probes during these shots. (Sample shot 145670, sweep rate may be revised based on thermal analysis) Perform at 50 MW/m^2 parallel heat flux. 5.Repeat OSP sweep at 75 MW/m^2 parallel heat flux. 6.Repeat OSP sweep at 100 MW/m^2 parallel heat flux. | ||
| Background: | The embedded thermocouple array has been used for several years to measure temperature change in the floor and divertor tiles of DIII-D. During the 2012 experimental campaign, 1D heat conduction calculations were made to determine the surface heat flux above the TCs. These heat flux measurements were then compared to the results collected by the divertor Langmuir probe array. The LP array provides plasma density and electron temperature at the divertor. These measurements can be used with the sheath power transmission factor (SPTF) to determine the heat flux reaching the divertor surface. A comparison of the heat flux measured by the LP array and the embedded TC array found that the theoretically predicted SPTF value was reasonably accurate. Future experiments will ideally bring in the heat flux measurements taken by the IR camera in order to ensure that the most accurate measurements are made by all available diagnostics. The role of the supporting thermal analysis is summarized below.
The general objectives for the thermal modeling on DIII-D divertor tiles are as follows. 1)Use 3-D thermal models to understand when FTCs can provide useful data. Confirm when simplifications such as single-value (RT) materials properties in 1-D and 2-D models are adequate. Quantify 3-D effects of interest: strike point close to tile edge; long shots; surface maps corrected for varying emissivity. Quantify limits for use of FTC data, e.g., signal too small, complex shot history, etc. Produce data to guide planning of experiments, e.g., dwell times for strike point and sweep rates that give useful FTC data, and thermal stresses for high power shots. 2)Use FTC data to improve diagnostics for real time 2-D analysis of heat flux. Typically the 2-D code THEODOR gets thermal profiles extracted from IRTV measurements of surface temperature (middle of tile) and assumes a surface layer of uniform thickness and uniform material across the diverter. The following can affect these analyses: tile edges, misalignments, gaps, varying properties of redeposited layers, camera motion, reflections from hot spots, ... 3)Understand what it is important to include in thermal models, e.g., temperature-dependent material properties, radiative losses, surface layers, etc. The use of temperature-dependent properties, compared with room temperature thermal properties, has a noticeable effect even in shots of 3 MW/m2 for 4s. 4)Integrate this information with other diagnostics (probes, TCs, IR, spectroscopy) to support teams investigating power exhaust, e.g., sheath transmission factors and interpretation of IR data. The general point here is that 3-D thermal modeling can provide insights for (a) valid use of the FTC data, (b) confirmation of conditions when less sophisticated 1-D and 2-D models routinely used determine heat flux are most accurate, and (c) optimizing conditions such as strike point sweep rates or dwell times to best utilize FTC data. | ||
| Resource Requirements: | The embedded FTC array is available, along with instrumentation provided by the divertor Langmuir probe array. The experiment would require the IR camera. Run time of approximately 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 the fast line scan mode of the IR camera (to improve time resolution during the x-point sweeps). Divertor Langmuir probes IR camera (preferable in line scan mode) Fast thermocouple array Divertor spectroscopy Magnetics for EFIT determination of field angles Zeff C02 interferometer Thomson scattering Fast filterscope channels viewing the lower divertor Other useful diagnostics Tile current array Bolometers DiMES Calorimeter Probe | ||
| Analysis Requirements: | Analysis of the Langmuir probe signals and IR data would be critical. Magnetics (EFIT) evaluation of the strike point locations and geometry changes would also be needed. Thermal conduction modeling of the TC data will be provided by Sandia. | ||
| Other Requirements: | -- | ||
| Title | 193: Magnetic perturbation amplitude scan to investigate RMP-induced damping of Zonal Flows | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Turbulence & Transport | Presentation time: | Requested |
| Co-Author(s): | M. Leconte*, P. Diamond**, Z. Yan
*NFRI, Korea **UCSD/WCI |
ITPA Joint Experiment : | No |
| Description: | Determine how RMPs modify the amplitude of Zonal Flows, in an L-mode plasma. [Michael Leconte (NFRI) will present remotely] | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish L-mode plasma, and then do a basic characterization of turbulence and Zonal Flows/GAMs using BES system resonant and non-resonant magnetic perturbations are applied. A reference L-mode plasma is shot number [ ]. BES would be used to measure turbulence characteristics (150L steady and flat is desired beam configuration, 150R not on). Then ramp up and/or modulate the I-coil (MPs with resonant spectrum at the plasma edge q_95~3.5) current to a fixed value and keep a flat top of about 1-2 s. Do a basic characterization of turbulence and Zonal Flows/GAMs with BES. Repeat the operation for different values of the I-coil current. | ||
| Background: | Experiments on TEXTOR [Y. Xu et al NF 2011] showed a damping of GAM Zonal Flows
by RMPs. A theory [Leconte and Diamond PoP 2012] predicts a linear scaling of Zonal Flow damping with the square of the RMP amplitude (assumed to be the vaccum amplitude i.e. I-coil) The proposed experiment investigates the ZF damping and the scaling. | ||
| Resource Requirements: | I-coils, n=3 even/odd parity configuration | ||
| Diagnostic Requirements: | BES, DBS, electrostatic probes | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 194: Correlation of confinement and fluctuations with electron heating in ITER Baseline Scenario | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | E.J. Doyle, T.L. Rhodes, C.C. Petty, B. Tobias, B. Grierson, L. Schmitz, G.R. McKee, Z. Yan, L. Zeng, M. Porkolab | ITPA Joint Experiment : | No |
| Description: | Extend work on dependence on confinement on Te/Ti in conventional ELMing H-mode (C.C. Petty, et al., PRL, Vol. 83, 3661, 1999), QH-mode (Schmitz, et al., 2012), hybrid discharges (Petty, IAEA 2008) to ITER Baseline Scenario discharges. Determine dominant turbulence modes in electron-heating dominated ITER Baseline Scenario. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Basic approach is to create matched pairs of ITER Baseline Scenario discharges, with beam heating only compared with ECH plus NBI. Adjust applied torque in NBI-only case to match rotation speed obtained in ECH case. Study two cases: (1) one with 'normal' levels of toroidal rotation speed (all co beams plus EC compared with adding enough counter-NBI in all NB case to match rotation, and (2) [low rotation] About zero applied torque NBI-only case compared with mostly EC plus about zero applied torque. In each case obtain detailed fluctuation measurements to enable later comparison of observed fluctuations and predicted dominant turbulence (TEM vs ITG, etc.) modes. Measure particle confinement with He puffing technique in all cases. Consider making perturbative measurements with modulation of EC compared with modulation of part of NB power. | ||
| Background: | Previous work (see citations in Description) on DIII-D has shown that confinement is very sensitive to Te/Ti in several regimes. In 2012 experiments in which ECH and/or FW direct electron heating was added to ITER Baseline Scenario discharges in order to move towards the dominant electron heating regime, it was observed that incremental confinement tended to be poor with either form of direct electron heating. These results are likely the result of the sensitivity of confinement to the ratio of electron and ion temperatures, as previously studied in detail on DIII-D in other plasma regimes. We propose to investigate explicitly the correlation between confinement and the presumed underlying fluctuations and turbulence in the ITER Baseline Scenario to test this picture. The 2012 work by Schmitz, et al. in QH-mode clearly showed an ITG-to-ETG mode dominated transition upon entrance to the electron-heating dominated regime. Measured fluctuations in Te at mid-radius significantly increased, while the density fluctuations (BES) did not change very strongly. Since ITER discharges will, almost by definition, be electron-heating dominated, it is quite important to study the confinement properties of IBS discharges as we transition from the typical hot-ion types of discharges towards the electron-heating dominated case that is of most interest. | ||
| Resource Requirements: | 1-2 day experiment. Needs 4-6 NBI sources (no off-axis beams needed, but will need counter-NBI along with mostly co-), all available gyrotrons (at least 6). Helium puffing in some discharges. ITER Baseline Scenario discharges. | ||
| Diagnostic Requirements: | In addition to all of the usual profile diagnostics (including EC-hardened profile reflectometers), all of the instruments for observing fluctuations in density and electron temperature. Consider trying to use the imaging EC system as a correlation instrument, to compare with CECE system. BES, DBS, etc. | ||
| Analysis Requirements: | Full-blown analysis with GYRO, TGLF, ONETWO, TRANSP to compare characteristics of observed fluctuations and available models. | ||
| Other Requirements: | Also submitted to Turbulence and Transport area. | ||
| Title | 195: Correlation of confinement and fluctuations with electron heating in ITER Baseline Scenario | ||
| Name: | Pinsker pinsker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | E.J. Doyle, T.L. Rhodes, C.C. Petty, B. Tobias, B. Grierson, L. Schmitz, G.R. McKee, Z. Yan, L. Zeng, M. Porkolab | ITPA Joint Experiment : | No |
| Description: | Extend work on dependence on confinement on Te/Ti in conventional ELMing H-mode (C.C. Petty, et al., PRL, Vol. 83, 3661, 1999), QH-mode (Schmitz, et al., 2012), hybrid discharges (Petty, IAEA 2008) to ITER Baseline Scenario discharges. Determine dominant turbulence modes in electron-heating dominated ITER Baseline Scenario. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Basic approach is to create matched pairs of ITER Baseline Scenario discharges, with beam heating only compared with ECH plus NBI. Adjust applied torque in NBI-only case to match rotation speed obtained in ECH case. Study two cases: (1) one with 'normal' levels of toroidal rotation speed (all co beams plus EC compared with adding enough counter-NBI in all NB case to match rotation, and (2) [low rotation] About zero applied torque NBI-only case compared with mostly EC plus about zero applied torque. In each case obtain detailed fluctuation measurements to enable later comparison of observed fluctuations and predicted dominant turbulence (TEM vs ITG, etc.) modes. Measure particle confinement with He puffing technique in all cases. Consider making perturbative measurements with modulation of EC compared with modulation of part of NB power. | ||
| Background: | Previous work (see citations in Description) on DIII-D has shown that confinement is very sensitive to Te/Ti in several regimes. In 2012 experiments in which ECH and/or FW direct electron heating was added to ITER Baseline Scenario discharges in order to move towards the dominant electron heating regime, it was observed that incremental confinement tended to be poor with either form of direct electron heating. These results are likely the result of the sensitivity of confinement to the ratio of electron and ion temperatures, as previously studied in detail on DIII-D in other plasma regimes. We propose to investigate explicitly the correlation between confinement and the presumed underlying fluctuations and turbulence in the ITER Baseline Scenario to test this picture. The 2012 work by Schmitz, et al. in QH-mode clearly showed an ITG-to-ETG mode dominated transition upon entrance to the electron-heating dominated regime. Measured fluctuations in Te at mid-radius significantly increased, while the density fluctuations (BES) did not change very strongly. Since ITER discharges will, almost by definition, be electron-heating dominated, it is quite important to study the confinement properties of IBS discharges as we transition from the typical hot-ion types of discharges towards the electron-heating dominated case that is of most interest. | ||
| Resource Requirements: | 1-2 day experiment. Needs 4-6 NBI sources (no off-axis beams needed, but will need counter-NBI along with mostly co-), all available gyrotrons (at least 6). Helium puffing in some discharges. ITER Baseline Scenario discharges. | ||
| Diagnostic Requirements: | In addition to all of the usual profile diagnostics (including EC-hardened profile reflectometers), all of the instruments for observing fluctuations in density and electron temperature. Consider trying to use the imaging EC system as a correlation instrument, to compare with CECE system. BES, DBS, etc. | ||
| Analysis Requirements: | Full-blown analysis with GYRO, TGLF, ONETWO, TRANSP to compare characteristics of observed fluctuations and available models. | ||
| Other Requirements: | Also submitted to Inductive Scenarios area. | ||
| Title | 196: Narrow heat flux widths and divertor power dissipation | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | The goal is to determine the required conditions for partial detachment as a function of q_parallel and possible limits to lambda_q and/or upstream density associated with the MHD stability of the SOL. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Plasma shape and q95 will be similar to the ITER baseline inductive scenario. A power scan from P_inj/P_LH of 1.5 to maximum power will be carried out. At each power level a density scan to partial detachment of the outer divertor and ultimately to the H-mode density limit. High quality profiles of upstream, midplane, pedestal and separatrix values of density, temperature and pressure will be key measurements. Divertor measurements of heat flux from IR cameras as well as divertor plasma measurements should also be made. Key issues to be addressed by the analysis include 1) divertor heat flux width and its spreading with increased dissipation, 2) upstream/downstream pressure and power balance, 3) the upstream separatrix density at detachment, and its relationship to the Greenwald density limit, and 4) variation of the upstream pressure gradient and its relationship to the MHD stability parameter. | ||
| Background: | A multi-machine scaling study involving 6 tokamaks conducted by the ITPA, yielded a scaling of the inter-ELM H-mode near-SOL characteristic heat flux widths, which extrapolates to a midplane 1 mm SOL width for the ITER baseline inductive scenario. An extremely narrow power width in ITER implies very high upstream parallel heat fluxes which may be challenging to dissipate in the ITER divertor whilst maintaining a reasonable H-mode operational window. The scaling has so far concentrated on strongly attached divertor plasmas, but must now be extended to more directly ITER-relevant dissipative divertor conditions. | ||
| Resource Requirements: | All available beam power | ||
| Diagnostic Requirements: | Core and divertor Thomson scattering, IR camera and divertor Langmuir probes. | ||
| Analysis Requirements: | Edge stability analysis | ||
| Other Requirements: | -- | ||
| Title | 197: Scaling of divertor detachment onset | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | The goal is to measure the 2D profile of the divertor plasma as it approaches detachment for testing the boundary plasma modelling codes. The profiles should be measured in L-mode, low power H-mode and high power H-mode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Experimental configuration should be LSN with divertor configuration optimized for divertor diagnostics, particularly divertor Thomson and the swing arm probe. There may be some sweeping of the divertor required for full coverage. A density scan should be carried out particularly near detachment. The density scan should also be carried out at different powers, L-mode, low power H-mode and high power H-mode. These data will be used for careful comparison with modelling of detachment with 2D fluid codes. | ||
| Background: | The 2D boundary plasma models currently used for prediction of ITER divertor heat flux control currently have difficulty reproducing many of the features observed. Particularly the operational space, or midplane separatrix density, at which divertor detachment occurs, is not properly predicted by the codes. A careful test of these models with detailed diagnostics is needed to isolate issues the models are not addressing correctly. A number of divertor diagnostics have recently been added, upgraded, or repaired, and include, divertor Thomson, swing arm probes, IR camera, carbon flow, and visible and IR periscope. Now is the time to revisit this divertor condition. | ||
| Resource Requirements: | LSN with co-beams | ||
| Diagnostic Requirements: | All divertor and boundary diagnostics | ||
| Analysis Requirements: | Careful reconstruction of divertor profiles and compared to 2D boundary modeling | ||
| Other Requirements: | -- | ||
| Title | 198: Compare the power threshold for pre-emptive NTM suppression vs Catch and Subdue | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Last year we showed that the pre-emptive NTM suppression needed very low (as low as 500 kw) power. However, we did not have time to do a power scan for Catch and Subdue. If we can show that the power thresholds are roughly the same, it would prove that the Catch and Subdue is the most power efficient NTM suppression strategy. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Same as last year. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 199: Reducing the tearing island detection time | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Last year, we showed that for tearing modes we suppressed before they reach the critical amplitude (knee level), it takes ~200 ms to suppress the mode. Once the size of the island is above this threshold, it takes an order of magnitude (~2 sec) longer to suppress the island. Also, late catch may impede the capability to stop the disruption.
Last year, we used only the simple Mirnov fft amplitude for mode detection. Three new capabilities are suggested to detect the mode earlier: 1. Combine Mirnov and ECE diagnostics to isolate 2/1 NTM from sawteeth and other noise: Currently, a new high resolution ECE is being developed. We want to connect this system to the real-time PCS and look at the q=2 surface to spatially identify the source of the mode which will increase the minimum detection level. 2. Use frequency windowing to better detect NTM: Use knowledge of the 2/1 mode frequency range to search the mode of interest. This will reduce the noise from side bands. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Same as last year | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 200: Multiple periods of NTM suppression | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Last year, we were only able to suppress one NTM in a shot. We did not have the capability to turn off the gyrotrons after the mode is suppressed. <br> <br>This year, we want to upgrade the PCS to enable shutting the ECCD when the mode is suppressed, wait for another mode to come up and use catch and subdue method when the mode emerges. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 201: NTM Control Upgrade: Capibility to deposit EC at multiple locations | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Plasma Control | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Currently we can only control the EC launchers to steer to a single q-surface. We want to upgrade PCS to enable multiple location EC deposition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
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| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 202: Error Field Correction Improvement: Adding BetaN dependence | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Tune and test the improved (newly developed) BetaN dependent EFC. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | There is a clear BetaN dependance of the EFC control parameters. Instead of multiplying the control by random constants which is the current way of operating at higher BetaN, last year we added a BetaN dependent term to EFC algorithm. This year, we want to use this capability. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 203: Dynamic plasma-wall equilibrium in ELMy H-mode | ||
| Name: | Pigarov apigarov@gmail.com | Affiliation: | CompX |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | -- | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 2-3 dedicated ELMy H-mode shots similar to 144977 and one high-density L-mode shot. Modulated gas puff when discharge reaches an equilibrium. | ||
| Background: | ELMy H-mode discharges with minimal NBI, short outer leg, no gas puff and weak pumping on DIII-D showed ~25% particle losses from pedestal during ELMs and an unsaturated recovery (almost linear)of discharge density up to 200ms. Preliminary modeling of ELM cycle based on new macroblob approach in UEGDE shows that without "wall" pumping of >50% of ELM particle losses, the pedestal density is recovered at very short time scale, 1-10ms.
Hypothesis is that (i)the "wall" is the main source of particles between ELMs, (ii) the ELMy dischage is in a dynamic plasma-wall equilibrium, and (iii) divertor is relatively transparent to neutrals. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | temporal evolution of plasma pedestal and divertors+gas balance analysis during and between the shots | ||
| Analysis Requirements: | 1)time-dependent UEDGE-MB
2)TGLF analysis for fluxes and if any inward pinch is involved. | ||
| Other Requirements: | -- | ||
| Title | 204: Suppression of Density Limit Disruptions with RMP Fields | ||
| Name: | Evans evans@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award | Presentation time: | Not requested |
| Co-Author(s): | Contributions from collaborators at DIII-D, ITER, AUG, KSTAR, LHD and TEXTOR | ITPA Joint Experiment : | No |
| Description: | ITER will operate at high pedestal density and will attempt to control the energy released by ELMs using pellet pacing. If the neutral particle load from the pellet pacing exceeds the ITER pumping throughput capability at high pellet frequencies the pedestal density may build up and exceed the edge radiation limit triggering a radiative collapse of the current profile. This will induce a voltage spike that will generate a massive runaway electron current. Very little work has been done in tokamaks to prevent or control the initial edge radiative instability that is responsible for this process. The goal of this experiment is to use a technique that was developed in Tore Supra using RMP fields to control radiative instabilities (T. E. Evans, et al., J, Nucl. Mater., 196-198 (1992) 421). The Tore Supra results suggest that the DIII-D I-coil will be able to prevent the radiative collapse and the subsequent disruption. A successful demonstration of this idea in ISS plasmas may allow ITER to operate above the Greenwald density limit, using their in-vessel ELM coil, without disruptions thus enhancing Q in various operating scenarios. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Two sets of discharges will be used to test this concept. Using inner wall circular limited discharges (ref. 146121) with q_a = 3, the density will be ramped up to the density limit with a continuous deuterium gas feed. Once conditions have been established for a reproducible density limit radiative collapse and disruption, the I-coil will be applied to hold the radiating zone between the edge plasma and the high field side wall as was done in the Tore Supra experiment. If the radiating zone can be robustly stabilized by the RMP field at the Greenwald limit, the density will be increased to see how far above the Greenwald limit the plasma can be pushed without triggering a disruption. After establishing the new upper bound on the density with the stabilized radiating zone, the NBI power will be dropped in order to quench the the radiation asymmetry. The next step is to repeat this procedure in a lower single null plasma and test its applicability to ITER. | ||
| Background: | In Tore Supra the ergodic divertor coil produced a dominate m,n = 18,6 mode spectrum on the q = 3 surface. The RMP coil was capable of operating with up 45 kA of current but it was found that the radiative instability could be stabilized and controlled with a coil current of only 18 kA and 3 MW of LH power. LHH was used to heat the plasma and produce a highly localized radiating zone between the edge of the plasma and the high-field side graphite wall. The DIII-D I-coil, when configured to produce an m,n = 9,3 resonant field on the q = 3 surface with ~ 7 kA, provides a delta_br approximately equivalent to that of the 30 kA Tore Supra RMP field implying that it will be possible to reproduce the Tore Supra results in DIII-D. In order to produce a significant radiation source needed to enhance the radiative instability and test the limits of the stabilizing effect of the RMP field, it may be necessary to add an impurity puff such as methane or neon. | ||
| Resource Requirements: | This experiment can be completed in one day. | ||
| Diagnostic Requirements: | IR camera, bolometer array, fast UCSD camera, SXRI and the full array of magnetic, fluctuation and pedestal profile diagnostics. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 205: SOL width in top-limited discharges | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Divertor & SOL Physics | 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 | 206: Studies of arcing on divertor PFC surfaces | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | R.P. Doerner, R.A. Moyer, C.P.C. Wong, C. Chrobak, V. Rohde (IPP) | ITPA Joint Experiment : | No |
| Description: | Arcing may contribute to PFC erosion and dust production in a tokamak. So far arc studies in DIII-D were limited mostly to analysis of campaign-integrated arc tracks on graphite tiles. We propose studying arcing on different material surfaces under controlled plasma conditions using DiMES. The focus will be on W surfaces with and without coatings. Post-exposure analysis of the arc tracks on the samples will be performed by SEM and profilometry. Fast arc imaging and time-resolved arc current measurements will be attempted. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose two 1/2 day experiments. In the first experiment we will expose multi-button DiMES holder with a few different 1/4" buttons including solid W, W coating on graphite, oxidized W, B-coated W, and W with fuzz grown in
PISCES. Fast imaging of arcs will be attempted. In the second experiment we will expose a specially built DiMES sample with a W disc partially coated with an isolating layer and isolated from the holder, thus allowing arc current measurements. | ||
| Background: | Recent work on ASDEX Upgrade has shown that in a machine with metallic PFCs arcing may be a dominant erosion mechanism. Earlier DiMES experiments in DIII-D showed that W is more prone to arcing than other metals such as V or
Be. There is also evidence pointing to isolating layers and fuzz on W surfaces increasing the probability of arcing. We will test arcing on different surfaces using DiMES, focusing on W as the most ITER-relevant material. | ||
| Resource Requirements: | 2 1/2 day experiments, ELMing LSN H-mode, 5-6 MW of NBI, OSP just inboard of DiMES. | ||
| Diagnostic Requirements: | DiMES, fast UCSD camera coupled to a view of DiMES, MDS, IR TV, filterscopes, divertor and core Thomson, SPRED, CER with W lines. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 207: Dust generation from deposited layers and leading edges | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | C. Wong, R. Moyer, N. Brooks, M. Fenstermacher, S. Krasheninnikov, C. Lasnier, R. Smirnov | 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 : | No |
| 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 called for exposure of tokamak generated deposits (carbon in the short term) 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, UCSD fast camera coupled to a view of DiMES, lower divertor tangential TVs, CER, Thomson (divertor and core), filterscopes, MDS, lower divertor Langmuir probes, SPRED, IR TV. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 208: qmin control | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | There has been work on the high dimensional multi-parameter current profile control. <br> <br>This is very hard problem and more importantly very hard to judge the improvements that are achieved from one experiment to another. <br> <br>I propose development of single output control for qmin (or li). This would focus the effort the most important single parameter of interest. <br> <br>The output and performance of this work can be easily understood by regular scientific stuff (it work, it does not work, it works under the following conditions,...). This would enable more feedback from the SS group and better final product within a shorter time frame. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Design a very simple control that can be understood by average scientist (no LQR, LQG,...). For example a proportional control can be designed to be a function of two plasma parameters. | ||
| Background: | John Ferron previously (2005-2006) studied the qmin control. This was a one off experiment. The resulting control was not used since then due to various issues.
Use gain scheduling Proportional control for qmin (i.e. proportional gain as function of important parameter). The PCS code to gain scheduling was written last year. Combine with Beta control to enable independent control of qmin and Beta. Now we have real-time steerable mirrors, off-axis beams and more capabilities in PCS. It should be possible to | ||
| Resource Requirements: | 1/2-1 day experimental time. PCS upgrades. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 209: RMP ELM-Control for Snowflake Configurations | ||
| Name: | Joseph joseph5@llnl.gov | Affiliation: | LLNL |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | SA Allen, BI Cohen, RH Cohen, ME Fenstermacher, E Koleman, TD Rognlien, DD Ryutov, VA Soukhanovskii, MV Umansky | ITPA Joint Experiment : | No |
| Description: | The combination of both the snowflake divertor and the use of external coils to generate RMP-ELM suppression could be a game-changer for future tokamak fusion reactors. This configuration possesses the unique ability to mitigate both steady-state and transient heat fluxes.
The snowflake divertor has significant advantages over the standard divertor in its ability to couple to RMPs. Because the snowflake has stronger magnetic shear and longer connection length: (i) the rational surfaces on which RMPs act are more closely packed near the separatrix (ii) the magnetic stochasticity criterion of island overlap is easier to satisfy (iii) it may be easier to place an island sufficiently close to the pedestal to suppress ELMs, e.g., to suppress ELMs more quickly after initiation of RMP coil currents (iv) the Kolmogorov length can be made much shorter than the connection length, which should lead to a relatively large edge region that is affected by enhanced effective perpendicular diffusion. For this experiment, it is important to clearly characterize the properties of the snowflake pedestal, including ELM stability and the relative magnitude of ExB and diamagnetic rotation. It is important to determine whether the perpendicular electron flow can be made to vanish sufficiently close to the steep gradient region of the pedestal. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce the successful snowflake discharges as an experimental control (with C-Coil error field control). Starting from this configuration, apply n=2 and 3 RMPs generated by I-Coils. Ramp I-coil current to 4KA-t to determine whether ELM mitigation or suppression can be observed. Use hi-res TS and CER to characterize pedestal before and after coils are energized.edestal before and after coils are energized. | ||
| Background: | Initial snowflake-minus divertor studies in DIII-D indicated that ELM energy was universally reduced in the snowflake configuration, and ELM peak heat flux in the divertor outer strike point region was significantly reduced in the snowflake configuration with deuterium seeding. For a recent review of RMP-ELM control, see I. Joseph, Contrib. Plasma Phys. 52, 326 (2012). | ||
| Resource Requirements: | Snowflake PCS, I-Coils, ,C-Coils | ||
| Diagnostic Requirements: | High-res pedestal profile diagnostics, core and edge impurity spectroscopy (CXR), divertor diagnostics (DTS, IR TV, spectroscopy) | ||
| Analysis Requirements: | TRANSP, EFIT, ELITE, UEDGE, TRIP3D, EMC3 | ||
| Other Requirements: | |||
| Title | 210: The Dependence of Poloidal Rotation on Turbulence | ||
| Name: | Chrystal chrystal@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | K.H. Burrell | ITPA Joint Experiment : | No |
| Description: | The goal is this experiment is to study poloidal rotation before and after the formation of an ITB. Turbulence will be minimized where the ITB forms and this will allow for the study of the effect of turbulence on poloidal rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Form ITB's of varying strengths and toroidal rotations, measure poloidal rotation with CER (using vertical and in/out tangential methods), and use fluctuation diagnostics to assess turbulence levels. The CER measurements require particular timing of 30LT or 30RT, 330LT and 210LT or 210RT, and BES requires one of the 150 neutral beams. Total beam power and torque will be varied to form ITB's of varying strength and therefore different levels of turbulence suppression. The main focus of the experiment is measurements of rotation and fluctuations in the time before the ITB forms and the time before the ITB collapses. | ||
| Background: | Poloidal rotation on DIII-D is not well understood and agreement between measurements and neoclassical theory is unpredictable. A relatively new poloidal rotation diagnostic using tangential CER chords on the high- and low-field side of the plasma is particularly useful for acquiring measurements to compare to neoclassical theory, though it can only make measurements in the core of the plasma. Results from this diagnostic on a few ITB shots from 2012 seem to indicate that agreement between measurement and theory improves after ITB formation, suggesting that the ITB's turbulence suppression allows the poloidal rotation to become neoclassical. Using the ITB as a tool to control turbulence while measuring poloidal rotation should allow the relationship between the two to be studied. | ||
| Resource Requirements: | 6-8 neutral beam sources, at least one source from the 30, 210 and 150 beam-lines. 330LT highly desirable. | ||
| Diagnostic Requirements: | Standard profile diagnostics and fluctuation diagnostics, especially BES. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 211: Control of experimentally simulated burning state | ||
| Name: | Suzuki suzuki.takahiro@qst.go.jp | Affiliation: | QST |
| Research Area: | Plasma Control | Presentation time: | Requested |
| Co-Author(s): | Nicholas Eidietis and Tim Luce | ITPA Joint Experiment : | No |
| Description: | To investigate controllability of alpha heating during a transient phase of confinement change, e.g. start of H-mode and subsequent burning state | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment focuses on the controllability of transient change into a burning state in the ITER standard operation scenario condition. That is, q95~3, Q=10 (Pα=2Pex) and at the timing of transition from L-mode regime to the ELMy H-mode (or ELM mitigated H-mode) regime. A group of external heating power, as a first step, is used as a control knob on the burning state. Alpha-particle heating is simulated through positive feedforward control of another group of heating power proportional to the neutron emission rate, or ni2<Ï?v>-equivalent quantity calculated using real-time data. The simulated alpha-particle heating power is controlled by feedback control of e.g. stored energy via the external heating power or density and density profile via fueling. | ||
| Background: | In burning plasma, pressure and heating power are strongly coupled through α-particle heating. This strong coupling can cause a thermal excursion without burn control in burning plasma. Reliable burn control is essential. Burn control simulation were performed in JET (IC) and JT-60U (NB), where heating power is divided into two groups, one group Pα simulating the α-heating (in proportional to neutron emission rate or ni2<Ï?v>, where <Ï?v>=f(Ti)) and the other Pex simulating external heating. In ELMy H-mode plasmas, control at effective Q=5Pα/Pex=15 and Q=5-30 were demonstrated in JET [1], and JT-60U [2,3], respectively.
[1] T.T.C. Jones et al., Proc. 28th EPS Conference, ECA 25A (2001) 1197. [2] H. Takenaga et al., Fusion Sci. Tech. 50 (2006) 76. [3] K. Shimomura et al., Fusion Eng. Des. 82 (2007) 953. | ||
| Resource Requirements: | feedback and feedforward control of heating or fueling systems | ||
| Diagnostic Requirements: | real-time evaluation of ni2<Ï?v> equivalent
(real-time Ti & ne (ni) or real-time neutron emission rate) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 212: Island seeding with saw teeth pacing: New method to control tearing modes | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | Basil Duval, Ian Chapman, Rob La Haye, Anders Welander, Dave Humphreys, Wayne Solomon | ITPA Joint Experiment : | No |
| Description: | Saw teeth pacing has been shown to seed islands. This is achieved by directing the EC on the q=1 surface and modulating the power. At a critical EC modulation frequency and duty cycle, saw teeth formatyion and crash times can be synchronized to the modulation frequency. 2/1 tearing modes are usually triggered by a saw teeth crash. When we know the crash time of the saw teeth, we can preemptively suppress the tearing mode by using the EC at the q=2 surface for a very short just before the crash occurs.
This method may bring more robust NTM suppression for the ITER (and ITER relevant targets at DIII-D) and reduce the EC power consumption. these saw teethAt a By adjusting the |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | 1/2-1 day experimental time. May need PCS upgrade if sawteeth triggering is sensitive to EC deposition distance to the q=1 surface. Then, multiple location EC deposition algorithm has be online. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 213: Real-time Ray Tracing for PCS | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Plasma Control | Presentation time: | Requested |
| Co-Author(s): | R. Prater | ITPA Joint Experiment : | No |
| Description: | Develop real-time ray tracing for PCS to trace the power deposition of EC in real-time and feedback on this information for better alignment of the EC with the q surface of interest. We have a Sneel's law simple calculator and a more computationally intensive. The goal is to include one of these in PCS (current plan is to use the simple one). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | Thursday night testing. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 214: Density (Greenwald) Limit and the Radiation Driven Islands | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | D. Gates, L. Delgado-Aparicio, E. Hollman, J. Ferron, M. Van Zeeland, N. Commaux, B. Tobias, B. Grierson | ITPA Joint Experiment : | No |
| Description: | A new theoretical model for the Greenwald Limit has been proposed that appears to be consistent with experimental observations based on radiation driven islands [D. A. Gates, L. Delgado-Aparicio, "On the origin of tokamak density limit scalings", PRL]. We propose to study and test this theory at DIII-D. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | One prediction of this theory is that direct heating of the rational surfaces that participate in the radiation driven island phenomena should suppress these islands and pass the density threshold, since this would avoid the shielding process described above.
1. Vary the heating power and location, and measure the island size and radiation a. Run at ECH cut-off density meets Greenwald density limit (~1 MA, 2 T). b. Measure the radiated power from the island to see if the theory predicted power balance criteria is met: High-resolution bolometer for time-resolved profiles of the radiated power density c. Use ECE-I, visible cameras, and MRI to get never before collected data at the density limit to understand the underlying physics of the density limit disruptions. Study the island formation process at different rational surfaces. Note that for high resolution visible camera working at high density will enable good measurements. 2. Seed different impurities (e.g. Argon) a. This will give us better measurement of the island b. Give more information on how the radiation from different species effect the island growth. 3. Look at the effect of heating additional surfaces if the ECH increases the limit. | ||
| Background: | -- | ||
| Resource Requirements: | 1 day. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 215: Control of tungsten PFC erosion by local gas injection | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | C.P.C. Wong, P.C. Stangeby, O. Buzhinskij | ITPA Joint Experiment : | No |
| Description: | Use local gas injection nearby DiMES to deposit a sacrificial low-Z coating on top of a tungsten PFC surface in order to protect tungsten from erosion. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A graphite DiMES sample with deposited W layer will be used for the experiment. Non-toxic, non-explosive metacarborane C2H12B10 will be injected through a gas line opening nearby DiMES. Partially detached H-mode plasma conditions will be used, that would normally result in measurable tungsten erosion during ELMs. The hope is that the coating deposited on the tungsten surface between ELMs will protect the surface from erosion. A second experiment without gas injection may be needed for comparison. | ||
| Background: | Tungsten is a leading PFC candidate material for ITER divertor and for devices beyond ITER. While erosion of tungsten under detached plasma conditions should be very small, transients such as large ELMs and disruptions can lead to a substantial erosion and surface damage including melting. Sputtered tungsten entering the plasma core would lead to unacceptably high radiation losses. If a sacrificial low-Z coating is deposited on top of W surface, it can protect surface from erosion and prevent core contamination with high-Z impurities. However, a thin coating will be quickly eroded, so it has to be renewable in-situ, preferably during a plasma discharge. We propose a proof-of-principle experiment, where a low-Z coating would be deposited on top of W-coated DiMES sample by local gas injection. | ||
| Resource Requirements: | One or two 1/2 day experiment(s), partially detached LSN H-mode with OSP near DiMES | ||
| Diagnostic Requirements: | All available lower divertor diagnostics, DiMES, core and divertor Thomson, CER with C and B lines. IRTV view of DiMES highly desirable. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 216: Campaign to study physics at the Greenwald limit | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | D. Gates, L. Delgado-Aparicio, E. Hollman, J. Ferron, M. Van Zeeland, N. Commaux, B. Tobias, B. Grierson | ITPA Joint Experiment : | No |
| Description: | There is growing interest at DIII-D physics close to the density limit. A new theoretical model for the Greenwald Limit has been proposed that appears to be consistent with experimental observations based on radiation driven islands by D. Gates. The study of the small pellet injection at the density limit is proposed by N. Commaux. Also, high density allows great conditions for high resolution cameras to study the plasma dynamics (e.g. Van Zeeland visible camera). <br> <br>We propose to make a scenario close to the density limit where all these interesting phenomena can be studied in a combined fashion. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop a scenario close to the density limit in which ECH can be deposited without cutoff or too much diffraction. Study all the physics of interest.
One prediction of this theory is that direct heating of the rational surfaces that participate in the radiation driven island phenomena should suppress these islands and pass the density threshold, since this would avoid the shielding process described above. 0. Study the effect of the small pellet injection at the density limit. 1. Vary the heating power and location, and measure the island size and radiation a. Run at ECH cut-off density meets Greenwald density limit (~1 MA, 2 T). b. Measure the radiated power from the island to see if the theory predicted power balance criteria is met: High-resolution bolometer for time-resolved profiles of the radiated power density c. Use ECE-I, visible cameras, and MRI to get never before collected data at the density limit to understand the underlying physics of the density limit disruptions. Study the island formation process at different rational surfaces. Note that for high resolution visible camera working at high density will enable good measurements. 2. Seed different impurities (e.g. Argon) a. This will give us better measurement of the island b. Give more information on how the radiation from different species effect the island growth. 3. Look at the effect of heating additional surfaces if the ECH increases the limit. | ||
| Background: | -- | ||
| Resource Requirements: | 1-2 days of experimental time. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 217: PID (proportional integral derivative)control of Error Field | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Current Error Field control algorithm is proportional only. Control Theory states that for a linear single input single output system PID (proportional integral derivative) gives as good as a control as any other more complicated system. In our case we have two inputs phase and magnitude but it is highly likely that PID for this case is very close to the most optimal control. The underlying reasoning behind this observation is that a big portion of the control effort is about the power supply and coil dynamics not the EF dynamics. The PID control have been avoided is due to the possible numerical noise in obtaining the derivative and felling that integral term is not of the greatest importance. There many control methods to overcome the issues related to derivative term. As for the integral term being not that important, it the combination of P, I and D terms that gives the optimal control effort.
This control can be designed for static or dynamic EF correction. A more ambitious experiment would use real-time PID auto-tuning via relay-feedback. In this case the tuning algorithm is turned on during various stages of the shot and the control is auto-tuned in real-time. This type of control tuning is standard practice in real life control systems such as chemical factories where the process can not be stopped and the system dynamically evolves. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use automatic relay-feedback experimental PID tuning method. Unlike previously used control development methods, this is a single shot tuning method without user interface (i.e. automatic). Under normal circumstances, we will be running a single for a tuning shot for regime of interest.
For real-time PID auto-tuning via relay-feedback, the tuning algorithm is turned on during various stages of the shot and the control is auto-tuned in real-time. Background: There are great number of experimental and data mining based methods to tune the PID control for systems. While, I will be developing PID controls based on data mining methods to get a first level answer, experience with previous experiments from NSTX show that the best approach is to use the automatic-experimental tuning. | ||
| Background: | |||
| Resource Requirements: | 1/2 day. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 218: Controlled Snowflake Divertor Study | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | V. Soukhanovskii, B. Duval, T. Petrie, S. Allen, SOL/Boundary Group. | ITPA Joint Experiment : | No |
| Description: | The snowflake configuration was studied with minimal control last year. Many of the parameter scans were not stable, snowflake + was not achievable with reliability, distance between the x-points was not an parameter we could adjust. <br> <br>We proposed to study controlled divertor snowflake this year. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Develop and test Snowflake control for lower and upper divertor.
2. After snowflake divertor configuration control is developed, perform pedestal and divertor measurements in two configurations: snowflake-plus and snowflake-minus. 3. Scan the distance between the x-points with the control. Note the plasma behavior. 4. Perform parameter scans, for input power, SOL/divertor collisionality with deuterium puffing). 5.Impurity injection (e.g. argon) scan. 6. Study effect use cryopumping being on and off. 7. Gas injection based on radiation feedback to keep detachment. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 219: Establishing an experimental basis for plasma effects on magnetic islands | ||
| Name: | Evans evans@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Jim Callen | ITPA Joint Experiment : | No |
| Description: | Measuring modifications of the properties of vacuum magnetic islands in H-mode discharges has not been possible due to diagnostic limitations in these highly rotating relatively hot plasmas. Understanding how the plasma transforms the width and internal structure of the islands (e..g, electric fields and flows) is essential for developing a theoretical basis for RMP ELM suppression that can be projected to ITER. The goal of this experiment is to use new imaging diagnostics (ECEI, SXRI, BESI, and high resolution CCD cameras) to obtain critical information on how edge magnetic islands are transformed by the plasma as the rotation and pressure is increased. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with inner wall limited plasmas and make measurements of magnetic islands with all available diagnostics (including fluctuation systems). Increase the power and torque to heat and rotate the plasma. If H-modes can not be obtained in this shape transition to a LSN ISS shape and apply various I-coil field combinations. Use all available diagnostics to search for islands. n=3 phase flipping will be use to modulate the islands. This is expected to increase the contrast for improved imaging. | ||
| Background: | In limited Ohmic and L-mode plasmas (e.g., Tore Supra, TEXTOR and TEXT) low edge temperatures have provided clear images of magnetic islands using visible emission line CCD cameras. Comparisons of these images with field line integration code calculations of vacuum magnetic islands have demonstrated that the plasma has very little if any affect on the size and locations of the islands under these operating conditions. Detailed measurements of the internal plasma properties of these islands have been made on TEXTOR, TEXT, CSTN-II and LHD using Langmuir probes, HIBPs and high resolution Thomson scattering. These reveal important processes that can have a significant impact on particle and energy confinement. This information is needed in H-mode plasmas to pin down how transport is affected by the RMPs and how ELMs are suppressed. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 220: Comparison of n=2 RMP field effects on ELMs and pedestal properties in KSTAR and DIII-D | ||
| Name: | Evans evans@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | YM Jeon, J. Kim, Y. Oh, J. Kwak, W. Kim, JY Kim. et al., and ITPA PEP-23+25 Group Members | ITPA Joint Experiment : | Yes |
| Description: | In 2011 KSTAR obtained ELM suppression using n=1 RMP fields but did not see suppression when using n=2 fields. In November 2012 ELM suppression was observed with n = 2 fields and the n = 1 results were reproduced with a longer suppression window and better density control. The goal of this experiments is to match as closely as possible, in DIII-D, the operating parameters used in n=2 ELM suppression experiments at KSTAR during 2012 and apply similar RMP fields. If ELM suppression is obtained we will vary the I-coil and discharge parameters to test the boundaries of the suppression window (e.g., I-coil current, q95, Pinj and shape). This data will be used as part of a worldwide n = 2 RMP study to quantify the parameters required for ELM suppression versus ELM mitigation with a variety of coil geometries and plasma conditions. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start with a plasma and RMP fields that are relatively well matched to to those used in KSTAR for their n=2 RMP experiments. Attempt to suppress ELMs with a Br spectrum, density, Pinj, and pedestal profiles similar to those in KSTAR. Acquire high resolution profile data with and without n=2 RMP fields and compare the differences with those seen at KSTAR. Adjust the RMP field plasma parameters until relatively long ELM suppressed conditions are obtained. Establish the boundaries on beta normal, rotation, density/collisionality, and q95 for good n=2 suppression and identify regions of ELM suppression parameter space that are compatible with KSTAR operations. | ||
| Background: | KSTAR has obtained n=1 and 2 ELM suppression with a significantly different type of RMP coil and plasma parameters than in DIII-D with either the n=2 or n=3 I-coil RMP fields. In addition, attempts to obtain ELM suppression in DIII-D with n=1 fields have not yet resulted in suppression or significant mitigation. Understanding the mechanisms involved in ELM suppression between the two machines will provide new insight into the key physics mechanisms involved in ELM suppression. Joint experiments between DIII-D and KSTAR can result in a rapid expansion of our ability to predict how the ITER ELM coils will behave and how best to optimize RMP coil designs to suppress ELMs while minimizing negative effects on H-mode performance and divertor operations. | ||
| Resource Requirements: | PCS shape control algorithms compatible with KSTAR shapes in addition to the standard RMP ELM control hardware and heating systems (NBI, ECH and ICRF). | ||
| Diagnostic Requirements: | Standard RMP ELM control diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | Scheduling of the experiment needs to accommodate the travel arrangement of the international participants. | ||
| Title | 221: Gas Injection Feedback with Radiation Measurements | ||
| Name: | Kolemen ekolemen@pppl.gov | Affiliation: | PPPL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | T. Petrie, N. Eidietis, D. Humphreys | ITPA Joint Experiment : | No |
| Description: | Measure the radiation at various location in the plasma (main, divertor, ...) using bolometer channels. By using these real-time measurements, adjust the gas injection of various species (Argon, Neon, Deuterium,...) to keep power exhaust, detachment, and many other parameters at desired values. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop and test the radiation control.
Use it to control power ?ux into the divertor, divertor temperature and the power load as estimated by the bolometric calculations of main chamber and divertor radiation | ||
| Background: | This feedback has been successfully used at Asdex-Upgrade with great utility. (See, A. Kallenbach et al 2012 Nucl. Fusion 52 122003 and reference within) | ||
| Resource Requirements: | PCS code upgrade and 1/2-1 day of experimental time to study the physics. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 222: Investigation of optimum pellet injection geometry for pellet triggering minimum throughput/PFC flux | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, N. Commaux, G. Huijsmans, S. Futatani | ITPA Joint Experiment : | No |
| Description: | This proposal aims at improving the determination of the minimum pellet penetration and pellet size for ELM triggering by injecting the pellets in a tangent direction to the plasma edge and to minimize fuel throughput and power fluxes to PFCs. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Develop a H-mode discharge with the lowest possible ELM frequency and optimum X-point position for tangential pellet injection from the R-2 line, which should be similar to that already developed for the super-X divertor experiments. Perform pellet injection in this configuration by adjusting velocity and pellet size until ELMs are triggered (or not triggered). Decrease the X-point height in several steps and repeat the scans of pellet plasma parameters. Repeat until pellets of all sizes and velocities achievable trigger ELMs. | ||
| Background: | ELM control by pellet pacing is one of the two ELM control schemes considered in the ITER baseline. The optimization of pellet pacing relies on the injection of the smallest possible pellets (to reduced throughput) but that provides a sufficiently large perturbation in the pedestal to trigger the ELM. | ||
| Resource Requirements: | NBI, super-X divertor plasma configuration. pellet injection | ||
| Diagnostic Requirements: | Core and pedestal plasma ne,Te, Ti, pellet diagnostics and IR divertor power fluxes during ELMs | ||
| Analysis Requirements: | Analyse the appropriateness of the existing Super-X divertor H-modes for these experiments | ||
| Other Requirements: | -- | ||
| Title | 223: Demonstration of integrated ELM-paced pellet fuelled H-mode plasma in ITER-relevant range. | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, N. Commaux, G. Huijsmans, S. Futatani | ITPA Joint Experiment : | Yes |
| Description: | This proposal aims at demonstrating that plasmas in which ELMs are controlled by pellet triggering to an ITER-like relevant level can be appropriately fuelled with fuelling pellets that have an ITER-like peripheral deposition. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Establish an ITER-like discharge with low ELM frequency and then increase the ELM frequency with the smallest possible LFS/X-point pacing pellets by a factor 4, 8, 12. Then introduce high field side pellets for fuelling at various frequencies and with characteristics that lead to a similar particle deposition as in ITER (i.e. rho_dep = 0.8-0.9) and determine effectiveness of pellet fuelling by comparing with gas fuelling replacing fuelling pellets. Repeat with increased input power (by 2 to increase edge temperature) and pellet controlled ELM frequency increased by 2,4 and 6 with respect to natural frequency in these conditions. | ||
| Background: | ELM control by pellet pacing is one of the control schemes included in the ITER baseline which also relies in pellet injection for plasma fuelling. Pellet penetration in ITER is also relatively shallow for fuelling pellets so that triggering of ELMs by pellet pacing could affect the efficiency of pellet fuelling. The compatibility of ELM control by pellet pacing with fuelling of plasmas by pellet fuelling with ITER-like peripheral deposition depth remains to be demonstrated. | ||
| Resource Requirements: | NBI, pellet injectors. | ||
| Diagnostic Requirements: | Core and pedestal plasma ne,Te, Ti, pellet diagnostics and IR divertor power fluxes during ELMs | ||
| Analysis Requirements: | Existing experiments with pellet pacing from LFS and fuelling pellets from HFS. | ||
| Other Requirements: | |||
| Title | 224: Determination of minimum confinement degradation for ELM suppression with I coils | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | T. Evans, O. Schmitz, G. Huijsmans | ITPA Joint Experiment : | No |
| Description: | To determine which is the minimum pedestal pressure drop required for ELM suppression with I-coils and the associated confinement degradation and its dependence on plasma shape, input power, coil current spectrum (n=2,3) and their detailed temporal shape when accessing the H-mode. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | To establish the minimum level of current in the I coils to get ELM suppression in a high delta with optimum q95 ~3.5, n=3 and a pre-I coil betaN_1.8 and the two absolute phases of the perturbation with respect to the known error field in DIII-D. Document the pressure decrease at ELM suppression and the associated plasma energy decrease. Optimize the waveform of I(t) and P_NBI(t) to access ELM suppression with minimum confinement degradation. Repeat at higher beta (beta_N = 2.2). Repeat the experiment with n=2. If feasible attempt the same experiment with a low delta configuration | ||
| Background: | ELM suppression can cause a decrease of the pedestal pressure and overall plasma energy but the level of decrease depends on plasma configuration, on the level of current in the I-coils, on the absolute phase of the perturbation and on the temporal waveform of the I coil and input power when accessing the H-mode and the ELM suppressed regime in ways that are not properly quantified nor understood. | ||
| Resource Requirements: | NBI injection, I coils in n=3 and n=2 configuration | ||
| Diagnostic Requirements: | Core and pedestal plasma diagnostics to document plasma parameter and confinement changes. MHD measurements to determine mode activity. | ||
| Analysis Requirements: | Edge stability analysis | ||
| Other Requirements: | |||
| Title | 225: Study of atomic and molecular processes in deep detachment conditions | ||
| Name: | McLean mclean@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | M. Groth, T. Carlstrom, B. Bray, S. Allen, V. Soukhanovskii | ITPA Joint Experiment : | No |
| Description: | Low temperature detached operation in the ITER divertor will lead to generation of recombination, and atomic and molecular interactions which will dominate plasma-surface interactions near the targets. DIII-D with its combination of the upgraded DTS and spectroscopic diagnostics is uniquely capable to study the existence of these processes and aide in validation of state-of-the-art boundary simulation codes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Density scan in low power L-mode and H-mode plasmas using a SAPP plasma configuration. Increase density in step-wise fashion to determine maximum density operating point with steady flattop. Repeat discharge for spectroscopy, imaging, and improved sample statistics. | ||
| Background: | Refurbishment of the DTS system has led to unprecedented capability for accurate measurement of low plasma temperatures, significantly below 1 eV. This capability will extend the uniqueness of the DIII-D DTS system to study ion recombination, and molecular dissociation, excitation, and interaction in low temperature conditions, and help elucidate their role in the detachment process. These are described in detail for Hydrogen and the CxHy,x=1-3 families at www.eirene.de/html/a_m_data.html with many processes including data in the 0.1-10 eV range, and have been incorporated into the Monte Carlo neutral modeling code EIRENE, but to date has very little in-situ data to compare/validate the models to with reliable local temperature measurements. | ||
| Resource Requirements: | LSN with OSP near the DiMES radius with sweeps for outer leg characterization. Gas puffing for high density operation near maximum capable n/nGW before MARFE formation. | ||
| Diagnostic Requirements: | Extended/enhanced DTS, high resolution MDS, NIR spectroscopy, filtered DiMES TV, filtered Tan TV, floor LPs, compact spectrometers | ||
| Analysis Requirements: | Plasma conditions from DTS and LPs, 2-D reconstructions of D and C emission profiles for UEDGE/OEDGE simulation with EIRENE neutral modeling. | ||
| Other Requirements: | -- | ||
| Title | 226: Compatibility of ELM suppressed regimes by 3-D fields with pellet injection for plasma fuelling | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, N. Commaux, T. Evans, O. Schmitz, G. Huijsmans, S. Futatani | ITPA Joint Experiment : | No |
| Description: | This experiment aims to study the ELM triggering and its avoidance of an ELM when a pellet with ITER-like particle deposition profiles is injected into ELM supressed regimes with 3-D fields (RMP and QH-mode)while maintaining a low collisionality plasma and to quantify how much lower is the particle confinement time for pellet injected particles in the presence of 3-d fields. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Stablish suppressed ELM discharges with RMP in n=2, n=3 and QH-mode with various levels of heating power to scan plasma edge temperatures (and effective pellet deposition depth). Inject fuelling pellets from the HFS at low frequency so that the plasma density and collisionality is not changed and determine if ELMs are triggered by the pellets. If they are, adjust I-coil current and evaluate if pellet triggered ELMs can be avoided for a given level of I-coils current. | ||
| Background: | The injection of shallow pellet can lead to the triggering of ELMs in suppressed ELM regimes and this remains an open compatibility issue of ELM suppression with pellet injection for plasma fuelling in ITER. In addition, 3-D fields can lead to a decreased edge particle confinement time that can increase fuelling requirements. It is thus important to quantify the edge particle confinement time in suppressed ELM regimes as this is important to quantify the throughput requirements associated with ELM suppression by in-vessel coils and potential QH-mode operation in ITER. | ||
| Resource Requirements: | NBI heating, I-coils in n=2 and n=3 configurations for RMP and QH-modes and pellet injection from HFS. | ||
| Diagnostic Requirements: | Core plasma and pedestal plasma measurements, divertor IR measurements for ELM fluxes and pellet measurements. | ||
| Analysis Requirements: | Analysis of edge stability, pedestal measurements, pellet measurements and divertor power fluxes. | ||
| Other Requirements: | |||
| Title | 227: Mechanisms leading to termination of the runaway discharges and effects on PFC energy deposition | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | E. Hollmann, J.R. Martin-Solis | ITPA Joint Experiment : | No |
| Description: | Determine characteristics of runaway deposition on PFCs (magnitude, timescales and area) dependence on processes leading to runaway discharge termination. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create large (200 kA+) RE beam and determine the runaway deposition characteristics on the DIII-D wall/divertor depending on the type of process leading to the termination of the discharge vertical upwards/downwards drift, radial inwards drift (or others) and on the speed of the drift. | ||
| Background: | Deposition characteristics of the RE plasma energy at their termination are poorly characterised for extrapolation to ITER (magnitude, timescales and areas). In part this is due to the RE beams having a significant magnetic energy that can be converted to magnetic energy in its termination. This conversion depends on the timescale of the runaway loss itself and this timescale is found to depend on the mechanism/details of the runaway discharge loss upwards/downards vertical drift and radial inwards drift which make extrapolation to ITER very uncertain. | ||
| Resource Requirements: | 1/2 to 1 run day. Reliable scheme of material injection for runaway discharge production. | ||
| Diagnostic Requirements: | Runaway diagnostics and IR | ||
| Analysis Requirements: | Existing experiments | ||
| Other Requirements: | |||
| Title | 228: High resolution study of collisionality and target structure in the transition to full detachment | ||
| Name: | McLean mclean@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | M. Groth, S. Allen | ITPA Joint Experiment : | No |
| Description: | Using DIII-D??s unique, world-leading diagnostic capabilities, carefully document target characteristics in the transition in recycling regimes to full detachment for validation of state-of-the-art edge modeling. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Determine divertor conditions with steady degree of detachment in semi and fully detached cases. At each, perform full characterization with strike point sweeping to full analyze the outer target with the suite of divertor diagnostics. | ||
| Background: | Accurate, fully physics-based validated modeling will be essential for predicting divertor conditions in ITER. Operation in detachment will be required to ensure the survivability of plasma facing components. While recent efforts to model detached conditions in DIII-D have led to significant discrepancies in predicted conditions, recent enhancements to edge diagnostic systems make DIII-D the pre-eminent machine for detailed study of a stepwise scan to fully detached conditions. | ||
| Resource Requirements: | LSN with OSP near the DiMES radius with sweeps for outer leg characterization. Gas puffing for high density operation near maximum capable n/nGW before MARFE formation. | ||
| Diagnostic Requirements: | Extended/enhanced DTS, high resolution MDS, filtered DiMES TV, filtered Tan TV, floor LPs, compact spectrometers | ||
| Analysis Requirements: | Plasma conditions from DTS and LPs, 2-D reconstructions of D and C emission profiles for UEDGE/OEDGE simulation with EIRENE neutral modeling. | ||
| Other Requirements: | -- | ||
| Title | 229: CQ de-confinement of REs using pulsed impurity injection | ||
| Name: | Loarte-Prieto Alberto.Loarte@iter.org | Affiliation: | ITER Organization |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | E. Hollmann, F. Saint-Laurent. N. Commaux, L. Baylor | ITPA Joint Experiment : | No |
| Description: | Demonstrate de-confinement of seed REs in initial phases of CQ with rapid impurity injection into CQ plasma edge. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Create disrupting target plasma with large RE seed in the CQ that evolves to a RE plateau of > 200 KA by injection of Ar (killer pellet, if possible) into low density IWL plasma. Then apply to this target plasma massive material injection pulse (from valves of shattered pellets) with various time lags with respect to the initial Ar injection and document effect on final runaway current. Perform variations of material amount of MMI in the CQ and gas specie (Ne, Ar + mixture with H or He?). | ||
| Background: | It has been proposed to de-confine RE seeds in ITER by using pulsed bursts of gas fired into the CQ plasma, thus shrinking the current channel and destabilizing tearing modes. This scheme has been tested experimentally in T-10 and Tore-Supra but results are not conclusive so far and needs systematic demonstration of its principle in tokamak experiments. We propose to test this proof of principle by using two DIII-D material injection systems one to generate the runaways and the other to de-confine them. | ||
| Resource Requirements: | 1 run day. Two material injection systems for creation of runaways and for their de-confinement. | ||
| Diagnostic Requirements: | Runaway diagnostics and IR cameras. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 230: Assessment of the degree of detachment using NIR spectroscopy | ||
| Name: | McLean mclean@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | V. Soukhanovskii | ITPA Joint Experiment : | No |
| Description: | The recent addition of a medium/high resolution near infrared (NIR) spectrometer by LLNL/PPPL to the suite of divertor diagnostics on DIII-D provides a valuable capability to diagnose spectral regions where significant emission corresponding to low temperature plasma and molecular emissions occur relevant to burning plasma studies. Key advantages of the NIR spectral region will be used to aide in development of indicators for determination of the degree of detachment similar to that of traditional emission intensity benchmarks such as the Da/Dg ratio. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using well controlled, repeat discharges with semi- and fully detached divertor configurations, NIR spectroscopy will be acquired, with focus on observing deuterium Paschen and diatomic deuterium emissions. | ||
| Background: | The NIR (0.8-2.2 micron) region has been poorly diagnosed in tokamak divertors due to the specialty of the camera and optical fiber technology required compared to more traditional visible-range spectroscopy. The NIR region, however, is especially relevant for the integrity of the Thomson scattering measurement (which will also apply in ITER), and for study of chemical vs. physical erosion sources. To this end, a new high resolution NIR spectrometer based on a high speed (up to 900 Hz) , InGaAs photodiode array has been installed on loan from LLNL/PPPL and available for regular diagnosis of the divertor region. | ||
| Resource Requirements: | LSN with OSP near the DiMES radius with sweeps for outer leg characterization. Gas puffing for high density operation near maximum capable n/nGW before MARFE formation. | ||
| Diagnostic Requirements: | NIR spectrometer connected to tangential MDS and DTS optical chords, Extended/enhanced DTS, high resolution MDS, filtered DiMES TV, filtered Tan TV, floor LPs, compact spectrometers | ||
| Analysis Requirements: | Plasma conditions from DTS and LPs, 2-D reconstructions of D and C emission profiles for UEDGE/OEDGE simulation with EIRENE neutral modeling. | ||
| Other Requirements: | |||
| Title | 231: Stabilization of NTMs with ECCD | ||
| Name: | Isayama isayama.akihiko@qst.go.jp | Affiliation: | QST |
| Research Area: | Stability & Disruption Avoidance | 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 and (c) ECCD width effect: investigate the stabilization effect for different ECCD deposition width by changing the toroidal injection angle. This proposal is mainly focused on comparison of the behavior and structure of magnetic islands in DIII-D and JT-60U. | 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 in 2012. After the discharge scenario for obtaining 2/1NTM 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: | neutral 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 | 232: Bifurcated helical core equilibrium - continued | ||
| Name: | Schmitz oschmitz@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | W.A.Cooper, B. Tobias, E. Lazarus, A. Turnbull, F. Turcon, T. Evans, L. Lao, M. Lanctot, O. Sauter | ITPA Joint Experiment : | No |
| Description: | Approach tackling to generate a bifurcated helical core equilibria from another standpoint - based on lessons learned during half day Torkil Jensen Award experiment in 2012. The experiment will be enhanced by looking into edge vs. core response. Recent ANIMEC modeling including RMP modes similar to the spectrum used for ELM control have shown a strong deformation at the edge with maintaned helical core. This can be an advantegeous scenario for ITER hybrid discharges. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Run highly elongated plasma with q_0 slightly above unity. Control broad q-profile by off axis ECCD and generate flat or best hollow q-profile with q_min at large radius. Let q profile evolve to lower q_0 from above 1 (which was not accomplished/attempted in initial experiment) to allow helical core take over before sawteeth set in. | ||
| Background: | In 2012 a Torkil Jensen Award experiment was executed in an international collaboartion to generate a bifurcated helical core plasma. The experimental sequence was based on ANIMEC modeling of 3D equilibria. We got a transient state with a long liced core mode which exhibits properties of the helical core, saturated kink like mode we are expecting. However, this mode occured only transiently and we concluded as a reason that we have to adapt the experimental strategy in two major points: (1) in order to achieve a q_0 at unity without letting sawteeth dominating the scene, we have to generate a plasma whcih drifts into q_0 around unity from above. (2) the plasma we run were L-mode plasmas with moderate elongation. These plasmas are not good for effective ECCD due to low edge temperatures and hence weak CD efficiency. To enhance this we need to attempt the experiment in H-mode. Last, we saw that increased elongation is beneficial and hence we want to run an H-mode plasma at increased elongation. Key is to get a flat or best even hollow q profile with q_o at unity avoiding sawteeth before q_0 goes towards this level. | ||
| Resource Requirements: | ECH for ECCD (all gyrotrons)
Highly elongated plasma (last time we used DN patch) H-mode in HFS limited configuration of maybe DN | ||
| Diagnostic Requirements: | toroidal and poloidal soft X-ray arrays
ECE-I Xpt soft Xray imaging n=3 in 240 degree phasing from Icoils and standard setup for ELM control | ||
| Analysis Requirements: | ANIMEC
kinetic EFITS | ||
| Other Requirements: | -- | ||
| Title | 233: C erosion and D up-take at high surface temperature | ||
| Name: | Wong wongc@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | C.P.C. Wong, D. Rudakov, P.C. Stangeby, J. Watkins, W. Wampler, D. Elder, J. Brooks, A. Chavez | ITPA Joint Experiment : | No |
| Description: | Compare graphite surface erosion and deuterium concentration at ambient and high surface (~700° C) temperature. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Two depth-marked graphite DiMES buttons will be used for the experiment. The graphite buttons will first be outgassed with minimum content of hydrogen, and the initial graphite thicknesses measured. Two DiMES button holder module will be used. The first DiMES module will be designed with the capability of heating the central button to ~800° C. The second DiMES module is a non-heated one. Partially detached H-mode plasma conditions will be used. The first part of the ½ day experiment will be exposing the heated sample to 4-5 well-characterized discharges, after which the heated DiMES module will be exchanged to the non-heated module. The second half of the ½ day experiment will be used to expose the ambient temperature module to another 4-5 well-characterized discharges. If these cannot be accomplished in one ½ day, a two ½ day experiment could be requested. | ||
| Background: | ITER will make decision on the divertor surface material for the initial phase of ITER operation by the end of 2013, between the use of W/C and all W divertor. A W/C divertor will be more tolerant to the ITER initial startup events including transient effects like ELMs and disruption. The critical potential drawback of the W/C divertor option is the high C erosion and the corresponding projected high tritium inventory in the carbon. Due to safety consideration the ITER PFC tritium inventory is limited to 0.7 kg. It is important to provide technical data on the graphite surface erosion and corresponding tritium inventory when the surface is operating at around 700° C. | ||
| Resource Requirements: | One, if necessary two ½ day experiments, partially detached LSN H-mode with OSP on DiMES. | ||
| Diagnostic Requirements: | All available lower divertor diagnostics, Langmuir probes, DiMES, core and divertor Thomson, CER with C lines. IRTV view of DiMES highly desirable. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 234: Study fishbone-like instability at the edge of QH-modes | ||
| Name: | Nave mfn@ipfn.ist.utl.pt | Affiliation: | Instituto Superior Tecnico, Lisboa, Portugal |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | W. Heidbrink, Y. Zhu, K. Burrell, N. Gorelenkov | ITPA Joint Experiment : | No |
| Description: | (Related research area: Energetic particles)<br>Investigate ehos in the form of chirping modes. Study their interaction with fast ions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat an old pulse where n=3 chirping eho modes are observed. Use the best diagnostics available for fast particles and fluctuations. The aim is to improve on the data available for existing pulses.
Possible reference discharges: pulse 118740 and pulse 131376. | ||
| Background: | QH-modes are characterized by the EHO, edge MHD activity that appears to delay ELMs by enhancing transport at the top of the pedestal. In some discharges the EHO appears in the form of repetitive bursts resembling fishbone activity, with typical mode numbers in the range n=3-6. These interesting edge mhd modes were observed in DIII-D QH-mode pulses up to 2008, when the QH-Mode used to be obtained with strong Neutral Beam Injection counter to the plasma current. The chirping nature of the bursts suggests the modes could be driven by NBI fast ions similar to internal n=1 fishbone bursts. Like in the internal fishbones, the eho bursts frequencies are close to the frequency of precession of fast ions. However, whether these edge modes interact with fast ions causing displacement or losses of fast ions remains inconclusive. | ||
| Resource Requirements: | Reverse Ip | ||
| Diagnostic Requirements: | Fast Particle Diagnostics, fluctuation diagnostics, profile diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 235: High resolution Te pedestal measurements in RMP ELM suppression experiments | ||
| Name: | Truong truongd@fusion.gat.com | Affiliation: | Sandia National Lab |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Max Austin, George McKee, ELM control aficionados | ITPA Joint Experiment : | No |
| Description: | The goal of the experiment is to measure the changes in the pedestal Te as RMP ELM suppression is applied using the new high-spatial-resolution (HR) ECE channels. The pedestal Te profile as well as the pedestal Te oscillations will be measured with microsecond time resolution using optically thick electron cyclotron emission. In addition, edge BES measurements will supply high resolution information about the density fluctuation changes on the same fast time scale. Significant increase in density fluctuations are observed with BES in the pedestal and interior region during RMP ELM-suppressed discharges. These measurements may help elucidate the q-dependence/resonance condition of the ELM suppression. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment can be carried out in any RMP ELM suppression set of discharges with Bt in the range 1.9-2.1 T. The goal will be to examine fast (fluctuation scale) and slow (equilibrium) changes in the edge temperature profile as the ELM-suppression resonance condition is entered and exited. | ||
| Background: | Up until now it was not possible to measure the edge Te profile with high resolution due to the limitations of the fixed frequency ECE radiometer channels. In 2013 eight new ECE channels with narrow radial separation and variable location capability will be added to the midplane ECE radiometer diagnostic. The channels will cover nominally 4 cm of radius with 5 mm resolution, with the actual resolution limited somewhat by relativistic broadening. Previous measurements of Te near the edge could only give a general trend of the Te gradient. With the HR channels it will be possible, for example, to observe the phase relationship of oscillations at various radii near the resonant q surface. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | ECE, BES | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 236: Wavenumber dependence of turbulence suppression in the L-H transition | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G. Tynan, G.R. McKee, Z. Yan, T.L. Rhodes | ITPA Joint Experiment : | No |
| Description: | Recent experimental evidence suggests that turbulence reduction in the edge transport barrier varies with the perpendicular turbulence wavenumber, such that intermediate wavenumber fluctuations couple more efficiently to LF flows and are more strongly suppressed (T. Estrada, EPS 2012, L. Schmitz, Nucl. Fus. 2009). Following the temporal turbulence evolution at different wave numbers would elucidate turbulence-flow coupling and provide evidence of the energy cascade. If additionally the total energy transfer is measured (via reciprocating probe insertion), the damping rate of turbulence-driven flow can be determined. Limit-cycle transitions allow acquiring all necessary data with improved SNR and statistics. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a discharge near LH-transition power threshold exhibiting extended limit cycle oscillations. Change DBS incidence angle as well as vertical plasma position to access the maximum possible range in poloidal turbulence wavenumber, and measure simultaneously turbulence level and flow evolution at each wavenumber. By scanning/optimizing shape/vertical position and DBS launch, we expect to be able to cover a wavenumber range of ~ 0.5 cm^-1 - 6 cm^-1. Data will be acquired for 4-5 wavenumber values within this range. | ||
| Background: | Limit cycle (LCO) L-H transitions have been previously investigated in detail via DBS and BES, but only limited reciprocating probe measurements have been obtained. Also, the DBS measurements have been limited to the lowest accessible turbulence wavenumber in the previous LH transition experiment (~3-4 cm^1), due to the limited number of available LCO shots. It is straightforward to assemble data across an extended wavenumber range with a sufficient number of repeat shots. | ||
| Resource Requirements: | 30,330,150 beams | ||
| Diagnostic Requirements: | DBS5,8 BES, PCI, Midplane reciprocating probe | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 237: Critical electric field for runaway electron growth and decay under quiescent conditions | ||
| Name: | Granetz granetz@mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Disruption Mitigation | Presentation time: | Requested |
| Co-Author(s): | C. Paz-Soldan, N. Eideitis, R. Granetz, E. Hollmann, A. Tronchin-James, J. Wesley | ITPA Joint Experiment : | Yes |
| Description: | This experiment aims to study the critical toroidal electric field (Ecrit) required for runaway electron (RE) growth under quiescent plasma conditions. A plasma target developed for extremely low-density operation (125010, n_e < 5E12 cm^-3) will be used. Quiescent RE plasmas are of fundamental interest because they are free from the complicating transient and impurity effects associated with disruption-generated RE beams.
To distinguish the observed RE dynamics from the Dreicer mechanism, the plasma current or density will be systematically increased and decreased during the discharge to assess hysteresis in the RE growth and decay (see background). These measurements will ultimately assess and clarify the anomalous RE loss mechanisms needed to match the data. Furthermore, the development of a robust and quiescent plasma target with a significant RE population could lead to significant advances in the science of RE control and measurement. Piggyback studies will use this experiment to improve equilibrium reconstructions of the RE beam. Depending on the robustness and magnitude of the RE beam, experiments suggested in other RoF proposals could also be applied to this target discharge in the future. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will begin by reproducing 125010, a discharge developed for error field control (EFC) measurements that was recently discovered to contain a measurable RE population that was quiescent for over 3 seconds without locked modes. Excellent EFC is critical to the success of this experiment, as the error field ultimately limits low-density operation via locked modes.
When low-density RE growth is observed, the plasma current (proportional to the toroidal electric field) will be slowly increased and decreased, in order to measure the levels of RE beam growth and decay. The density will then be decreased and increased slowly and the same measurements made. | ||
| Background: | The required Ecrit for RE growth is the dominant parameter for gauging whether collisional suppression of the RE beam is likely during MGI. However, observations on C-mod and DIII-D suggest that Ecrit is anomalously large when compared to existing collisional theories [R. Granetz, ITPA 2012], both in quiescent and post-disruption plasmas.
Furthermore, the Ecrit required for the suppression of an already-avalanching RE beam can be measured by decreasing the toroidal electric field (Ephi) until RE decay is observed. It is expected that the Ecrit for RE decay will be smaller than for RE formation. Equivalently, the required density for RE formation is expected to be lower than that for RE decay at constant Ephi. | ||
| Resource Requirements: | Mostly Ohmic only operation. Beam blips for MSE, 1 gyrotron for possible RE seed generation. 4 SPAs operational. | ||
| Diagnostic Requirements: | ZNS or FPLASTIC in low gain mode. Thomson, MSE, SXR, ECE, UCSD fast camera, CO2 and 288 GHz interferometers | ||
| Analysis Requirements: | May enable development of new tools for RE equilibrium measurement | ||
| Other Requirements: | |||
| Title | 238: Direct Measurement of NTV Torque | ||
| Name: | Logan ncl2128@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | E.J. Strait, J.K. Park, A. Cole | ITPA Joint Experiment : | No |
| Description: | This proposal is for the direct measurement of NTV torque using the integrated Maxwell stress tensor at the vessel wall.
The goal is to test the hypothesis that the fundamental mechanism of NTV momentum exchange between the plasma and surrounding world is electromagnetic. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The idea is to use the extensive set of new magnetic diagnostics in DIII-D to measure the poloidal and radial field structure across the vessel surface during application of large n=3 and/or n=2 error fields. The variation in the phase difference between Br and Bp from 90 degrees gives a stress, which can be integrated to find the total electromagnetic torque exerted on the plasma.
The approach should be as simple and clear as possible to address the fundamental question at hand. 1) Reproduce previously successful high beta NTV discharge scenario at low rotation (ex: 131861). 2) Turn on step function NTV error field while using constant torque NB feedback to maintain constant beta and pumping/gas-puffing to maintain constant density. This will allow the plasma rotation to evolve to a offset value, and thus distinguish between any resonant (T~1/vÏ?) and non-resonant (T~vÏ?) EM torque. 3) If necessary, rotate non-resonant error field, providing synchronous detection for clearest magnetics data (may also provide information about intrinsic error field contribution to non-resonant torques). | ||
| Background: | Discussion of the fundamental mechanism of NTV momentum exchange between the plasma and outside world has been contradictory over the last decade [Boozer Phys. Plas. 2009, Pustovitov Nuc. Fus. 2011].
The DIII-D magnetic diagnostics upgrade has made possible n=1-3 Maxwell stress tensor measurements that will directly validate/invalidate the interpretation that non-ambipolar transport leads to J�B forces between external coils and the plasma. Although this method has been used to measure EM torque on large n=1 tearing modes [Logan Plas. Phys. Cont. Fus. 2010], it has never before been used to study n>1 or non-resonant breaking effects. | ||
| Resource Requirements: | n=3 I-coils on top of EFC for low rotation. Co and counter neutral beams. | ||
| Diagnostic Requirements: | 3D magnetics. This experiment should be done after calibration experiments exploring poloidal structure of plasma response to 3D fields for a better understanding of diagnostic capabilities.
CER poloidal rotation (Ï?E) and toroidal rotation. Standard profile diagnostics. | ||
| Analysis Requirements: | Sufficient plasma response signal in the magnetics and decent rotation scan must be ensured. The integral torque from the maxwell stress tensor can be checked between shots for an eye-ball sense of how clean the data looks. | ||
| Other Requirements: | |||
| Title | 239: Continuous Scan of NTV Torque Across Low Rotation Peak | ||
| Name: | Logan ncl2128@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | E.J. Strait, J.K. Park, A. Cole | ITPA Joint Experiment : | No |
| Description: | The proposal is to perform a peak scan on NTV torque using the direct measurement of the integrated Maxwell stress tensor to obtain a complete trace for T(Ï?Ï?).
The goal is to improve observation of peak NTV in DIII-D at low rotation when the plasma enters the super banana plateau regime. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment should be performed after a passive (constant NBI torque) rotation scan establishes the capability of new 3D magnetics to measure electromagnetic torque imparted by non-resonant error fields.
1) Reproduce discharge parameters known to be near the previously observed peak (ex 138574) 2) Switch on non-kink-resonant error field, maintaining as much as possible plasma parameters (NB feedback for beta, gas-puffing/pumping, etc.). Rotate fields if necessary for clear magnetic signal of plasma response. 3) Use mix of co/counter NB to slowly - compared to a transport time scale ~[(vti/R0)(δ B/B0)2]-1 - scan plasma rotation past the NTV peak at low counter rotation. | ||
| Background: | Recent experiments have emphasized key resonances physics in NTV theories that can result in peaked torque as a function of plasma rotation. These have used shot-by-shot rotation scan techniques. The Maxwell stress tensor provides a continuous torque measurement, allowing for a direct trace of T(Ï?Ï?) that can be used to more accurately test theoretical models. In addition, poloidal rotation data will help reduce crucial degrees of freedom in the models (see [Cole, et al. PRL 2011]). | ||
| Resource Requirements: | Co and counter beam feedback as well as careful density control. n=3 I-coils. | ||
| Diagnostic Requirements: | 3D magnetics (Maxwell Stress). CER poloidal rotation (Ï?E) and toroidal rotation. Standard profile diagnostics. | ||
| Analysis Requirements: | Initial Maxwell stress tensor integrated torques can be calculated in the time between shots to give a sense of how things look. TRANSP for all shots. | ||
| Other Requirements: | |||
| Title | 240: Optimization of Applied Error Field Spectrum for NTV | ||
| Name: | Logan ncl2128@columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | E.J. Strait, J.K. Park, C. Paz-Soldan | ITPA Joint Experiment : | No |
| Description: | This proposal is the experimental optimization of NTV torque using DIII-D 3D field coil sets.
The goal is to test the hypothesis that a single spectrum can be found to best couple to the plasma in such a way as to induce NTV torque. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The basic plan is to perform a scan of combined I and C coil currents to optimize the poloidal spectrum for n=3 non-resonant breaking. Assuming no (or corrected) intrinsic error field contributions and monotonic increase of NTV with applied amplitude of a given spectrum, the degrees of freedom are reduced to relative phases and amplitudes. Due to the limited phasing options of the six-coil coil sets applying n of 3 fields, the full operating space can thus be represented on a 2D grid of I vs. C coil current amplitude (where negative I-coil current implies odd parity, and negative C-current implies a similar parity with respect to the bottom I-coil).
The trick here is the ability to follow a contour of constant area (or equivalently energy) norm for the applied plasma surface mode. These contours will need to be created in the two dimensional I and C-coil amplitude space prior to the experiment. This is not challenging using DCON. If the theory is valid, the trace of a single such contour will provide the optimum configuration. To validate this, however, a minimum of 3 shots should be used: 1) Establish equilibrium and shape identical to selected NTV reference shot (ex. 131861). 2) Energizing n=3 I-coils in odd parity, scan from positive to negative amplitude adding C-coil field as necessary to follow contour of applied energy norm spectra. During this stage NBI and gas/pumping should be on feedback to maintain as closely as possible plasma rotation, beta, density and shape. The NTV torque as a function of relative coil amplitudes will thus be available from the required NBI torque, and directly from the electromagnetic torque measured at the vessel wall by the new 3D magnetic diagnostics (see relevant devoted proposals). 3) Repeat step 2 for multiple initial amplitudes so as to effectively confirm the maximum torque occurs at a consistent value of I vs C current. If scans find optima with differing spectra, an empirical optimum must be found by filling the space more completely. If scans find optima with same spectra and time permits, non-resonant error field effect should be checked by comparing the two available phasing of the n=3. | ||
| Background: | Recent success of NTV experiments on DIII-D have provided a tool to effectively control plasma rotation, improving access to QH mode. They have also inspired significant advances in the theory and modeling of these non-resonant torques. This experiment will seek to optimize NTV torque as an experimental tool by choosing the optimal spectrum. This (as far as the author knows) has not been attempted beyond taking advantage of the n^2 dependence and choosing between even or odd parity based on pitch/kink resonance calculations.
Further, the experiment seeks to validate qualitative details of the NTV models used in support of experiment. The Ideal Perturbed Equilibrium Code (IPEC) can currently be used to calculate the dominant (and ith-dominant) external field (with energy norm) for a given equilibrium maximizing the sum of the squared pitch resonant field at each rational surface. The code is being upgraded to perform a nonlinear optimization of the normalized external field for inducing NTV torque with the hypothesis a dominant mode structure will be found. The experiment would thus validate or invalidate the theories ability to predict optimized I/C-coil configurations for NTV torque experiments (impacting lower n non-resonant EFC). Measurement of the Maxwell stress tensor (and thus integral electromagnetic torque) have never before been available for n=3 error field experiments in DIII-D. If they are successful in measuring torque from non-resonant electromagnetic fields this will be an ideal candidate to showcase the new ability. The author expects that low m modes will dominate the NTV spectrum, allowing clear reconstruction of the plasma response across a large area of the vessel wall in the optimal configuration. | ||
| Resource Requirements: | Requires careful pre-determined mixing of I and C-Coil amplitudes, along with ability to phase shift at least 2 sets (i.e. C and top I coils). Also requires co and counter beam mixing and feedback to ensure profile / plasma shape changes do not obscure the NTV dependence on the poloidal profile. | ||
| Diagnostic Requirements: | 3D magnetics. This experiment should be done after calibration experiments exploring poloidal structure of plasma response to 3D fields for a better understanding of diagnostic capabilities (i.e. do we need to rotate the applied fields for clean signal?).
CER poloidal rotation (Ï?E) and toroidal rotation. Standard profile diagnostics. | ||
| Analysis Requirements: | The experiment requires some theoretical preparation: DCON analysis of energy norm contours in coil-amplitude space, and IPEC prediction of optimized coil fields with the corresponding falloff (ratio of NTV for the ith-dominant modes) to guide experiment in choosing intelligent scans through the operating space.
Control room analysis of NTV (especially EM torque measurements) would guide the experiment as to how densely each operating space needed to be filled. | ||
| Other Requirements: | |||
| Title | 241: C-13 and background carbon | ||
| Name: | Wong wongc@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | C.P.C. Wong, D. Rudakov, P.C. Stangeby, J. Watkins, W. Wampler, D. Elder, J. Brooks, A. McLean | ITPA Joint Experiment : | No |
| Description: | Quantify the eroded C from sample versus the carbon background from DIII-D | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Expose a depth marked C-13 coating on DiMES and at the same time, collect the C-13 and background carbon deposit in a down stream Si wall slot. After exposure to a few detached plasma discharges we will be able to distinguish both eroded C-13 from the sample and background plasma in DIII-D. | ||
| Background: | DIII-D is a carbon machine. In quantifying net and gross-erosion from a carbon wall, it is impossible to distinguish the net and gross erosion impacts from the sample surface itself and from the background carbon. With a C-13 surface, the distinction can be quantified at least for the prescribed plasma discharge. This will aid the benchmarking of modeling surface material erosion and re-deposition with better quantification of the C background in DIII-D. We have experience in designing the slot on DiMES for the collection of carbon in DIII-D. | ||
| Resource Requirements: | One ½ day experiments, ITER relevant partially detached LSN H-mode with OSP on DiMES and with as many repeated shot as time allow. | ||
| Diagnostic Requirements: | All available lower divertor diagnostics, Langmuir probes, DiMES, core and divertor Thomson, CER with C lines. IRTV view of DiMES highly desirable. | ||
| Analysis Requirements: | Analysis of collected diagnostics data. | ||
| Other Requirements: | |||
| Title | 242: q95 < 2 Tokamak Operation Via Dynamic Error Field Correction | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Attempt to sustain plasma operation at q95<2 by using RWM feedback for dynamic error field correction, similar to what was done ~10 years ago to show for the first time sustainment of plasma operation at beta above the no-wall limit. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat Aug. 2012 experiment with I-coil on SPAs, follow recipe for DEFC:
Preprogram I-coil currents matching time average of feedback-driven currents in Aug. 2012 experiment Apply RWM feedback (poloidal field sensors) on top of preprogrammed offset currents Use proportional gain only (large) Use long time constant Update preprogrammed currents shot-to-shot | ||
| Background: | Aug. 2012 experiment (Miniproposal # 2012-09-02) failed to maintain q95<2 tokamak operation. The experiment performed a C-coil compass scan to determine the optimal EFC currents. Session Leader Summary states "we believe we eliminated the possibility of residual n=1 error fields affecting plasmas or feedback". However, simple analysis of the q95<2 experiment shows I-coil currents driven by feedback overlay shot after shot, showing the presence of a significant uncorrected error field. This behavior is similar to what was observed in pioneering RWM stabilization experiments of ~10 years ago, where RWM feedback was shown to maintain stability by successfully find the optimal error field correction. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 243: 256: SOL width and divertor peak heat flux in QH-mode plasmas with/without n=3 fields | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, M. Fenstermacher, W. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to compare the SOL width and the divertor peak heat flux observed in QH-mode plasmas to the values expected from scalings based on ELMing H-modes. The effect of rotation and the effect of 3D fields will also be investigated. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using the same tokamak set-up first used on Nov. 1, 2011, we can generate both counter-rotating QH-modes and co-rotating ELMing H-modes, for direct comparisons.
By operating soon after a boronization (as on Nov.1) we can maintain the low-density required for QH-mode even with the lower strike point on top of the baffle for documentation with the IR camera. | ||
| Background: | QH-mode is an attractive mode of operation for ITER because it enables high confinement without ELMs. Recent experiments have shown that the use of nonresonant magnetic fields (NRMFs) allows QH-mode even with the low co-Ip NBI torque expected in ITER. Furthermore, the energy confinement is higher at lower rotation, unlike in other regimes. Thus, QH-mode provides solution simultaneously to the ELM problem and the usual problems associated with low NBI torque, e.g. low confinement and locked modes.
Another major ITER issue is the high peak heat flux expected on the divertor during H-mode operation. The characteristics of the SOL and divertor heat flux for QH-mode plasmas have not been studied yet, mostly because to study the divertor heat flux the strike point needs to be off pumping position, and QH-modes normally require pumping. Experiments on Tuesday Nov. 1, 2011, showed that right after boronization we could get very low density QH-modes without pumping and without the pump-out effect from n=3 fields. This suggests that we should be able to obtain QH-modes in ITER shape with strike point above the lower baffle plate, as long as such experiment is done shortly after boronization. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 244: Study NRMF driven torque in ECH-only heated plasma (ELMing H-modes) | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, J. deGrassie, W. Solomon | ITPA Joint Experiment : | No |
| Description: | Is the NRMF torque due to loss of fast ions, or is it a neoclassical effect on thermal ions? To conclusively answer this question, we need to clearly observe the counter torque from NRMFs in ECH H-modes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In previous attempts which used odd parity n=3 I-coil fields (May 13, 2009), the plasmas were severely affected by the density pumpout associated with the applied NRMFs. Recent experiments have shown that the C-coil provides even larger NRMF torque than the I-coil, without the density pump-out.
We propose to repeat the ECH H-mode experiments of May 2009 (e.g. discharge 137225) applying NRMFs using the C-coil, instead of the I-coil. | ||
| Background: | Experiment 20090513: Effect of n=3 fields on ECH H-modes was observed for the first time. Only odd parity n=3 I-coil fields were used. Very strong density pump-out was observed, accompanied by strong reduction of beta and rotation. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 245: Investigate the physics of NTM suppression by large externally applied helical fields | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use stable, locked plasmas observed in QH-mode experiments with n=3 NRMFs in July 2009, to investigate the physics of NTM suppression by large externally applied helical fields [Q. Yu, S. Gunter, K. Lackner, PRL (2000); La Haye, et al., PoP (2002)]. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce locked QH-mode plasmas like 138611.
Document the locking by modulating the NBI torque at varying amplitude. Investigate the presence of locked n=1 islands by providing an additional n=1 seed and then rotating it. Investigate the Yu-Gunter-Lackner model by applying ECH pulses in plasmas with and without the external n=3 field. Compare transport across rational surfaces. | ||
| Background: | Theory [Q. Yu, S. Gunter, K. Lackner, PRL (2000)] and DIII-D experimental results [La Haye, et al., PoP (2002)] suggest that the application of 3D fields to a high beta plasma can provide a stabilizing effect on NTMs. The hypothesis is that the applied helical field enhances the perpendicular transport across the NTM's key rational surface, thus weakening the destabilizing effect of the helically perturbed bootstrap current.
July 2009 experiments on low NBI torque QH-mode with n=3 NRMFs have shown cases of high beta plasma with rotation locked to zero at the q=2 and 3 surfaces, suggesting the formation of (2:1) and (3:1) islands. These islands remain stable until the n=3 field is removed. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 246: Develop low-torque, high normalized fusion performance QH-mode for ITER baseline scenario | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, W. Solomon, M. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | Demonstrate QH-mode operation at ITER-relevant torque, ITER-similar shape, and G>0.4 (betan>2.0, H89>2.0, q95<3.3) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Re-create discharges such as 149220 (reversed Ip) with normal-Ip operation.
Modify the short Bt ramp down in 149220 to reach q95~3.2. Dynamic EFC may be necessary at this low q95 value, to correct possible n=1 error field introduced by a large amplitude n=3 C-coil field. | ||
| Background: | Reversed Ip experiemnt of June 19, 2012 (Miniproposal No. 2012-93-2).
D1 power supply was connected to the C-coils, which gave us 7.1 kA current capability in those coils. Using the C-coil with that current to produce an n=3 nonaxisymmetric magnetic field, we were able to create stationary counter-rotating QH-mode operation with co-Ip torque of +1 Nm for about 2 seconds, limited only by hardware constraints. These shots had excellent confinement with H98y2=1.3. Normalized fusion performance parameter is sustained at G=0.2 with q95~4.7, which projects to G=0.4 with q95 =3.3. In this experiment, we used a PCS controlled, brief toroidal field ramp in order to suppresses the n=1 EHO. The Bt is held constant once the n=1 EHO disappears. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 247: Develop low-torque, high normalized fusion performance QH-mode for FNSF-AT scenario | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Demonstrate QH-mode operation at high values of beta and betaN, and with reactor relevant values of the NBI net torque (i.e. near zero). We plan to use the magnetic counter-Ip torque from C-coil fields to produce the edge rotation shear required for QH-mode, and the off-axis current drive from a BT ramp and tilted NBI to produce and sustain a broad current profile favorable for MHD stability at low q95. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The plan is to further improve on the Nov. 2011 experiments [D3DMP No.: 2012-31-01] by exploring higher betan, and zero-net NBI torque. | ||
| Background: | Experiments in Nov. 2011 have shown QH-mode plasmas with low net NBI torque (<1Nm counter-Ip) and normalized fusion performance reaching G=0.4, that is the target needed for Q=10 in ITER. This result was obtained using: a Bt ramp down simultaneous to the Ip ramp up to reach q95~3.4, a large n=3 field from the C-coil to apply the counter-Ip torque necessary for QH-mode operation, co+counter NBI to increase betaN at low torque after Ip and Bt flattop.
Further scenario improvements (in duration and performance) were foiled by hardware issue and limited run time. Attempts to follow up on these experiments in June 2012 did not succeed because of poor wall conditions, possibly because following a boronization which had problems with the pumping system. | ||
| Resource Requirements: | n order to maximize the NBCD from the tilted beamline, forward IP with reversed BT will be used. We will use the I/C-coil configuration with the C-coil connected to the D1 supply for up to 7 kA n=3 operation, and the I-coil connected to C-supplies for up to ~6.5 kA for n=1 correction plus n=3 operation. To optimize the error field correction, we will use 120 deg. quartet (even parity). | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 248: Investigate elongation limit in low li discharges with/without applied 3D magnetic fields | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Torkil Jensen Award | 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. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 249: Steady-state high beta with NCS and qmin>2 | ||
| Name: | Garofalo garofalo@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | J. Kinsey, C. Holland | ITPA Joint Experiment : | No |
| Description: | This is a re-entering of 2009 ROF proposal#376, which was a re-entering of 2005 ROF proposal #1118. Utilize new off-axis current drive capabilities and new lower divertor to extend duration and maximum betan of 2005 discharge 122959, which achieved betan=4 with qmin>2 and surface voltage=0 during plasma current flattop. This discharge used a Bt ramp-down technique for off-axis current drive. With tilted NBI and more gyrotron power this year, we may be able to replace the Bt ramp-down with steady-state compatible ECCD.<br>The Bt ramp scenario also provides a unique scenario for several high priority FNSF studies:<br>- study dependence of global energy confinement on negative shear of core q-profile, since Bt ramp discharges are unique in achieving very good confinement quality at high and very high qmin values. <br>- divertor heat flux mitigation studies, since Bt ramp discharges are unique in maintaining very good confinement quality despite strong gas puffing leading to high upstream density value. <br>- study of dependence of MHD stability limit and confinement quality on radius of qmin, since Bt ramp down is the most powerful tool available for current drive at rho~0.7<br>- producing optimal target for potential helicon off-axis current drive, since Bt ramp discharges are unique in the achievement of high electron beta at large minor radius | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Restore discharge 122959. Test effect of new lower divertor on the ability to control the density. Investigate effect of varying the density on the evolution of the q-profile and the pressure profile (lower density is favorable for better ECCD efficiency). Use ECCD for co-Ip current drive at radius larger than rho(qmin). Use tilted beamline for off-axis current drive. Some counter-NBI may be used for counter-current drive at small radius, as long as lower rotation does not reduce confinement severely. Initially, use Bt ramp down as in 122959. Work toward reduced or no Bt ramp rate. Initially, keep betaN from exceeding 4 by using NBI feedback control. Work toward higher betan with improved noninductive current profile alignment. | ||
| Background: | Analysis of discharge 122959 shows that the high beta phase is terminated by a kink mode destabilized by qmin dropping below 2 because of noninductive current overdrive near rho of qmin. With the new lower divertor we should be able to keep the density low, slow down the qmin evolution and improve the ECCD efficiency. With higher available ECCD power and FW power we should be able to improve the noninductive current alignment, avoid the rapid qmin drop observed in 122959, and extend the duration at betaN=4 and qmin>2. The availabily of counter-Ip NBI provides an additional tool to oppose the noninductive current overdrive inside rho of qmin, and improve the noninductive current alignment. | ||
| Resource Requirements: | Normal Ip, reversed Bt for off-axis NBI CD.
All gyrotron power | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 250: Plateau Ecrit with MGI | ||
| Name: | Wesley wesley@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Do the MGI into mature plateau experiment proposed for 2012. Requires successful recovery of long-duration RE plateau following ArKP, eg14584x. Gas options = D2, He, Ne, Ar. Yields steady-state effective Ecrit + assessment of dynamics to reach fully stationary equilibrium state. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | See last year's miniproposal | ||
| Background: | |||
| Resource Requirements: | ArKP injector, low-elongation ECH target, 2011 or better RE equilibrium control | ||
| Diagnostic Requirements: | As much as can be available | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 251: New Study of Alfven Eigenmodes Induced Fast Ion Transport/Loss | ||
| Name: | Chen chenxi@fusion.gat.com | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | the EP group | ITPA Joint Experiment : | No |
| Description: | To study the fast ion transport/loss induced by Alfven eignmodes by using our new prompt-loss "light-ion beam probe" which can measure the radial excursion of fast ion induced by individual mode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use plasma discharges similar to shot146096, with changes such as:
- with constant 30L before 450ms (insert one 5ms 330L beam blip while 30L turned off to confirm it's "prompt" loss and couple 330L beam blips (30L off is not required) for diagnostic purpose) and then after 450ms, switch to constant 330L (insert one 5ms 30L beam blip while 330L turned off) - vary beam power to get a more controlled variation in mode amplitude - modify the current ramp (program a flat spot in current) to keep the prompt loss orbit close to FILD longer - puff gas to change the edge ionization gradient - swap constant 330R with constant 150R (to get BES measurement all the time) | ||
| Background: | Alfvén eigenmodes (AE) induced prompt loss of beam ions have been observed. The loss amplitude shows a linear relationship with the mode amplitude. The scale is different for different mode. By relating the AE induced prompt loss to the mode amplitude and the edge ionization rate(in practice, we use density scale length), the radial excursion of fast ion caused by individual mode can in inferred. It has been found that a single RSAE with amplitude of B_tw/|B| ~10E-3 can cause as large as 10cm radial kick.
With better diagnostics setup, such as avoiding FILD PMT signal saturation, better ECEI optimal detection location, 150R for BES, we want to make a new study the AE induced fast ion transport/loss using this new prompt-loss "light-ion beam probe". | ||
| Resource Requirements: | normal BT and Ip direction, 30L, 150R, 330L | ||
| Diagnostic Requirements: | FILD2&1, CER, BES, ECEI, fast magnetics, FIDA, Thomson,â?¦ | ||
| Analysis Requirements: | Kinetic EFIT, TRANSP, NOVA-K, SPIRAL, â?¦ | ||
| Other Requirements: | |||
| Title | 252: Implement finite-state Off Normal Fault Response system | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Control | Presentation time: | Not requested |
| Co-Author(s): | Humphreys, Walker | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to test the new Matlab/Stateflow finite state system for off normal fault response (ONFR) on d3d. The goal is to both exercise the basic mechanics of implementation and to demonstrate a comprehensive locked mode response scenario | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Implement present dud system in new architecture.
Implement LM detection/avoidance/mitigation setup/mitigation sequence in new architecture. Piggy-back (no control) to ensure that logic is working correctly. In dedicated session, create locked modes w/ very low density plasmas, and allow logic to try to avoid disruption. If avoidance is too successful, purposely "break" avoidance response so that system has to proceed to mitigation setup and initiation. | ||
| Background: | ITER will require an off-normal handling system that can handle ALL off-normal events. At the same time, time is very expensive on ITER, so simply injecting impurities at the first sign of trouble is not an option. The ONFR must be able to decide to recover the shot if possible, avoid disruption without mitigation, or initiate impurity mitigation if necessary. As of now, there is no design for the ITER ONFR system, but its architecture is expected to be finalized late 2013 anyways. D3D is trying to establish some real-world experimental basis for these decisions. | ||
| Resource Requirements: | -Programming ONFR system n Stateflow & exporting to PCS
-PCS control of mitigation injection systems | ||
| Diagnostic Requirements: | locked mode detector | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 253: Study QH-mode with advanced fast ion diagnostics and codes | ||
| Name: | Chen chenxi@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | Bill Heidbrink, Raffi Nazikian, Gerrit Kramer, Keith Burrell, the EP group | ITPA Joint Experiment : | No |
| Description: | Fast ion losses associated with EHO during ELM-free QH-mode have been detected on DIII-D. But the previous measurements and interpretations are limited by the fast ion diagnostics and codes. Now, we have pitch and energy resolving Fast Ion Loss Detector (FILD) with bandwidth up to 500kHz and fast-FIDA (up to 200kHz), and full orbit calculation code SPIRAL, we can get better fast ion measurements for study of EHO, e.g. the relation with fast ion loss, the mode location, the EHO<->ELM transition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | Previous study in W.P. West et al. Jounal of Nuclear Materials, Volume 337, p. 420-424 | ||
| Resource Requirements: | piggyback of QH-mode exp. with normal BT and Ip direction and 210R | ||
| Diagnostic Requirements: | FILD1&2, FIDA and other | ||
| Analysis Requirements: | Kinetic EFIT, TRANSP, NOVA-K, SPIRAL and other | ||
| Other Requirements: | |||
| Title | 254: Neon SPI injection into early CQ to stunt RE formation | ||
| Name: | Eidietis eidietis@fusion.gat.com | Affiliation: | GA |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | P. Parks | ITPA Joint Experiment : | No |
| Description: | Inject Hi-Z (Neon) SPI immediately after the TQ of an RE-generating small argon pellet shot in order to stunt the growth of a very young RE beam. The idea here is that immediately after the TQ there is no hot thermal plasma to ionize the SPI, so the impurities will ONLY deposit within the small forming RE core, perhaps allowing the Rosenbluth density to be reached within the small volume of that RE core. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Initiate RE-producing disruption with small Ar pellets or Ar MGI into limited, low density target. After 1-2 shots to determine injection->TQ timing, inject Neon SPI at such a time that the neon will hit the plasma withing ~ 1ms after thermal quench. Observe if RE current is substantially reduced from case with no secondary injection. | ||
| Background: | It appears unlikely that any single impurity mitigation strategy will properly handle the TQ and suppress runaway electrons. As an alternative, we propose a 2-step strategy for partial RE suppression. The fundamental problem with RE suppression using just one step is that the impurities become ionized by the hot thermal plasma before they come near the core where they are needed to collisionally suppress RE production. It is possible that immediately AFTER the TQ, when there is no hot thermal plasma, a second high-Z SPI (or large pellet) injection will pass untouched through the cold thermal plasma and ONLY deposit in the newly forming RE core, creating sufficient density only in the core where it is needed to suppress further growth of the RE bea,. | ||
| Resource Requirements: | Argon pellet injector or Argon MGI, Neon SPI | ||
| Diagnostic Requirements: | neutron counters, SXR, fast camera | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 255: Super H-Mode | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | K. Burrell, R. Groebner, T. Osborne | ITPA Joint Experiment : | No |
| Description: | Background: <br>Global tokamak fusion performance is strongly tied to the H-mode pedestal height. Because fusion power scales with pressure^2, while instabilities are driven by gradients, it is possible to consider tokamak optimization via moving much of the gradient region as far radially outward as possible. This also allows optimal plasma shaping in the gradient region as well as strong wall stabilization. <br> <br>As our understanding of both pedestal physics and core transport & stability continue to improve, it is of interest to try to make best use of this understanding (even if tentative) to explore methodologies to qualitatively improve potential fusion performance. Because extensive optimization studies have already been done, we can expect that new, qualitatively improved regimes will be characterized by difficult access. However, in some cases, theory can provide guidance into possible approaches to accessing such regime, and provide motivation via the substantial predicted benefits of the regime itself. <br> <br>One such predicted regime is what is sometimes referred to as "Super H-Mode". The existence of this regime is predicted by pedestal stability studies and by the EPED pedestal model. Theory predicts that, in strongly shaped discharges, it should be possible to access very high pedestal pressure at high density. However, starting at high density results in a high collisionality which suppresses bootstrap current and prevents access to high pressure (resulting in a relatively low pedestal pressure and ELMs). But there is a predicted parameter trajectory, starting at low density and later increasing density, that should allow access to this Super H-mode regime. With very strong shaping this can lead to markedly higher pedestal pressure. Access to this regime may be optimal via starting in QH-mode, but it is also possible to consider access via an ELM-ing AT regime. <br> <br>The EPED model predicts that access to a Super H-Mode edge should be possible in both DIII-D and ITER. Initial studies in DIII-D have suggested that it is possible to go at least partway into this regime, but that wall conditions and impurity concentration are very important issues for going further. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Approach:
In a clean machine, ideally shortly after boronization, attain very strongly shaped plasmas, in a configuration optimized via EPED calculations, but with a LSN in order to minimize impurity accumulation. We'd like to consider two approaches: 1) Start in counter rotating QH-mode at low density, and increase density by reducing torque and core pellet fueling. After deuterium density increases have reached their limit, explore puffing of Neon (or other low-Z gas) to raise Zeff and increase collisionality at fixed density. Adjust density and Zeff together to optimize access to Super H regime. 2) Start in a low density co-rotating AT-like plasma - after reaching an initial ELMing steady state, steadily increase density via core pellet fueling. Then explore increases of Zeff using low-Z impurities to optimize approach to Super H-mode. In both cases, optimize shape and wall coupling to achieve very high pedestal pressure. Employ EC to stabilize core tearing modes and attempt to achieve very high global betaN. | ||
| Background: | -- | ||
| Resource Requirements: | core pellet fuelling, low-Z impurity source (eg Neon) | ||
| Diagnostic Requirements: | Thomson, CER, fluctuation diagnostics across pedestal | ||
| Analysis Requirements: | EPED runs prior to expt | ||
| Other Requirements: | -- | ||
| Title | 256: Fine scale stepwise q-scan: role of low-order rationales and resonance condition for ELM suppression | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Evans, Nazikian, Wade, Yan | ITPA Joint Experiment : | No |
| Description: | Determine how transport and turbulence in the near edge region (0.8 < r/a< 0.98) vary as q95 is varied in very small increments inside and outside of the ELM suppression window (q95~3.5), and if and how this affects the ELM-suppression process. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish ELM-suppressed plasma conditions, and then ramp q95/Ip in a stepwise fashion to establish quasi stationary conditions at each step. The goal is to perform these steps in and around the ELM-suppression windows in q95, e.g., q95~3.5. A reference condition is 145019, but Ip would not be ramped continuously as it was in these discharges. It will be stepped with delta_q95~0.02-0.05 and held constant for approximately 250 ms at each step to establish a quasi stationary condition for pedestal profiles, and allow for comprehensive documentation of turbulence amplitude and spectra, turbulence and ExB flow profile, and ELM response. BES and DBS would be used to measure turbulence characteristics (150L flat and steady is desired beam configuration, no 150R) | ||
| Background: | During q95/Ip ramp discharges in RMP ELM-control experiments, turbulence in the near outer plasma and pedestal region is found to change sharply and locally in response to varying q95. These observations were made with BES covering the 0.8 < r/a < 1+ region and observing temporal variation in local turbulence characteristics. These sharp jumps are sometimes, but not always, associated with changes into or out of ELM-suppressed conditions, and may be associated with low-order rational q95 values in the edge and pedestal region. The plasma profiles and conditions were not stationary and naturally evolved in response to change q95 and ELMing condiion. The goal of this new experiment would be to establish quasi-stationary conditions at each q95 step so that turbulence and profiles could be evaluate in these stationary conditions. The goal would be to raise and lower q95 in steps of about 0.05 near ELM suppression windows and document these changes. | ||
| Resource Requirements: | I-coils in standard RMP ELM-suppression configuration: even-parity n=3 | ||
| Diagnostic Requirements: | BES (8x8), DBS, usual edge profile diagnostics, high-speed CER, UF-CHERS | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 257: Transport Investigation During Slow Current Ramp-Up and Ramp-Down [for ITER] | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | G. Jackson, T. Casper, D. Mikkelsen | ITPA Joint Experiment : | Yes |
| Description: | Ascertain the evolution of turbulence, electron temperature and current density profiles during ITER-similar ramp-up and ramp-down plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish ITER-similar plasmas (ISS shape) and equivalent ramp-rates (similar to Jackson experiments from a couple years ago). It will be critically important to obtain thorough documentation of turbulence and transport (esp. Te) profiles during the ramp-phases, from as low an Ip as feasible. This will require a neutral beam (150 L or R), derated as necessary, with full MSE-documentation (may require repeat, reproducible discharges for full diagnostic coverage) | ||
| Background: | Accurately modeling the profile evolution during the current ramp-up and ramp-down in ITER will be critical to control and performance. Recent work (Jackson, Casper et al.) have shown discrepancies between models of plasma temperature evolution during current ramp up/down experiments on DIII-D. The electron temperature was observed to be modestly cooler than predicted (Casper-NF-20??), and current diffusion correspondingly faster. This is important to stabile operation and control of the the poloidal field supplies. The dynamics of the how profiles evolve will thus be important to the providing the proper ohmic power. This discrepancy has general similarities with the observed "L-mode shortfall", whereby transport simulations underpredict heat fluxes and turbulence and thus over-predict temperatures. This problem may be exacerbated at high q95, just the conditions during ramp-up. The dynamic evolution of the coupled temperature, turbulence and current density will be critical and thoroughly documented in this experiment. This is an ITPA-relevant topic and of strong interest to ITER (or should be.) | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | BES, DBS, PCI, CECE, ECEI, ECE, CER, MSE | ||
| Analysis Requirements: | TGLF | ||
| Other Requirements: | |||
| Title | 258: Main-ion and impurity rotation in helium plasmas | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | K.H. Burrell, J.A. Boedo, C. Chrystal, J.S. deGrassie, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | Recent measurements of main-ion toroidal rotation and inferred poloidal rotation in deuterium plasmas has displayed significant disagreement with neoclassical theory. The discrepancy with neoclassical is most significant at low collisionality, where ITER will operate. The aim of this proposal is to perform a sequence of discharges at low torque (intrinsic) with ECH H-modes, and simultaneously measure the helium and carbon toroidal and poloidal rotation with the standard CER system (carbon) and main-ion system (helium) for toroidal rotation, and the vertical and IN/OUT systems for poloidal rotation. This significantly extends the work of Grierson and deGrassie (PoP 2007). Theory by Diamond, McDevitt et. al. have established the framework for turbulent enhancement of poloidal rotation, and the extensive DIII-D fluctuation diagnostic set will be used to characterize the turbulence in these conditions and compare to theory of turbulent driven poloidal flow at low collisionality. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish dominantly helium plasmas and use deuterium balanced NBI blips for measurements. To be done in a sequence of shots varying Ip and Bt where measurements are done in ohmic and H-mode plasmas to obtain a range of collisionality. | ||
| Background: | Many measurements on DIII-D have displayed anomalous poloidal rotation of both impurities and main-ions (Solomon, Grierson, deGrassie). We need to determine if the magnitude of, and type of turbulence in these plasmas is sufficient to explain this historical discrepancy. | ||
| Resource Requirements: | One day. Helium plasmas, NBI 30LT, 210RT, 330LT. Turbulence diagnostics. | ||
| Diagnostic Requirements: | Complete CER coverage and full profiles. Turbulence characterization highly desirable. | ||
| Analysis Requirements: | CERFIT and complete kEFIT for modeling codes NCLASS, NEO, GTC-NEO. GYRO for turbulence levels. | ||
| Other Requirements: | |||
| Title | 259: Helium Campaign (D beams) | ||
| Name: | Grierson grierson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Take the time to make dominantly helium plasmas for testing RMP ELM suppression, scenarios, L-H and rotation physics for ITER non-nuclear phase. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make helium plasmas and give each area at least one day to perform experiments designed to test current theories, techniques, scenarios, etc... in helium plasmas. | ||
| Background: | ITER will have a non-nuclear phase of operation, and we need an understanding of L-H transition, ELM suppression and disruption mitigation. | ||
| Resource Requirements: | Multiple days and multiple area leaders coordinating experiments. | ||
| Diagnostic Requirements: | All. Particularly CER preparation for making robust helium measurements. | ||
| Analysis Requirements: | Most tools should work. | ||
| Other Requirements: | |||
| Title | 260: ITER Baseline Scenario Demonstration | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | DIII-D has demonstrated stable operating conditions similar to the planned baseline scenario for Q=10 operation of ITER. The similarities are in poloidal cross-section, I/aB (q95), and beta_N. The areas of difference where there is the largest concern are applied torque leading to strong rotation (potentially modifies confinement, stability, and LH threshold) and temperature ratio (potentially modifies confinement, but the resulting profiles could also affect stability and heating method could affect LH threshold). There has been confusion in the past between the existence of a stationary solution and access to that solution. It appears that stationary solutions exist in a broad range of beta_N, li, applied torque, and heating scheme, but the access conditions may be more challenging under the ITER conditions. The purpose of this proposal is to explore the access conditions and find the sufficient physics conditions for reaching stationary solutions with low torque and strong electron heating. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | Continuation and extension of work begun in 2011-2.
This would contribute to ITPA joint experiment IOS-1.1 | ||
| Resource Requirements: | Need ctr-NBI for control of applied torque. Maximum EC power available. | ||
| Diagnostic Requirements: | Need excellent measurements of current profile and kinetic profiles. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 261: Demonstration of stationary inductive solutions at low and moderate q95 | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Previous experiments on DIII-D showed stationary inductive solutions at q95~3 for beta_N=1.6-2.2 and beta_N=2.5-2.8 using different startup methods. At q95~4, there a similar range (up to beta_N=3.5) was found also with a gap in beta_N. It appears that the gap is an access issue and not the absence of stable stationary solutions in the gap. This proposal seeks to demonstrate that inductive operation is possible at all beta_N between the LH back-transition at the low end and the no-wall beta_N limit (or higher) at the high end. Demonstration of this continuum would have two effects--it would demonstrate the real operational range of inductive scenarios available for tokamak design and it would finally enable the investigation of the long-asked question of "what is different between advanced inductive and standard H mode?". | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | Builds on experiments on ITER baseline scenario and advanced inductive development over the last 10 years. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 262: Test Pedestal Structure Model Where Paleoclassical Transport Is Dominant | ||
| Name: | Callen jdcallen@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | J. Canik, T. Osborne, S. Smith | ITPA Joint Experiment : | No |
| Description: | The main objective of this ROF proposal is to test the pedestal structure model (see Background section below) predictions in regimes where paleoclassical plasma transport processes are predicted to be dominant -- mainly high n_e and low T_e. During the 2012 DIII-D campaign, studies of the slow evolution of the pedestal just above the L-H transition power produced some potentially relevant discharges for these studies. However, the parameter ranges of these pedestals were limited mainly by the emphasis on regimes where the density was low and they mainly did not use NBI heating -- so plunging probes could be inserted into the edge plasma. Further studies of the paleoclassical-based pedestal structure model are needed to: <br>1) determine if paleoclassical processes do indeed provide the minimum transport level in the pedestal; <br>2) explore the maximum T_e and n_e gradients that can be obtained in H-mode pedestals; <br>3) provide a base transport model for the higher collisionality pedestals in DIII-D, AUG, NSTX and MAST before RMPs are applied to such pedestals; and <br>4) further explore the properties of H-mode pedestals at power levels just above the L-H threshold which are likely to be the most important situations in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The basic approach would be similar to the types of discharges in the 2012 campaign (e.g., shot 148692) in which the ECH power was initially just above the L-H transition power and then reduced slightly to keep the plasma just barely in the H-mode. The difference from those shots would be to do them at higher collisionality via higher n_e and hence lower T_e, and perhaps increased Z_eff. Also, it would be desirable to do them at the highest possible B_t and I_p where the pedestal beta is smallest and KBMs are least likely. Finally, long nearly steady ELM-free periods would be desirable to facilitate good Thomson pedestal profile measurements, with good pumping to limit the pedestal density buildup. | ||
| Background: | A pedestal structure model based on paleoclassical processes providing the minimum level of plasma transport in H-mode pedestals has been developed and published recently:
1) J.D. Callen, J.M. Canik and S.P. Smith, "Pedestal Structure Model," Phys. Rev. Lett. 108, 245003 (2012). It had been shown earlier that this pedestal structure model provided reasonably good agreement with the chi_e and n_e profiles in NSTX H-mode pedestals both with and without Lithium coated plasma facing components: 2) J.M. Canik et al., "Edge transport and turbulence reduction with lithium coated plasma facing components in the National Spherical Torus Experiment," Phys. Plasmas 18, 056118 (2011). Extensive, relatively successful comparisons (factor ~< 2 for pedestal T_e gradient and n_e) have been made of this pedestal structure model with 158 pedestals from the DIII-D 2010 and 2011 pedestal databases: 3) S.P. Smith et al., "Comparisons of paleoclassical based pedestal model predictions of electron quantities to measured DIII-D H-mode profiles," Nucl. Fusion 52, 114016 (2012). The last reference indicated that paleoclassical transport processes may be dominant at low beta and high collisionality, while other processes (e.g., KBMs) are likely dominant at high beta and low collisionality. However, more dedicated and extensive studies with higher collisionality pedestals are needed to firm up this conclusion. | ||
| Resource Requirements: | Mainly low NBI power, high B_t, I_p low recycling, well pumped discharges held just barely in the H-mode regime for long periods without ELMs. | ||
| Diagnostic Requirements: | Mainly good Thomson n_e and T_e profiles. But also good CER data so the Z_eff profile in the pedestal can be determined. | ||
| Analysis Requirements: | Good kinetic EFITs, ONETWO and SOLPS transport analyses. | ||
| Other Requirements: | -- | ||
| Title | 263: Radiative divertor solutions in the ITER baseline scenario | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | The ITER baseline scenario envisions introduction of radiating impurities (e.g., nitrogen, neon, argon) to reduce the amount of power entering the SOL (radiating mantle) and the divertor (radiating divertor). No machine can replicate the conditions of ITER, but the key issues can be explored. These are: accumulation of impurities, impact on the pedestal and global confinement, methods to enter and exit the desired radiating state. This work is a necessary pre-requisite to work on the ITER urgent task of development of detachment control. The experiment should compare directly conditions with no added impurities, trace impurities, and sufficient impurity to reduce the power measured in the divertor to <50% of the total power. The stability of the radiating scheme for all candidate gases should be demonstrated. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | |||
| Background: | This contributes to ITPA joint experiment IOS-1.2. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Need to measure the steady-state and transient heat flux under intrinsic and radiating conditions. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 264: Radiative divertor solutions in the ITER baseline scenario | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | See proposal 263 under Dynamics and Control-Inductive Scenarios | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 265: o Experiments using the centerpost Swing-probes | ||
| Name: | Tsui C7tsui@ucsd.edu | Affiliation: | UCSD |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | C Tsui | ITPA Joint Experiment : | No |
| Description: | Each of the centerpost swing-probes swing out a mach-probe on a 20cm arc, which is long enough to reach the LCFS in common plasma shapes. The swing-probes should be employed with the rest of DIII-Dâ??s Scrape-off layer diagnostics to characterize a simple-as-possible-plasma as well as possible. A well characterized plasma with the additional measurements by the centerpost on the under-diagnosed inboard side of the SOL will be very useful for testing and improving computational edge models. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | â?¢Sheath limited ohmic, to confirm probe measurements are correct. This should be scheduled early in the campaign or during the startup period, so that if any unforeseen problems with the assemblies arise, there is a change to repair them if there is a mid-campaign vent. This can be run in piggyback so long as the plasma shape and conditions are appropriate.
oParticle Transport expt This can be combined with an impurity flow speed experiment, or a PMI experiment. This will require multiple shots with sweeping to characterize a simple-as-possible-plasma. (only one swing-probe, the upper or the lower, can be operated during any given shot) | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | For the SAPP characterization, Langmuir probes, X-point probe and reciprocating probes, tangential TV, divertor and core Thompson, and filterscopes. Fewer diagnostics will be needed for the proof of concept exeriments. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 266: Understand resistive wall mode control physics in low-q95 plasmas | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | J. Bialek, A. Garofalo, G. Jackson, M. J. Lanctot, E. Lazarus, J. P. Levesque, P. Martin, M. E. Mauel, G. A. Navratil, M. Okabayashi, C. Paz-Soldan, P. Piovesan, E. Strait, F. Turco, A. Turnbull | ITPA Joint Experiment : | No |
| Description: | Unstable resistive wall modes (RWMs) and the loss of active RWM control are problems that lead to disruptions in plasmas with q95 < 2. This experiment would address the control issues, and if successful, lead to a feedback-stabilized q95 < 2 discharge or a determination of the feedback limit (eg, the maximum open-loop growth rate or minimum q95 that can be stabilized using feedback). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (a) Measure latency and transfer function of the RWM feedback system (sensors, PCS, amplifiers, coils) during maintenance or plasma start-up time (no plasma required). These measurements will serve as the basis for RWM control modeling and controller design, to be performed prior to (c) â?? (e).
(b) Perform companion experiments on the HBT-EP device to address aspects of control at low q. For example, investigate the impacts of latency and saturation, test advanced control algorithms. The goal is to leverage the flexibility of HBT-EP to develop control techniques and experience that help inform the DIII-D experiment. (c) Reproduce shot 150593 without feedback to establish a baseline reference (d) Test for a possible additional uncorrected error field by applying slow RWM feedback. Follow the established strategy of iterating over several shots, using the feedback currents from the previous shot as feedforward waveforms for the next shot. (e) Instability onset is still expected as q95 is decreased. If no unstable modes are encountered, attempt a further reduction in q95 and steady operation with q95 = const < 2. In the likely event that a discharge-terminating instability is encountered, apply RWM feedback using the algorithm and settings motivated by (a) and (b). | ||
| Background: | The low-q95 RWM could become a â??standard candleâ?? for developing RWM control knowledge, similar to the low-density locked mode used for error field control studies. Recent attempts to access q95 < 2 in diverted DIII-D plasmas were met with apparent unstable resistive wall modes (RWMs) when q95 was ramped to ~2. Applying RWM feedback control facilitated temporary access to q95 < 2, for a duration of about 400 ms in the most successful case. However, feedback control was ultimately lost when control coil power supplies reached their voltage limits, allowing an unstable RWM to grow and cause a disruption. Preliminary analyses indicate that a slowly-evolving error field and pickup from higher frequency plasma modes are two issues that affected feedback performance in these discharges. Additional experimental time is needed to confirm or rule out the presence of an uncorrected error field, and to demonstrate robust stability control in this regime.
Two innovative RWM control algorithms have been developed and implemented in the DIII-D PCS. The first compensates feedback sensor measurements for known ac vacuum coil couplings. An improved feedback transfer function should result, because many of the ac couplings have non-trivial frequency dependent phase-shifts. (At present dc-only compensation is used.) The second algorithm is based on a 3d VALEN model including the DIII-D wall, coils, sensors, and plasma mode [J. Bialek, et al, Phys. Plasmas 8 (2001) 2170]. The model serves as a basis for a state-space linear quadratic Gaussian (LQG) control algorithm. The LQG algorithm exploits the knowledge contained in the model in two ways: (1) sensor measurements are reconciled with model predictions to reduce the impact of noisy signals, and (2) the optimal gain for minimizing feedback effort and mode amplitude is used. The HBT-EP experiment is a high aspect ratio, circular cross-section tokamak specifically designed for MHD control studies [D. A. Maurer, et al. Plasma Phys. Control. Fusion 53 (2011) 074016]. Low-q regimes with current-driven instabilities can be accessed using plasma current ramps. The mode control system consists of a powerful GPU-based controller and arrays of in-vessel poloidal sensors and radial coils, similar to sensors and I-coil actuators used for RWM feedback on DIII-D. HBT-EPâ??s high availability and flexible control system make it an ideal device for developing fusion-relevant mode control techniques. | ||
| Resource Requirements: | 1 â?? 2 co-Ip NBI sources, audio amplifiers on I-coils, SPA supplies on C-coils. | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | Determination of RWM feedback system transfer function and feedback modeling, prior to experiment. | ||
| Other Requirements: | |||
| Title | 267: Understand and control resistive wall mode stability in high-qmin plasmas | ||
| Name: | Hanson hansonjm@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | J. Berkery, M. Lanctot, G. Navratil, S. Sabbagh, E. Strait, F. Turco | ITPA Joint Experiment : | No |
| Description: | The goals of this experiment are to understand and control the stability of the resistive wall modes (RWMs) encountered in 2012 experiments with betan ~ 3 and qmin ~ 3. The following questions will be addressed:
(a) Is dependence of the RWM growth rate on plasma parameters, such as rotation, consistent with previously measured dependencies of the driven plasma response and the stability theory incorporating kinetic modifications to ideal MHD? (b) How do the driven plasma response and plasma rotation behave as the marginal stability boundary is crossed? (c) Can the transition to instability and associated beta collapses be avoided by controlling the plasma response using NBI feedback? |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce DIII-D shot 150301, which suffers an unstable n=1 mode and beta collapse starting at t=2265 ms. Re-optimize error field correction using slow RWM feedback. Vary the injected NBI torque to determine the sensitivity of the stability to plasma rotation. Document the change in plasma response as marginal stability is approached using active MHD spectroscopy. Attempt to avoid beta collapse by maintaining a safe level of plasma response using NBI feedback. | ||
| Background: | RWMs and other MHD modes have a significant impact on transport at high qmin, often leading to performance-limiting beta collapses. Thus, this proposal addresses a key directive from the 2013 run time guidance memo, on "Understanding and improving transport in steady state high qmin scenarios".
Previous experiments have shown that the rotation-dependence of the driven plasma response is consistent with the predictions of a theory that includes kinetic modifications to ideal MHD [H. Reimerdes, et al, Phys. Rev. Lett. 106 (2011) 215002], and it is expected that unstable RWMs will exhibit a similar sensitivity. This experiment will provide crucial stability threshold data for comparison with predictions of kinetic stability codes such as MISK and MARS-K. Control of the driven plasma response has been demonstrated using NBI feedback below the no-wall beta limit [J. M. Hanson, et al, Nucl. Fusion 52 (2012) 013003]. However, control above the no-wall limit is expected to be more challenging due to the increased importance of kinetic effects, and the non-linear dependence of the plasma response on plasma stored energy. An assessment of this control technique in a regime near the RWMâ??s marginal stability point will establish its usefulness for disruption avoidance. | ||
| Resource Requirements: | At least 7/8 NBI sources, including both 210 sources | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | Kinetic equilibrium reconstructions, ideal MHD and kinetic stability calculations | ||
| Other Requirements: | |||
| Title | 268: nu* dimensionless scaling | ||
| Name: | Joffrin none | Affiliation: | CEA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Luce, C. Challis | ITPA Joint Experiment : | Yes |
| Description: | JET and ASDEX are planning to include in their campaigns in 2013. In JET, dedicated pulses to this studies are planned in both the baseline scenario (q95~3, betaN~2) and the advance inductive scenario (q95~4, betaN~3). It is proposed to produce in DIII-D 2 pairs of points at 2 different beta at a rho* that can be accessible by JET. A match in rotation should be also achieved in this experiment by changing the beam energy. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In this type of experiment 2 pairs of points at slightly different field need to be obtained at same density same bea and same rho*. Only a small variation in field is needed (20% in BT, makes a change in collisionality of more than a factor of 2). DIII-D has already produced identity discharges with JET in the rho* scan experiment (see Politzer et al IAEA 2010). To minimise, the set-up time, it is proposed here to redo and use the best references at low rho* as a pivotal point for the nu* scan and infer the "baseline" point (q95~3; betaN~2) from this. Then to lower the field by ~15 to 20% to get the nu* step in DIII-D. | ||
| Background: | Analysis of the ITPA advanced inductive scenario performance database indicated that there is a substantial increase in H98y2 with decreasing collisionality. Dedicated experiments are required to assess the impact on projection to ITER, since even a weak variation of confinement with collisionality can have a substantial influence, due to ITER lying at more than an order of magnitude lower collisionality than present-day experiments. The dependence in nu* is particularly important for the extrapolation to ITER and the particle transport extrapolation which is now crucial also for impurities.
Both JET and ASDEX Uprade are planning this experiment in their programme in 2013. | ||
| Resource Requirements: | Neutral beam at different voltage
ECRH for adjusting Ti/Te ratio Max toroidal field JET shape (already achieved in 2010 and 2011 | ||
| Diagnostic Requirements: | - charge exchange
- Thomson scattering - ECE - MSE if possible to check q profile differences. | ||
| Analysis Requirements: | - EFIT/TRANSP in a first phase
- ONETWO | ||
| Other Requirements: | |||
| Title | 270: L-mode Turbulence and L-H Transition Characterization in Snowflake Divertor Plasmas | ||
| Name: | Umansky none | Affiliation: | LLNL |
| Research Area: | Divertor & SOL Physics | Presentation time: | Requested |
| Co-Author(s): | B. I. Cohen, D. Ryutov, V. Soukhanovskii, S. L. Allen, J. Boedo, R. Groebner, M. Makowski, G. McKee, T. Petrie | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to assess the influence of the Snowflake divertor configuration on characteristics of L-mode turbulence in the edge/SOL region and L-H transition. The experiments proposed would address the influence of the Snowflake divertor configuration on edge/SOL turbulence and compare to BOUT simulations extended to the Snowflake configuration. This builds on a successful campaign that used data from a well-characterized series of DIII-D L-mode plasmas in standard divertor configuration to validate BOUT nonlinear simulations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use standard L-mode shot with 1-1.5 MW ECH heating and do the snowflake transition and fast probe measurements and NBI blips for beam-assisted diagnostics (CER, BES). Measure turbulent fluctuations in the edge/SOL both in the outboard mid-plane and in the region of the X point(s) at the bottom of the device with scanning Langmuir probe and BES. Then raise the power further to transition to H-mode and monitor the power threshold and other characteristics of the L-H transition. | ||
| Background: | Recent experimental work on the Snowflake divertor configuration at DIII-D, TCV, and NSTX have demonstrated the beneficial effects of the Snowflake on reducing the heat flux density on the divertor plates by a higher flux expansion and flux splitting between a larger number of strike points. DIII-D snowflake experiments so far were done by first establishing an H-mode, and then transitioning to snowflake, so there is no L-mode snowflake data available so far. Recent BOUT fluid simulations and validation on DIII-D edge/SOL measurements (Cohen et al, APS DPP 2012, submitted to Phys. Plasmas) have been successful in characterizing and understanding L-mode edge/SOL turbulence and anomalous transport, but the influence of the Snowflake divertor on edge/SOL turbulence remains to be measured and simulated. The L-H transition power was found to be insensitive to Snowflake in TCV although ELM characteristics were affected (Piras et al, PRL 105, 155003 (2010)). The physics perhaps is that ELMs have low toroidal mode numbers and they can feel the null-point region while L-mode edge fluctuations have higher mode numbers and cannot extend poloidally to the null-point. | ||
| Resource Requirements: | Standard L-mode plasma with standard and Snowflake divertor configurations, SOL and divertor measurements, all fast edge, SOL and divertor diagnostics. This could be done in a half-day experiment. | ||
| Diagnostic Requirements: | Fast probe measurements and NBI blips for beam-assisted diagnostics (CER, BES). Measure turbulent fluctuations in the edge/SOL both in the outboard mid-plane and in the region of the X point(s) at the bottom of the device with scanning Langmuir probe and BES. | ||
| Analysis Requirements: | EFIT to reconstruct plasma profiles. BOUT simulations (LLNL) to compare to experiment. | ||
| Other Requirements: | |||
| Title | 271: Investigation of Momentum transport with Modulated ECH and ECH resonance layer sweeping | ||
| Name: | Shi none | Affiliation: | National Fusion Research Institute (NFRI), Korea |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | W.H.Ko, K.Ida, J.M.Kwon, P.H.Diamond, S.H.Koh, S.H.Hahn , W. Solomon, G.McKee, J. DeGrassie | ITPA Joint Experiment : | Yes |
| Description: | Radial scan of ECH resonance location during a single shot
Goal is to optimize V_Ï?=0 radial location Involve turbulence measurement to explore trapped particle population effects on TEM and rotation Comparison with KSTAR modulation experiments Related core poloidal rotation anomaly: reversal upon ITG â??â?? TEM transition, as predicted by simulation and theory. Modulation Obtain T_e,n_e,V_Ï? profile responses (especially density profile effect) Determine relative hysteresis â??n_e vs V_Ï? , â??T_e vs â??V_Ï? , â??T_e (axis),â??n_e (axis) vs â??V_Ï? (pivot). What is fundamental to VÏ? flattening , â??n_e or â??Te ? How does change in rotation profile evolve? Analysis of momentum flux in detail: For momentum flux given in Î =-Ï?_Ï? â??V_Ï?+V_TEP V_Ï?+Î _Resid, find Ï?_Ï? and Î _resid, given use of model for VTEP. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The background plasma for this proposal is heated with balanced NBI for minimum external momentum torque, which is similar to the low external torque situation of ITER. We donâ??t need too high NBI power, which should be hold at fixed power at the level to sustain H-mode. ECH is injected at the flat top of H-mode phase (quasi steady-state). During injection of ECH, the plasma should be well controlled and main parameters very stable. This proposal includes two steps (ECH resonance layer sweeping and ECH modulation). Firstly, we will do ECH resonance layer sweeping. The ECH layer should be swept from on-axis (Ï?~0) to off-axis (Ï?~0.6) during the flattop of one discharge. If the flattop is long enough, ECH layer can be swept back from off-axis to on-axis. Based on the ECH layer sweeping results, we can find the optimized ECH layer (maximum change of rotation and turbulence transition) for the modulation experiment. The modulation frequency of ECH is about 20Hz and the duty cycle is 50%. For modulation experiment, the first EC beam is aimed at optimized layer based on layer sweeping experiment. The second EC beam is aimed at off-axis (Ï?~0.6). The first EC beam should be turn on first for 10 pulses, and turn off. Then, the second EC beam turn on for 20 cycles if flattop is long enough. The first EC beam should be turn on again during the last 10 cycles of the second EC beam. | ||
| Background: | Toroidal rotation is important for control of stability and transport in tokamaks. While NBI is used widely to control rotation in todayâ??s tokamaks, it is not a feasible approach for ITER. Moreover, the tendency of confined tokamak plasmas to self-accelerate to a state of intrinsic rotation has been identified and related to the state of plasma confinement. Intrinsic rotation is self-generated by ambient turbulence via the non-diffusive residual stress. This, then, motivates the question of how macroscopic rotation profiles will evolve in response to changes in the ambient micro-turbulence. One â??control knobâ?? for the micro-turbulence population is the heating mix of NBI (heats ion, and so drives ITG) and ECH (heats electrons, especially at lower density, and so can drive CTEM). On KSTAR, both XICS and CES confirm that the core toroidal rotation dramatically decreases when modest amount of on-axis ECH is injected to H-mode plasmas. Both the change of rotation and its gradient have a close relation to the change of electron temperature and its gradient in the core plasma. The change in rotation toward the counter-direction by ECH in KSTAR is explained by the turbulence change from ITG to CTEM. Gyrokinetic simulations support aspect of an ITG-TEM transition with peaked density profiles during ECH injection (there is no density profile available in KSTAR). In KSTAR, we also found the decrease of V_Ï? also depends on the ECRH deposition location â?? i.e. |Î?V_Ï? | is larger for on axis deposition than that for off-axis ECRH. DIII-D has powerful NBI, ECH, all relevant profiles and fluctuation diagnostics. Due to the unique balance NBI injection on DIII-D, we hope to make a rotation profiles of the form: counter-current direction (counter intrinsic torque by ECH) in core , co-current direction in edge (co- intrinsic torque due to H-mode pedestal). Here, we want to change the V_Ï?=0 location with a scan of the ECH resonance layer. Turbulence diagnostics should be applied to explore trapped particle population effects on TEM and rotation during ECH resonance layer scan. The poloidal rotation anomaly is predicted to reversal when ITGâ?? TEM transition occurs. For ECH modulation proposal, we expect to obtain Ï?_Ï? and Î _resid directly from experiment. And we found a close relation between Te, ne and VÏ? in KSTAR ECH experimental. With ECH modulation, more detail and accurate between T_e,n_e,V_Ï? profile responses (especially density profile effect) to ECH can obtained. Futhermore, the relative hysteresis of â??n_e vs V_Ï? , â??T_e vs â??V_Ï? , â??T_e (axis),â??n_e (axis) vs â??V_Ï? (pivot) will also be investigated. | ||
| Resource Requirements: | NBIs, ECH, 5 shots might be desired | ||
| Diagnostic Requirements: | All profiles and all fluctuation diagnostics, especially core BES and ECE-I, high speed CER for toroidal and poloidal rotation of full profile, Thomson scattering and microwave reflectometry for density profile, DBS, PCI, | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 272: Dither injection for closed-loop system identification of vacuum and plasma response | ||
| Name: | Frassinetti none | Affiliation: | KTH |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | K.E.J. Olofsson (Columbia), F.A. Volpe (Columbia)
P.R. Brunsell (KTH), J.R. Drake (KTH) |
ITPA Joint Experiment : | No |
| Description: | The proposal aims at using the I-coils for Feedback control of RWMs and simultaneous measurement of their growth rates by means of â??dither injectionâ?? (an automatic control technique largely used in the industry). The technique has the advantage of estimating the growth rates in a non-disruptive way while inherently taking into consideration any 3D effect of the wall conductive structures on the plasma.
If successful on DIII D, the technique can be proposed as a valid tool in ITER and in future steady state fusion reactors to monitor in real time the RWM growth rate while simultaneously suppressing the instability with the feedback coils. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The envisaged application of dither injection to RWM control at DIII-D is as follows:
- the I-coils will be used to control the RWM with one of the algorithms commonly used at DIII-D - a small perturbation (random in time and space) will be applied to the voltages that drive the I-coils current. - the system response (plasma plus any conductive structure present in the wall) will be measured by the sensor coils - Post processing analysis will allow the RWM growth. The technique will initially be applied to vacuum shots, both as a preliminary test as well as to obtain information on the vacuum response. Dithering will be applied to both the I- and C-coils. In particular using the external C-coils will allow estimating the wall diffusion time [Olofsson et al., 2012, Fus. Eng. Design, in press]. The technique will then be tested on few plasma shots using a feedback algorithm normally deployed at DIII-D [for example as in Okabayashi et al., 2009 Nucl. Fusion 49, 125003]. If successful, we propose to then apply the dithering to several plasma shots with different beta in order to study in closed-loop the RWM growth-rate dependence. The technique inherently takes into consideration the 3D effects of the conductive structure. This allows a non-disruptive experimental cross-check of VALEN predictions [Okabayashi et al., 2005 Nucl. Fusion 45, 1715]. | ||
| Background: | Dither injection is a well-known technique in automatic control and it is largely applied in industry [e.g. Y. Zhu, Multivariable System Identification for Process Control (Elsevier Science 2001)] where quantitative models are required for the improvement of plant operations. It consists in the application of random perturbations to the signals in input to a physical system. The perturbations are usually generated as the output of a (designed) low-pass filter fed with white noise. By analyzing the response of the system (i.e. the relation between the input signals and output signals) it may be possible to identify the physical properties of the system. The technique is in general useful for the assessment of linear model applicability and prospecting the time-domain prediction horizon of linear models (potentially useful for routine plasma control).
Recently, dither injection was successfully applied to the system-identification of RWM growth rates in the EXTRAP T2R reversed field pinch [Olofsson et al., 2012, Fus. Eng. Design, in press]. Preliminary tests were also carried out on RFX-mod. In EXTRAP T2R, dither injection allowed suppressing a large range of unstable RWMs avoiding disruptions and prolonging the discharge while simultaneously identifying the RWM growth rate. The estimated growth rates are in very good agreement with theoretical expectations, proving the validity of the method [Olofsson et al., 2011 Plasma Phys. Controll. Fusion 53, 084003]. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 273: Sawtooth control via n=1 external magnetic perturbation | ||
| Name: | Martin none | Affiliation: | Consorzio RFX and Padova U |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | J. Bialek, D. Bonfiglio. A. Garofalo, G. Jackson, J. Hanson, M. J. Lanctot, E. Lazarus, G. A. Navratil, M. Okabayashi, C. Paz-Soldan, P.Piovesan, E. Strait, F. Turco, A. Turnbull | ITPA Joint Experiment : | No |
| Description: | Sawtooth control is a major challenge for tokamak. While on the one hand sawtooth might be desirable for example for the removal of helium ash, on the other hand large sawtooth crashes, like those that happen in presence of fast ions â?? with their stabilizing effect -, may trigger NTMs and eventually lead to disruptions. For this reason several techniques are used to reduce sawtooth amplitude making them more frequent. ECRH and ECCD are effective tools for this purpose, since they can tailor locally the current density profile. But they rely on rather complex and expensive heating and current drive systems.
With this experiment we propose to demonstrate for the first time in a large tokamak that the amplitude of sawteeth can be drastically and reproducibly reduced by means of a small amplitude n=1 magnetic perturbation. This proposal is based on a successful experiment realized in RFX-mod run as a tokamak (e.g. Martin et al, IAEA Overview 2012, Bonfiglio and Piovesan APS-DPP invited talk 2012). In one of the experiments where the RFX tokamak was run at qedge<2 by means of active feedback on the (2,1) RWM, the amplitude b_21 of the (2,1) mode was kept at a low but finite value (â??0.05% of the main field). While with full cancellation of the b_21 mode the plasma was sawtoothing as usual, keeping b_21 at small finite amplitude via the feedback system allowed to run a flat-topped plasma with very small sawteeth. In this new condition a quasi-continuous helical core replaces the sawtoothing (1,1) internal mode. Experiments also reproducibly showed that the amplitude of sawtooth is inversely proportional to the b_21 amplitude. This behavior is reproduced in toroidal geometry by the non-linear MHD code PIXIE3D (Bonfiglio, et al APS-DPP 2012). PIXIE3D shows a bifurcation to a quasi-stationary helical equilibrium with q(o)~1 and sawtooth motigation. The code also shows that what matters most for the success of the mitigation is the n=1 character of the magnetic perturbation, and indicates that a (1,1) perturbation is indeed more effective than a (2,1) perturbation. The (2,1) is coupled due to tooridal effects. Results agree with recent 3D VMEC equilibrium calculations in toroidal geometry by Lazarus (IAEA 2012), who shows the existence of an equilibrium with q(0)<<1 and a saturated core toroidal kink. For DIII-D we plan to apply a n=1 external perturbation (with I- or C-coils), starting from very low value, and monitor the sawtooth behavior. This will be done starting in L-mode, but H-mode will be targeted as well if L-mode works. In particular, it is planned to create the conditions for sawteeth of various amplitude tailoring the current profile with ECRH/ECCD and to test the proposed control method also with large sawteeth. This experiment has the potential to be transformational, since it would provide a simple and robust tool for the control and tailoring of sawteeth in large tokamaks, which in principle might be considered also for ITER |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Even if the original RFX experiment was made in a qedge<2 feedback stabilized regime, exploiting the residual amplitude of the unstable (2,1) mode as perturbation, this might not be the best approach to start with in DIII-D.
L-mode The experiment will proceed according to the following steps. a) target plasma. q95<2 operation in DIII-D has in fact been obtained, but not yet in stationary conditions. A more robust target plasma is that developed in 2012 by Turco et al. for the q95<2 campaign, corresponding for example to #150578. These L-mode plasmas have q95=2.2 and are appear to be rather stable, with a nice flat-top. To this target plasma we plan to apply a controlled n=1 external magnetic perturbation and monitor the sawtooth behavior (amplitude and period). b) quick scan with perturbation ramp. Applying an n=1 perturbation ramp, which starts from zero, is a quick approach to check if the effect is present and, if yes, to isolate a range of perturbation amplitude that is effective and not degrade plasma performance. The relatively outer q=1 surface might facilitate the coupling of the external perturbation. c) flat-topped discharges with constant values of the perturbation. Once this is done and an effect similar to RFX is seen, one performs several discharges applying various levels of the perturbation amplitude to demonstrate that flat-topped plasma with very low or no sawtooth can be run in DIII-D. d) q95 scan. In the final step of the L-mode approach, the experiment is repeated using target plasmas with different (and increasing) values of q95, to check whether the process is reproducible as a function of q95. H-mode If the experiment is successful in L-mode, a second stage will aim at reproducing it in H-mode. a) target plasma. if possible, a target plasma with q95=2.2 similar to L-mode will be used. Otherwise higher q95 will be the starting point. b) quick scan with perturbation ramp. As in L-mode. c) flat-topped discharges with constant values of the perturbation. As in L-mode d) giant sawteeth regime. If previous steps are successful, a target plasma with ITER-like equilibria will be produced. Sawteeth of different period/amplitude will be produced controlling the current density profile via ECRH/ECCD, and we will check whether the method is applicable also in this case. | ||
| Background: | The experiments in RFX-mod, the simulations by Bonfiglio and Lazarus (see Sect. 4).
This proposal is linked to proposal #56 "sawtooth control" submitted to the ROF 2013 by Lazarus et al and detailed in the miniproposal 2012-99-50 | ||
| Resource Requirements: | 1 â?? 2 co-Ip NBI sources, audio amplifiers on I-coils, SPA supplies on C-coils.
NBI and ECCD/ECRH for H-mode | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, ECE radiometer, density interferometer, Thomson scattering SXR tomography. | ||
| Analysis Requirements: | Tomography reconstruction. | ||
| Other Requirements: | |||
| Title | 274: Investigation of Momentum transport with Modulated ECH and ECH resonance layer sweeping | ||
| Name: | Shi none | Affiliation: | National Fusion Research Institute |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | W.H.Ko, K.Ida, J.M.Kwon, P.H.Diamond, S.H.Koh, S.H.Hahn , W. Solomon, G.McKee, J. DeGrassie | ITPA Joint Experiment : | Yes |
| Description: | First target of this proposal is Radial scan of ECH resonance location during a single shot. There four purpose of radial scan of ECH.
(1)Goal is to optimize V_Ï?=0 radial location (2)Involve turbulence measurement to explore trapped particle population effects on TEM and rotation (3)Comparison with KSTAR modulation experiments (4)Related core poloidal rotation anomaly: reversal upon ITG-TEM transition, as predicted by simulation and theory. The second target is ECH modulation. There three items for modulation experiment. (1)Obtain T_e,n_e,V_Ï? profile responses (especially density profile effect) (2)Determine relative hysteresis â??n_e vs V_Ï? , â??T_e vs â??V_Ï? , â??T_e (axis),â??n_e (axis) vs â??V_Ï? (pivot). What is fundamental to V_Ï? flattening , â??n_e or â??Te ? How does change in rotation profile evolve? (3)Analysis of momentum flux in detail: For momentum flux given in Î =-Ï?_Ï?*â??V_Ï?+V_TEP* V_Ï?+Î _Resid, find Ï?_Ï? and Î _resid, given use of model for V_TEP. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The background plasma for this proposal is heated with balanced NBI for minimum external momentum torque, which is similar to the low external torque situation of ITER. We donâ??t need too high NBI power, which should be hold at fixed power at the level to sustain H-mode. ECH is injected at the flat top of H-mode phase (quasi steady-state). During injection of ECH, the plasma should be well controlled and main parameters very stable. This proposal includes two steps (ECH resonance layer sweeping and ECH modulation). Firstly, we will do ECH resonance layer sweeping. The ECH layer should be swept from on-axis (ï?²~0) to off-axis (ï?²~0.6) during the flattop of one discharge. If the flattop is long enough, ECH layer can be swept back from off-axis to on-axis. Based on the ECH layer sweeping results, we can find the optimized ECH layer (maximum change of rotation and turbulence transition) for the modulation experiment. The modulation frequency of ECH is about 20Hz and the duty cycle is 50%. For modulation experiment, the first EC beam is aimed at optimized layer based on layer sweeping experiment. The second EC beam is aimed at off-axis (ï?²~0.6). The first EC beam should be turn on first for 10 pulses, and turn off. Then, the second EC beam turn on for 20 cycles if flattop is long enough. The first EC beam should be turn on again during the last 10 cycles of the second EC beam. | ||
| Background: | Toroidal rotation is important for control of stability and transport in tokamaks. While NBI is used widely to control rotation in todayâ??s tokamaks, it is not a feasible approach for ITER. Moreover, the tendency of confined tokamak plasmas to self-accelerate to a state of intrinsic rotation has been identified and related to the state of plasma confinement. Intrinsic rotation is self-generated by ambient turbulence via the non-diffusive residual stress. This, then, motivates the question of how macroscopic rotation profiles will evolve in response to changes in the ambient micro-turbulence. One â??control knobâ?? for the micro-turbulence population is the heating mix of NBI (heats ion, and so drives ITG) and ECH (heats electrons, especially at lower density, and so can drive CTEM). On KSTAR, both XICS and CES confirm that the core toroidal rotation dramatically decreases when modest amount of on-axis ECH is injected to H-mode plasmas. Both the change of rotation and its gradient have a close relation to the change of electron temperature and its gradient in the core plasma. The change in rotation toward the counter-direction by ECH in KSTAR is explained by the turbulence change from ITG to CTEM. Gyrokinetic simulations support aspect of an ITG-TEM transition with peaked density profiles during ECH injection (there is no density profile available in KSTAR). In KSTAR, we also found the decrease of V_Ï? also depends on the ECRH deposition location â?? i.e. |Î?V_Ï? | is larger for on axis deposition than that for off-axis ECRH. DIII-D has powerful NBI, ECH, all relevant profiles and fluctuation diagnostics. Due to the unique balance NBI injection on DIII-D, we hope to make a rotation profiles of the form: counter-current direction (counter intrinsic torque by ECH) in core , co-current direction in edge (co- intrinsic torque due to H-mode pedestal). Here, we want to change the V_Ï?=0 location with a scan of the ECH resonance layer. Turbulence diagnostics should be applied to explore trapped particle population effects on TEM and rotation during ECH resonance layer scan. The poloidal rotation anomaly is predicted to reversal when ITGï? TEM transition occurs. For ECH modulation proposal, we expect to obtain Ï?_Ï? and Î _resid directly from experiment. And we found a close relation between Te, ne and Vï?¦ in KSTAR ECH experimental. With ECH modulation, more detail and accurate between T_e,n_e,V_Ï? profile responses (especially density profile effect) to ECH can obtained. Futhermore, the relative hysteresis of â??n_e vs V_Ï? , â??T_e vs â??V_Ï? , â??T_e (axis),â??n_e (axis) vs â??V_Ï? (pivot) will also be investigated. | ||
| Resource Requirements: | NBIs,
ECH, 5 shots might be desired | ||
| Diagnostic Requirements: | All profiles and all fluctuation diagnostics, especially core BES and ECE-I, high speed CER for toroidal and poloidal rotation of full profile, Thomson scattering and microwave reflectometry for density profile, DBS, PCI, | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 275: Investigation of Momentum transport with Modulated ECH and ECH resonance layer sweeping | ||
| Name: | Shi none | Affiliation: | National Fusion Research Institute |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | W.H.Ko, K.Ida, J.M.Kwon, P.H.Diamond, S.H.Koh, S.H.Hahn , W. Solomon, G.McKee, J. DeGrassie | ITPA Joint Experiment : | Yes |
| Description: | First target of this proposal is Radial scan of ECH resonance location during a single shot. There four purpose of radial scan of ECH.
(1)Goal is to optimize V_Ï?=0 radial location (2)Involve turbulence measurement to explore trapped particle population effects on TEM and rotation (3)Comparison with KSTAR modulation experiments (4)Related core poloidal rotation anomaly: reversal upon ITG-TEM transition, as predicted by simulation and theory. The second target is ECH modulation. There three items for modulation experiment. (1)Obtain T_e,n_e,V_Ï? profile responses (especially density profile effect) (2)Determine relative hysteresis â??n_e vs V_Ï? , â??T_e vs â??V_Ï? , â??T_e (axis),â??n_e (axis) vs â??V_Ï? (pivot). What is fundamental to V_Ï? flattening , â??n_e or â??Te ? How does change in rotation profile evolve? (3)Analysis of momentum flux in detail: For momentum flux given in Î =-Ï?_Ï?*â??V_Ï?+V_TEP* V_Ï?+Î _Resid, find Ï?_Ï? and Î _resid, given use of model for V_TEP. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The background plasma for this proposal is heated with balanced NBI for minimum external momentum torque, which is similar to the low external torque situation of ITER. We donâ??t need too high NBI power, which should be hold at fixed power at the level to sustain H-mode. ECH is injected at the flat top of H-mode phase (quasi steady-state). During injection of ECH, the plasma should be well controlled and main parameters very stable. This proposal includes two steps (ECH resonance layer sweeping and ECH modulation). Firstly, we will do ECH resonance layer sweeping. The ECH layer should be swept from on-axis (ï?²~0) to off-axis (ï?²~0.6) during the flattop of one discharge. If the flattop is long enough, ECH layer can be swept back from off-axis to on-axis. Based on the ECH layer sweeping results, we can find the optimized ECH layer (maximum change of rotation and turbulence transition) for the modulation experiment. The modulation frequency of ECH is about 20Hz and the duty cycle is 50%. For modulation experiment, the first EC beam is aimed at optimized layer based on layer sweeping experiment. The second EC beam is aimed at off-axis (ï?²~0.6). The first EC beam should be turn on first for 10 pulses, and turn off. Then, the second EC beam turn on for 20 cycles if flattop is long enough. The first EC beam should be turn on again during the last 10 cycles of the second EC beam. | ||
| Background: | Toroidal rotation is important for control of stability and transport in tokamaks. While NBI is used widely to control rotation in todayâ??s tokamaks, it is not a feasible approach for ITER. Moreover, the tendency of confined tokamak plasmas to self-accelerate to a state of intrinsic rotation has been identified and related to the state of plasma confinement. Intrinsic rotation is self-generated by ambient turbulence via the non-diffusive residual stress. This, then, motivates the question of how macroscopic rotation profiles will evolve in response to changes in the ambient micro-turbulence. One â??control knobâ?? for the micro-turbulence population is the heating mix of NBI (heats ion, and so drives ITG) and ECH (heats electrons, especially at lower density, and so can drive CTEM). On KSTAR, both XICS and CES confirm that the core toroidal rotation dramatically decreases when modest amount of on-axis ECH is injected to H-mode plasmas. Both the change of rotation and its gradient have a close relation to the change of electron temperature and its gradient in the core plasma. The change in rotation toward the counter-direction by ECH in KSTAR is explained by the turbulence change from ITG to CTEM. Gyrokinetic simulations support aspect of an ITG-TEM transition with peaked density profiles during ECH injection (there is no density profile available in KSTAR). In KSTAR, we also found the decrease of V_Ï? also depends on the ECRH deposition location â?? i.e. |Î?V_Ï? | is larger for on axis deposition than that for off-axis ECRH. DIII-D has powerful NBI, ECH, all relevant profiles and fluctuation diagnostics. Due to the unique balance NBI injection on DIII-D, we hope to make a rotation profiles of the form: counter-current direction (counter intrinsic torque by ECH) in core , co-current direction in edge (co- intrinsic torque due to H-mode pedestal). Here, we want to change the V_Ï?=0 location with a scan of the ECH resonance layer. Turbulence diagnostics should be applied to explore trapped particle population effects on TEM and rotation during ECH resonance layer scan. The poloidal rotation anomaly is predicted to reversal when ITGï? TEM transition occurs. For ECH modulation proposal, we expect to obtain Ï?_Ï? and Î _resid directly from experiment. And we found a close relation between Te, ne and V_Ï? in KSTAR ECH experimental. With ECH modulation, more detail and accurate between T_e,n_e,V_Ï? profile responses (especially density profile effect) to ECH can obtained. Futhermore, the relative hysteresis of â??n_e vs V_Ï? , â??T_e vs â??V_Ï? , â??T_e (axis),â??n_e (axis) vs â??V_Ï? (pivot) will also be investigated. | ||
| Resource Requirements: | NBIs,
ECH, 5 shots might be desired | ||
| Diagnostic Requirements: | All profiles and all fluctuation diagnostics, especially core BES and ECE-I, high speed CER for toroidal and poloidal rotation of full profile, Thomson scattering and microwave reflectometry for density profile, DBS, PCI, | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 276: Investigation of Momentum transport with power ratio P_ECH/P_NBI and torque ratio scan | ||
| Name: | Ida none | Affiliation: | National Institute for Fusion Science, Toki, Japan |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | J.M.Kwon,W.H.Ko,S.H.Hahn,Y.J.Shi, P.H.Diamond, S.H.Koh, W. Solomon, B. Grierson,G.Tynan, G.McKee | ITPA Joint Experiment : | Yes |
| Description: | (A)scan of P_ECH with fixed Ï?_NBI=0
a.1)low or almost zero torque in ITER, where V_Ï? null point falls with ECH? a.2)scan of T_e/T_i , and effect on intrinsic torque a.3)scan of core intrinsic torque / pedestal intrinsic torque ration a.4)turbulence studies focused on studies mixed ITG/TEM state, transport analysis to obtain residual stress using Ï?_Ï?~Ï?_i and TEP pinch model a.5)study of poloidal rotation anomaly in higher P_ECH (B) scan of Ï?_NBI/P_ECH with fixed P_NBI b.1)large Ï?_NBI/P_ECH was covered in KSTAR, how about low Ï?_NBI/P_ECH in DIII-D? b.2)fluctuation studies: ITG ï? TEM transition b.3)extraction of Î _resid from transport analysis (see above) |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This proposal includes two parts: power ratio(P_ECH/P_NBI) with balanced NBI, and torque ratio(Ï?_NBI/P_ECH) with fixed NBI power and variable ECH power. The background plasma for the first part of the proposal is heated with balance NBI for minimum external momentum torque, which is similar to low external torque situation of ITER. ECH is injected at the flap top of H-mode phase (quasi steady-state). For the first step, the resonance layer of all EC beam should be aimed at optimized position, based on previous results. In order to achieve the power ratio PECH/PNBI scan in one discharge, the six gyrotrons should be turn on step by step. The time interval is 100ms. So we need at least 600ms flattop to do the power scan. If the flattop is long enough, the six gyrotrons can be turn off step by step. If we can get more discharges, we will do the PECH/PNBI scan at other ECH resonance layers. For the torque ratio scan, the power of NBI and ECH are fixed. All gyrotrons should be turn on simultaneously (aimed at the same resonance layer). In order to achieve the torque ratio scan in one discharge, the three co-NBI ion sources and two ctr-NBI ion sources will be used. Three equal power ion sources should be always turn for prososal. For the first 200ms time interval, three co-NBI ion sources should be turn on, for next 200ms, one co-NBI ion sources will be turn off and one ctr-NBI source will be turn on. For last 200ms, two co-NBI ion sources will be turn off and two ctr-NBI source will be turn on. | ||
| Background: | Many investigations of ECH effects on rotation exist, but a coordinated P_ECH/P_NBI and Ï?_NBI/ P_ECH scans, with fluctuation measurement, is not available up to now. | ||
| Resource Requirements: | NBIs,
ECH, 6 shots might be desired | ||
| Diagnostic Requirements: | All profiles and all fluctuation diagnostics, especially core BES and ECE-I, high speed CER for toroidal and poloidal rotation of full profile, Thomson scattering and microwave reflectometry for density profile, DBS, PCI, | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 277: ECH effects on low or balanced torque discharges in DIII-D | ||
| Name: | Ko none | Affiliation: | National Fusion Research Institute (NFRI), Korea |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | Y.J.Shi, K. Ida, W. Solomon, B. Grierson, P.H.Diamond, J.M.Kwon, S. H.Ko, S. H. Hahn, G. Tynan | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to measure the variation of rotation, density (peaking or pump-out), and turbulence characteristics due to the effect of ECH heating (H-mode plasma with low or no torque i.e. balanced NBI). We focus on turbulence studies in possible mixed ITG/TEM state and transport analysis to identify residual stress using Ï?Ï?~Ï?i and a TEP pinch model. We focus on counter-current flow induced by ECH in NBI heated H-mode plasma without torque or with low torque. Specifically, we will explore the effect of ECH injection into
i) A â??cancellationâ?? state â?? a base state formed by counter NBI + pedestal intrinsic torque cancellation. Here Vâ??0 (Solomon â??07) ECH will be applied, and the resulting residual rotation profile VÏ?(r) will be compared to Î?VÏ?(r) â?? the change in VÏ? induced in a co-NBI H-mode rotation profile when on-axis ECH is added. A key question is to understand how do residual VÏ? (r) and Î?VÏ?(r) compare and relate. ii) An OH state (a) first, with no ECH, scan density across LOC->SOC transition (and back) to determine correlations between reversal and TEM ->ITG->TEM evolution. This experiment effectively combines the Rettig, et. al. 2002 fluctuation study with OH reversal measurements. (b) next, with ECH, scan density to explore the change in sign of ECH-induced increment as one progresses from TEM to ITG dominated base state. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use balanced NBI and gyrotron to measure rotation and fluctuation effect on H-modes plasma
1) ECH into cancellation state (H-mode): initially counter-NBI + pedestal torque (=zero net rotation) -> after ECH, what VÏ?(r) results, and how does it compare to the increment? Use LOC/SOC (i.e. density scan) and different shape to measure rotation and fluctuation changes on Ohmic plasma with ECH heating 2) ECH into OH (CER measurement with balanced NBI blip) a) LOC -> SOC scan: rotation + fluctuation measurement (TEM -> ITG). Does reversal correlate with TEM<->ITG? b) ECH applied during density scan (i.e. LOC -> SOC base state) combined with different boundary conditions i.e. LSN, USN, DN. Recall TCV found boundary condition effect in reversal occurrence. | ||
| Background: | For ITER and future reactors, the input torque from NBI will be very low or nonexistent and cannot produce the needed rotation. As a result, there is a need to develop alternative or complementary methods for driving plasma rotation. Significant intrinsic rotation has been observed on a number of tokamaks and the core rotation profiles, which are centrally peaked in the pure NBI heating phase, flatten when ECH is injected, while the edge pedestal is unchanged in KSTAR.
The measurement of the rotation induced by ECH without external momentum input is important in order to understand the mechanism and prospect of intrinsic torque and so to improve understanding of the physics basis of rotation in low torque regimes. | ||
| Resource Requirements: | 1.0 run day might be desired.
Need all gyrotrons and balanced NBI Need balanced NBI blip for Ohmic discharge | ||
| Diagnostic Requirements: | Standard (ne, Ti, Te, rotation, Er profiles)
Nice to have main ion CER Density and temperature fluctuation diagnostics (BES and ECEI) High speed CER for toroidal and poloidal rotation Thomson scattering and microwave reflectometry for density profile, DBS, PCI | ||
| Analysis Requirements: | none | ||
| Other Requirements: | none | ||
| Title | 278: Possibility of â??Off-axis-Fishbone Mode and ELM Pacingâ?? due to Direct Energetic Particle Coupling | ||
| Name: | Matsunaga none | Affiliation: | JAEA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | M. Okabayashi, J. Ferron, C. Holcomb, T. Luce, D. Pace, F. Turco | ITPA Joint Experiment : | No |
| Description: | One important question of â??Off-axis-Fishbone Mode(OFM)-ELM pacingâ?? is whether the ELM is triggered directly by Energetic Particle(EP) or indirectly through pressure or current profile change caused by enhanced EP transport. In this proposal, we plan to look into a possible hypothesis of direct EP coupling to ELM. This hypothesis is based on the EP loss observation near the edge by BES and onset of ELM within relatively-short time period after higher toroidal harmonic appearance in OFM waveform.
Firstly, as pre-requisite, fundamental study of enhanced EP loss mechanism by OFM will be documented by Full EP diagnostics and following series of EP orbit calculation. Secondly, the role of EP is examined by enhancing EP component near the edge as much as possible, simultaneously reducing OFM activity. EP amount near edge will be controlled by various combinations of Off-axis NBI parameters. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To observe the role of EP near edge, we will utilize configurations more stable to OFM. Here, as pre-requisite, we carry out systematic numerical study of MHD stability analogous to classical Fishbone activity reduction. Our initial trial is higher triangularity with various q_profile flatness. Then, off-axis-NBI parametric scan injection-angle / energy will be included. One uncertainty will remain how to estimate the divertor functioning compared with the present configuration.
In the experiment, first step of fundamental study of enhanced EP loss by OFM will be documented by full EP diagnostics and series of EP orbit calculation. This operation requires the normal Bt operation for full use of EP diagnostics. Second step of the role of EP component near the edge is investigated using a configuration more stable to internal modes based on numerical studies discussed above. the EP amount near edge will be controlled by various combinations of Off-axis NBI parameters. In addition, Go Matsunaga is considering to carry out the ICH edge heating in AUG as independent assessment of EP near the edge contribution. | ||
| Background: | Last several years, the understanding of global MHD behavior in relation to the overall plasma performance has been greatly advanced experimentally and theoretically.
However, recent SSI experiments suggest that EPs play a significant role in such way that OFM events induce global MHD modes, which impact critically on the plasma performance. More importantly, the relation of EP-driven OFM to the RWM / ELM is identical in DIII-D and JT60U devices even with completely-different NBI arrangement. Even waveform develops in time non-linearly in an identical manner in these two devices. This suggests important fundamental process takes place in the relation between EP, EP-driven OFM and global MHD modes. EP pressure likely remains as major plasma content even up to ignition (like 1 MeV in ITER NBI). Thus, it is important to investigate how the EP couples to global mode. This proposal is planned to look into possible mechanism of recent unique observed â??OFM-ELM pacingâ??. The details are presented at IAEA 2012. There are two ways to hypothesize the OFM coupled to ELM. One thought is that the OFM causes the PTh or JTh near the edge, which increases the local gradient and induce. In this proposal, we plan to look into a hypothesis of direct EP coupling to ELM, as we discussed above. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 279: ECH effects on low or balanced torque discharges in DIII-D | ||
| Name: | Ko none | Affiliation: | National Fusion Research Institute (NFRI), Korea |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | Y.J.Shi, K.Ida, W.Solomon, B.Grierson, P.H.Diamond, J.M.Kwon, S.H.Ko, S.H.Hahn and G.Tynan | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to measure the variation of rotation, density (peaking or pump-out), and turbulence characteristics due to the effect of ECH heating (H-mode plasma with low or no torque i.e. balanced NBI). We focus on turbulence studies in possible mixed ITG/TEM state and transport analysis to identify residual stress using Ï?Ï?~Ï?i and a TEP pinch model. We focus on counter-current flow induced by ECH in NBI heated H-mode plasma without torque or with low torque. Specifically, we will explore the effect of ECH injection into
A â??cancellationâ?? state â?? a base state formed by counter NBI + pedestal intrinsic torque cancellation. Here Vâ??0 (Solomon â??07) ECH will be applied, and the resulting residual rotation profile VÏ? (r) will be compared to Î?VÏ? (r) â?? the change in VÏ? induced in a co-NBI H-mode rotation profile when on-axis ECH is added. A key question is to understand how do residual VÏ? (r) and Î?VÏ?(r) compare and relate. An OH state (a) first, with no ECH, scan density across LOC->SOC transition (and back) to determine correlations between reversal and TEM ->ITG->TEM evolution. This experiment effectively combines the Rettig, et. al. 2002 fluctuation study with OH reversal measurements. (b) next, with ECH, scan density to explore the change in sign of ECH-induced increment as one progresses from TEM to ITG dominated base state. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use balanced NBI and gyrotron to measure rotation and fluctuation effect on H-modes plasma
ECH into cancellation state (H-mode): initially counter-NBI + pedestal torque (=zero net rotation) -> after ECH, what V_Ï? (r) results, and how does it compare to the increment? Use LOC/SOC (i.e. density scan) and different shape to measure rotation and fluctuation changes on Ohmic plasma with ECH heating ECH into OH (CER measurement with balanced NBI blip) LOC ï? SOC scan: rotation + fluctuation measurement (TEM -> ITG). Does reversal correlate with TEM<->ITG? ECH applied during density scan (i.e. LOC -> SOC base state) combined with different boundary conditions i.e. LSN, USN, DN. Recall TCV found boundary condition effect in reversal occurrence. | ||
| Background: | For ITER and future reactors, the input torque from NBI will be very low or nonexistent and cannot produce the needed rotation. As a result, there is a need to develop alternative or complementary methods for driving plasma rotation. Significant intrinsic rotation has been observed on a number of tokamaks and the core rotation profiles, which are centrally peaked in the pure NBI heating phase, flatten when ECH is injected, while the edge pedestal is unchanged in KSTAR.
The measurement of the rotation induced by ECH without external momentum input is important in order to understand the mechanism and prospect of intrinsic torque and so to improve understanding of the physics basis of rotation in low torque regimes. | ||
| Resource Requirements: | 1.0 run day might be desired.
Need all gyrotrons and balanced NBI Need balanced NBI blip for Ohmic discharge | ||
| Diagnostic Requirements: | Standard (ne, Ti, Te, rotation, Er profiles)
Nice to have main ion CER Density and temperature fluctuation diagnostics (BES and ECEI) High speed CER for toroidal and poloidal rotation CER, Thomson scattering and microwave reflectometry for density profile, DBS, PCI | ||
| Analysis Requirements: | none | ||
| Other Requirements: | none | ||
| Title | 280: : Trigger-less onset of NTMs with PEC and NTV effects on rotation regime | ||
| Name: | Lazzaro none | Affiliation: | Istituto di Fisica del Plasma, Euratom-ENEA-CNR |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | E.Lazzaro,O.Sauter,S.Nowak,et al | ITPA Joint Experiment : | No |
| Description: | Test that modified rotation at rational surfaces in the regime predicted by the latest version of the theory [CONNOR, J.W. WAELBROECK F. L. and WILSON, H. R. Phys. Plasmas 8 (2001) 2835], causes NTMs even in the absence of triggers. This would be the first genuinely trigger-less mechanism for NTM onset and has an obvious relevance to ITER. Experiments on TCV show that NTMs can occur without seed island formation, in regimes of non-steady rotation caused by ECH near central power absorption and non-steady pressure and current density profiles [E. Lazzaro, S. Nowak, O. Sauter et al,Proc. FEC-IAEA 2012 , paper EX/P4-32]. The onset of the NTM is also apparently consistent and concomitant with the instability condition associated with the ion polarisation current (IPC). The growing tearing modes contribute a nonlinear magnetic braking that consistently modifies the rotation profile. This would be the first genuinely trigger-less mechanism for NTM onset and has an obvious relevance to ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Prepare a sawtooth-free target marginally stable (i.e., with beta slightly below critical) against 2/1 and 3/2 NTMs. Use n=3 RMPs to suppress ELMs, if necessary. Also make sure that fishbones or other NTM triggers are absent. Ramp down the toroidal rotation by means of balanced injection or magnetic braking. If successful, repeat with reverse Bt (but normal Ip), which would decouple the toroidal and poloidal rotation effects.
As recently observed on TCV (Lausanne) injection of PEC upstream of the q=3/2 surface, may both drive a 3/2 mode and a strong effect on plasma rotation. Even in absence of triggers subsequently a 2/1 NTM can appear, associated with local flattening of the rotation shear, and modification of 2/1 rotation that enters the IPC unstable regime. [LA HAYE R. J., PETTY., C. C., STRAIT E. J., WAELBROECK F. L., AND WILSON H. R., Phys. Plasmas 10 (2003) 3644] Therefore it is proposed to test: a) A direct effect of PEC (~ POhm) on (N)TM stability through modification of J profile (upstream of q=m/n) (1 good shot with ECH , 1 good shot with ECCD) b) An indirect effect of PEC on (N)TM stability through direct effect on rotation A simple theory involving mode rotation in the presence of NTV effects (due to the static applied RMP and the TM modes themselves) will help designing the magnetic braking aspects of the experiment. Note that braking could be in place anyway, if RMPs will be used for ELM control. | ||
| Background: | The transverse ion polarization current (IPC) can be stabilizing or destabilizing for NTMs, depending whether the rotation is faster or slower than a reference â??naturalâ?? value of the order of the ion diamagnetic drift frequency [E. Lazzaro, Proc.36th EPS Conference, Sofia (Bulgaria) 2009, paper P1.125].
Thus,Varying down the rotation of a q=m/n surface around that value can cause NTM onset even in absence of triggers (sawteeth, ELMs , fishbones...). | ||
| Resource Requirements: | In case of success of the first day or half day, possibility of a second day or half day with reverse Bt or Ip. | ||
| Diagnostic Requirements: | Magnetics, MSE(optional), ECE (both radiometer and Michelson), TS, Interferometer, CER. | ||
| Analysis Requirements: | EFIT, Tearing mode stability codes, optional reconstruction of NBI beam momentum deposition (TRANSP?), CXSR measurement of rotation. | ||
| Other Requirements: | Possible support of other codes: ECW ray tracing, and transport codes | ||
| Title | 281: ECH effects on low or balanced torque discharges in DIII-D | ||
| Name: | Ko none | Affiliation: | National Fusion Research Institute (NFRI), Korea |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | Y.J.Shi, K. Ida, W. Solomon, B. Grierson, P.H.Diamond, J.M.Kwon, S. H.Ko, S. H. Hahn, G. Tynan | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to measure the variation of rotation, density (peaking or pump-out), and turbulence characteristics due to the effect of ECH heating (H-mode plasma with low or no torque i.e. balanced NBI). We focus on turbulence studies in possible mixed ITG/TEM state and transport analysis to identify residual stress using Ï?_Ï?~Ï?_i and a TEP pinch model. We focus on counter-current flow induced by ECH in NBI heated H-mode plasma without torque or with low torque. Specifically, we will explore the effect of ECH injection into
A â??cancellationâ?? state â?? a base state formed by counter NBI + pedestal intrinsic torque cancellation. Here Vâ??0 (Solomon â??07) ECH will be applied, and the resulting residual rotation profile V_Ï?^r (r) will be compared to ã??Î?Vã??_Ï? (r) â?? the change in V_Ï? induced in a co-NBI H-mode rotation profile when on-axis ECH is added. A key question is to understand how do V_Ï?^r (r) and ã??Î?Vã??_Ï? (r) compare and relate. An OH state (a) first, with no ECH, scan density across LOCâ??SOC transition (and back) to determine correlations between reversal and TEM â??ITGâ??TEM evolution. This experiment effectively combines the Rettig, et. al. 2002 fluctuation study with OH reversal measurements. (b) next, with ECH, scan density to explore the change in sign of ECH-induced increment as one progresses from TEM to ITG dominated base state. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use balanced NBI and gyrotron to measure rotation and fluctuation effect on H-modes plasma
ECH into cancellation state (H-mode): initially counter-NBI + pedestal torque (=zero net rotation) â?? after ECH, what V_Ï? (r) results, and how does it compare to the increment? Use LOC/SOC (i.e. density scan) and different shape to measure rotation and fluctuation changes on Ohmic plasma with ECH heating ECH into OH (CER measurement with balanced NBI blip) LOC â?? SOC scan: rotation + fluctuation measurement (TEM â?? ITG). Does reversal correlate with TEM<->ITG? ECH applied during density scan (i.e. LOC â?? SOC base state) combined with different boundary conditions i.e. LSN, USN, DN. Recall TCV found boundary condition effect in reversal occurrence. | ||
| Background: | For ITER and future reactors, the input torque from NBI will be very low or nonexistent and cannot produce the needed rotation. As a result, there is a need to develop alternative or complementary methods for driving plasma rotation. Significant intrinsic rotation has been observed on a number of tokamaks and the core rotation profiles, which are centrally peaked in the pure NBI heating phase, flatten when ECH is injected, while the edge pedestal is unchanged in KSTAR.
The measurement of the rotation induced by ECH without external momentum input is important in order to understand the mechanism and prospect of intrinsic torque and so to improve understanding of the physics basis of rotation in low torque regimes. | ||
| Resource Requirements: | 1.0 run day might be desired.
Need all gyrotrons and balanced NBI Need balanced NBI blip for Ohmic discharge | ||
| Diagnostic Requirements: | Standard (ne, Ti, Te, rotation, Er profiles)
Nice to have main ion CER Density and temperature fluctuation diagnostics (BES and ECEI) High speed CER for toroidal and poloidal rotation Thomson scattering and microwave reflectometry for density profile, DBS, PCI | ||
| Analysis Requirements: | |||
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| Title | 282: Investigation of Momentum transport with power ratio PECH/PNBI and torque ratio Ï?NBI/ PECH scan | ||
| Name: | Ida none | Affiliation: | National Institute for Fusion Science, Toki, Japan |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | J.M.Kwon,W.H.Ko,S.H.Hahn,Y.J.Shi, P.H.Diamond, S.H.Koh, W. Solomon, B. Grierson,G.Tynan, G.McKee, | ITPA Joint Experiment : | Yes |
| Description: | scan of P_ECH with fixed Ï?_NBI=0
a.1)low or almost zero torque in ITER, where V_Ï? null point falls with ECH? a.2)scan of T_e/T_i , and effect on intrinsic torque a.3)scan of core intrinsic torque / pedestal intrinsic torque ration a.4)turbulence studies focused on studies mixed ITG/TEM state, transport analysis to obtain residual stress using Ï?_Ï?~Ï?_i and TEP pinch model a.5)study of poloidal rotation anomaly in higher P_ECH (B) scan of Ï?_NBI/P_ECH with fixed P_NBI b.1)large Ï?_NBI/P_ECH was covered in KSTAR, how about low Ï?_NBI/P_ECH in DIII-D? b.2)fluctuation studies: ITG â?? TEM transition b.3)extraction of Î _resid from transport analysis (see above) |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This proposal includes two parts: power ratio(PECH/PNBI) with balanced NBI, and torque ratio(Ï?NBI/ PECH) with fixed NBI power and variable ECH power. The background plasma for the first part of the proposal is heated with balance NBI for minimum external momentum torque, which is similar to low external torque situation of ITER. ECH is injected at the flap top of H-mode phase (quasi steady-state). For the first step, the resonance layer of all EC beam should be aimed at optimized position, based on previous results. In order to achieve the power ratio PECH/PNBI scan in one discharge, the six gyrotrons should be turn on step by step. The time interval is 100ms. So we need at least 600ms flattop to do the power scan. If the flattop is long enough, the six gyrotrons can be turn off step by step. If we can get more discharges, we will do the PECH/PNBI scan at other ECH resonance layers. For the torque ratio scan, the power of NBI and ECH are fixed. All gyrotrons should be turn on simultaneously (aimed at the same resonance layer). In order to achieve the torque ratio scan in one discharge, the three co-NBI ion sources and two ctr-NBI ion sources will be used. Three equal power ion sources should be always turn for prososal. For the first 200ms time interval, three co-NBI ion sources should be turn on, for next 200ms, one co-NBI ion sources will be turn off and one ctr-NBI source will be turn on. For last 200ms, two co-NBI ion sources will be turn off and two ctr-NBI source will be turn on. | ||
| Background: | Many investigations of ECH effects on rotation exist, but a coordinated PECH/PNBI and Ï?NBI/ PECH scans, with fluctuation measurement, is not available up to now. | ||
| Resource Requirements: | NBIs, ECH, 6 shots might be desired | ||
| Diagnostic Requirements: | All profiles and all fluctuation diagnostics, especially core BES and ECE-I, high speed CER for toroidal and poloidal rotation of full profile, Thomson scattering and microwave reflectometry for density profile, DBS, PCI, | ||
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| Other Requirements: | |||
| Title | 283: ITER Baseline Scenario Operation near the LH Threshold | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | The proposed operating point of the ITER baseline scenario for Q=10 operation has a loss power quite close to the LH power threshold scaling. This is known to lead to issues with density control, giant ELMs, and tearing instability. DIII-D is unique in being able to run q95=3 over a factor of 4 in B, which varies the loss power and the LH threshold in different ways. The hidden variable is density, which is typically also linear with B, so that the loss power and LH threshold power scaling vary somewhat closer than may be expected. However, the density scaling is uncertain and DIII-D can pump a shape nearly the same as the ITER shape, so the degeneracy can be broken. The proposal is to run the ITER baseline scenario over a range of B=0.5-2.1 T at the natural density and ~half that to see how the operating point changes. Time permitting, the points will be repeated at low applied torque. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | A two-shot look at this was done during the ITER baseline scenario experiment in 2012 and saw dramatic difference in the plasma behavior. This work contributes to the ITPA joint experiment IOS-1.3 (new for 2013). | ||
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| Title | 284: Access conditions for steady-state operation in ITER | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | A joint database is being compiled of advanced scenario plasmas from JET, DIII-D, AUG, MAST, and JT-60U to determine necessary and sufficient conditions for accessing current profiles demonstrated to be favorable for steady-state tokamak operation. Following the analysis, experiments will be proposed to test the conclusions of the analysis directly on various tokamaks. This would be a good application of the current profile controller for the current rise phase. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | This contributes to ITPA joint experiment IOS-3.2 | ||
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| Title | 285: Access to advanced inductive operation in ITER | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | A database of current rise data from AUG, DIII-D, JET, JT-60U, and MAST is being analyzed for determination of the necessary and sufficient conditions for access to advanced inductive operation in ITER. Following the analysis, experiments will be proposed to test the conclusions of the analysis by direct experiments on various tokamaks. A good example of this activity is the application of the current overshoot technique in JET. Because JET cannot apply auxiliary heating in the current rise like AUG and DIII-D typically do, they had to find another way to form the q profile (presumably) sufficient for access to advanced inductive regimes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | This experiment will contribute to ITPA joint experiment IOS-4.1 | ||
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| Title | 286: Collisionality scaling of confinement in advanced inductive regimes | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Confinement data in the ITPA joint database for advanced inductive plasmas indicates a strong variation of H98y2 with normalized collision frequency. From the database, it is not possible to determine if this is a real trend or an artifact of the operating space of the contributing tokamaks. However, some serendipitous scans in DIII-D indicate the trend is real. This would be very favorable for ITER, but also call into question projections to ITER using the IPB98y2 scaling. Therefore, this is an important issue for projecting present-day experiments to ITER. A classical collisionality scaling experiment is proposed (n constant as I and T are varied in proportion to B and B^2 respectively). We have successfully repeated such scans in L mode and H mode in the last 2 years, so it is expected that this is straightforward operationally. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | This contributes to ITPA joint experiment IOS-4.3 | ||
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| Title | 287: High li steady-state scenario development | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Continue 2012 experiments on high li with focus on mitigating the effect of the first ELM and making a smoother transition to the ELMing regime. Look for dependence on q95 of beta limit and confinement. See what current sources are needed for fully non-inductive operation after the first ELM. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | Continuation of successful experiments in 2012. | ||
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| Title | 288: Amplifying the Geodesic Acoustic Mode via Resonant Radial Field Amplification: New Methods | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | K. Hallatschek, A. Garofalo, J. Hanson, G. Jackson, Z. Yan | ITPA Joint Experiment : | No |
| Description: | NOTE: This is a follow-up and continuation of an 2011 TJA experiment
Amplify the naturally-occuring Geodesic Acoustic Mode, to control and suppress turbulence and associated transport near the plasma edge region while maintaining a non-ELMing L-mode condition. The goal is to achieve enhanced 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).
Establish a moderate q95 condition, e.g., Ip=1.0 MA, B_t=2.0 MA, q95~6.5 (144872). The I-Coil will be configured in an n=0, m=0 configuration (upper and lower coils in phase) and run near 15 kHz using the high-frequency Audio Amplifiers connected to the I-Coils. Also try at lower field/current (1.0 T, 0.5 MA) to enhance the relative amplitude of applied field to plasma fields: increase B_r/B_T. 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. | ||
| 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 is to generate a 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. An initial attempt of this experiment was performed in 2011 as the Torkil Jensen Award [McKee-May, 2011]. While the GAM was found and a radial field applied at the resonant frequency, no clear enhancement of the GAM was observed due to the resonant field. This may have been because the field amplitude was too low, or didn't have the best spectral mode structure. K. Hallatschek has since then performed simulations which suggest that the concept is viable, given adequate field amplitude, and suggests application of a different mode structure: even parity. | ||
| Resource Requirements: | I-coils configured in n=0, even parity configuration, connected to Audio Amps 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 | ||
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| Title | 289: Demonstrate access to beta_N=5 with min(q)>2 | ||
| Name: | Luce tim.luce@iter.org | Affiliation: | ITER Organization |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The motivation for going to min(q)>2 in steady-state scenario development is the prediction of access to high beta_N (>5) due to wall stabilization when li is very low. This key assumption has yet to be observed experimentally. As a proof of principle, use a B ramp to q95 between 3.5 and 4 to make the current profiles needed for the demonstration. Try to limit the end value of B to be >1.3 T to allow MSE measurements for stability analysis of reconstructed equilibria. But, do what is needed to get to beta_N>5. Propose unpumped double-null to maximize the beta_N limit and confinement, and minimize li. If successful, reduce current to give q95~4-4.5, then 5-5.5 to see if high beta_N is still possible. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
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| Title | 290: Effect of n=2 Poloidal Spectrum on NTV Torque | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | C. Paz-Soldan, S. Haskey, N. Logan | ITPA Joint Experiment : | No |
| Description: | Document the effect of the n=2 poloidal spectrum on NTV torque. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure plasma response and rotation changes as a function I-coil phasing with and without C-coil field. Measured rotation changes will be used to order poloidal spectra. Compare to theoretical predictions. Document fast ion losses. | ||
| Background: | Odd parity n=3 I-coil fields have been used to modify the edge rotation shear in QH-mode experiments. The inability to vary the poloidal spectrum of the n=3 field makes it impossible to validate NTV theory by looking at the effect of the poloidal spectrum on NTV torque. Such an experiment is possible using the n=2 configuration for C and I-coil. This experiment could most likely be done at low beta so that n=1 EFC may not be necessary, leaving the power supplies to drive only n=2 fields.
Extensive simulations of the plasma response using the MARS code have been completed to understand the effect of I-coil phasing, betan, and q95 on the plasma response. These will be used to guide the experiment and for validating the plasma response calculations. A potential benefit is the possibility that a net increase in the torque related to 3D fields (compared to n=3) could be realized by maximizing the NTV effect while minimizing magnetic braking from pitch-resonant harmonics. Such a result would inform future a variety of studies including co-injected QH-modes. This experiment parallels a variety of other ROF proposals so it can likely be combined with other efforts including efforts to understand the plasma response in RMP ELM suppressed H-modes. | ||
| Resource Requirements: | I&C coil in n=2 configuration. All C-supplies and SPAs | ||
| Diagnostic Requirements: | 3D magnetics. LLNL IR camera on periscope. Both BES diagnostics (UW and UCSD). Profile reflectometer. | ||
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| Other Requirements: | |||
| Title | 291: Trigger-less onset of NTMs with P_EC and NTV effects on rotation regime | ||
| Name: | Nave mfn@ipfn.ist.utl.pt | Affiliation: | Instituto Superior Tecnico, Lisboa, Portugal |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | E.Lazzaro,O.Sauter, S.Nowak, M.F. Nave et al. | ITPA Joint Experiment : | No |
| Description: | Test that modified rotationat rational surfaces in the regime predicted by the latest version of the theory [CONNOR, J.W. WAELBROECK F. L. and WILSON, H. R. Phys. Plasmas 8 (2001) 2835], causes NTMs even in the absence of triggers. This would be the first genuinely trigger-less mechanism for NTM onset and has an obvious relevance to ITER. Experiments on TCV show that NTMs can occur without seed island formation, in regimes of non-steady rotation caused by ECH near central power absorption and non-steady pressure and current density profiles [E. Lazzaro, S. Nowak, O. Sauter et al,Proc. FEC-IAEA 2012 , paper EX/P4-32]. The onset of the NTM is also apparently consistent and concomitant with the instability condition associated with the ion polarisation current (IPC). The growing tearing modes contribute a nonlinear magnetic braking that consistently modifies the rotation profile. This would be the first genuinely trigger-less mechanism for NTM onset and has an obvious relevance to ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Prepare a sawtooth-free target marginally stable (i.e., with beta slightly below critical) against 2/1 and 3/2 NTMs. Use n=3 RMPs to suppress ELMs, if necessary. Also make sure that fishbones or other NTM triggers are absent. Ramp down the toroidal rotation by means of balanced injection or magnetic braking. If successful, repeat with reverse Bt (but normal Ip), which would decouple the toroidal and poloidal rotation effects.
As recently observed on TCV (Lausanne) injection of P_EC upstream of the q=3/2 surface, may both drive a 3/2 mode and a strong effect on plasma rotation. Even in absence of triggers subsequently a 2/1 NTM can appear, associated with local flattening of the rotation shear, and modification of 2/1 rotation that enters the IPC unstable regime. [LA HAYE R. J., PETTY., C. C., STRAIT E. J., WAELBROECK F. L., AND WILSON H. R., Phys. Plasmas 10 (2003) 3644] Therefore it is proposed to test: a)A direct effect of P_EC (~ P_Ohm) on (N)TM stability through modification of J profile (upstream of q=m/n) (1 good shot with ECH , 1 good shot with ECCD) b)An indirect effect of P-EC on (N)TM stability through direct effect on rotation. A simple theory involving mode rotation in the presence of NTV effects (due to the static applied RMP and the TM modes themselves) will help designing the magnetic braking aspects of the experiment. Note that braking could be in place anyway, if RMPs will be used for ELM control. | ||
| Background: | The transverse ion polarization current (IPC) can be stabilizing or destabilizing for NTMs, depending whether the rotation is faster or slower than a reference "natural' value of the order of the ion diamagnetic drift frequency [E. Lazzaro, Proc.36th EPS Conference, Sofia (Bulgaria) 2009, paper P1.125].
Thus,varying the rotation of a q=m/n surface around that value can cause NTM onset even in absence of triggers (sawteeth, ELMs , fishbones...). | ||
| Resource Requirements: | In case of success of the first day or half day, possibility of a second day or half day with reverse Bt or Ip. | ||
| Diagnostic Requirements: | Magnetics, MSE(optional), ECE (both radiometer and Michelson), TS, Interferometer, CER. | ||
| Analysis Requirements: | EFIT, Tearing mode stability codes, optional reconstruction of NBI beam momentum deposition (TRANSP?), CXSR measurement of rotation. | ||
| Other Requirements: | -- | ||
| Title | 292: Optimize electron pedestal pressure in low-torque ITER baseline plasmas | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | G. Jackson, T. Luce, W. Solomon, F. Turco, K.H. Burrell, E. Doyle, T.L. Rhodes, C. Holland | ITPA Joint Experiment : | No |
| Description: | Low torque ITER-similar discharges were successfully achieved in the 2012 campaign using a combination of ECH with NBI co- and counter-injection at low- or moderate power. Building on these achievements, this proposal seeks to optimize pedestal electron pressure to achieve a higher pedestal electron temperature and Te/Ti > 1 in the outer core plasma. Combined with a beam torque scan this will allow investigating the dependence of thermal and momentum transport on Te/Ti in the ITER-like shape while decreasing electron-ion collisional coupling to some extent. Increased counter beam torque can increase the pedestal ExB shear and width of the shear layer potentially improve the edge transport barrier/pedestal pressure. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish ITER-similar plasma with moderate co-NBI and full ECH power (Reference shot #150400). Improve pedestal Te by varying the beam torque (combined co/counter-injection), and scanning the ECH deposition location between r/a=0.25-0.5. Obtain detailed fluctuation measurements in the outer core plasma and pedestal, using BES, DBS and CECE (the latter will require blocking ELM events during analysis to increase sensitivity). Core DBS will be possible (using O-mode) if the outer core density gradient is not inverted or flat ( Increasing the density gradient in the outer core via deep fueling/pellet injection may be explored if it does not com promise the pedestal temperature). | ||
| Background: | In lower density QH-mode experiments withy ECH run in 2012 the achieved increment in Te and the pedestal top electron temperature with ECH were very sensitive to he ECH deposition radius. A strong dependence of the toroidal rotation profile on deposition location was also observed. Pedestal electron temperatures up to 3 keV were achieved. The present proposal seeks to explore to what extent the pedestal rotation profile, the outer core momentum transport, and the ration of electron to ion thermal transport can be tailored to improve pedestal pressure and Te/Ti at higher density in ITER-like (ElMing H-mode) plasmas | ||
| Resource Requirements: | All beams , ECH (6 gyrotrons), possibly pellet injection. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Transport analysis/TGLF/GYRO to assess instability growth rates; the ITG/TEM transition, and nonlinear turbulence evolution. | ||
| Other Requirements: | |||
| Title | 293: Test of ETG zeff supression | ||
| Name: | Staebler staebler@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Inject neon into a hybrid regime plasma and look for changes in the electron thermal transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | probably need to inject the neon gradually in order to prevent sudden changes to the current chanel. Start with an established Hybrid regime and add incresing amounts of neon, CER on carbon and Neon. Need full profile diagnostics. High-k fluctuation measurements desireable. | ||
| Background: | Linear stability shows that kinetic impurity species have a stabilizing effect on the high-k ETG mode. The hybrid regime has been shown to have electron thermal transport domianted by ETG modes with TGLF. Previous modeling has shown a large difference in electron transport is the carbon is treated as simple dillution or as a fully kinetic species. It is important to validate the predicted effect of impurities on the ETG transport with experiement. It is easier to control neon cconcentrations in the plasma then carbon. An improvment in H-modes has been observed with neon on DIII-D in the past. (IH-mode) | ||
| Resource Requirements: | 1/2 day would be sufficient using a well established hybrid target. Slow gass puff of neon after the current flatop with a duration increased from shot to shot. One wasted shot to find the disruption limit then 3-4 good shots at lower neon levels. | ||
| Diagnostic Requirements: | need CER for both carbon and neon, all plasma profile measurments. As many fluctuation measurments as the target conditions allow. High-k a priority. | ||
| Analysis Requirements: | ONETWO or TRANSP analysis. XPTOR predictive runs with TGLF+NEO. | ||
| Other Requirements: | |||
| Title | 294: Simulated Alpha particle dillution | ||
| Name: | Staebler staebler@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use Helium NBI in a low density ITER Baseline H-mode target to simulate the impact of alpha particles on the energy transport in a burning plasma. The fast helium ion density would be varied to see how the fast helium ion dillution impacts the thermal transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with the DIII-D similarity discharge for the ITER baseline H-mode in Deuterium. Add Helium NBI to produce a population of Helium fast ions. Balanced injection is preferable to take out the rotation impact on confinement. Compare with discharges heated with central ECH to the same power. Compare with discharges heated with deuterium NBI to the same power from previous experiments. | ||
| Background: | The impact of helium dillution on fusion power is usually computed by assuming the electron density and ion temperatures are unchanged as the helium added so that just the dillution of the fuel is used. This overestimates the impact of heluim dilution since the fast alpha particles will be stabilizing the the ITG modes. Hence increasing the alpha density will both increase the ion temperature and dilute the fuel. The increse in ion temperature compensates somewhat for the reduced fuel density so the fusion power does not go down as much as the fixed temperature model would predict. This experiment aims to determine how much the ion temperature increases. | ||
| Resource Requirements: | ITER baseline H-mode resources + two beam boxes in Helium co+cnt. | ||
| Diagnostic Requirements: | full plasma profile and fast particle diagnostics | ||
| Analysis Requirements: | ONETWO or TRANSP analysis. TGLF transport predictions with XPTOR or TGYRO, GYRO simulations of selected cases. | ||
| Other Requirements: | |||
| Title | 295: preliminary simulated burn control | ||
| Name: | Staebler staebler@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This experiment would establish the feedback control requirments for a simulation of fusion burn control on DIII-D | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | use two NBI boxes in helium, one co the other counter to simulate the heating due to alpha-particles. Ultimately the power of these beams in balanced pairs would be on feedbackc control based on a sclaed calculation of the alpha-particle heating due to fusion reaction determined by the EFIT thermal stored energy. pre-programmed ramps of the balance helium beam power would be used to establish the response of the plasma so that a feedback algorithm could be developed off line for latter use. | ||
| Background: | ITER needs this. It cannot be developed on ITER due to disruption risk. | ||
| Resource Requirements: | ITER baseline H-mode resources + two beam boxes in Helium.
(same at 294) | ||
| Diagnostic Requirements: | full profile and magnetics | ||
| Analysis Requirements: | analysis of plasma response for feedback algorithm. | ||
| Other Requirements: | |||
| Title | 296: Determine dependence of RWM marginal stability on rotation and energetic particles | ||
| Name: | Sabbagh sabbagh@pppl.gov | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J.W. Berkery, J.M. Hanson, F. Turco, J.M. Bialek, B. Grierson, G. Jackson, R. La Haye, E. Hollman, M.J. Lanctot, E.J. Strait, B. Tobias, et al. | ITPA Joint Experiment : | Yes |
| Description: | Goals:
1. Determine the effect of rotation profile and energetic particles on plasma very near to the RWM marginal stability point a) scan Vphi profile and magnitude b) scan EP fraction and profile 2. Test present kinetic RWM stability theory for confident extrapolation to ITER Approach: The approach of this experiment would be to vary plasma rotation and energetic particle fraction in a plasma very close to the RWM marginal stability point. The effect of both the plasma rotation *profile* (not simply magnitude) and energetic particle fraction and profile affects RWM stability, but the stability gradient can change as a function of normalized beta. Past DIII-D experiments, including MP2012-83-02 have shown the RWM stability based on low frequency MHD spectroscopy at relatively low normalized beta ~ 2.4. The present experiment aims to focus on varying plasma rotation and energetic particle fraction *near* the RWM marginal stability point. Variations of V_phi and EP profile/magnitude that drive the RWM unstable will best guarantee that the RWM stability diagnosis is accurate, and will assure that the stability gradient is the most meaningful â?? being near the point of marginal stability. Significant progress was made in last yearâ??s DIII-D experiment MP2012-83-02 to determine the effect of plasma rotation on RWM stability in DIII-D. A sub-goal of that experiment was to create a target near the RWM marginal stability point. A few discharges were specifically created with normalized beta substantially higher than in past experiments ~ 3.5, and are determined to be close to marginal based on low frequency MHD spectroscopy (e.g. shot 149782). An issue making the next logical step and simply using this target to complete the Vphi and Ep scans is that most, or all NBI power was required to reach this target. Another issue is that tearing modes appear in the high beta, and the target plasma is not a candidate for NTM suppression with ECH. Tearing modes will generally preclude the onset of an unstable RWM (e.g., based on NSTX experience). Based on ideas of the co-authors, several new approaches have been discussed for the target plasma: 1) Use the present target from MP2012-83-02, and stabilize (using ECH) or avoid the tearing mode 2) Switch to another NBI target, such as the high qmin, â??steady-stateâ?? target (e.g. shot 150301 suggest by J. Hanson, at normalized beta ~ 3.1), which appears to be at RWM marginal stability, but does so with lower NBI power, allowing leeway for the beams to alter Vphi and EP fraction at the RWM marginal point (off-axis NBI might be used here) 3) Consider an RF-heated target. Shot 150840, suggested by F. Turco, might be used as an initial target (betaN ~ 1.8), and heated by just a few NBI sources to provide a greater change in Vphi and EP profile / level. The ideal deltaW would be driven as high as possible, using Ip ramps, and/or other means. (continued in "Experimental Approach/Plan" box) |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | (see "Description" box above, plus the following):
At present, approach 2) is arguably the one that would yield the highest probability of success, and is arguably more appropriate for ITER applicability. However, approach 3) has significant merit in the ability to more greatly alter the EP distribution (profile, and in velocity space). 5. Experimental approach/plan a. Reach RWM marginal stability point in one, or more of the suggested target plasmas, starting from the target requiring the least amount of development b. Maximize the ideal deltaW to bring the plasma closest to the RWM marginal stability point (via reduced triangularity, current ramping, lower q95, etc.) c. Vary rotation via NBI mix, and possibly by non-resonant NTV by application of n = 2 or 3 fields with I and C coils d. Vary EP profile and distribution by OANB, varied Ip/Bt at fixed q, amount of NBI vs RF (in possible RF target), etc. | ||
| Background: | As mentioned above, this experiment stems from MP2012-83-02. Significant research has been conducted on NSTX in this regard, and DIII-D/NSTX have conducted ITPA joint experiments (for ITPA MDC-2) to determine RWM stability physics. The present experiment can be seen as reaching a *significant* milestone in the joint experiment in which the (i) RWM marginal stability boundary is convincingly demonstrated, and (ii) variations from this point ensure that the stability gradient in the region of interest is being evaluated, rather than at values significantly lower than the marginal stability point (which have already been studied and published based on DIII-D data). | ||
| Resource Requirements: | At least 6 NBI sources â?? all 8 highly preferred. For the RF target, 6 gyrotrons at max power and max duration. | ||
| Diagnostic Requirements: | Magnetics, RWM sensors, MSE and CER, Thomson scattering, SXR, ECE/I, and full suite of available macroscopic stability-related diagnostics as applicable to the stated target plasma | ||
| Analysis Requirements: | - kinetic EFIT reconstructions
- transport calculations including fast particle distribution - MISK kinetic RWM stability calculations (provided by our group) - MARS-K calculations, with comparison to MISK (provided by our group) | ||
| Other Requirements: | |||
| Title | 297: ITER baseline scenario with low NBI torque | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | F. Turco, W. Solomon, T. Luce | ITPA Joint Experiment : | No |
| Description: | Explore low torque operation in ITER Baseline Scenarios plasmas | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Re-establish a stable low torque discharge, e.g. 149682 (Tinj = 0.36 Nm)
2. Increase T in small steps. Varying li (via dIp/dt). Then decrease to 0.3 Nm (2/1 unstable) and add EC (including mirror tracking) to stabilize | ||
| Background: | Operating range for ITER baseline scenario discharges in 2012 showed a rather narrow window at low torque (<1 Nm). However this range was not fully explored. The goal is to map a more robust operating range. | ||
| Resource Requirements: | NB, ECH and ITER shape | ||
| Diagnostic Requirements: | Standard diagnostics only | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 298: Soft landing with 2/1 TM in the ITER shape | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | F. Turco, A. Hyatt, W. Solomon, D. Humphreys | ITPA Joint Experiment : | No |
| Description: | Find a Scenario to successfully meet ITER rampdown requirements with a 2/1 TM | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Trigger a 2/1 TM (probably easiest with low density) in the ITER shape. Holding the ITER shape, ramp down without disrupting. This can be done in piggyback. Add ECH if necessary in the rampdown. | ||
| Background: | Disruptions are problematic for ITER, hence a valid rampdown scheme needs to be developed, even in the presence of a 2/1 TM and/or locked mode. | ||
| Resource Requirements: | ITER shape | ||
| Diagnostic Requirements: | standard, but may need to extend clocks, EFITs, etc. to cover rampdown. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 299: Aspect ratio scaling of turbulence and transport: DIII-D and NSTX-U comparison | ||
| Name: | Smith none | Affiliation: | Univ. Wisconsin (at PPPL) |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | George McKee | ITPA Joint Experiment : | No |
| Description: | Measure the aspect ratio scaling of turbulence and transport in DIII-D and NSTX-U discharges with similar dimensionless parameters (q profile, Te/Ti, beta, rho*), and compare and contrast turbulence characteristics and behavior in tokamaks vs. spherical torus. Parameters that are unlikely to match, such as nu*, can provide additional scaling results. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | An aspect ratio scaling experiment at ~1 T will require low field operation on DIII-D and high field operation on NSTX-U. In addition, neutral beam sources will be tuned to best match Te/Ti, beta, gradients, and rotation in the outer core region where temperature gradients are finite and turbulence amplitudes are adequate. Fueling, pumping, and wall conditioning are additional levers to control density and beta. Matching collision frequencies normalized to bounce frequencies (nu*) may not be feasible, but the challenge can be an opportunity to investigate nu* scalings. Target DIII-D H-mode plasmas will be similar to well-behaved NSTX plasmas scaled up to higher field, current and power as expected for NSTX-U plasmas. BES systems, available on both DIII-D and NSTX-U, will be the primary fluctuation diagnostics, along with a range of other high and low-k fluctuation systems. Rho* variations can establish a range of plasma conditions for future comparison to complimentary NSTX-U plasmas. Relatively high-rotation plasmas will be developed via co-current NB injection since NSTX-U will have full co-NBI systems. | ||
| Background: | The aspect ratio (A=R/a) is known to be a key dimensionless turbulence parameter along with rho*, nu*, Te/Ti, and beta. The aspect ratio can influence turbulence and transport through trapped particle fraction, curvature favorability, and zonal flow dynamics. While higher A reduces the trapped particle fraction that gives rise to TEM turbulence, higher A also expands bad curvature regions on flux surfaces. According to Rewolt et al (1996), the net result is higher micro-instability growth rates at higher A. On the other hand, other results point to reduced growth rates or transport coefficients at higher A (ITG: Kotschenreuther et al (1995), TEM: Lang et al (2007), ETG: Jenko et al (2001)). In addition, higher A reduces the zonal flow potential due to enhanced neoclassical polarization shielding (Diamond et al (2005)). An aspect ratio scaling experiment can provide insight into a key dimensionless parameter for turbulence, but such an experiment was previously not feasible on DIII-D and NSTX due to a large rho* discrepancy. The higher field and current capabilities on NSTX-U will make such an experiment feasible and exploit the similar minor radii and comparable fields/currents on DIII-D and NSTX-U. | ||
| Resource Requirements: | co-NBI | ||
| Diagnostic Requirements: | BES | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 300: Effect of triangularity in ITER shape and improved shaping | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | A. Hyatt, D. Humphreys, W. Solomon | ITPA Joint Experiment : | No |
| Description: | Examine effect of triangularity on stability in the approximate ITER baseline shape, especially with the improved ITER shape. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce an IBS discharge, e.g 147044. Vary lower trinagularity from appx 0.64 - 0.78. Do this without cryopumping (ne ~ 6e19)shortly after a boronization for uniform recycling. If lithium is used, this might also provide more uniform reycycling.
Use standard IBS parameters (betaN=1.8, q95=3.1, etc.), but full Co-beams. Map stability region by varying li (dIp/dt). | ||
| Background: | Some evidence exists that the stability region is reduced at lower triangularity in IBS discharges, but no systematic scan has been performed. The DIII-D scenario has not exactly reproduced the ITER shape, especially the lower outer squareness due to limitations with return current. This is being addressed in FY13 with shape development and should allow a wider range of triangularity, approaching the ITER baseline shape. The one impediment is that part of this scan would be on the baffle, the other on the floor, which has traditionally had cryo pumping. This might be obviated by a fresh boronization or lithium injection. | ||
| Resource Requirements: | New boronization (after 1 day), possibly lithium if the dropper has been commissioned. Both of these could provide more uniform recycling as triangularity is moved onto the baffle. | ||
| Diagnostic Requirements: | standard | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 301: Access to the small ELM regime with a shape control | ||
| Name: | Ahn jahn@pppl.gov | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | R. Maingi | ITPA Joint Experiment : | No |
| Description: | This experimental proposal seeks to access the small ELM regime, possibly with wider operation window than the existing small ELM regimes such as type-II and grassy ELMs, by the means of plasma shape control. This is based on the empirical observation of small ELMs at KSTAR in the last two consecutive campaigns; small ELM phases were observed in a rather wide operating windows, i.e. lower densities (ne/nG <0.4) and various configurations (DN, USN, LSN, limited). From the experience at KSTAR, it is expected that the adjustment of three shape parameters could lead to the access of a small ELM regime with good confinement, i.e. higher squareness, non-conventional inner separatrix shape (more straight or even slightly concave), and smaller drsep (|drsep| < 1 - 1.5cm).
A stability analysis at DIII-D and experimental results (Leonard, NF 2007) indicate that higher squareness reduces pedestal pressure with smaller ELMs, and is beneficial for maintenance of a steady internal 3/2 TM and good confinement. This result is consistent with the experimental observation at KSTAR that higher squareness led to small ELMs. There is no recent ELM stability analysis regarding the non-conventional inner separatrix shape to the authors' knowledge, but there are early works on the bean/crescent shape in mid-80's which showed that this configuration will offer an easier access to the 2nd stability regime. As for the smaller drsep, recent stability analyses, e.g. Saibene NF 2005 (JET) and Saarelma PPCF 2009 (MAST), demonstrated that this configuration can help, along with high collisionality, improve stability against type-I ELM triggering P-B modes and instead hit the stability boundary set by the high-n ballooning modes, therefore small ELMs. In fact, small ELMs at KSTAR did not require high density and we speculate that this might be due to the combined effect of non-conventional inner separatrix shape. This might also explain why the drsep values for KSATR small ELMs (|drsep| < 1 - 1.5cm) are noticeably larger than the conventional drsep requirement for type-II ELMs (|drsep| < 0.5cm). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Three shape parameters will be varied; plasma squareness, shape of inner separatrix, and drsep. As the plasma squareness has been already shown to change the ELM size at DIII-D (Leonard, NF 2007), a maximum allowable squareness (to be determined through discussion with the GA staff) will be used in combination with varied inner separatrix shape and drsep. This will save the number of discharges needed for the experiment.
Ideally it will be best to change the inner separatrix shape in a single discharge to check its impact on ELMs, e.g. convex --> straight --> concave, so is the drsep (LSN --> DN --> USN). The final goal is to find the optimal combination of these parameters for the most preferred small ELM regime. | ||
| Background: | Naturally occurring small/no ELM regimes are a preferred H-mode operation mode for ITER as well as the RMP ELM suppression. However, the operating window has been historically known quite narrow. We propose here to explore the possibility of accessing small ELM regime with a significantly wider parameter window, i.e. density and magnetic configuration, by controlling the shape of plasma. | ||
| Resource Requirements: | Major auxiliary heatings (NBI and ECH), EFIT reconstruction | ||
| Diagnostic Requirements: | TS, CER, magnetics, ECEI, BES, fast IR camera, edge turbulence diagnostics, divertor spectroscopy, etc. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 302: RMP ELM Suppression at the NTV Offset Rotation | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | 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: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 303: Effect on lithium wall conditioning for Co-QH-mode and edge pedestal characteristics | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, A. Garofalo, C. Chrobak, R. Maingi | ITPA Joint Experiment : | No |
| Description: | Use the new lithium dropper to reduce recycling and assess performance of Co-QH mode and modified H-mode edge pedestal parameters | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | MP 2012-94-02 was submitted but was not approved before the end of FY12 operations. The idea is to establish an USN discharge (147354) and inject a series of shots with lithium powder, evaluating changes (hopefully improvements) in QH mode. It is anticipated that the end of these discharges can also be used to examine H-mode pedestal physics, specifically any changes to the density pedestal width and other parameters. | ||
| Background: | Lithium conditioning has been very effective on NSTX in modifying edge pedestal parameters, and even producing ELM free H-mode. In DIII-D, the low recycling may also lead to a better EHO, particularly co-injected EHO. | ||
| Resource Requirements: | Lithium dropper (anticipated installation is spring 2013). Rev. Bt preferred. | ||
| Diagnostic Requirements: | Visible cameras viewing lithium injection port (tentatively 285 V+1) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 304: Birth of ITER: the world's largest resistor | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Physics | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Evaluate conditions for ITER first plasma and validate with modeling | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Before any plasma attempts in 2013, make a "dirty" plasma with no burnthrough, resembling the ITER first plasma (ITER parameters: 100 kA, 100ms, 0.3 V/m). DIII-D parameters: Ip=10-20 kA, Vloop = 3V, normal field nulls. ECH pre-ionization power scan to assess effectiveness. | ||
| Background: | ITER has very low expectations for first plasma. In fact, if it were to burnthrough and go to 1 MA or greater there is a concern that it might contact other parts of the unprotected vessel such as the outer wall and damage components. So there is an opportunity to study the opposite challenge that DIII-D usually faces: Can we reliably produce a plasma that doesn't burn through? We can then extrapolate this to ITER. Since ITER conditions will probably be poor vaccum and little conditioning, our startup provides a unique opportunity to assess this before applying any conditioning except baking. | ||
| Resource Requirements: | ECH pre-ionization and Vloop control (similar to ITER startup experiments) | ||
| Diagnostic Requirements: | Need visible cameras, magnetics, interferometers, Thomson, SPRED operating for this first plasma attempt | ||
| Analysis Requirements: | Develop breakdown models, and compare to startup model by Kim (JET). MHD equilibria at low current may require special attention to the magnetics. | ||
| Other Requirements: | -- | ||
| Title | 305: Scaling Of Pedestal Plasma Transport With RMP I-coil Current | ||
| Name: | Callen jdcallen@wisc.edu | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | S. Smith, N. Ferraro, S. Mordijck, R. Moyer | ITPA Joint Experiment : | No |
| Description: | The basic proposal is to update and expand the exploration, as a function of I-coil current, of the RMP suppression of ELMs in low collisionality DIII-D ISS discharges in the Evans et al, Nucl. Fusion 48, 024002 (2008) paper. Such a set of experiments is very important for developing a better and more quantitative understanding of the effects of RMPs on pedestal plasma transport -- see Background discussion below. Revisiting these experiments is warranted at this time because of the recent upgrading of the edge Thomson scattering system and other diagnostics plus the recent more comprehensive characterization of RMP ELM-suppressed regimes. A new set of I-coil scaling data will provide the key impetus for a new, much more precise round of interpretive transport modeling (via M3D-C1, ONETWO and the flutter transport model module) of RMP effects on pedestal plasma transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In the previous I-coil current scaling discharges (126435-126443) the I-coil current was held constant through the discharge. An alternate approach for determining the I-coil current scaling would be to change the I-coil current in steps, throughout a given discharge, at constant beam power. In addition to fiducial discharges with no RMPs, the I-coil current could also be initiated at a low level (2 kA?) and then stepped up into RMP-suppressed regimes. It would be important to look for any abrupt changes in the plasma transport characteristics in the pedestal region, particularly in the carbon toroidal and poloidal flows and hence the implied radial electric field there, as the ELM suppression I-coil current threshold is exceeded. | ||
| Background: | The best, most systematic and comprehensive set of data on RMP suppression of ELMs in low collisionality DIII-D pedestals as a function of I-coil current was provided in the Evans et al, Nucl. Fusion 48, 024002 (2008) paper. This set of data was critical for many papers on pedestal density transport and in particular the development of the RMP-flutter-induced plasma transport model:
1) Saskia Mordijck's Ph.D. thesis, "Particle transport as a result of Resonant Magnetic Perturbations," UCSD, January 2011. 2) J.D. Callen, A.J. Cole, C.C. Hegna, S. Mordijck and R.A. Moyer, "Resonant magnetic perturbation effects on pedestal structure and ELMs," Nucl. Fusion 52, 114005 (2012). 3) J.D. Callen, A.J. Cole and C.C. Hegna, "Resonant-magnetic-perturbation-induced plasma transport in H-mode pedestals," Phys. Plasmas 19, 112505 (2012). 4) J.D. Callen, C.C. Hegna and A.J. Cole, "RMP-Flutter-Induced Pedestal Plasma Transport," San Diego IAEA FEC paper TH/P4-20, 8-13 October 2012. 5) P.T. Raum, S.P. Smith, J.D. Callen, N.M. Ferraro et al., "Comparison of flutter model with DIII-D RMP data," paper currently being written for submission to Nucl. Fusion. The last reference in particular shows relatively good agreement between the ONETWO interpretive results for discharges 126006 and 126443 and the flutter model predictions for the radial electron heat diffusivity and T_e profile at the top of the pedestal when the RMP-induced magnetic perturbations calculated by M3D-C1 which include two-fluid plasma response effects are used. The recent upgrades in the diagnostics and better characterization of ELM-suppression regimes in DIII-D can facilitate the development of a modern, much more precise and comprehensive set of I-coil current scaling data for plasma transport interpretive modeling via M3D-C1, ONETWO and the flutter transport model module that has been developed. A new set of I-coil current scan data should also facilitate exploring the role of the electric field and its effect on the toroidal plasma rotation -- and whether changes in them are critical factors in obtaining ELM suppression, as suggested in Ref. 4) above and Rick Moyer's Providence APS-DPP invited talk. | ||
| Resource Requirements: | Mainly a set of similar parameter ELM-suppressed discharges are needed with increasing I-coil current from 0 (for fiducial discharges) to about 6 kA in small steps -- say in units of 1 kA (or less around the threshold for ELM suppression?). Discharges in which the ELM suppression I-coil threshold current is about 2 or 3 kA would probably be best. It would be critical to hold the pedestal in about the same relevant parameter regime (beta, in center of q_95 resonance, core heating etc.) during the changes in I-coil current. | ||
| Diagnostic Requirements: | Good Thomson T_e and n_e measurements and CER carbon rotation measurements of the plasma parameters at the pedestal top (0.9 < Psi_N < 0.98) in response to various I-coil currents are critical. More generally, all the critical diagnostics that are needed to facilitate developing good kinetic EFITs should be operative. Measurement of the current profile in the pedestal region with the recently revitalized Lithium beam diagnostic would also be useful both for developing good kinetic EFITs and for exploring the sensitivity of the RMP-induced magnetic perturbations calculated by M3D-C1 to the current density profile in the pedestal. | ||
| Analysis Requirements: | Good kinetic EFITs, ONETWO interpretive analysis of the electron density and thermal transport plus plasma toroidal rotation as a function of the I-coil current in ELM-suppressed discharges. Also, M3D-C1 analysis of the magnetic perturbations including flow screening effects will be needed to quantify estimates of the corresponding flutter model predictions for these same discharges. | ||
| Other Requirements: | |||
| Title | 306: Edge turbulence and blob generation | ||
| Name: | Smith none | Affiliation: | Univ. Wisconsin |
| Research Area: | Divertor & SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Observe and characterize the spatial transition from drift-wave/interchange turbulence in or near the pedestal to intermittent blobs in the SOL. Directly measure density fluctuations in the pedestal and near-SOL regions with BES to compliment theory models of blob generation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Target stationary L-mode (low edge flow shear) and H-mode (high edge flow shear) plasmas with minimal MHD activity. Scan 2D BES measurements across the pedestal and near-SOL regions. If feasible, augment BES measurements with plunging edge probe measurements. While maintaining stationary discharges, modify edge pressure profiles with fueling, heating, and rotation strategies. | ||
| Background: | Theory models attribute SOL blob generation to the flow shear decorrelation of radial streamers associated with drift-wave/interchange turbulence in or near the LCFS (Myra et al (2006), Russell et al (2011), D'Ippolito et al (2011)). In addition, models and simulations indicate blob size is related to the pressure gradient in the blob birth zone, and nonlinear saturation mechanisms for turbulence govern blob birth rates. BES density measurements of the turbulence-blob interplay near the LCFS will be useful for validation of blob generation models and simulations. | ||
| Resource Requirements: | NBI | ||
| Diagnostic Requirements: | BES | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 307: Role of turbulent-driven and Ion pressure gradient-sustained ExB flows in Triggering H-mode | ||
| Name: | Tynan none | Affiliation: | UCSD |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | L. Schmitz, G.R. McKee, Z. Yan, P.H. Diamond, L. Zeng,
J.A. Boedo, T.L. Rhodes, E.J. Doyle, D. Eldon |
ITPA Joint Experiment : | No |
| Description: | This experiment is in follow up to results from the 2011 campaign, presented at IAEA 2012, which show evidence that turbulent driven sheared ExB flows and turbulent stresses become large at the transition from L-mode to the intermediate LCO regime. Taken together with earlier results by Schmitz et al (Schmitz, PRL 2012) which show that the ion pressure gradient then slowly builds up during the LCO regime, these new results point towards the important role that ZF/GAMs play in first accessing the regime of improved confinement, and then then pressure gradient term then â??locks inâ?? the improved confinement state, resulting in the onset of the H-mode regime. This experiment will aim to make all the measurements needed to definitively show this sequence. Currently the picture has been pieced together from experiments performed across the 2010-2011 campaigns. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose to reproduce the experimental results of the Dec 2011 L-H half-day experiment (see e.g. shot 147725) but would bring to bear a full diagnostic suite of measurements (no fast profile reflectometry or burst-mode edge TS was available during that day of experiments). We would establish the target conditions with an extended LCO regime and then plunge the midplane probe with all tips operating to obtain Reynolds stress and full fluctuation field (Ne, Te, Phi, V_parallel) measurements. The probe would plunge to ~1cm inside the LCFS and capture L-mode to LCO transition, early/mid/late LCO regime, and LCO-to-Hmode transition data. Fast profile reflectometry and burst-mode edge TS would obtain high time resolution edge profile evolution data, and 2D BES imaging would be used to obtain turbulence imaging across the LCFS. If possible, some BES channels should be configured into a 1D radial chord to obtain time-space data on fluctuation amplitude evolution aross the outer ~20-30% of the minor radius during the sequence. | ||
| Background: | The results from the 2010-2011 campaigns have provided the most detailed measurements of the L-mode/LCO/H-mode transition physics, and the picture discussed above has emerged from that work. However, we do not have a complete dataset of Reynolds stress, fluctuation amplitude, and density, temperature, velocity and E field profile evolution during the transition sequence during the same discharge. This proposal would address this important missing element, and would be aimed at providing the dataset needed to clearly and convincingly demonstrate the important dual roles of the turbulent driven shear flow, the ion pressure gradient driven flow, and the onset of the edge pedestal. | ||
| Resource Requirements: | Resource Requirements:30,330,150 Beams, ECH | ||
| Diagnostic Requirements: | Midplane probe, BES large array centered on pedestal region, DBS-5,DBS-8, fast reflectometry, CER, edge TS in burst mode. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 308: RMP (Resonant Magnetic Perturbation) code validation via UCLA 288 GHz polarimeter | ||
| Name: | Zhang xyzhangj@physics.ucla.edu | Affiliation: | UC, Los Angeles |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Tony Peebles, Troy Carter, Neal Crocker, Edward Doyle, Terry Rhodes, Todd Evans, Guiding Wang, Mike Van Zeeland, Lei Zeng | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to validate the code predictions, e.g., TRIP3D and M3D-C1, for plasma response to RMP. The TRIP3D calculates the magnetic field produced by I-coils in vacuum, while M3D-C1 includes other plasma response, predicting the magnetic and density perturbations. The magnetic perturbations predicted by these two codes, however, are quite different. The recently commissioned UCLA 288 GHz polarimeter can provide an experimental measurement of the internal magnetic field perturbation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Low electron density (ne < 4E19 m-3) discharges are preferred for optimal performance of the polarimeter. The plasma center is positioned at +7.5 cm, the height of horizontal polarimeter diagnostic beam, to reduce refraction due to density gradient transverse the beam. The I-coils are operated with either n=2 or n=1 configuration. Operational details, for instance, plasma shape, the I-coil frequency, magnitude, and on/off period, etc., are under active discussion. Target shots will be developed during the startup of the campaign.
Due to the line-integral nature of the polarimetry measurements, density perturbation measurements should simultaneously be acquired with profile reflectometer, BES, Lithium beam (when applicable). A synthetic diagnostic code, which has been developed to calculate the polarimeter response with inputs of density and magnetic profiles, will be utilized to assist measurement interpretation. Toroidal field will be scanned, i.e. BT = 0.75 T, 1.5 T, and 2.0 T. This will vary the contribution from the Cotton-Mouton effects to the polarimeter measurements, and hence the density perturbations. | ||
| Background: | The ELMs (Edge Localized Modes) cause large, fast heat and particle impulses to the divertor target plates that can exceed the transient thermal capacity of the target plates and limit the divertor lifetime. The RMPs on DIII-D have achieved successful ELM suppression, while the underlying physics is still unclear. There are also some outstanding issues, e.g., the kink response, resonant field screening and amplification, etc.
This experiment, if successful, will become a data set for Mr. Jie Zhangâ??s PhD thesis. | ||
| Resource Requirements: | NBI for MSE, CER and BES diagnostics | ||
| Diagnostic Requirements: | Whenever possible, profile diagnostics (Thomson Scattering, Mirnov coils, MSE (ï?³ï? ï?±ï?®ï?²ï? ï??), profile reflectometer, CER)
CO2 interferometer BES ECE (ï?³ï? ï?±ï?®ï?¶ï? ï??) DBS CECE fast magnetic coils Lithium beam | ||
| Analysis Requirements: | EFITs, TRIP3D, M3D-C1 | ||
| Other Requirements: | |||
| Title | 309: Full 3D Rotation optimization of H mode | ||
| Name: | Buttery buttery@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | All the people I speak to in stability | ITPA Joint Experiment : | Yes |
| Description: | The crucial limiting effect of error fields in H-modes comes through braking, which either opens the door to inherent tearing instability, or arrests rotation to drive locked modes directly. this most likely acts through NTV, with different 3D field sources of NTV possibly adding independently. Thus an approach to optimize the error field by rotation optimization, exploring the optimization in different I coils (or I-coil combinations) independently. A combined optimal EFC would result from combining these individual responses. Slow rotation feedback would be a good way to achieve each optimization | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | At medium betan, ramp current in an I coil to optimize rotation. If no discernible optimization raise betan. Introduce a slow feedback approach to rotation optimization to optimize current in coil. Repeat shot to shot with different coils. Once all coils done, then combine correction of individual coils, scaling current across all coils up and down together for the rotation optimization. (These are the principles, some compromises may be needed in implementation give available PSs and PS limits; might use some combinations of audio amps to add degrees of freedom - otherwise a lot of patch changes). (The rotation feedback optimization might be further improved using Frasinetti's proposed dithering technique) | ||
| Background: | TBM studies show that H mode, traditional error field correction only captures 25% recovery. | ||
| Resource Requirements: | Magnetic feedback control development. All possible RMP power supplies, augmented by suitable combination of audio amps. I coils. | ||
| Diagnostic Requirements: | Rotation - CER in real time | ||
| Analysis Requirements: | Some effort to implement rotation feedback linkage into 3D coils, and develop algorithm. | ||
| Other Requirements: | Someone to run this XP | ||
| Title | 310: Energy Transport During Electron-Dominated Heating of ITER-Relevant H-Mode Discharges | ||
| Name: | Taylor gtaylor@pppl.gov | Affiliation: | PPPL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | N. Bertelli, J.C. Hosea, R.J. Perkins, C.K. Phillips, P.M. Ryan, D.R. Smith, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | This experiment will study electron transport and plasma turbulence in DIII-D Advanced Inductive (AI) and ITER Baseline Scenario (IBS) H-mode discharges that are predominantly heated by electron cyclotron (EC) power and it aims to identify the dominant mechanism(s) responsible for enhanced electron transport when EC power is applied to these ITER-relevant scenarios, especially as produced on DIII-D with NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | All discharges should be run with balanced NBI to minimize the applied torque, and they should be run with no beta feedback on NBI power so that the NBI power remains constant.
The run plan is as follows: 1.Begin with an AI discharge similar to shot 146571 (the outer gap can be larger since there will be no fast-wave (FW) heating), with sufficient NBI power to transition to and sustain an H-mode. 2.Apply an EC heating pulse that is considerably shorter than the NBI pulse (~1 s). Vary the EC pulse timing and duration. Measure the change in stored energy at the turn-on and turn-off of the EC pulse. Substitute ECH power with similar level of NBI power. 3.Repeat 2 with increasing EC power (eg. 2, 4 and 6 gyrotrons). 4.Repeat 3 with one or possibly two gyrotrons modulated to study electron transport with ECE etc. 5.Repeat 1-4 for an IBS discharge similar to shot 150840 (once again, the outer gap can be larger since there will be no FW heating). Typically the IBS discharges in 2012 had to be run at much higher densities than the AI discharges (~5.5x1019 m-3 for IBS compared to ~ 3.5x1019 m-3 for AI) to avoid NTMs. The higher density in the IBS discharges caused the plasma to go overdense for second harmonic ECE, so the density should be lowered to get core ECE data. If NTM??s appear some of the gyrotron launchers should be configured for ECCD in order to stabilize NTM??s. [NOTE: For the 2013 campaign it is hoped to upgrade at least one, possibly more, of the EC launchers so that they can be changed from ECCD to ECH orientation in 0.5-1 s, compared to ~ 5 s at present.] 6.Compare heating efficiencies for ECCD and ECH for best heating case of 5. Couple EC power from mirrors configured for ECCD followed by coupling power from the mirrors configured for ECH and vice versa. Perform in successive shots, or if possible, using two ECH pulses in the same shot with fast mirror movement or with power configured for ECH and ECCD. 7.Run AI and IBS H-mode discharges with EC heating only (no NBI) at the highest gyrotron power available to assess how well EC heats an ECH-only H-mode discharge. | ||
| Background: | ITER will utilize virtually torque-free, fuelling-free, dominant electron heating to generate and sustain plasmas in the H-mode regime. EC heating will play a major role in generating H-mode discharges in ITER. However there is growing evidence for significantly increased electron transport when EC heating is applied to DIII-D ITER-relevant H-mode discharges produced with NBI. Comparison of EC and FW heating of AI H-mode discharges in 2011 showed similar core electron heating and heating efficiency based on the time evolution of stored energy for the first ECH pulse and the FW heating pulse applied on top of the first ECH pulse, whereas very little heating and increase in stored energy was obtained with a second ECH pulse on top of the first (eg. shots 146571 and 146574). In 2012 a saturation of stored energy was observed in DIII-D IBS discharges when increasing levels of EC heating were applied (eg. shots 150840 and 150821). The radiated power observed when the EC power was applied to either the AI or the IBS discharge scenarios was essentially proportional to the total power, suggesting that the observed behavior is due to enhanced electron transport in these scenarios. It is imperative that the dominant mechanism(s) causing the enhanced electron transport are identified so that ITER-relevant scenarios with reduced electron transport during EC heating can be developed. | ||
| Resource Requirements: | Machine Time: 1-1.5 days (Steps 1-4 of the run plan for the AI target discharges can be completed in about 0.5 days, similarly Step 5 for the IBS target discharges can be completed in about 0.5 days, and the remaining steps of the run plan may take another 0.5 days)
Number of gyrotrons: 6 (7 if available) Number of neutral beam sources: 4, plus beam blips for BES and MSE | ||
| Diagnostic Requirements: | ECE, BES, CHERS, , MSE, UCLA reflectometry for oblique angles, and other diagnostics for measuring turbulence | ||
| Analysis Requirements: | TORAY, GENRAY, TORBEAM, TRANSP | ||
| Other Requirements: | Also submitted to Inductive Scenarios and Turbulence & Transport - please discuss placement with that group | ||
| Title | 311: Transport with qmin>2 | ||
| Name: | Holcomb holcomb@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Ferron, Heidbrink | ITPA Joint Experiment : | No |
| Description: | This proposal adds to ideas presented in Heidbrink's ROF #39. Scan various quantities and make good diagnostic measurements to clarify the reasons for poor energy confinement in plasmas with qmin>2. After identifying the best conditions for confinement in controlled scans, apply maximum power to test betaN limit. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Standard double null shape. Bt=+1.75 T. Flattop betaN=2.7. In all scans obtaining FIDA, neutrons, standard profiles (n, T, etc), and fluctuation data is a high priority.
1. 2 point scan in qmin (~1.5, 2.1) at fixed q95 to recover H89 trend with qmin. 2. Then 3 pt. scan in q95 (4.5, 5.5, 6.5) with qmin=2.1. 3. Density scan with fixed q. 4. 2 pt. ECCD deposition scan with fixed q.(maybe?) 5. Scan rhoqmin out by starting with max Bt and ramping down to Bt=1.75 T at one value of q95. 6. Identify best conditions for confinement and maximize betaN with qmin>2. | ||
| Background: | Plasmas with off-axis NBI, ECCD, and qmin>2 in 2011 & 2012 typically had H89<2, making access to betaN>4 with the available power difficult. Indirect evidence suggests enhanced fast ion transport occurs at the highest qmin. Thermal transport is likely also affected, but questionable ne and Te measurements mean considerable uncertainty remains. No FIDA or thermal turbulence measurements were made. Several shots with Bt-ramps to broaden the current profile did have periods of time with qmin>2 and H89>2, but it is unclear if this was due to the q-profile (i.e. increased volume of low shear) or lack of ECCD because of low field operation. | ||
| Resource Requirements: | All NBI except 210lt. Ability to go to higher power. 150 at max tilt angle. Consistent gyrotron power at least 3 MW. Clean machine. | ||
| Diagnostic Requirements: | MSE, TS, ECE, CER, Spred
FIDA! Neutrons Desire Reflectometer Desire fluctuation diagnostics (PCI, ECEI, polarimeter, DBS, etc.) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 312: Limits to pedestal radiation high beta discharges | ||
| Name: | Leonard leonard@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Document the limits to pedestal/mantle radiation for heat flux control in high beta discharges | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Set up high beta discharge currently under study as candidate for FNSF/DEMO. Inject neon or argon for mantle radiation. Use radiated power feedback to control the radiation level. On a shot to shot basis increase the pedestal radiation level until significant degradation of the pedestal and confinement. The primary issue is how much power above the LH threshold must cross the separatrix in order to maintain confinement. Other issues could include the effect of increased Zeff and collisionality on the operational regime. | ||
| Background: | FNSF and DEMO scenarios for handling divertor heat flux rely on radiating as much power as possible inside the separatrix. A radiating mantle could this, but it is uncertain how much power can be radiated before the pedestal degrades, affecting confinement. It is generally thought that 1.5xP_LH needs to cross the separatrix, but little data exists to support this. Additionally, the addition of radiating species, such as neon or argon, could have additional effects, including Zeff and collisionality. How this will affect high beta scenarios is also not known. | ||
| Resource Requirements: | Radiated power control of neon or argon puffing | ||
| Diagnostic Requirements: | Bolometer feedback into the PCS | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 313: Hybrids with co-NBI QH-mode edge | ||
| Name: | Petty petty@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | This experiment has both a physics goal and a programatic goal. The physics goal is to determine whether poloidal magnetic flux pumping in hybrids (the phenomena that maintains qmin>1) is due to a coupling between ELMs and the 3/2 mode. If this is the case, a hybrid with QH-mode edge should evolve to qmin<1. We can test whether the 3/2 mode is still modifying the q profile in this new regime by using ECCD aimed at the q=1.5 surface to suppress the 3/2 mode on demand. The programatic goal is to develop a ELM-suppressed hybrid plasma that can be a viable scenario for ITER. A QH-mode edge hybrid would be advantageous over an RMP case as the latter tends to develop locked modes at high beta as the tearing modes interact with the RMP. (On the other hand, I acknowledge that obtaining the required amount of edge co-rotation for QH-mode in ITER is "challenging".) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The key to obtaining a co-rotation QH-mode is to spin the toroidal rotation to a high value before the L-H transition. To achieve this, we will use the standard hybrid startup with NBI pre-heating, but we will keep the density as low as possible and delay the L-H transition until well after Ip flat top by keeping the plasma shape strongly biased to an upper X-point. The large step-up in NBI power at the start of Ip flat top (which normally triggers the L-H transition) will instead lead to the formation of a strong ITB in and L-mode edge plasma. If allowed to continue for too long, which will eventually lead to a beta limit as the pressure profile is too peaked. However, just before we reach this limit we will trigger the L-H transition by jogging the plasma shape to a lower X-point, which will broaden the pressure profile as the H-mode pedestal forms. As the ITB relaxes to a hybrid-like regime, we expect to form a QH-mode edge owing to the strong co-rotational shear near the plasma edge. Note that after the L-H transition we will return the plasma shape to an upper X-point because that configuration gives the highest toroidal rotation rates.
It is important to have q95>4 so that we are in the regime where sawteeth are normally suppressed in hybrid discharges. Since the QH-mode is dependent on q95, we can make some adjustments to give us the type of EHO that we want. The Ip flat top should expend to at least 6 s to allow the current profile time to fully relax. Besides demonstrating a hybrid plasma with a co-rotation QH-mode edge, we want to probe the stability limits by scanning beta_N from a low of 2.5 to above 3 (typically the no-wall limit is around beta_N=3.2). The last phase of the experiment explores the sensitivity of this new hybrid regime to the presence of the 3/2 mode. This will be done by using co-ECCD aimed at the q=1.5 surface to suppress the 3/2 mode on demand. In the normal ELMy hybrid regime the disappearance of the 3/2 mode results in a drop in qmin below 1. We want to see if the q profile responds to the presence of the 3/2 mode in this ELM suppressed regime. | ||
| Background: | The hybrid scenario with an ELMy H-mode edge is well known to have an anomalously broad current profile with qmin>1. This is a beneficial feature that helps to avoid triggering the 2/1 mode (owing to the lack of sawteeth and fishbones) and perhaps improves transport as well. Analysis of MSE data indicated that the anomalously broad current profile is due to poloidal magnetic flux pumping that occurs during an interaction between the 3/2 mode and ELMs [CC Petty et al., Phys. Rev. Lett. 102 (2009) 045005]. Thus, if this is correct, we should expect the current profile to behave classically (or neoclassically) in an ELM-suppressed hybrid plasma.
Previous DIII-D experiments successfully suppressed ELMs with an n=3 RMP at q95=3.6 in the hybrid scenario at moderate beta values (beta_N<2.5) [CC Petty et al., Nucl. Fusion 50 (2010) 022002]. However, these plasmas were likely to slow down and lock owing to an interaction between the tearing modes and the RMP. Thus, it would be programatically beneficial to create an ELM-suppressed hybrid regime without RMP, which will likely be able to achieve higher beta_N. The QH-mode edge with co-NBI is a candidate for this. | ||
| Resource Requirements: | All co-NBI sources are required.
Four gyrotrons are required. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 314: Optimization of castellation for a new W-divertor of ITER: studies of shaped castellation in DIII-D | ||
| Name: | Rudakov rudakov@fusion.gat.com | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | A. Litnovsky (FZJ), M. Hellwig (FZJ), V. Philipps (FZJ), C.P.C. Wong (GA), R.
Boivin(GA), N. Brooks (GA), J. Watkins (SNL), P. Stangeby (Univ. of Toronto), A.Mclean (GA), Y. Krasikov (FZJ), J. Boedo (UCSD), R. Moyer (UCSD), D. Matveev (FZJ), M. Komm (IPP Prague), G. De Temmerman (DIFFER), R. Pitts (ITER), M. van den Berg (DIFFER) |
ITPA Joint Experiment : | Yes |
| Description: | The aim of this experiment is to evaluate the efficiency of deposition mitigation in the shaped castellation proposed for a new tungsten divertor of ITER. Castellation cells having roof-like shape with rounded edges 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 whereas rounded edges should increase the necessary plasma-wetted area of the castellation around the gaps. Metallic plates below the castellation should provide the
information on deposition at the bottom of the gaps. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Impurity deposition and undesirable fuel accumulation in the gaps of castellated structures represent known safety issue for ITER. Recent studies revealed significant deposition at the bottom of the gaps which is yet to be understood and reliably modeled. Shaping of castellated structures is the most direct way for reduction of the impurity deposition and fuel
accumulation in the gaps of castellation. Experiments in TEXTOR and DIII-D have proven the expected advantages of shaped castellation. It is planned to elaborate these experiments by making the long-term piggyback exposure using DiMES system. A conventional (rectangular) and new shaped tungsten castellation will be exposed simultaneously to allow for a direct comparison. To obtain the representative deposition patterns in the gaps a piggyback one-week exposure preferably in LSN configuration, is requested for the castellation. The metal plates installed below the castellation will be used to collect the deposited material at the bottom of the castellation. The gaps and instrumented plates will be analyzed at FZJ. | ||
| Background: | The castellated armor of the first wall and divertor in ITER will be used to maintain the durability of the machine under the thermal excursions during plasma operation and to alleviate the forces caused by induced currents. However, the impurity deposition and fuel accumulation in the gaps of castellated structures represent 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.
Mitigation of the fuel accumulation in the gaps by the gap shaping and study of material migration towards the bottom of gaps are among key topics of a task DSOL 27 of the IEA-ITPA Joint Experiments Program. Within this task the comparative modeling studies of conventional and shaped castellation were made. In particular, the PIC code SPICE2 results predict a full suppression of the ion flux in the gaps of shaped castellation accompanied with a drastic decrease of impurity deposition as modeled with Monte-Carlo 3D GAPS code. To validate these results experimentally dedicated multi-machine investigations are ongoing on several tokamaks worldwide. The same design of a castellation will be used in DIII-D, ASDEX-Upgrade, EAST, KSTAR, LHD 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. 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. Another advantage of this experiment is the possibility of a direct comparison with experimental results from the other major tokamaks involved in multi-machine studies. | ||
| Resource Requirements: | 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 | 315: Current profile control model requirements | ||
| Name: | Walker walker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Control | Presentation time: | Not requested |
| Co-Author(s): | Eugenio Schuster, Didier Moreau | ITPA Joint Experiment : | No |
| Description: | Develop criteria for defining requirements on models for development of current profile controllers. Evaluate existing models based on those criteria. The near-term purpose of this work is to establish the range of applicability of controllers for DIII-D experiments and to provide confidence in using profile control for routine experimental operation. To support longer term efforts, we would like to identify predictive accuracy targets for transport models. â??Good enough" transport models may be achievable in time to use for current profile control in ITER. Previous experiments on profile control provide some useful data to begin analysis. However, previous and ongoing modeling work emphasize creating models that are as predictive as possible. In addition, the profile control experiments conducted intentionally limit excursions from nominal plasma conditions. We wish to obtain a few shots in which plasma conditions are varied substantially to allow understanding of the loss of predictive accuracy and the consequences on control. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Work would begin by simulating profile control using controllers that were tested experimentally, to evaluate control degradation with variations in the plasma model from "nominalâ?? and control performance when regulating to a target operating point different from nominal conditions. These simulations would be compared with experimental data whenever the experiment and simulation conditions can be matched. When experimental data with matching conditions is not available, this data would need to be generated by piggybacking on other current profile control development activities. Since profile control work this year will likely focus on a single pre-selected DIII-D target scenario, allowed variations of nominal plasma conditions may be limited. Thus wider variation of plasma conditions may the subject of future work. | ||
| Background: | Development of controllers capable of full current profile control consists of two parts: (1) generating a model capable of predicting future profile evolution from actuator inputs and (2) using that model to build a controller. Good full-profile control (item (2)) was demonstrated on DIII-D in experiments conducted in 2011-12. Two different models used for developing the controllers (item (1)) both relied on some "identification" from data. One modeling approach used a data fit to a linearized â??physics structureâ?? model. A second approach used a physics-based nonlinear poloidal flux evolution model combined with empirical actuator response models. We prefer, in general, to generate predictive model from basic physics principles because this allows confident extrapolation to other plasma regimes (in DIII-D) and to other devices (e.g., ITER, FNSF). Experience with DIII-D control experiments suggest that the models used in 2011-12 were "good enough" even when somewhat outside the regime used for controller designs. The objective of the proposed effort is to support the further development and validation of profile control models by defining objective metrics for required model prediction accuracy, i.e. to better characterize what "good enough" means. | ||
| Resource Requirements: | An occasional shot or portion of a shot during allocated current profile experiments to collect data needed to evaluate models.
All neutral beams and all gyrotrons active (but not necessarily all in any given test shot). | ||
| Diagnostic Requirements: | In realtime: MSE, Thomson, line average density, standard magnetics (processed through realtime MSE efit), beam powers, ECH powers, CER.
For offline analysis: ECE, density profile, toroidal rotation profile, ion and electron temperature profiles | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | |||
| Title | 316: Model-based control of the current profile and βN for steady state scenarios | ||
| Name: | Moreau didier.moreau@cea.fr | Affiliation: | CEA Cadarache |
| Research Area: | Plasma Control | Presentation time: | Not requested |
| Co-Author(s): | Michael Walker (GA), Eugenio Schuster (Lehigh University) | ITPA Joint Experiment : | No |
| Description: | This proposal aims at continuing the development of model-based current profile control in steady state scenarios started in 2011/2012. The goal is to obtain, in a reproducible manner, various requested target q-profiles and βN values for the high-βN phase of the steady state scenarios, by applying the control as early as possible during the ramp-up phase. Magnetic and kinetic profile control algorithms have been implemented in 2011 in the DIII-D PCS. The control actuators are (i) co-current on-axis NBI power, (ii) co-current off-axis NBI power, (iii) counter-current NBI power, (iv) total ECCD power from all gyrotrons in a fixed off-axis current drive configuration, and (v) loop voltage (ohmic coil). Closed-loop experiments were performed successfully in 2011 and 2012 [1] (see background). Now the method must be validated with a broader variation of the target plasma configurations and profiles with the aim of making profile control routinely used as an experimental tool. For this purpose, dedicated non-linear simulations tools are being developed in order to adapt the controller models to changes in the actuators and plasma configuration, and also to choose the optimal controller parameters before each experiment. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The control of the poloidal flux, Ï?(x), and iota (i(x) =1/q(x)) profiles (x is the normalized radius) will be tested with various consistent [Ï?(x), i(x)] targets, starting control at t = 1 s (during ramp-up) for 5 seconds. First we shall reproduce the i(x) control obtained during current ramp-up in 2012 with target profiles based on steady state shot # 147634 (qmin=1.7) and extend it beyond the ramp-up phase, up to t = 6 s, using different profile controller parameters, possibly optimized through dedicated non-linear simulations that are being developed. Then, we shall track different [Ï?(x), i(x)] targets such as those obtained in the steady state experiments with various qmin. If time allows, we shall finally add simultaneous βN control with the same [Ï?(x), i(x)] targets and various βN targets. | ||
| Background: | Real-time control of the plasma current profile and βN is important to achieve stable and reproducible operation of tokamaks in the advanced steady state regime. The ability to regulate the requested normalized plasma pressure and the current profile is of great potential interest for physics studies in which they play an essential role, as it would require much less experimental time to obtain the adequate actuator waveforms in order to reach a particular goal. A multi-variable approach based on a semi-empirical dynamical plasma model has been developed 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 [2]. For integrated magnetic and kinetic control, the 2-time-scale controller design uses singular perturbation techniques and is quite generic (i.e. independent of the tokamak) so it can be extrapolated to ITER. It takes into account the strong coupling between the kinetic parameters (such as βN) or profiles and the current profile. The control-oriented (semi-empirical) models to be used to determine the controller matrices have been obtained from system identification experiments performed on DIII-D in 2009. The models were shown to provide excellent fits to the experimental data [3], 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 shows the relevance of the linearized data-driven models for control applications. The proof-of-principle experimental tests of this control method performed in 2011/2012 were successful and the results were presented at the 2012 IAEA Fusion Energy Conference [1].
References: [1] D. Moreau et al., â??Integrated Magnetic and Kinetic Control of Advanced Tokamak Scenarios Based on Data-Driven Modelsâ??, IAEA Fusion Energy Conf., San Diego, 2012, paper ITR/P1-20. [2] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. [3] D. Moreau et al., Nucl. Fus. 51 (2011) 063009. | ||
| Resource Requirements: | The ongoing current profile control development will require half a day in 2013. Prior to this, some dedicated PCS tests and possibly a few model identification/validation experiments should be conducted in a couple of 2-hour preliminary sessions. NBI at full power will be 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 will also be required. | ||
| 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 / data stored in mdsplus. | ||
| Other Requirements: | |||
| Title | 317: Study on sawteeth in the neutral beam (NB) heated plasmas | ||
| Name: | Bak jgbak@nfri.re.kr | Affiliation: | Korea National Fusion Research Center |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | J.H. Kim, E.J. Strait, R.J. La Haye, B. Tobias, E. Hollmann, W. Heidbrink, G. McKee, C. Holcomb, R.J. Buttery, S.G. Lee, K.D. Lee, W.H. Ko, H.S. Kim, W.C. Kim, Y.S. Bae | ITPA Joint Experiment : | No |
| Description: | Long period sawteeth, due to fast ions, triggered the NTMs degrading plasma confinement and causing plasma disruption, so control of the sawtooth period is required for reducing the NTM occurrence due to sawteeth. In this experiment, the study on the sawteeth in the NB heated plasma will carried out for the sawteeth control by using NB. There are two items for the study.
Item#1 : Investigation of sawteeth behavior in co/counter NB heating - Counter on-axis NB power for achieving minimum sawtooth period in L-mode / H-mode plasmas - Growth rate of n=1 internal kink mode vs. sawtooth period - Effect of fast ion on the period near q=1 surface - Core impurity accumulation as increasing the period - Coupling between internal kink mode and other MHDs by poloidal density asymmetry in high core-rotation due to co-NB heating - Dependence of counter on-axis NB power for achieving minimum sawtooth period on plasma parameters and plasma shape Item #2 : Investigation of Sawteeth behavior in off-axis co-NB heating - Reduction of sawtooth period due to off-axis co-NB heating - Effect of fast ion on the period near q=1 surface Some investigations from this experiment will be compared with KSTAR. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Item #1 : For fixed counter on-axis NB power (210 deg. beam line) of ~ 2 MW (or optimum power?), co on-axis NB powers (three beam line) are varied scanned from 0 to 10 MW (or max power) during L-mode and H mode discharge (Ip : 0.8 ~ 1.5 MA, BT: 1 ~ 2.5 T, ne : 2 ~ 6 x 1019 m-3, q95 : 3~4 )
Item #2 : For fixed co on-axis NB powers (two beam lines) of ~ 2 MW ( or optimum power ?), co off-axis NB power (150 deg. beam line, optimum tilt angle ?) is varied from 0 to 3 (max. power ?) during L-mode discharge. Note that the 210 deg. beam line is turn off in this experiment. ( Ip : 0.8 ~ 1.5 MA, BT: 1 ~ 2.5 T, ne : 2 ~ 6 x 1019 m-3, q95 : 3~4) | ||
| Background: | Experiments for studies on the sawteeth controls were done in MAST, JET, TEXTOR and AUG; 1) Asymmetric stabilization of sawteeth by NBI heating: sawtooth period became longer as co-NB power increases, and minimum value of the period was observed in the counter-NB regime. The minimum value was shorter than the period in Ohmically heated plasmas, 2) Destabilization of sawteeth by off-axis co-NB heating: The sawtooth period became shorter when heated with off-axis co-NB than with on-axis co-NB beams NB. The stabilization and destabilizations of sawteeth in the NBI heated plasmas are strongly related to the parameters such as toroidal flows due to NB heating, fast ion distribution near q=1 surface, ion diamagnetic drift frequency. We will carry out the investigation of sawteeth behaviors in the NB heated plasmas during L and H-mode discharges for the sawtooth control by the NB heating | ||
| Resource Requirements: | NBI ( counter-NB, off-axis NB ), ECH | ||
| Diagnostic Requirements: | ECE radiometer and SXR with low and high sampling rate, ECEI, magnetic diagnostics, CER, MSE, Diagnostics for impurity measurement (Ar at core), BES, Thomson scattering, density interferometer, reflectometer, Diagnostics for fast ion (distribution) detection, BES | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 318: Current profile control for stable ITER baseline scenario plasmas | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | D. Humphreys, G. Jackson, T. Luce, W. Solomon, M. Walker | ITPA Joint Experiment : | No |
| Description: | Analysis of ~100 ITER b.s. discharges has indicates that detailed features of the current profile in the outer half of the plasma are closely linked to the triggering of tearing modes that limit the performance and duration of this scenario. Global changes in the current profile are likely responsible for changes in the classical tearing stability term (Î?â??), that destabilize the 2/1 tearing mode. While direct stabilization of an island has already been demonstrated, this analysis and experiment is aimed at modifying the global current profile to avoid having to directly stabilize any island and without pre-emptive stabilization.
This experiment would provide â?¢ a direct test of the hypothesis postulated from a 2-yr analysis (see background section) using a controlled environment where the current profile changes are systematic and externally controlled â?¢ a method to control the current profile in the region where it matters for tearing stability, to steer the plasma away from tearing stability boundaries, without sacrificing betaN, performance or rotation |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Run an ITER discharge placed on the stability boundary (possibly low torque) and monitor the MSE pitch angle time history of the stable and unstable branch in the region of interest. Correlate with RT-efit reconstructions of J. Place single gyrotrons at the locations between MSE channels 11-44, roughly between rho=0.5 and rho=0.9.
b. Checkout the new MSE-ECCD feedback algorithm (take data in the PCS and check the response, without activating the algorithm â?? described under analysis requirements), while using feed-forward, shot-to-shot control of J: for the unstable case, repeat adding ECCD at the location that shows narrowing of the MSE channels (supposed to lead to instability). Create cases that become unstable or remain stable â??at willâ??, by using co- and counter-ECCD at different locations. c. Use feedback to control MSE channels, and hence J: link the difference between 2 contiguous MSE channels signals to the gyrotron(s) aimed at the location between those two MSE channels - when the difference falls below a threshold (TBD), turn on the corresponding gyrotron(s). Check the threshold values, the timing of the feedback, the necessary gyrotron power (modulated ECCD power ok?). d. Run discharges with the feedback on, in all the locations of the stability boundary that would not be stable otherwise (low/zero torque, low density, lower li, etc). The algorithm will measure and control the MSE pitch angles (channels 11-44), by modifying the current profile by means of ECCD, to reduce or increase the MSE pitch angles differences locally (which represent the amount of plasma current present at that location). | ||
| Background: | The ITER baseline scenario is know for being prone to the triggering of m=2,n=1 tearing modes, that spoil the performance, perturb the current profile in a way that is not recoverable, and often lock to the wall forcing an early termination of the discharge. It was recently demonstrated that the evolution of the current profile is the main effect responsible for the triggering of this instability[1]. Analysis of >100 ITER baseline scenario shots, and new experiments in the 2012 DIII-D campaign have explored the tearing stability boundaries in betaN, li, torque, and density. The statistical analysis shows that all the discharges that have â??j95nâ??>0.5 (average normalized current at 95% of the poloidal flux) are stable, while with lowed j95n only high density shots are observed to be stable. Independent analysis of raw MSE pitch angles time histories for a small sample of the discharges reveals that the statistical study is consistent with the trends observed in the raw data. In the unstable cases, the MSE pitch angles evolve in such a way that less current is detected at rho~0.8 and more current builds up at rho~0.6 (the total Ip is kept constant). The q=2 surface is located between these two rho values. This points to a rather simple explanation for the triggering of the limiting instability in these discharges: when less current is present outside the q=2 surface (whatever the physical mechanisms that caused this), more current will build up on the inner side because the total current is kept constant. This causes a higher current gradient to be created at the rational surface, which is the main destabilizing effect for tearing modes.
Since the raw MSE time histories provide direct information about the local current profile, independently from any equilibrium reconstruction, this study gives an independent confirmation of the statistical observations, which are based on reconstructed efit equilibria. We propose to directly check if the physics hypothesis is correct, by controlling, off-line, in feedforward and in feedback, the local features of the current profile, that were observed to be related to the tearing stability of the discharges. Note that this method does not rely on direct stabilization of an existing island, nor on the continuous ECCD power to pre-emptively stabilize tearing mode, but instead it constitutes a physics basis for tailoring the current profile to avoid having to use direct stabilization at all. This will also directly provides a simple way to steer the discharges away from instability, without sacrificing performance (i.e. not lowering betaN), and potentially with minimal extra injected power during the discharge. [1] F. Turco and T.C. Luce, Nucl. Fusion 50 (2010) 095010 | ||
| Resource Requirements: | 30 and 330 NBI sources, both 210 NBI sources. 6 gyrotrons at max power (stand-by power capability needed). | ||
| Diagnostic Requirements: | Magnetics, MSE, CER, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | Create and implement MSE-ECCD feedback control algorithm in the PCS before the experiment. Basic logic needed:
- measure RT-MSE pitch angles and have at least channels 11-44 available in the PCS algorithm. Compute differences between contiguous channels (40-11, 41-40, etc). Connect this category to a dud trip that switches phases in the gyrotron power category. - Assign one (or 2, depending on the configuration, TBD) gyrotron power signal to one MSE difference-signal (gyrotrons have fixed positions, the assignments are done before the experiments, off-line) - Set a threshold â??Tâ?? for the MSE differences (TBD): when 1 or more (TBD) of the signals go below T, turn on the corresponding gyrotron(s) - When the MSE-difference signal returns below T, turn off the corresponding gyrotron. | ||
| Other Requirements: | |||
| Title | 319: Impact of fast-ions on the RWM stability boundary | ||
| Name: | Turco turcof@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | General Physics | Presentation time: | Not requested |
| Co-Author(s): | J. Berkery, J. Bialek, J. Hanson, M. J. Lanctot, G. A. Navratil, M. Okabayashi, C. Paz-Soldan, S. Sabbagh, T. Strait, A. Turnbull | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to study the physics of the RWM in the absence of fast-ions, in order to determine if the apparent stability of DIII-D discharges to the destabilization of the RWM is due to the presence of stabilizing fast particles. We propose to eliminate the presence of fast-ions completely, and asses the RWM stability under those conditions, by producing ECH-only plasmas (similar discharge obtained in the 2012 campaign, plus an extra 0.9 MW of ECH available in 2013) and then ramping the plasma current up quickly to produce an unstable current-driven RWM. The goal is to map the stability boundary in the complete absence of fast-ions, then progressively add NBI power (and hence adding fast-ions to the scenario) to determine whether the stability boundary moves, and ultimately if the RWM is stabilized in the presence of fast-ions, and by what amount and distribution of particles. When adding NBI power, care must be taken to use balanced co-counter injection to obtain the same level of rotation with and without beams.
Even though the proposed scenario has intrinsically rather low betaN (~1.8), the physics of the destabilization of the RWM with and without fast-ions is universal and this experiment would provide crucial information on whether RWM stability poses a threat to the future machines that will not rely on fast-particle-producing heating systems. Moreover, if a way is found to run the scenario with qmin>1 for long enough, the experiment would provide the perfect (and only) platform to benchmark the stability codes (MARS-K, MISK, etc) that are needed to predict the MHD characteristics of future machines. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | a. Produce an ECH-only, diverted H-mode discharge (a starting point is 150840), with all the available ECH power (3.6 MW?). Insert a fast enough Ip ramp to destabilize a current-driven RWM.
b. Obtain some data on the Ip ramp-rate necessary to destabilize the mode, and attempt to modify the scenario to obtain qmin>1 (early heating, co-ECCD off-axis, counter-ECCD on axis?). c. When the RWM is systematically reproducible, add balanced NBI power, as much as it can be used while maintaining a fixed betaN level (estimated 1-3 MW NBI), with on-axis or off-axis sources. Observe how (and if) the RWM stability boundary changes: is the original discharge stabilized? Is a faster/slower Ip ramp necessary to destabilize the mode? How does the stability change with on-axis vs off-axis fast-ion distributions? | ||
| Background: | Theoretical models suggest that the RWM stability is strongly affected by the presence of fast particles. In particular, is has been postulated that the RWM is stabilized in the DIII-D scenarios because of the fast ions produced by the NBI power, and machines that do not rely on NBI for heating and current drive may have issues with unstable RWMs when operating above the no-wall limit. Studies to assess the effects of fast-ions on the RWM stability have been preformed in DIII-D. However, while it was possible to modify the localization of the fast-ions and the fast-ion distribution function by means of the OA-NBI system in the previous campaign, this approach has proven difficult and the interpretation of the results hard to determine. This experiment proposes to tackle the problem in a different way, focusing on nailing down the physics in a scenario that allows for significant changes in the amount of fast-ions, and for a simple way to assess the RWM stability. The ECH-only shot would have no fast particles, and the discharges with different levels of NBI power and different injection angles will provide the direct knob to determine the effect of the amount and distribution of fast-ions. This test will allow us to create a map of the RWM stability under controlled conditions for fast-particles and rotation profiles. | ||
| Resource Requirements: | 30 and 330 NBI sources, both 210 NBI sources. 6 gyrotrons at max power and max duration. | ||
| Diagnostic Requirements: | Magnetics, MSE and CER when NBI usage allows, Thomson scattering, ECE radiometer, density interferometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 320: RWM excitation in the SSI target by reducing triangularity" | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Francesca Turco, | ITPA Joint Experiment : | No |
| Description: | Recently, the SSI group has developed betan AT plasmas ( 1.5 < q_min < 2) bn~4 near MHD limit. In these discharges, we observed EP-driven RWM and ELMs. Another observation was slow n=1 bursting mode without any D-alpha spike and mirnov MHD activity ( like n=1 magnetic bubble). <br> <br>One hypothesis is that these discharges are caused by physics processes relate dto the ideal MHD driving force. <br>The traditional approach for stabilizing high beta plasmas can be useful for improving the stability. For example, It is quite possible to increase betan by 10-20% by modifying the triangularity. <br> <br> This proposal is related to the ROF #138 | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 321: Access to low torque QH-mode using high beta low torque AI target with NRMF | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | AM Garofalo | ITPA Joint Experiment : | No |
| Description: | The aim of the experiment is to try to integrate ELM-free operation into the low torque advanced inductive scenario by utilizing the QH-mode edge. The aim would be to combine the favorable characteristics of QH-mode at low rotation, with demonstrated stationary, low torque operation above betaN>3 from the advanced inductive. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a low torque AI like #145338 and working on the early density to allow access to QH-mode, and applying NRMF to try to establish the edge ExB shear believed needed for QH-mode. This discharge is unstable to 2/1 NTMs without EC power, and lower density will certainly exacerbate this, so EC will likely be needed. This, may be unfavorable for QH-mode operation. Still it may be possible to remove the EC once the NRMF is turned on. | ||
| Background: | In FY11, advanced inductive discharges were successfully initiated and sustained with low torque and normalized fusion performance approaching the requirements for ITER Q=10 operation. One obvious missing element of this regime as an operating scenario for ITER is ELM-free operation. Integrating advanced inductive operation with a QH-mode edge may offer a solution to this, while potentially also enabling a recovery of the lost confinement associated with low torque operation to date. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons, 210 beams, I-coils for nRMF, possible counter-Ip | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 322: ELM pacing in low torque, electron heated ITER baseline | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | G. Jackson, F. Turco, N. Commaux, L. Baylor | ITPA Joint Experiment : | No |
| Description: | The aim of this experiment is to test the performance of pellet pacing in low torque, electron dominant heated ITER baseline plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This should be an extension to eg proposal #118, which starts with an ECH dominant IBS plasma (progressing from #150480) and extends to lower q95. If this is successful, pellet pacing should be deployed to integrate an additional important aspect of the ITER baseline scenario (ELM control). The impact on confinement and stability should be investigated. | ||
| Background: | Pellet pacing has successfully been used in ITER like discharges. However, these have been highly rotating, NBI heated discharges. Work in 2012 has established low torque IBS plasmas, and has begun to explore dominant EC heating. A logical extension is to include ELM control. RMP suppression has proven challenging at low rotation, presumably owing to undesirable field penetration deep in the plasma. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons with NTM control, 210 beams, pellets | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 323: Access to QH-mode with low NBI torque | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | KH Burrell, AM. Garofalo | ITPA Joint Experiment : | No |
| Description: | The aim of this experiment is to determine the torque and rotation requirements for access to QH-mode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This is an addendum to ROF #53, and would utilize work performed in ROF #118 and #322. One would start with a low torque ITER baseline discharge and add NRMF before the L-H transition to try to drive sufficient rotation shear to enter QH-mode directly, ideally with no ELMs. If the early NRMF is deemed undesirable (eg increase L-H power threshold), then delay the NRMF until after the L-H transition, and use pellet pacing to mitigate the early ELMs until a transition to QH-mode is achieved. If QH-mode is not achieved in this way, go to maximum counter NBI for the startup and then work back toward ITER relevant torque. | ||
| Background: | The torque from NRMFs have been successfully used to enable operation of QH-mode at low torque. However, to date, these fields have only been applied after the high rotation shear conditions needed for QH-mode have been established with neutral beam torque. Since the torque from the NRMF shows a significant beta dependence, it is not known whether the torque will be adequate to allow a direct transition from L-mode to QH-mode. In addition, the possible impact of NRMFs on the H-mode power threshold has not been documented for low collisionality QH-mode conditions. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons, 210 beams, pellets, n=3 NRMF | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 324: Document turbulence change with rotation in AI, and compare with theory | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | GR McKee | ITPA Joint Experiment : | No |
| Description: | The aim of this experiment is to document changes in turbulence levels associated with controlled changes in the rotation and q profiles, for benchmarking and validation against transport codes including GYRO and TGLF. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat discharges based on torque ramp down shots, and halt the ramp down at different rotation levels. Avoid ECH for mode control until necessary (lowest rotation levels). Repeat for different q95 and q-shear. | ||
| Background: | A significant reduction in confinement has been observed for AI discharges at low rotation, and while changes in turbulence resulting from the reduced ExB shear has been implicated, this has never been directly confirmed. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons with NTM control, 210 beams | ||
| Diagnostic Requirements: | BES, FIR, DBS and all fluctuation diagnostics | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 325: Document turbulence change with rotation in AI, and compare with theory | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | GR McKee | ITPA Joint Experiment : | No |
| Description: | (Duplicate of #324) The aim of this experiment is to document changes in turbulence levels associated with controlled changes in the rotation and q profiles, for benchmarking and validation against transport codes including GYRO and TGLF. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat discharges based on torque ramp down shots, and halt the ramp down at different rotation levels. Avoid ECH for mode control until necessary (lowest rotation levels). Repeat for different q95 and q-shear. | ||
| Background: | A significant reduction in confinement has been observed for AI discharges at low rotation, and while changes in turbulence resulting from the reduced ExB shear has been implicated, this has never been directly confirmed. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons with NTM control, 210 beams | ||
| Diagnostic Requirements: | BES, FIR, DBS and all fluctuation diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 326: Understand mechanisms by which ECH influences NTM stablility | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R. Prater | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to document and better understand the means and conditions whereby ECH (with or without current drive) can lead to improved stability against 2/1 NTMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Conditions for 2/1 NTM onset will be measured by ramping the torque down at different levels of betaN and at different q95's. During these ramp downs, ECCD should be applied and scanned shot-to-shot inward from the q=2 surface toward the axis. This should also be repeated in a nominally heating only configuration, and varying between narrow and broad radial deposition. Document which conditions ECH is successful in expanding into the otherwise unstable operating regions. Use density feedback control to maintain constant density, choose a shape for optimal edge TS for good edge BS measurement, and optimize beams for best MSE. For consistency, check some key points at fixed torque with betaN ramp up. | ||
| Background: | Experiments in FY11 showed that EC power was typically needed to access high beta, low torque states in advanced inductive plasmas. In most cases, the EC was configured to drive current *near* (but without any optimal alignment) at the q=2 surface. This, however, was realized as a limitation in terms of being able to go to lower field and higher betaN. Since we were not especially well-aligned in any case, we decided to do a deposition scan. To our surprise, EC power could be deposited well inside the q=2 surface, and even configured for heating instead of current drive, yet the same benefits to stability were realized. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 327: rho* scaling of intrinsic torque between DIII-D and JET | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Plasma Rotation | 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 (steps and low frequency modulation) to infer intrinsic torque as described eg in Solomon et al PoP 2010, and combine with modulation techniques to simultaneously measure chi_phi and vpinch. Since edge intrinsic torque is believed to depend on pedestal gradients and edge temperature, perform combinations of betaN and Ip scans around the identity match 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. By combining the rho* scans from DIII-D and JET (starting from the matched pair), a wide enough scan is produced so that ITER extrapolation becomes reliable. This experiment has been prepared and approved on JET. | ||
| Resource Requirements: | 1 day experiment, modulated co and counter NBI | ||
| Diagnostic Requirements: | Full profile diagnostics, especially CER, plus FIDA and main ion CER. | ||
| Analysis Requirements: | TRANSP, intrinsic torque + modulated transport analysis | ||
| Other Requirements: | |||
| Title | 328: Error field correction at low rotation | ||
| Name: | Solomon solomon@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | AM. Garofalo | ITPA Joint Experiment : | No |
| Description: | The aim of this experiment is to determine if there is a preferable error field correction to use at high betaN and low rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We should begin by utilizing the dynamic error field correction (DEFC) system to search for an improved correction, similar to what was attempted previously in the low torque AI experiments. Different to that attempt, we should perform the beta ramps needed for DEFC in a low torque/rotation discharges, stabilized against 2/1 NTMs with ECH. In addition, some effort should be made to perform a specific correction arising from the B-coil current feed. This might best be done using independent feedback on the upper and lower I-coils using the new magnetics. We should then remove the ECH stabilization and compare the accessibility to these low torque states between the different EFCs. If an improved EFC is obtained that allows stable, high betaN, low torque operation, then the torque should be ramped up to large co-NBI. Measurements last year indicated that it was difficult to spin the plasma up from the low torque state, perhaps indicative of an edge island - improved EFC may help avoid opening this island and remove the apparent hysteresis in rotation and confinement. | ||
| Background: | Experiments in FY11 attempted used DEFC to determine an improved EFC, based on a beta ramp at high rotation. The DEFC approached a solution that used approximately 50% higher coil currents than the standard EFC algorithm, a result that has often been noted for high beta plasmas. However, when these new multipliers were applied for use during torque ramp downs at fixed betaN, we did not realize any benefit in terms of the lowest achievable torque before 2/1 onset. The question remains whether the plasma becomes more sensitive to different error fields at low torque (for example, increased sensitivity to the localized B-coil error), which a different EFC optimized at low rotation may better deal with. | ||
| Resource Requirements: | 1 day expt, 6 gyrotrons, 210 beams, DEFC capability (independent upper/lower I-coils, and C-coils) | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 329: Discussion of plasma rotation transient process during OFM/ELM-driven RWM w/wo feedback usin | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Yueqinag Liu | ITPA Joint Experiment : | No |
| Description: | Current hypothesis of RWM stabilization in high beta plasmas is that sufficient energetic particles (EP) and high plasma rotation can increase the stabilizing effect over the external kink driving source. <br> <br>However, the major contributor EPs potentially can cause deleterious effect by inducing various types of MHD activities. In particular, onset of EP-driven off-axis fishbone modes (OFM) causes rapid decrease of EP population. This results in decrease of another stabilizing effect of plasma rotation within a short time period before these properties can be restored by adjusting auxiliary heating system [1]. Here, the rotation drop is partly due to non-ambipolar Er buildup with rapid EP losses. In addition, the increase of mode amplitude reduces the rotation due to increase of dissipative Electro magnetic(EM) torque. <br> <br>Thus, the discussion based on the steady state criterion only may become irrelevant once plasma reaches near the ideal MHD stability limit. <br> <br>In addition, when the feedback is used, the applied feedback field can induce non-resonant effect modifying rotation profile due to NTV. <br> <br>Recently, a time dependent MARS-Q code has been developed [2] to include the time evolution of mode amplitude and rotational profile. Various viscous effects, such as Electro magnetic(EM) torque and NTV are included in this code. <br> <br>Plasma condition <br> <br>Recently, the SSI group has developed betan AT plasmas ( 1.5 < q_min < 2) bn~4 near MHD limit. In these discharges, we observed EP-driven RWMs and ELMs, and slow n=1 magnetic burst without any D-alpha spike. These discharges are well suited for studying RWM stability. <br>The initial objective of this experiment is to document the rotation and Ti profile evolution with fast time scaling to assess the adequacy of MARS-Q applicability. Then, we use the ode as a simulator to improve the plasma performance. <br> <br> Ref <br>(1) M. Okabayashi et al, Phys. Plasmas 18, 056112 (2011); <br>(2) Y.Q. Liu et al., Plasma Phys. Control. Fusion, 2012, 54124013 <br>(3) M. G. Matsunaga et al., IAEA 2012 EX5-1 | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 330: QH-mode with TBM | ||
| Name: | Smith smithsp@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal is to produce a QH-mode plasma with the TBM in and energized. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Starting from 145117 or similar QH-mode, turn on the TBM at various levels and reestablish QH-mode conditions. | ||
| Background: | QH-mode provides a possible ELM-free high performance regime in which ITER could operate. The TBM mockup in DIII-D could provide counter Ip NTV torque to keep the plasma counter rotating. | ||
| Resource Requirements: | Typical setup for QH-mode (reversed current). TBM. 3D fields coils (I & C). | ||
| Diagnostic Requirements: | Magnetics. All profile diagnostics. Fluctuations would be nice if the plasma is in the right parameter space. | ||
| Analysis Requirements: | Power balance. CERFIT. Plasma magnetic perturbation analysis (M3D-C1). | ||
| Other Requirements: | |||
| Title | 331: QH-mode with TBM (dup 330) | ||
| Name: | Smith smithsp@fusion.gat.com | Affiliation: | GA |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal is to produce a QH-mode plasma with the TBM in and energized. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Starting from 145117 or similar QH-mode, turn on the TBM at various levels and reestablish QH-mode conditions. | ||
| Background: | QH-mode provides a possible ELM-free high performance regime in which ITER could operate. The TBM mockup in DIII-D could provide counter Ip NTV torque to keep the plasma counter rotating. | ||
| Resource Requirements: | Typical setup for QH-mode (reversed current). TBM. 3D fields coils (I & C). | ||
| Diagnostic Requirements: | Magnetics. All profile diagnostics. Fluctuations would be nice if the plasma is in the right parameter space. | ||
| Analysis Requirements: | Power balance. CERFIT. Plasma magnetic perturbation analysis (M3D-C1). | ||
| Other Requirements: | |||
| Title | 332: Demonstration of noninductive Q=5 scenario in ITER shape using off-axis NBI | ||
| Name: | Park parkjm@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate fully non-inductive and Q=5 condition simultaneously in ITER shape using off-axis beam and higher power ECH; This is resubmission of 2012 proposal 103. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start at relatively high q95 (~6). Develop higher qmin scenario with a larger radius of qmin 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. Optimize q95: move to higher (lower) q95 if fNI<1 (Q>5). Document differences between ITER SS demonstration discharges and "DIII-D standard" SS scenario in double null (DN) shape. | ||
| 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]. This experiment aims at simultaneous achievement of the fNI=1 and Q=5 goals using off-axis beam and high power ECH. TGLF simulation suggests that a larger radius for the minimum of q helps to increase both fNI and fusion performance by maximally utilizing the benefits of low magnetic shear. As in the 2011~2012 experiments in DN shape, it is very promising to develop high qmin (~2) scenario in ITER shape using off-axis beam and high power ECH. Importantly, the power requirement for the goal is lower than the 2011 high qmin steady-state discharges in DN shape. | ||
| Resource Requirements: | All neutral beam sources with 150 beams at maximum tilt angle. All available gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 333: Far off-axis NBCD | ||
| Name: | Park parkjm@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Evaluate far off-axis NBCD physics; This is resubmission of 2012 proposal 92. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use circular plasma and/or vertically shifted small plasmas with off-axis beams at maximum tilt angle. Drive off-axis NBCD with a peak location at rho > 0.7. Measure off-axis NBCD, beam ion density/energy distribution, and fast ion loss. | ||
| Background: | Driving current far off-axis ( rho > 0.7) is crucial in testing a potential of high bootstrap fraction, steady-state operation with a broad current profile for the future tokamak reactor beyond ITER. DIII-D experiment and modeling show that off-axis NBCD efficiency is as good as on-axis NBCD because the increased fraction of trapped electrons reduces the electron shielding of the injected ion current IN CONTRAST WITH electron current drive schemes where the trapping of electrons degrades the efficiency. 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 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 microturbulence on fast ion confinement since the background turbulent fluctuation is in general larger when moving to outer radius region. | ||
| Resource Requirements: | All neutral beam sources with 150 beams at maximum tilt angle. All available gyrotrons to make variation of discharge conditions. | ||
| Diagnostic Requirements: | MSE, Neutrons, FIDA spectrometers & cameras, Core spectrometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 334: Measurement of neoclassical response of off-axis beam ions | ||
| Name: | Park parkjm@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the local profile of electron shielding of beam ion current for comparison with theory. This is resubmission of 2012 proposal 105. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Primarily compare two discharges with different Zeff otherwise in similar condition. Reproduce 144265 (off-axis NBCD measurement shot with full power of off-axis beams at max tilt angle) Inject Ne to increase Zeff. Apply ECH for high Zeff case to match beam ion slowing down time. Repeat using on-axis beams
(2) Modulate impurity injection (Ne, 10 Hz-10 msec injection). Find the electron shielding profile directly form the periodic response of the MSE signals to the modulation of Zeff created by pulsed impurity injections. We should be able to separate electron response even in the presence of change of beam ion slowing down time due to Zeff modulation (much longer time scale than electron response). (3) Repeat with ECH to increase/decrease Te to test collisionality dependancy of electron shielding | ||
| Background: | Measurement of the neoclassical electron response to beam ions is an important next step to complete a validation of the classical model against DIII-D experiments. Slowing down of fast ions by collisions with electrons causes toroidal drift of electrons, which cancels part of the fast ion current (electron shielding). The off-axis NBCD efficiency is as good as on-axis NBCD mainly because the increased fraction of trapped electrons reduces the electron shielding in outer radius region. Nonetheless, the precise profile of electron shielding has never been measured and benchmarked with the models, in particular, the analytic models implemented in NUBEAM. The electron shielding depends mainly on Zeff; JNBCD = Jf [1-(Zb/Zeff)(1-G)], where JNBCD = net NBCD current, Jf = fast ion current, Zb = charge number of beam ion, G = trapped electron correction that depends on Zeff and trapped electron fraction. In principle, the electron shielding profile can be determined by comparing two discharges with different Zeff or by the periodic response of the MSE signals to the modulation of Zeff. | ||
| Resource Requirements: | All neutral beam sources with 150 beams at maximum tilt angle. All available gyrotrons to make variation of discharge conditions. | ||
| Diagnostic Requirements: | MSE, Neutrons, FIDA spectrometers & cameras, Core spectrometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 335: Off-axis NBCD measurement in steady state scenario | ||
| Name: | Park parkjm@fusion.gat.com | Affiliation: | ORNL |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Measure off-axis NBCD profiles in high beta steady-state scenario. | ITPA Joint Experiment : | No |
| Description: | Measure local NBCD profiles at betaN ~ 2.8 with qmin ~ 1.0 and ~ 2.0 by comparing two discharges with and without OANB. Focus will be placed on obtaining stationary discharges at high beta as long as possible for accurate differential current drive analysis.<br><br>(1) low qmin without OANB: Restore 147379 with the following changes: i) 30L & 30R in timcon, 330L & 330R in PCS betaN feedback, ii) take out 150LR completely, iii) apply 210RT at 70 kV, 10 on, 90 off starting after H-mode achieved<br>(2) low qmin with OANB: Reproduce (1) with the following changes: i) put 150L & 150R in PCS control and prioritize over 330's.<br>(3) high qmin (~2) without OANB: Restore 147394 with the same changes in (1)<br>(4) high qmin (~2) with OANB: Reproduce (3) with the following changes: i) put 150L & 150R in PCS control and prioritize over 330's.<br><br>This 4-point scan will also provide data set to investigate transport characteristics of high qmin discharges in great detail. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 336: Routine current profile control in operation | ||
| Name: | Walker walker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Eugenio Schuster, Didier Moreau | ITPA Joint Experiment : | No |
| Description: | Reduce to practice the previously demonstrated full current profile methods and apply to relevant ongoing physics experiments. Explore and expand the range of plasma scenarios able to use current profile control. Develop the tools and procedures needed to provide current profile control as a routine part of the capabilities supported by the DIII-D PCS. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The work would start by focusing on a single physics experiment that would benefit from being able to control the full current profile either in this experimental campaign or in subsequent campaigns. Given a target plasma scenario, model and controller development would focus on that scenario, first using piggybacks and control simulations and then Thursday evening experiments. Control of the nominal scenario would be first demonstrated in Thursday evening experiments, then using one or two shots during the target physics experiment(s) to explore its range of controllability. If the session leader desires, the profile control could be turned on in other shots during an experimental day. | ||
| Background: | Experimental demonstrations of current profile control in 2011-12 showed that controllers can be designed to provide good control of the full plasma current profile to a pre-specified target. These experiments also showed that some of these controllers provide control that is rather robust to variations in plasma parameters and control targets. The current profile work appears to have matured sufficiently that the process of reducing this control to routine experimental use should now be undertaken. | ||
| Resource Requirements: | Obtain commitment from a target physics experiment allowing use of a small number of shots, assuming capability to control the target profile of that experiment has been demonstrated in Thursday evening sessions.
Occasional piggybacks on similar experiments to collect data for model validation and development of simulations. Two thursday evening sessions. An occasional shot or portion of a shot during the target physics experiment days. All neutral beams and all gyrotrons active. | ||
| Diagnostic Requirements: | In realtime: MSE, Thomson, line average density, standard magnetics (processed through realtime MSE efit), beam powers, ECH powers, CER.
For offline analysis: ECE, density profile, toroidal rotation profile, ion and electron temperature profiles | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | |||
| Title | 337: Routine current profile control in operation | ||
| Name: | Walker walker@fusion.gat.com | Affiliation: | GA |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | Eugenio Schuster, Didier Moreau | ITPA Joint Experiment : | No |
| Description: | Reduce to practice the previously demonstrated full current profile methods and apply to relevant ongoing physics experiments. Explore and expand the range of plasma scenarios able to use current profile control. Develop the tools and procedures needed to provide current profile control as a routine part of the capabilities supported by the DIII-D PCS. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The work would start by focusing on a single physics experiment that would benefit from being able to control the full current profile either in this experimental campaign or in subsequent campaigns. Given a target plasma scenario, model and controller development would focus on that scenario, first using piggybacks and control simulations and then Thursday evening experiments. Control of the nominal scenario would be first demonstrated in Thursday evening experiments, then using one or two shots during the target physics experiment(s) to explore its range of controllability. If the session leader desires, the profile control could be turned on in other shots during an experimental day. | ||
| Background: | Experimental demonstrations of current profile control in 2011-12 showed that controllers can be designed to provide good control of the full plasma current profile to a pre-specified target. These experiments also showed that some of these controllers provide control that is rather robust to variations in plasma parameters and control targets. The current profile work appears to have matured sufficiently that the process of reducing this control to routine experimental use should now be undertaken. | ||
| Resource Requirements: | Obtain commitment from a target physics experiment allowing use of a small number of shots, assuming capability to control the target profile of that experiment has been demonstrated in Thursday evening sessions.
Occasional piggybacks on similar experiments to collect data for model validation and development of simulations. Two thursday evening sessions. An occasional shot or portion of a shot during the target physics experiment days. All neutral beams and all gyrotrons active. | ||
| Diagnostic Requirements: | In realtime: MSE, Thomson, line average density, standard magnetics (processed through realtime MSE efit), beam powers, ECH powers, CER.
For offline analysis: ECE, density profile, toroidal rotation profile, ion and electron temperature profiles | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | |||
| Title | 338: Local correction of the TBM error field | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye, M. Lanctot, E. Strait, and the TBM group | ITPA Joint Experiment : | No |
| Description: | A brief experiment is proposed to study the efficacy of local correction of the TBM error field. Local correction may prove superior to prior schemes due to the ability of the local coilset to correct several different toroidal harmonics (n) of the error field simultaneously. Thus, this experiment will shed light on the role of higher n fields on important parameters such as rotation degradation and low-density operation limits.
Local correction is achieved by energizing the two C-coil elements adjacent to the TBM (C319 & C259). It is proposed to measure the standard metrics of error field control in both low-beta (Ohmic) and high-beta plasmas as these fields are applied. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | In low beta (Ohmic) plasmas, the current in the local coil pair will be ramped in each polarity until a locked mode is induced. An offset in the required current will likely be found, and this offset will define the optimum. A low-density rampdown discharge will then be performed at the optimum level and compared to the prior low-density limit results from 2009. The reference shot is 149499.
In the high beta target plasma, the local coil currents will again be ramped at each polarity, beginning from the offset identified in the low-beta case. Rotation, angular momentum, plasma response, and the plasma-wall Maxwell stress will be measured throughout the ramps. It is again expected that these metrics will be optimized at a non-zero value of local coil current. The degree of metric optimization can then be compared to rotation degradation results from 2011, with the reference shot being 147131. Throughout the experiment, n=1 error field control will be maintained via the I-coils. It will likely be necessary to de-rate the n=1 I-coil currents in order to compensate for the n=1 pollution by the local correction coils, thus isolating the experimental effects to higher n as much as possible. This experiment could be accomplished in 3 (Ohmic) + 3 (H-mode) good shots. | ||
| Background: | TBM experiments in 2011 have found optimum levels of I-coil currents needed for correction of the n=1 TBM error field. However, applying this optimal correction only recovered 25% of the rotation degradation introduced by the TBM. This modest result clearly demonstrates that there is room for improvement in the development of the best 3D field for correction of the TBM field. One such improvement may be the use of a local field (in addition to the optimal n=1) correction, as discussed in this proposal. | ||
| Resource Requirements: | TBM installed on DIII-D
All 4 SPAs functional | ||
| Diagnostic Requirements: | Ohmic: Thomson, 3D magnetics, CO2 and 288 GHz interferometers
H-mode: Thomson, ECE, CER, MSE, 3D magnetics | ||
| Analysis Requirements: | L-mode: Standard locked mode analysis tools already in existence.
H-mode: TRANSP runs for all discharges to keep track of total angular momentum. | ||
| Other Requirements: | |||
| Title | 339: Combined effect of TBM and NTM fields on fast-ion confinement | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Matt Lanctot, EP group | ITPA Joint Experiment : | Yes |
| Description: | Measure the fast-ion transport in a beam-heated plasma with a large-amplitude NTM with & without TBM fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make a plasma with a reasonably steady, large amplitude (2,1) NTM. On a repeat shot, pulse on the TBM coils after the mode is established. Use ECE to measure the mode structure and all fast-ion diagnostics to infer the fast-ion transport. | ||
| Background: | A concern for ITER is that core MHD will transport alphas to the edge where TBM fields will cause concentrated losses at the wall. This proposal is one of three to address these possible synergistic effects. | ||
| Resource Requirements: | Beams; TBM mock-up coil | ||
| Diagnostic Requirements: | Fast-ion loss diagnostics (especially IRTV on TBM coils) | ||
| Analysis Requirements: | NUBEAM, FIDASIM.
Best case analyzed with SPIRAL, OFMC, and ASCOT by EP experts. | ||
| Other Requirements: | |||
| Title | 340: Combined effect of TBM and Sawtooth fields on fast-ion confinement | ||
| Name: | Heidbrink heidbrink@fusion.gat.com | Affiliation: | UC, Irvine |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Matt Lanctot, EP group | ITPA Joint Experiment : | Yes |
| Description: | Measure the fast-ion transport in a beam-heated plasma with large sawtooth crashes with & without TBM fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make a plasma with large sawtooth crashes (both in amplitude and inversion radius). This will probably be at low q95 and also will use beam combinations and ECCD that promote sawtooth stability. On a repeat shot, pulse on the TBM coils during the large sawtooth phase. Use ECE to measure the mode structure and all fast-ion diagnostics to infer the fast-ion transport. | ||
| Background: | A concern for ITER is that core MHD will transport alphas to the edge where TBM fields will cause concentrated losses at the wall. This proposal is one of three to address these possible synergistic effects. | ||
| Resource Requirements: | Beams, ECCD, TBM coil | ||
| Diagnostic Requirements: | All fast-ion diagnostics, especially IR camera viewing the TBM tiles. | ||
| Analysis Requirements: | Modelling with SPIRAL, ASCOT, and OFMC by EP experts | ||
| Other Requirements: | |||
| Title | 341: Investigation of Momentum Transport with Power ratio PECH/PNBI and torque ratio Ï?NBI/PECH scan | ||
| Name: | Ida none | Affiliation: | National Institute for Fusion Science, Toki, Japan |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | J.M.Kwon, W.H.Ko, S.H.Hahn, Y.J.Shi, P.H.Diamond, B. Grierson, W. Solomon | ITPA Joint Experiment : | Yes |
| Description: | This is a revised re-submission of ROF #276 and #282 by the same primary author (K. Ida):
This experiment will i. use PECH scan at fixed NBI torque to elucidate the effect of Te/Ti and density profile structure on intrinsic torque, and thereby also obtain insight into global rotation profile structure. ii. use a scan of Ï?NBI /PECH at fixed PNBI to undertake fluctuation studies which elucidate the relation between ITG-TEM evolution and intrinsic torque. This study will also unify the body of studies of ECH effects on rotation. iii. study the poloidal rotation anomaly during the scans, especially for higher PECH. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This proposal includes two parts: the power ratio scan (PECH/PNBI) with balanced NBI and the torque scan (Ï?NBI/PECH) with fixed NBI power. For the first part, we assume a H-mode plasma with minimal external torque input. ECH is injected at the flat top of the discharges (i.e. quasi-steady state). In order to achieve the power ratio scan in a single discharge, the six gyrotrons should be turned on and off step by step like Figure1 with NBI modulation. We repeat this discharge with different P_NBI=2,4,6 MW. (three successful shots are required)
For the second part, i.e. the torque scan (Ï?NBI/PECH), we fix PECH and turn on and off the all 5 NBIs step by step as shown in Figure 2. ECH power is modulated as shown in the figure. So, we put 3 co-NBIs first and put 2 counter-NBIs later to achieve the torque scan in a single discharge. We repeat this scan with different values of PECH =2,3,4 MW (three successful shots are required). | ||
| Background: | For future tokamaks like ITER with much lower external torque input and higher Te/Ti ratio than present day machines, the role of intrinsic torque will be highly prominent in determining the global profile of plasma rotations. We propose to study the effects of these parameters (PECH/PNBI, Ï?NBI/PECH) on intrinsic rotation with systematic fluctuation studies. By performing the P_ECH scan, we want to cover a range of Te/Ti and see its effects on intrinsic rotation. It is necessary to identify ITG and TEM population change during the scan. This will reveal how this change is related to the changes of intrinsic torque. The power scan (P_ECH) can naturally cover the scan of the ratio of core and pedestal intrinsic torque, which will provide good physical insight how these intrinsic torques contribute to the global rotation profile formation. There have been several previous studies covering some low and high values of these scans in other devices like AUG and KSTAR. But weâ??d like to emphasize that there has been no attempt to systematically scan these values in a single device. DIII-D has a unique capability to perform these scans covering wide ranges of these values with careful control of other parameters. | ||
| Resource Requirements: | 6 Gyrotrons, 3 co-NBIs and 2 counter-NBIs
half run day might be desired. | ||
| Diagnostic Requirements: | Standard profile diagnostics (ne, Ti, Te, rotation, Er profiles)
Nice to have main ion CER Density and temperature fluctuation diagnostics (BES and ECEI) High speed CER for toroidal and poloidal rotation Thomson scattering and microwave reflectometry for density profile, DBS, PCI | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 342: Investigation of Core Poloidal Rotation Anomaly Change at ITG -> TEM transition | ||
| Name: | Ko none | Affiliation: | National Fusion Research Institute (NFRI) |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | B. Grierson, S.H. Ko, J.M.Kwon, Y.J.Shi, P.H.Diamond, S.H.Hahn, W. Solomon | ITPA Joint Experiment : | Yes |
| Description: | This experiment will study the change of the core poloidal rotation anomaly at ITG ->TEM transition and the role of turbulence driven poloidal Reynolds stress in the anomaly change. With strong ECH heating on H-mode plasmas with nulled toroidal rotation, it is expected that turbulence population changes from ITG to TEM. Measurement of poloidal rotation will reveal the departure of the rotation value from the neoclassical prediction and how the departure changes according to the turbulence population change. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | This proposal assumes H-mode plasma with balanced NBI as a base state to minimize external torque effects on plasma rotation. Applying strong ECH heating on the plasma core, we trigger ITG -> TEM transition and measure the poloidal rotation change with detailed fluctuation measurement. With measured profiles of T_i,T_e,n_e, we can calculate the neoclassical poloidal rotation values and compare these with measurements by poloidal CES. Detailed fluctuation measurements can provide poloidal Reynolds stress changes during the ITG â?? TEM transition. Then, we can compare the poloidal rotation anomaly and its change with measured poloidal Reynolds stress change. To study the effects of fluctuation level and the mode population, we propose to scan P_ECH with particular emphasis on the highest values of P_ECH. | ||
| Background: | Recent DIII-D experiments found the departure of poloidal rotation from the neoclassical prediction [B. Grierson et al, Phys. Plasmas 19, 056107(2012)]. One of the possible candidates causing the departure is turbulence driven poloidal Reynolds stress. The equation for poloidal rotation evolution can be written as
â??/â??tâ?©V_θâ?ª=-e Ì?_θâ??â??Ï? â?¡_neo-â??_r â?©v Ì?_r v Ì?_θâ?ª, where the first and the second term in the right hand side correspond to the contributions from neoclassical viscosity and turbulence driven Reynolds stress, respectively. In the absence of the Reynolds stress, the steady state poloidal rotation yields the neoclassical value V_θ^neo=Kâ??T/â??r, where K depends on the ion collisionality. Quasi-linear theory of turbulence predicts the dependence of the poloidal Reynolds stress on the mode propagation direction, i.e. sign flip upon ITG â?? TEM transition. The Reynolds stress is proportional to the group velocity of fluctuation V_gr, â?©v Ì?_r v Ì?_θ â?ªâ??V_gr â?©V_E^' â?ª â??/â??k_r â?©N_k â?ª, and V_gr flips its sign during ITG â?? TEM transition. Also, a recent global gyrokinetic simulation study demonstrated that the reversal of the global structure of poloidal EÃ?B flow occurs during ITG â?? TEM transition (Figure 1). So, we can expect similar changes of poloidal rotation if the component of fluctuation driven EÃ?B flow is strong enough. In this proposal, we investigate the dependence of the poloidal rotation anomaly on the fluctuation population change. The poloidal EÃ?B rotation plays very important role in the fluctuation symmetry breaking, which is necessary to generate toroidal intrinsic torque. Therefore, this study will provide very important insight on the physics of intrinsic torque and the formation of global toroidal rotation profile in future burning plasma experiments such as ITER. | ||
| Resource Requirements: | 6 Gyrotrons, balanced NBIs
half run day might be desired. | ||
| Diagnostic Requirements: | Standard profile diagnostics (ne, Ti, Te, rotation, Er profiles)
Nice to have main ion CER Density and temperature fluctuation diagnostics (BES and ECEI) High speed CER for toroidal and poloidal rotation Thomson scattering and microwave reflectometry for density profile, DBS, PCI | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 343: Evaluating Physics of RMP Stationary Enhanced Confinement Regime Without ELMs for the 2013 JRT | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, A. Hubbard, D. Whyte, S. Gerhardt, R. Maingi | ITPA Joint Experiment : | No |
| Description: | The goal of these experiments is to address the requirements of the 2013 Joint Facilities Research Target (JRT) milestone and DIII-D milestone 183 on stationary enhanced confinement regimes without ELMS. The JRT work involves both 1) understanding the physics mechanisms allowing control of edge particle transport while maintaining a robust edge thermal transport barrier and 2) assessing and understanding the operational space for several ELM control regimes. The ultimate goal is to strengthen the physics basis for extrapolation of ELM control regimes to ITER. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | One focus of the 2013 JRT is on understanding the physics mechanisms that can regulate edge particle transport while allowing a robust edge thermal transport barrier in stationary plasmas without ELMs. The JRT charge asserts that plasma operation with continuous edge plasma modes or externally applied 3D fields can be candidates for â??plasma laboratoriesâ?? to investigate this physics. For these experiments at least 3 different research lines could be attempted, vis: 1) in QH-mode plasmas the experiments would be targeted to understand the effect of the EHO on regulation of the edge particle transport in such a way to produce the wide, high pressure pedestals characteristic of high performance QH-mode operation, 2) in H-mode plasmas with applied RMPs the experiments would attempt to understand the physics mechanism responsible for â??density pumpoutâ?? observed after RMP application, its dependence on RMP spectrum, perturbation amplitude and q95, and whether it is tied to the appearance of mitigated high frequency ELMs with RMP applied, and 3) in I-mode plasmas DIII-D diagnostics and torque variation capability would be applied to understand the effect of a high frequency weakly coherent mode on the pedestal particle transport. Various aspects of this part of the JRT work have been proposed in other ROF submissions, vis.: for (1) above see proposals 13 and 14 by Burrell and proposal 41 by Diallo proposed in the Inductive Scenarios WG and ROF # 93 by Lanctot in the Plasma Control WG. In the RMP area (2) above see proposals 25 (Boedo), 45 (Petrie), 84 (Jakubowski), 145 (Moyer), 184 (Evans et al.), 224 (Loarte), and 305 (Callen) proposed in the ELM Control WG. For work on mechanisms for particle transport control in I-mode see proposal 37 (White). Other proposals targeting physics that could contribute to these goals of the JRT include 104 (Sontag) and 153 (Canik).
Another focus of the 2013 JRT work is to assess and understand the operating space of various ELM control regimes. Many proposals have been submitted toward this goal, vis. for RMP see proposals 3, 19, 23 (DeGrassie), 24 (Battaglia), 55 (Lazarus), 77 Mordijck), 80 (King), 101 (J=K Park), 121 Shafer), 143, 144, 146 (Moyer), 161 (Nazikian), 185 (Chen), 189, 190 (Paz-Soltan), 220 (Evans), 226 (Loarte), 256 (McKee), and 302 (Petty). For QH-mode see proposals 9,10,15,16,53,115,120 (Burrell et al.), 127 (Yu), 234 (Nave), 243, 246, 247 (Garofalo), 253 (Chen), 303 (Jackson), 313 (Petty) 321, 323 Solomon, and 330 (Smith) submitted to Inductive Scenarios, and for I-mode see proposals 83 (Whyte et al.), and 111 (Hubbard et al.) submitted to the ELM Control WG. | ||
| Background: | The text of the 2013 JRT is: Conduct experiments and analysis on major fusion facilities, to evaluate stationary enhanced confinement regimes without large Edge Localized Modes (ELMs), and to improve understanding of the underlying physical mechanisms that allow acceptable edge particle transport while maintaining a strong thermal transport barrier. Mechanisms to be investigated can include intrinsic continuous edge plasma modes and externally applied 3D fields. Candidate regimes and techniques have been pioneered by each of the three major US facilities (C-Mod, D3D and NSTX). Coordinated experiments, measurements, and analysis will be carried out to assess and understand the operational space for the regimes. Exploiting the complementary parameters and tools of the devices, joint teams will aim to more closely approach key dimensionless parameters of ITER, and to identify correlations between edge fluctuations and transport. The role of rotation will be investigated. The research will strengthen the basis for extrapolation of stationary regimes which combine high energy confinement with good particle and impurity control, to ITER and other future fusion facilities for which avoidance of large ELMs is a critical issue.
The title of 2013 DIII-D Milestone 183 is: Assess the physics mechanisms allowing stationary enhanced performance H-mode plasmas without large ELMs. Supports FY2013 FES Joint Research Target. | ||
| Resource Requirements: | Minimum run time needed to make new contributions to both goals of the 2013 JRT: detailed physics of edge particle transport and understanding of the operating space in QH-mode, RMP ELM control and I-mode, is two days for each regime, especially to respond to the JRT goal of assessing and understanding the operating space for the regimes. Joint discussions between the ELM Control and Inductive Scenarios groups should focus on determining the best plasma laboratory in which to study the regulation of edge particle transport in a stationary regime without ELMs. All of these experiments will require moderated NB power in the 5 â?? 10 MW range, with some torque control capability (co-vs counter beams) and good LSN shape control. | ||
| Diagnostic Requirements: | All pedestal, SOL and divertor diagnostics especially fast measurements including all fluctuation diagnostics capable of pedestal measurements (BES, ECEI, CECE, DBS, profile reflectometer etc.) | ||
| Analysis Requirements: | Kinetic profile analysis and kinetic EFITS will be needed in preparation for ELITE pedestal stability analysis. Edge transport analysis will be needed (eg. TRANSP, ONETWO or other) to determine the effect of parameter variations on edge particle transport. | ||
| Other Requirements: | |||
| Title | 344: Evaluating Physics of QH-mode Stationary Enhanced Confinement Regime Without ELMs for the 2013 J | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, A. Hubbard, D. Whyte, S. Gerhardt, R. Maingi | ITPA Joint Experiment : | No |
| Description: | The goal of these experiments is to address the requirements of the 2013 Joint Facilities Research Target (JRT) milestone and DIII-D milestone 183 on stationary enhanced confinement regimes without ELMS. The JRT work involves both 1) understanding the physics mechanisms allowing control of edge particle transport while maintaining a robust edge thermal transport barrier and 2) assessing and understanding the operational space for several ELM control regimes. The ultimate goal is to strengthen the physics basis for extrapolation of ELM control regimes to ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | One focus of the 2013 JRT is on understanding the physics mechanisms that can regulate edge particle transport while allowing a robust edge thermal transport barrier in stationary plasmas without ELMs. The JRT charge asserts that plasma operation with continuous edge plasma modes or externally applied 3D fields can be candidates for â??plasma laboratoriesâ?? to investigate this physics. For these experiments at least 3 different research lines could be attempted, vis: 1) in QH-mode plasmas the experiments would be targeted to understand the effect of the EHO on regulation of the edge particle transport in such a way to produce the wide, high pressure pedestals characteristic of high performance QH-mode operation, 2) in H-mode plasmas with applied RMPs the experiments would attempt to understand the physics mechanism responsible for â??density pumpoutâ?? observed after RMP application, its dependence on RMP spectrum, perturbation amplitude and q95, and whether it is tied to the appearance of mitigated high frequency ELMs with RMP applied, and 3) in I-mode plasmas DIII-D diagnostics and torque variation capability would be applied to understand the effect of a high frequency weakly coherent mode on the pedestal particle transport. Various aspects of this part of the JRT work have been proposed in other ROF submissions, vis.: for (1) above see proposals 13 and 14 by Burrell and proposal 41 by Diallo proposed in the Inductive Scenarios WG and ROF # 93 by Lanctot in the Plasma Control WG. In the RMP area (2) above see proposals 25 (Boedo), 45 (Petrie), 84 (Jakubowski), 145 (Moyer), 184 (Evans et al.), 224 (Loarte), and 305 (Callen) proposed in the ELM Control WG. For work on mechanisms for particle transport control in I-mode see proposal 37 (White). Other proposals targeting physics that could contribute to these goals of the JRT include 104 (Sontag) and 153 (Canik).
Another focus of the 2013 JRT work is to assess and understand the operating space of various ELM control regimes. Many proposals have been submitted toward this goal, vis. for RMP see proposals 3, 19, 23 (DeGrassie), 24 (Battaglia), 55 (Lazarus), 77 Mordijck), 80 (King), 101 (J=K Park), 121 Shafer), 143, 144, 146 (Moyer), 161 (Nazikian), 185 (Chen), 189, 190 (Paz-Soltan), 220 (Evans), 226 (Loarte), 256 (McKee), and 302 (Petty). For QH-mode see proposals 9,10,15,16,53,115,120 (Burrell et al.), 127 (Yu), 234 (Nave), 243, 246, 247 (Garofalo), 253 (Chen), 303 (Jackson), 313 (Petty) 321, 323 Solomon, and 330 (Smith) submitted to Inductive Scenarios, and for I-mode see proposals 83 (Whyte et al.), and 111 (Hubbard et al.) submitted to the ELM Control WG. | ||
| Background: | The text of the 2013 JRT is: Conduct experiments and analysis on major fusion facilities, to evaluate stationary enhanced confinement regimes without large Edge Localized Modes (ELMs), and to improve understanding of the underlying physical mechanisms that allow acceptable edge particle transport while maintaining a strong thermal transport barrier. Mechanisms to be investigated can include intrinsic continuous edge plasma modes and externally applied 3D fields. Candidate regimes and techniques have been pioneered by each of the three major US facilities (C-Mod, D3D and NSTX). Coordinated experiments, measurements, and analysis will be carried out to assess and understand the operational space for the regimes. Exploiting the complementary parameters and tools of the devices, joint teams will aim to more closely approach key dimensionless parameters of ITER, and to identify correlations between edge fluctuations and transport. The role of rotation will be investigated. The research will strengthen the basis for extrapolation of stationary regimes which combine high energy confinement with good particle and impurity control, to ITER and other future fusion facilities for which avoidance of large ELMs is a critical issue.
The title of 2013 DIII-D Milestone 183 is: Assess the physics mechanisms allowing stationary enhanced performance H-mode plasmas without large ELMs. Supports FY2013 FES Joint Research Target. | ||
| Resource Requirements: | Minimum run time needed to make new contributions to both goals of the 2013 JRT: detailed physics of edge particle transport and understanding of the operating space in QH-mode, RMP ELM control and I-mode, is two days for each regime, especially to respond to the JRT goal of assessing and understanding the operating space for the regimes. Joint discussions between the ELM Control and Inductive Scenarios groups should focus on determining the best plasma laboratory in which to study the regulation of edge particle transport in a stationary regime without ELMs. All of these experiments will require moderated NB power in the 5 â?? 10 MW range, with some torque control capability (co-vs counter beams) and good LSN shape control. | ||
| Diagnostic Requirements: | All pedestal, SOL and divertor diagnostics especially fast measurements including all fluctuation diagnostics capable of pedestal measurements (BES, ECEI, CECE, DBS, profile reflectometer etc.) | ||
| Analysis Requirements: | Kinetic profile analysis and kinetic EFITS will be needed in preparation for ELITE pedestal stability analysis. Edge transport analysis will be needed (eg. TRANSP, ONETWO or other) to determine the effect of parameter variations on edge particle transport. | ||
| Other Requirements: | |||
| Title | 345: Evaluating Physics of I-mode Stationary Enhanced Confinement Regime Without ELMs for the 2013 JR | ||
| Name: | Fenstermacher fenstermacher@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, A. Hubbard, D. Whyte, S. Gerhardt, R. Maingi | ITPA Joint Experiment : | No |
| Description: | The goal of these experiments is to address the requirements of the 2013 Joint Facilities Research Target (JRT) milestone and DIII-D milestone 183 on stationary enhanced confinement regimes without ELMS. The JRT work involves both 1) understanding the physics mechanisms allowing control of edge particle transport while maintaining a robust edge thermal transport barrier and 2) assessing and understanding the operational space for several ELM control regimes. The ultimate goal is to strengthen the physics basis for extrapolation of ELM control regimes to ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | One focus of the 2013 JRT is on understanding the physics mechanisms that can regulate edge particle transport while allowing a robust edge thermal transport barrier in stationary plasmas without ELMs. The JRT charge asserts that plasma operation with continuous edge plasma modes or externally applied 3D fields can be candidates for â??plasma laboratoriesâ?? to investigate this physics. For these experiments at least 3 different research lines could be attempted, vis: 1) in QH-mode plasmas the experiments would be targeted to understand the effect of the EHO on regulation of the edge particle transport in such a way to produce the wide, high pressure pedestals characteristic of high performance QH-mode operation, 2) in H-mode plasmas with applied RMPs the experiments would attempt to understand the physics mechanism responsible for â??density pumpoutâ?? observed after RMP application, its dependence on RMP spectrum, perturbation amplitude and q95, and whether it is tied to the appearance of mitigated high frequency ELMs with RMP applied, and 3) in I-mode plasmas DIII-D diagnostics and torque variation capability would be applied to understand the effect of a high frequency weakly coherent mode on the pedestal particle transport. Various aspects of this part of the JRT work have been proposed in other ROF submissions, vis.: for (1) above see proposals 13 and 14 by Burrell and proposal 41 by Diallo proposed in the Inductive Scenarios WG and ROF # 93 by Lanctot in the Plasma Control WG. In the RMP area (2) above see proposals 25 (Boedo), 45 (Petrie), 84 (Jakubowski), 145 (Moyer), 184 (Evans et al.), 224 (Loarte), and 305 (Callen) proposed in the ELM Control WG. For work on mechanisms for particle transport control in I-mode see proposal 37 (White). Other proposals targeting physics that could contribute to these goals of the JRT include 104 (Sontag) and 153 (Canik).
Another focus of the 2013 JRT work is to assess and understand the operating space of various ELM control regimes. Many proposals have been submitted toward this goal, vis. for RMP see proposals 3, 19, 23 (DeGrassie), 24 (Battaglia), 55 (Lazarus), 77 Mordijck), 80 (King), 101 (J=K Park), 121 Shafer), 143, 144, 146 (Moyer), 161 (Nazikian), 185 (Chen), 189, 190 (Paz-Soltan), 220 (Evans), 226 (Loarte), 256 (McKee), and 302 (Petty). For QH-mode see proposals 9,10,15,16,53,115,120 (Burrell et al.), 127 (Yu), 234 (Nave), 243, 246, 247 (Garofalo), 253 (Chen), 303 (Jackson), 313 (Petty) 321, 323 Solomon, and 330 (Smith) submitted to Inductive Scenarios, and for I-mode see proposals 83 (Whyte et al.), and 111 (Hubbard et al.) submitted to the ELM Control WG. | ||
| Background: | The text of the 2013 JRT is: Conduct experiments and analysis on major fusion facilities, to evaluate stationary enhanced confinement regimes without large Edge Localized Modes (ELMs), and to improve understanding of the underlying physical mechanisms that allow acceptable edge particle transport while maintaining a strong thermal transport barrier. Mechanisms to be investigated can include intrinsic continuous edge plasma modes and externally applied 3D fields. Candidate regimes and techniques have been pioneered by each of the three major US facilities (C-Mod, D3D and NSTX). Coordinated experiments, measurements, and analysis will be carried out to assess and understand the operational space for the regimes. Exploiting the complementary parameters and tools of the devices, joint teams will aim to more closely approach key dimensionless parameters of ITER, and to identify correlations between edge fluctuations and transport. The role of rotation will be investigated. The research will strengthen the basis for extrapolation of stationary regimes which combine high energy confinement with good particle and impurity control, to ITER and other future fusion facilities for which avoidance of large ELMs is a critical issue.
The title of 2013 DIII-D Milestone 183 is: Assess the physics mechanisms allowing stationary enhanced performance H-mode plasmas without large ELMs. Supports FY2013 FES Joint Research Target. | ||
| Resource Requirements: | Minimum run time needed to make new contributions to both goals of the 2013 JRT: detailed physics of edge particle transport and understanding of the operating space in QH-mode, RMP ELM control and I-mode, is two days for each regime, especially to respond to the JRT goal of assessing and understanding the operating space for the regimes. Joint discussions between the ELM Control and Inductive Scenarios groups should focus on determining the best plasma laboratory in which to study the regulation of edge particle transport in a stationary regime without ELMs. All of these experiments will require moderated NB power in the 5 â?? 10 MW range, with some torque control capability (co-vs counter beams) and good LSN shape control. | ||
| Diagnostic Requirements: | All pedestal, SOL and divertor diagnostics especially fast measurements including all fluctuation diagnostics capable of pedestal measurements (BES, ECEI, CECE, DBS, profile reflectometer etc.) | ||
| Analysis Requirements: | Kinetic profile analysis and kinetic EFITS will be needed in preparation for ELITE pedestal stability analysis. Edge transport analysis will be needed (eg. TRANSP, ONETWO or other) to determine the effect of parameter variations on edge particle transport | ||
| Other Requirements: | |||
| Title | 346: Control of H-L back transition via applied RMP fields | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G. McKee, Z. Yan, G. Tynan, T.L Rhodes, L. Zeng, P.H. Diamond | ITPA Joint Experiment : | No |
| Description: | Energy release and inductivity changes during the H-L back transition are a concern for ITER, as they can place undue stress on the poloidal field system and lead to excessive divertor heat load excursions. Control of the transition timing/temporal evolution is therefore important. This experiment will explore control of the back transition dynamics via RMP fields. Addition of RMP accomplishes two important modifications: <br>1) The upper pedestal density gradient can be limited/mitigated, possibly allowing trigger of the H-L back transition via NBI power ramp-down without a preceding large ELM; 2) Addition of RMPs may allow us to exceed the transition power hysteresis and allow precise control of the H-L back-transition timing. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In H-mode plasmas in ITER-similar shape, a gradual NBI power ramp down will be performed in the presence of even parity n=3 I-coil perturbations, using (static) I-coil currents of 1-4 kA, with q95 in- and outside the ELM-suppression window.The experiment will then be repeated using odd parity. An additional experiment will be carried out with P_NBI adjusted to the power needed to maintain H-mode after the L-H transition, and adding a modulated I-coil pulse (at the fastest possible rate) to induce the back-transition. In this manner intermediate states similar to limit cycle oscillations may be accessed (LCOs have previously observed spontaneously during back transition in a low triangularity plasma) | ||
| Background: | H-L back transitions with extended limit cycles were previously observed in a low triangularity plasma. These transitions were induced via a step-down/shut-off in NBI power and typically were triggered via type-I ELMs. Control of the upper pedestal density/pressure profile via RMPs may allow us to trigger the back-transition independently of ELMs, and may also allow control of the detailed transition dynamics and energy release rate. | ||
| Resource Requirements: | All beams, pulsed/modulated I-coil (details to be determined) | ||
| Diagnostic Requirements: | BES, DBS, reciprocating probe, profile reflectometry, edge Thomson, | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 347: Do differences in co/cnter GAM/ZF behavior affect deficit in gyrokinetic predicted Lmode transport? | ||
| Name: | Rhodes rhodes@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | usual suspects - too many to list | ITPA Joint Experiment : | No |
| Description: | Differences have been observed in co/counter GAM and ZF behavior (J Hillesheim). Detailed measurements and comparisons to gyro-kinetic simulations would reveal if these differences affect the observed deficit in gyro-kinetic predicted transport towards edge of Lmode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | L-mode plasmas. One expt day. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 348: Can we obtain ELM suppression without substantial loss of density? | ||
| Name: | Rhodes rhodes@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | T Evans | ITPA Joint Experiment : | No |
| Description: | Explore higher density ELM suppressed Hmode regime similar to 147170. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Shot 147170 saw a secular increase in density (to near the pre RMP value) after ELM suppression. This shot was an n=3 RMP phase varied from 0 to 60. Starting from this condition explore physics and extension of this regime. | ||
| Background: | |||
| Resource Requirements: | Hmode, RMP, ELM suppressed. One expt day. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 349: Density/current Scaling of Intrinsic Rotation and ITG/TEM Transition | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, T.L. Rhodes, L. Zeng, J. deGrassie, K.H. Burrell, B. Grierson | ITPA Joint Experiment : | No |
| Description: | Toroidal rotation reversal and a transition from LOC to SOC confinement have been observed in C-Mod [1] and ascribed to a transition from a TEM-dominated to an ITG-dominated regime and concomitant reversal of the (fluctuation-driven) residual stress.<br>The goal of this experiment is to elucidate the current/density scaling of the residual stress via detailed measurement of the turbulence characteristics (in particular the poloidal wavenumber spectrum and radial correlation length) at the ITG-TEM transition, using purely ECH-heated L-mode and diverted H-mode plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Transitions from a mixed mode to a TEM dominated regime have been observed in ECH L-mode plasmas during previous experiments where the electron transport stiffness was investigated. The L-mode part of the proposed experiment would be based on a similar reference discharge possibly using IWL plasma to avoid an early H-mode transition. Depending on the ECH deposition location the electron temperature gradient can be sufficiently modified to access the TEM regime. A current/density scan will be performed to obtain Fluctuation/profile data across the ITG/TEM boundary will be obtained during a current/density scan in order to extract information on rotation reversal and residual stress. Transitioning to diverted shape, an H-mode transition later during the same series of shots will be used to check the dependence of rotation and fluctuation characteristics on the edge electron temperature and density gradient. DBS can measure the radial profile of (wavenumber-resolved) low-k and intermediate k density fluctuations, and CECE can measure the radial profile of electron temperature fluctuations. In addition PCI will be used to obtain chord-averaged fluctuation spectra.
The main ion toroidal velocity will be estimated from high spatial resolution measurements of the ExB velocity obtained via DBS. This requires knowledge of the main ion pressure gradient (obtained via profile reflectometry, using the CER-measured carbon ion temperature (beam blips). This method has been successfully employed previously in low-Z_eff plasmas. A caveat of this method is that main ion poloidal rotation would have to be either neglected or estimated from theoretical models. | ||
| Background: | Intriguing work on the current/density scaling of intrinsic rotation, and rotation reversals ascribed to ITG-TEM transitions has been performed in C-mod [1,2]. iIn DIII-D, clear transitions from ITG/mixed mode dominated to TEM-dominated regimes have been observed in L-mode and QH-mode plasmas recently with ECH, but the concomitant changes in momentum transport and intrinsic rotation drive have not been systematically investigated. With the present high resolution diagnostic capabilities DIII-D is in a unique position to address changes in the fluctuation dynamics and outer core/edge density, temperature and rotation profiles during ITG/TEM transitions. The H-mode part of the experiment would be best performed during the initial ELM-free phase to maximize diagnostic resolution, although low-frequency ELMs may be acceptable.
[1] J. Rice et al., PRL May 2011. [2] J. Rice at al., PRL Dec. 2011. This proposal has also been submitted to the ":Rotation" category | ||
| Resource Requirements: | All beams, 6 gyrotrons | ||
| Diagnostic Requirements: | DBS 5, DBS8, CECE, PCI | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 350: Current/Density Scaling of Intrinsic Rotation and ITG/TEM Transition | ||
| Name: | Schmitz schmitzl@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, T.L. Rhodes, L. Zeng, J. deGrassie, K.H. Burrell, B. Grierson | ITPA Joint Experiment : | No |
| Description: | Toroidal rotation reversal and a transition from LOC to SOC confinement have been observed in C-Mod [1] and ascribed to a transition from a TEM-dominated to an ITG-dominated regime and concomitant reversal of the (fluctuation-driven) residual stress. <br>The goal of this experiment is to elucidate the current/density scaling of the residual stress via detailed measurement of the turbulence characteristics (in particular the poloidal wavenumber spectrum and radial correlation length) at the ITG-TEM transition, using purely ECH-heated L-mode and diverted H-mode plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Transitions from a mixed mode to a TEM dominated regime have been observed in ECH L-mode plasmas during previous experiments where the electron transport stiffness was investigated. The L-mode part of the proposed experiment would be based on a similar reference discharge possibly using IWL plasma to avoid an early H-mode transition. Depending on the ECH deposition location the electron temperature gradient can be sufficiently modified to access the TEM regime. A current/density scan will be performed to obtain Fluctuation/profile data across the ITG/TEM boundary will be obtained during a current/density scan in order to extract information on rotation reversal and residual stress. Transitioning to diverted shape, an H-mode transition later during the same series of shots will be used to check the dependence of rotation and fluctuation characteristics on the edge electron temperature and density gradient. DBS can measure the radial profile of (wavenumber-resolved) low-k and intermediate k density fluctuations, and CECE can measure the radial profile of electron temperature fluctuations. In addition PCI will be used to obtain chord-averaged fluctuation spectra.
The main ion toroidal velocity will be estimated from high spatial resolution measurements of the ExB velocity obtained via DBS. This requires knowledge of the main ion pressure gradient (obtained via profile reflectometry, using the CER-measured carbon ion temperature (beam blips). This method has been successfully employed previously in low-Z_eff plasmas. A caveat of this method is that main ion poloidal rotation would have to be either neglected or estimated from theoretical models. | ||
| Background: | Intriguing work on the current/density scaling of intrinsic rotation, and rotation reversals ascribed to ITG-TEM transitions has been performed in C-mod [1,2]. iIn DIII-D, clear transitions from ITG/mixed mode dominated to TEM-dominated regimes have been observed in L-mode and QH-mode plasmas recently with ECH, but the concomitant changes in momentum transport and intrinsic rotation drive have not been systematically investigated. With the present high resolution diagnostic capabilities DIII-D is in a unique position to address changes in the fluctuation dynamics and outer core/edge density, temperature and rotation profiles during ITG/TEM transitions. The H-mode part of the experiment would be best performed during the initial ELM-free phase to maximize diagnostic resolution, although low-frequency ELMs may be acceptable.
[1] J. Rice et al., PRL May 2011. [2] J. Rice at al., PRL Dec. 2011. This proposal has also been submitted to the ":Turbulence/Transport" category | ||
| Resource Requirements: | All beams, 6 gyrotrons | ||
| Diagnostic Requirements: | DBS5,8,PCI,CECE | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 351: Confinement enhancement via impurity-seeding of H-mode plasmas | ||
| Name: | McKee mckee@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | C. Holland, L. Schmitz, S. Smith, T. Rhodes, 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: | All fluctuation and profile diagnostics | ||
| Diagnostic Requirements: | BES (8x8), UF-CHERS, DBS, CECE, FIR, PCI, etc. | ||
| Analysis Requirements: | lots | ||
| Other Requirements: | |||
| Title | 352: Does increased turbulence in H-mode degrade agreement of GK transport/turbulence simulations? | ||
| Name: | Rhodes rhodes@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Starting with previously obtained Hmode conditions (eg, from stiffness expt), change core profiles/turbulence via combination of off-axis NBI and/or ECH. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform power and heating location scans to modify drive and turbulence. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 353: Correlation length of electron temperature turbulence measurement and comparison with simulations | ||
| Name: | Wang wangg@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | Rhodes, Peebles, Holland | ITPA Joint Experiment : | No |
| Description: | Investigate radial correlation length of electron temperature turbulence using CECE system and compare with gyrokinetic simulations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Steady-state L-modes, density scan and heating power scan | ||
| 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. A dedicated experiment will allow for a detailed investigation. | ||
| Resource Requirements: | most gyrotrons and NB sources | ||
| Diagnostic Requirements: | CECE and all other turbulence and profile diagnostics | ||
| Analysis Requirements: | lots | ||
| Other Requirements: | |||
| Title | 354: Control of Major Disruptions in DIII-D | ||
| Name: | Sen amiya@ee.columbia.edu | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Robert Granitz, MIT; Rob Lahaye,DIII-D | ITPA Joint Experiment : | No |
| Description: | It is argued that major disruptions in ITER can be avoided by the feedback control of the causative MHD precursors. The sensors will be 2D-arrays of ECE detectors and the suppressors will be modulated ECH beams injected radially to produce non-thermal radial pressures to counter the radial dynamics of MHD modes. The appropriate amplitude and phase of this signal can stabilize the relevant MHD modes and prevent their evolution to a major disruption. For multimode MHD precursors, an optimal feedback scheme with a Kalman filter is discussed. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose a novel non-magnetic suppressor in the
form of a radially injected and modulated (at MHD frequency in laboratory frame) ECH beam. In this case, there will be no current drive, but the transverse energy of electrons will nearly instantaneously increase with a resonance response. This implies a prompt local increase in electron non-thermal pressure, which at an appropriate phase can push the MHD mode radially inward. For hardware issues, the ECH beam will be used in pulsed mode (on or off), which will suffice. | ||
| Background: | The background is available in:
"Feedback control of major disruptions in ITER" PHYSICS OF PLASMAS 18, 082502 (2011) | ||
| Resource Requirements: | Gating of ECH beams. | ||
| Diagnostic Requirements: | ECE imaging system based on shotky barrier diodes. | ||
| Analysis Requirements: | Image reconstruction and localization of the rough centroid of the relative MHD modes. | ||
| Other Requirements: | No | ||
| Title | 355: q dependence of L-H transition power threshold | ||
| Name: | Wang wangg@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | Peebles, Rhodes, Doyle, McKee, Yan | ITPA Joint Experiment : | No |
| Description: | To study the role of edge turbulence and zonal flows in the L-H transition via modifying zonal flow damping using q scan to observe L-H transition power threshold variation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | scan q but match other parameters (line-averaged density, Bt, plasma shape,temperature, rotation) to investigate edge q dependence of the L-H transition power threshold | ||
| Background: | Although the empirical scaling of the L-H transition power threshold evolves only global parameters, the L-H transition is an edge phenomenon believed to be related to edge turbulence and transport. The edge turbulence is a zonal flow-drift wave self-organized system. Zonal flows suffer Landau damping which has a strong q-dependence. | ||
| Resource Requirements: | ECH and NBI | ||
| Diagnostic Requirements: | all turbulence and profile diagnostics | ||
| Analysis Requirements: | lots | ||
| Other Requirements: | |||
| Title | 356: GAM eigenmode excitation mechanism | ||
| Name: | Wang wangg@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | Peebles, Rhodes, McKee, Yan | ITPA Joint Experiment : | No |
| Description: | To understand the Geodesic Acoustic eigenmode excitation mechanism by diagnosing the transition from a continuum GAM to an eigenmode and/or vice versa | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | L-mode plasma with heating power ramp | ||
| Background: | Both GAM eigenmodes and continuum have been oberved on DIII-D. Generally a continuum exists in Ohmic or low heating power L-mode plasmas, and eigenmodes exist in higher heating power L-modes. Characterization of the transition between these two is desired for a better understanding of the eigenmode excitation mechanism. | ||
| Resource Requirements: | ECH and NBI. | ||
| Diagnostic Requirements: | all turbulence and profile diagnostics. | ||
| Analysis Requirements: | lots | ||
| Other Requirements: | |||
| Title | 357: GAM eigenmode 2D structure | ||
| Name: | Wang wangg@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | Peebles, Rhodes, McKee, Yan | ITPA Joint Experiment : | No |
| Description: | To obtain radial and poloidal amplitude and propagation structure of GAMs in poloidal ExB flow, density, and electron temperature in L-mode plasmas, and compare with radial eigenmode theories | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | L-mode, reduced size plasma vertically scanned over 40 cm | ||
| Background: | GAM is an important player in edge turbulence and transport. GAM eigenmodes have been observed on DIII-D. A 2D (radial and poloidal) characterization is needed for a better understanding of GAM and its interaction with turbulence. | ||
| Resource Requirements: | ECH and NBI. | ||
| Diagnostic Requirements: | all turbulence and profile diagnostics. | ||
| Analysis Requirements: | lots | ||
| Other Requirements: | |||
| Title | 358: Investigation of Momentum transport with Modulated ECH and ECH resonance layer scan | ||
| Name: | Shi none | Affiliation: | National Fusion Research Institute |
| Research Area: | Plasma Rotation | Presentation time: | Requested |
| Co-Author(s): | W.H.Ko, K.Ida, J.M.Kwon, P.H.Diamond, S.H.Koh, S.H.Hahn , W. Solomon, G.McKee, J. DeGrassie | ITPA Joint Experiment : | Yes |
| Description: | The is revised version for ROF#274,275.
There two experiments in this proposal. One is radial scan of ECH resonance layer (A). The other is ECH modulation (B). (A)For radial scan of ECH resonance layer (balanced NBI), the purpose is as following: Motivation: (1)Previous DIII-D studies (J.deGrassie 2007) have noted that ECH results in the appearance of a V_Ï?=0 null point (core intrinsic counter rotation + pedestal intrinsic co-rotation). What sets null point location and proximity to low q resonance? (2)Add fluctuation studies and density scan to explore sensitivity of ï¼?TEM drivenï¼? intrinsic torque to trapped electron friction, collisionality,â?¦â?¦ Specific goals: (1) How does Vphi=0 point location vary? What sets the point position ? (2) How does ECH induced counter-torque depend on population of trapped electrons?Collisionality? (3) For equal P_ECH ,does â??electron characterâ?? of fluctuations increase with minor radius? Can we relate this to collisionality dependence via a density scan? (4) Using plan B (ECH modulation), does magnitude of Î _Resid increase with minor radius within the null point? Anticipated results: (1)Improved understanding of what sets Vphi=0 location (2)Insight into trapped particle fraction, collisionality dependence of ECH-driven intrinsic torque (B) ECH Modulation Motivation: (1)Study causality of ECH induced rotation flattening--> relative hysteresis between â??n_e, â??T_e, â??V_Ï?, etc. (2)Deduce the profiles of rotation convection (pinch) V and residual stress Î _Resid from modulations Specific goals: (1)Compare relative hysteresis: â??n_e vs V_Ï? , â??T_e vs â??V_Ï? ,which one is fundamental to V_Ï? flattening?? (2) Compare relative hysteresis: â??T_e (axis) vs â??V_Ï? (pivot), â??n_e (axis) vs â??V_Ï? (pivot) (3) Take Ï?_Ï?=Ï?_i (PB), then use to Î =-Ï?_Ï? (â??V_Ï?+(V_ V_Ï?)/Ï?_Ï? +Î _Resid/Ï?_Ï? ) to determine V(ï?²) and Î _Resid(ï?²) (4) simultaneously, ECH modulation --> Particle D_n and V_n --> Physics of density peaking in ECH rotation flattening discharges Anticipated results: (1)Determine rotation convection V ï?¦ and Î _Resid profiles (2)Determine relative importance of â??n_e, â??T_e in ECH/TEM intrinsic torque by comparing relative hysteresis (3)Determine particle convection V_n from modulation, compare V_n / V_ï?¦ to theory and assess density profile effects on intrinsic torque |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The background plasma for this proposal is heated with balanced NBI for minimum external momentum torque, which is similar to the low external torque situation of ITER. We donâ??t need too high NBI power, which should be hold at fixed power at the level to sustain H-mode. ECH is injected at the flat top of H-mode phase (quasi steady-state). During injection of ECH, the plasma should be well controlled and main parameters very stable. This proposal includes two steps (ECH resonance layer sweeping and ECH modulation). Firstly, we will do ECH resonance layer sweeping. The ECH layer should be swept from on-axis (ï?²~0) to off-axis (ï?²~0.6) during the flattop of one discharge. If the flattop is long enough, ECH layer can be swept back from off-axis to on-axis. We want repeat the ECH layer scan at several background density values. If ECH cannot scan in one shot, we will scan the density in one shot and change the ECH layer shot by shot.
Based on the ECH layer sweeping results, we can find the optimized ECH layer (maximum change of rotation and turbulence transition) for the modulation experiment. The modulation frequency of ECH is about 20Hz and the duty cycle is 50%. For modulation experiment, the first EC beam is aimed at optimized layer based on layer sweeping experiment. The second EC beam is aimed at off-axis (ï?²~0.6). The first EC beam should be turn on first for 10 pulses, and turn off. Then, the second EC beam turn on for 20 cycles if flattop is long enough. The first EC beam should be turn on again during the last 10 cycles of the second EC beam. We want repeat the ECH layer scan at several background density values. | ||
| Background: | Toroidal rotation is important for control of stability and transport in tokamaks. While NBI is used widely to control rotation in todayâ??s tokamaks, it is not a feasible approach for ITER. Moreover, the tendency of confined tokamak plasmas to self-accelerate to a state of intrinsic rotation has been identified and related to the state of plasma confinement. Intrinsic rotation is self-generated by ambient turbulence via the non-diffusive residual stress. This, then, motivates the question of how macroscopic rotation profiles will evolve in response to changes in the ambient micro-turbulence. One â??control knobâ?? for the micro-turbulence population is the heating mix of NBI (heats ion, and so drives ITG) and ECH (heats electrons, especially at lower density, and so can drive CTEM). On KSTAR, both XICS and CES confirm that the core toroidal rotation dramatically decreases when modest amount of on-axis ECH is injected to H-mode plasmas. Both the change of rotation and its gradient have a close relation to the change of electron temperature and its gradient in the core plasma. The change in rotation toward the counter-direction by ECH in KSTAR is explained by the turbulence change from ITG to CTEM. Gyrokinetic simulations support aspect of an ITG-TEM transition with peaked density profiles during ECH injection (there is no density profile available in KSTAR). In KSTAR, we also found the decrease of V_Ï? also depends on the ECRH deposition location â?? i.e. |Î?V_Ï? | is larger for on axis deposition than that for off-axis ECRH. DIII-D has powerful NBI, ECH, all relevant profiles and fluctuation diagnostics. Due to the unique balance NBI injection on DIII-D, we hope to make a rotation profiles of the form: counter-current direction (counter intrinsic torque by ECH) in core , co-current direction in edge (co- intrinsic torque due to H-mode pedestal). Here, we want to change the V_Ï?=0 location with a scan of the ECH resonance layer. Turbulence diagnostics should be applied to explore trapped particle population effects on TEM and rotation during ECH resonance layer scan. The poloidal rotation anomaly is predicted to reversal when ITGï? TEM transition occurs. For ECH modulation proposal, we expect to obtain Ï?_Ï? and Î _resid directly from experiment. And we found a close relation between Te, ne and Vï?¦ in KSTAR ECH experimental. With ECH modulation, more detail and accurate between T_e,n_e,V_Ï?profile responses (especially density profile effect) to ECH can obtain. Furthermore, the relative hysteresis of â??n_e vs V_Ï? , â??T_e vs â??V_Ï? , â??T_e (axis),â??n_e (axis) vs â??V_Ï? (pivot) will also be investigated. | ||
| Resource Requirements: | NBIs,
ECH, 6 shots might be desired | ||
| Diagnostic Requirements: | All profiles and all fluctuation diagnostics, especially core fluctuation measurement (BES, FIR scattering, and ECE-I), high speed CER for toroidal and poloidal rotation of full profile, Thomson scattering and microwave reflectometry for density profile.
If possible, DBS and PCI in core region | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 359: q95 scaling of the coupled turbulence/zonal flow during LH transition | ||
| Name: | Yan yanz@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | George McKee, L. Schmitz, George Tynan, P. Diamond, Jose Boedo, Dimitry Rudakov | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the q95 scaling of the coupled turbulence/zonal flow system before, during and after the L-H transition. Try to understand the L-H transition and the transition power threshold scaling physics. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main idea is to vary q95 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, UF-CHERS, DBS, CECE, etc 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 may relate to the mechanism for L- to H- mode transition [1]. It is shown that the zonal flow has strong q95 dependence.
So far no q95 scan has been done. By completing these it will help understanding the underlying physics and the scaling of power threshold, which is a key issue for ITER. | ||
| 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 | 360: density scaling and X-point position scan of LH transition power threshold | ||
| Name: | Yan yanz@fusion.gat.com | Affiliation: | U of Wisconsin |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | George McKee, L. Schmitz, George Tynan, P. Diamond, Jose Boedo, Dimitry Rudakov | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the physics behind the density scaling of the L-H transition power threshold, especially the density minimal for the power threshold. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main idea is to vary electron density at different X-point position in balanced beam injection plasma. Start with the plasma condition and shape used in the last L-H expt. in 2012 campaign lowest density, then reduce density more to try to find the region where L-H power threshold increases with decreasing density. If no density minimal can be found in that condition, then adjust X-point position relative to the pumping for another density scan. The beam power will be kept low to favor the Langmuir probe measurement of Reynolds stress at plasma edge. BES, UF-CHERS, DBS, CECE, etc will be used to measure plasma turbulence. | ||
| Background: | In 2012 campaign L-H expt. no density minimal of the L-H power threshold had been observed. This contrasts to the observations made years before. The X-point position could be a factor for this difference. | ||
| 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 | 361: Dither injection for closed-loop system identification of vacuum and plasma response | ||
| Name: | Olofsson olofsson@fusion.gat.com | Affiliation: | Consultant |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | L. Frassinetti, F.A.G. Volpe, P.R. Brunsell, J.R. Drake | ITPA Joint Experiment : | No |
| Description: | The proposal aims at using the I-coils for Feedback control of RWMs and simultaneous measurement of their structure and growth-rates (full 3d-response) by means of â??dither injectionâ?? (an automatic control technique largely used in the industry). The technique has the advantage of estimating the growth rates in a non-disruptive way since the plasma is simultaneously perturbed and stabilized. The data acquired during these type of experiments is furthermore of general interest since it is information-rich with respect to RWM code validation/unfalsification. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The envisaged application of dither injection to RWM control at DIII-D is as follows:
- the I-coils will be used to control the RWM with one of the algorithms commonly used at DIII-D - a small perturbation (random in time and space) will be applied to the voltages that drive the I-coils current. - the system response (plasma plus any conductive structure present in the wall) will be measured by the sensor coils. - Post processing analysis will allow the RWM growth to be determined. The technique will initially be applied to vacuum shots (DIII-D without plasma), both as a preliminary test as well as to obtain information on the 3D vacuum diffusion response. Dithering will be applied to both the I- and C-coils. In particular using the external C-coils will allow estimating the wall diffusion time [Olofsson et al., 2012, Fus. Eng. Design 87 pp 1926-1929]. The technique will then be tested on few plasma shots using a feedback algorithm normally deployed at DIII-D [for example as in Okabayashi et al., 2009 Nucl. Fusion 49, 125003]. If successful, we propose to then apply the dithering to several plasma shots with different beta in order to study in closed-loop the RWM growth-rate dependence. The technique inherently takes into consideration the 3D effects of the conductive structure. This allows a non-disruptive experimental cross-check of VALEN predictions [Okabayashi et al., 2005 Nucl. Fusion 45, 1715]. | ||
| Background: | Dither injection is a well-known technique in automatic control and it is largely applied in industry [e.g. Y. Zhu, Multivariable System Identification for Process Control (Elsevier Science 2001)] where quantitative models are required for the improvement of plant operations. It consists in the application of random perturbations to the signals in input to a physical system. The perturbations are usually generated as the output of a (designed) low-pass filter fed with white noise. By analyzing the response of the system (i.e. the relation between the input signals and output signals) it may be possible to identify the physical properties of the system. The technique is in general useful for the assessment of linear model applicability and prospecting the time-domain prediction horizon of linear models (potentially useful for routine plasma control).
Recently, dither injection was successfully applied to the system-identification of RWM growth rates in the EXTRAP T2R reversed field pinch. Preliminary tests were also carried out on RFX-mod. In EXTRAP T2R, dither injection allowed suppressing a large range of unstable RWMs avoiding disruptions and prolonging the discharge while simultaneously identifying the RWM growth rate. The estimated growth rates are in very good agreement with theoretical expectations, proving the validity of the method [Olofsson et al., 2011 Plasma Phys. Controll. Fusion 53, 084003]. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Magnetics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 362: Investigation of GAMs during I-mode on D3D | ||
| Name: | Hubbard hubbard@psfc.mit.edu | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | Anne White, Dennis Whyte (MIT)
George Tynan, Istvan Cziegler, J. Boedo (Center for Momentum Transport and Flow Organization, UCSD) G.M. McKee, Z. Yan (U. Wisconsin), L. Schmitz, E. Doyle, L. Zeng (UCLA). |
ITPA Joint Experiment : | Yes |
| Description: | The I-mode regime is extremely attractive for fusion in that it combines a thermal barrier and high energy confinement with high particle transport, preventing accumulation of impurities and removing the need for ELMs. It is also of interest for understanding the physics of edge turbulence, flows and transport barriers, particularly the separation of thermal and particle transport. Recent observations on C-Mod, reported by I. Cziegler at APS, that GAMs are present continuously during, and only during, I-modes and interact with the other turbulence are important in this regard. We propose to study these fluctuations and their role during I-modes on D3D. This will strongly support the 2013 Joint Research Target, and also contribute to ITPA Joint experiments TC-19 and PEP-31. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment should be coordinated with Dennis Whyteâ??s Idea 83 and/or Anne Whiteâ??s idea 37, demonstrating robust I-mode with LSN, reversed BT, standard Ip (co-NBI), and obtaining broadband turbulence measurements. Additional discharges in Hubbardâ??s companion Idea in reversed Bt will be valuable in improving BES resolution. We propose to focus available edge flow diagnostics to measure fluctuating and mean flows, in particular to look for possible GAMs at a few kHz. It will also be of interest to look for correlations and phase differences between fluctuations in different quantities (density, temperature, potential). As noted, it may be needed to repeat shots to optimize measurements. | ||
| Background: | I-mode is a stationary, high performance regime without ELMs â?? attractive in many respects for fusion, including ITER. It has been robustly obtained over wide parameter ranges on
C-Mod, and also observed on AUG over several years. Initial assessments of extrapolation to ITER, by Dennis Whyte, look promising. C-Mod measurements have shown that key edge turbulence signatures include a decrease in mid-freq turbulence (~ 100 kHz), which correlates with decreasing thermal transport (Hubbard PoP 2010) and usually a higher frequency â??weakly coherent modeâ?? (200-400 kHz), whose amplitude correlates with particle flux (Dominguez Ph.D. 2012). CORE turbulence has also been found to decrease in I-mode vs L-mode. An exciting new development, presented by I. Cziegler in an APS invited talk and submitted for publication, is the observation of a fluctuating zonal flow, at the GAM frequency, which appears during all I-modes examined to date. It exchanges energy with higher frequency turbulence, playing a role in broadening the WCM and perhaps in depleting the mid-frequency turbulence. In contrast to results at the L-H transition, on D3D and elsewhere, it is CONTINUOUS, not a transient phenomenon. However, the physics seems likely to be connected. Observations on D3D could be very important to make these connections and make progress in understanding turbulence and transport in L, I and H-mode regimes! | ||
| Resource Requirements: | NBI, co or balanced injection. And/or ECH (preferred due to ease of power programming). 3 MW total.
150L solid on (or perhaps 150R) would be required for BES; 330R for edge CER data (30L is best for core CER data). We may need to repeat shots to optimize different diagnostics, or may to alternate with 150 L/R or usebeam blips for CER beams to limit the total NBI power. | ||
| Diagnostic Requirements: | Essential: Edge fluctuation and flow diagnostics. Including Doppler Backscattering diagnostics (DBS-5, DBS-8) for GAMs, BES, reflectometry, Langmuir probes.
Also edge profile diagnostics (Edge TS, Profile reflectometry, CER) | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 363: TBM mockup simulating ITER GDC with ferritic materials | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Physics | Presentation time: | Not requested |
| Co-Author(s): | M. Shimada (ITER IO) | ITPA Joint Experiment : | No |
| Description: | Initiate and maintain a glow discharge in the presence of the TBM magnetic field | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. At specified helium pressure, turn on the TBM for â?? 5s, or as long as coil heating allows and apply glow voltage to attempt to initiate the GDC. Scan fill pressures and compare to 0 TBM current.
2. With a steady GDC, ramp up the TBM until the glow is extinguished. Measure any toroidal uniformity using 2nd glow anode as a langmuir probe. | ||
| Background: | A concern for ITER is obtaining an effective GDC to condition the walls in the presence of ferritic materials which have a residual magnetic field. Anecdotal evidence and some dedicated experiments have shown that fields of the order of â??100 G can inhibit breakdown and/or cause non-uniform conditioning of the wall. | ||
| Resource Requirements: | Diii-D glow system and TBM. GDC anodes at the wall (not inserted). Anode #1 nearest TBM operates normally. Anode #2 grounded through a 1 kOhm resistor to measure toroidal current uniformity. | ||
| Diagnostic Requirements: | Glow parameters (I,V, possibly PCM) digitized in sync with normal shot cycle. Possibly periscope camera, though the glow is very diffuse. Anode #2 configured to measure wall current (â?? 2.5V at 10 uA/cm^2 of wall current) | ||
| Analysis Requirements: | |||
| Other Requirements: | Need to sync glow diagnostics with d3 shot cycle. Probably requires adding Iglow and Vglow to Ptdata. | ||
| Title | 364: Absolute plasma response null measurement by multiple betaN steps | ||
| Name: | Paz-Soldan paz-soldan@fusion.gat.com | Affiliation: | Columbia U |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | S & DA team | ITPA Joint Experiment : | No |
| Description: | This proposal aims to demonstrate a new technique to experimentally determine optimum error field control (EFC) currents that is compatible with high-performance plasma scenarios. The measured optimum current levels will be compared to results with existing techniques. <br> <br>The technique is predicted to be able to determine the absolute plasma response null (which has been previously shown to be near optimum correction levels) using multiple betaN steps. The coil current prior to the betaN step is varied (per step), and the coil current is quickly scanned after the betaN step to discover the relative plasma response null. Because the structure of the measured plasma response after the step is proportional to the coil current before the step, repeating the scan with different initial conditions allows an experimental determination of the absolute plasma response null. <br> <br>It is proposed to do this experiment with and without a known (proxy) error field. The proxy field case will serve as a benchmark of the technique and ensure the plasma response null is where it is predicted by modeling (and the experiment described in RoF#188, if it occurs). The proxy-off case can then be confidently interpreted and yield the required current to null the plasma response to the machine intrinsic error. It is possible to skip the proxy step and jump straight to the intrinsic only case, though at a cost of reduced confidence in the final result. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The execution of the experiment requires stepping a stable plasma from a lower beta to a higher beta, if possible several times within a discharge. The coil currents before and after the step will be pre-programmed to map out the plasma response in the compass scan space after each step. This information will yield the absolute plasma response null, and thus a prediction of the optimal error field correction currents for that equilibrium.
This proposal will complement #188, where a rotation optimization will be used to independently determine optimal correction currents. The discharge scenario should be similar to that of the 2011 TBM experiment so that the results from the two may be compared. The reference shot is 147135, although extra betaN steps will be added. These measurements may also be useful for the new 3D magnetics diagnostic as the plasma response will be synchronously detectable. This could also be tried for n=2 if the n=1 experiment is a success. | ||
| Background: | The current technique used most often used to optimize error field correction in high-performance scenarios is dynamic error field correction (DEFC), which minimizes the change in the plasma response to a step in betaN using feedback. This process, however, is fundamentally iterative and requires several dedicated discharges. Furthermore, there is no transparent way to ascertain if the results are indeed optimal or if a feedback error took has taken.
The present technique is theoretically possible to accomplish in a single (long-pulse) discharge, requires no deleterious instability, and relies on our knowledge of the plasma response functional form. Furthermore, it is entirely pre-programed so could become a routine tool for the future with minimal control room tweaking. | ||
| Resource Requirements: | BetaN control is critical. Six 3D power supplies will need to be available (4 SPAs and 2 Cs). | ||
| Diagnostic Requirements: | ECE, CER, MSE, SXR, Thomson, 3D magnetics. | ||
| Analysis Requirements: | Normal suite of 3D field analysis tools. | ||
| Other Requirements: | -- | ||
| Title | 365: Lithium dropper commissioning and ELM modification | ||
| Name: | Jackson jackson@fusion.gat.com | Affiliation: | GA |
| Research Area: | General Physics | Presentation time: | Not requested |
| Co-Author(s): | C. Chrobak | ITPA Joint Experiment : | No |
| Description: | Commission the lithium dropper on DIII-D and evaluate the effect of lithium on ELMs | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In the startup phase of DIII-D in FY13, during beam conditioning, drop lithium and check out the functioning of the lithium dropper. As a secondary goal, look for any changes in edge conditions. This is the first part of MP2012-94-02 | ||
| Background: | The lithium dropper has been tested in the DIII-D lab and will be ready for installation on DIII-D before startup. | ||
| Resource Requirements: | lithium dropper. To mitigate any concern about window contamination the amount of lithium during startup will be less than 0.4 gm, which was the total of all lithium previously injected using the DIII-D lithium pellet injector (no adverse effects were noted in the past). | ||
| Diagnostic Requirements: | density, magnetics, filterscopes, zeff array, bolometer, Thomson and CER if shutters have been opened | ||
| Analysis Requirements: | |||
| Other Requirements: | This is listed as an ROF proposal (not just commissioning), because we expect to obtain some data on the effect of lithium on the plasma edge parameters. | ||
| Title | 366: Up/down Asymmetric EFC | ||
| Name: | Lanctot matthew.lanctot@science.doe.gov | Affiliation: | Department of Energy |
| Research Area: | Stability & Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo | ITPA Joint Experiment : | No |
| Description: | Develop up-down asymmetric error field correction using the I-coil. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start from a H-mode case with n=1 EFC optimized via plasma rotation maximization and then vary upper vs. lower I-coil row amplitudes at fixed phase and note rotation changes. Find maximum rotation. | ||
| Background: | The intrinsic error field and most plasma discharges are up-down asymmetric. Therefore, the optimal error field correction currents are also likely up-down asymmetric. Presently, all existing EFC algorithms are up/down symmetric and up/down asymmetric algorithms have not been investigated systematically. | ||
| Resource Requirements: | All four SPAs | ||
| Diagnostic Requirements: | CER | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 367: JET/DIII-D similarity experiment on rotation in plasmas with low momentum input | ||
| Name: | Nave mfn@ipfn.ist.utl.pt | Affiliation: | Instituto Superior Tecnico, Lisboa, Portugal |
| Research Area: | Plasma Rotation | Presentation time: | Not requested |
| Co-Author(s): | J. de Grassie | ITPA Joint Experiment : | Yes |
| Description: | A JET/DIII-D similarity experiment on intrinsic rotation has been proposed for JET (at the moment accepted as a back-up experiment, expected to be executed in C32 depending on the availability of rotation measurements). If the JET experiment goes ahead, it will be necessary to produce plasmas with similar shapes at DIII-D for measurements of intrinsic rotation.<br>The main goal of this experiment is the determination of the size scaling for intrinsic rotation, which is crucial for extrapolation for ITER. The JET/DIII-D similarity experiment will complement existing and proposed DIII-D/C-Mod similarity experiments.<br>This experiment is part of the ITPA JEX TC-9. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | JET intrinsic rotation experiments are planned to use a low triangularity configuration, low plasma currents (1-1.5 MA) and toroidal fields (1.1-1.7T) to achieve H-modes with high normalised beta values (1-2) using ICRH powers up to 4 MW. It is proposed to match JET shape and beta normalised values in DIII-D plasmas with ECH. Rotation should be measured during short NBI blips. We would like to compare JET and DIII-D intrinsic rotation during Ohmic and ECH phases. | ||
| Background: | -- | ||
| Resource Requirements: | ECH + NBI blips | ||
| Diagnostic Requirements: | standard profile measurements, in particular CER | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 368: Investigation of extreme outer wall heating caused by the DIII-D TBM | ||
| Name: | McLean mclean@fusion.gat.com | Affiliation: | LLNL |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | A. McLean, M. Lanctot, G. Kramer, C. Lasnier, J-W. Ahn | ITPA Joint Experiment : | Yes |
| Description: | An experiment using the TBM and magnetically insensitive ORNL fast infrared (IR) camera is proposed to explore heating of the TBM tile surface with respect to plasma parameters and shaping configurations not investigated in previous TBM experiments (2010 and 2011). | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Repeat target discharge from 2011 TBM (147603) and vary outer gap, heating, etc. to create validation cases for simulations and attempt to mitigate high sustained heat flux seen in the 2011 experiment. | ||
| Background: | Thermal data acquired using the ORNL fast IR camera monitoring the protective tiles on the DIII-D test blanket module (TBM) in 2011 revealed extreme, sustained heat flux of >10 MW/m2. This observation, largely confirmed by extensive fast ion modeling, has major detrimental implications for the ITER outer Be wall which, under this level of heat flux, would melt in <1 second. | ||
| Resource Requirements: | TBM installed in DIII-D, 270R0 port, beams, ECH | ||
| Diagnostic Requirements: | ORNL fast IR camera installed on support boom at 150R+1, TBM thermocouples, fast ion diagnostics | ||
| Analysis Requirements: | Plasma conditions from Thomson scattering, kinetic EFIT magnetic equilibriums, fast ion simulation/modeling for surface thermal deposition (SPIRAL, OFMC, and ASCOT) | ||
| Other Requirements: | Fast IR camera calibrated to high temperature beyond the current capability of 1050 degC using a 30+ year old blackbody source owned by GA; ideally 2000 degC. Temperature of the graphite tiles on the TBM was found to go well beyond this upper calibration and thus was dependent on an estimated calibration curve. | ||
| Title | 369: First-principles-driven Model-based Current-profile Control in H-mode Discharges | ||
| Name: | Schuster schuster@lehigh.edu | Affiliation: | Lehigh U |
| Research Area: | Plasma Control | Presentation time: | Not requested |
| Co-Author(s): | John Ferron, Tim Luce, Mike Walker, Dave Humphreys â?? General Atomics | ITPA Joint Experiment : | No |
| Description: | Establishing a suitable current profile has been demonstrated to be a key condition for the achievement of advanced tokamak scenarios with improved confinement and possible steady-state operation. The present 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. Previous experiments on DIII-D showed that the high dimensionality of the problem, and the strong coupling between magnetic and kinetic variables, call for the design of a model-based, multi-variable controller that takes into account the dynamic response of the full current profile to the different actuators.<br><br>The objective of this experiment is twofold. First, a control-oriented first-principles-driven model for the current profile dynamics in H-mode discharges will be developed and validated in DIII-D. Second, based on the developed and validated control-oriented first-principles-driven model, controllers for the regulation of the current profile in H-mode discharges will be designed and tested in DIII-D. Unique characteristics of the control approach are (i) the use of first-principles-driven models for the control synthesis, (ii) the integration of both static and dynamic plasma response models into the design of the feedback controllers, and (iii) the possibility of capturing the nonlinear dynamics of the system during the control synthesis. <br><br>This experiment is a natural extension of the successful experiments on first-principles-driven model-based current profile control in L-mode carried out in 2011. Extending control scheme to H-mode first requires model extension. In the to-be-developed H-mode current-profile response model: i- the electron and density profile model must include edge transport barrier; ii- non-inductive current drive and heating systems must be modeled individually (not together); iii- the effect of bootstrap current must be included. The controllers developed from first-principles models in L-mode discharges have used so far three actuators - plasma current, beam total power and line-averaged density. By adding EC H&CD as an actuator and grouping the beams in different categories we intend to improve controllability for simultaneous current profile and beta_N regulation in H-mode discharges. The developed control-oriented nonlinear model for current profile response in H-mode discharges will be used to design feedback controllers, which will be tested in scenarios relevant to the Steady State Scenarios and Inductive Scenarios thrusts. <br><br>It is important to emphasize that with the development of the DIII-D/LU profile control algorithm carried out in 2011, the PCS (plasma control system) at DIII-D does have now the necessary infrastructure for implementing such advanced profile 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, non-inductive current drive (NBI) power and heating (EC) power. The closed-loop controller will regulate in real-time these actuators based on real-time measurements of the q profile. 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. One additional goal of the controllers is to avoid MHD activity in the form of NTMs. Therefore, beta_N will also be regulated in closed-loop. Different initial and target profiles will be considered. The first-principles-driven, model-based, current-profile control experiment in H-mode will require two 2-hour evening sessions and at least one half-day session.
It is important to emphasize that the to-be-developed nonlinear control-oriented model could find applications beyond feedback control design. First, this model could be used for feedforward control design (scenario planning). Determining whether a particular current profile is achievable given the initial conditions and actuators constraints, and eventually finding the actuators trajectories that are necessary to achieve a particular achievable current profile are two very important problems arising in tokamak operation that could find solutions by exploiting the developed nonlinear control-oriented model. Second, this model could be used for state estimation and prediction in real time. Noise could be separated from the actual plasma state (current profile) by using the developed nonlinear control-oriented model, which would filter any component of the estimated current profile not predicted by the model. Moreover, the plasma state (current profile) could be estimated in real time from a limited set of diagnostic (not including MSE for instance) by exploiting the prediction by the developed nonlinear control-oriented model. Finally, this model could be used as a simulation testbed. Controllers designed based on more simplified models, including identified linear plasma response models arising in data-driven modeling, could be tested in closed-loop simulations based on the developed nonlinear control-oriented model (controllers developed as part of proposal #316 could greatly benefit from this simulation capability). This will provide the opportunity of systematically comparing first-principles-driven and data-driven approaches to profile control. | ||
| Background: | The Plasma Control Group at Lehigh University (LU) headed by Prof. Eugenio Schuster has been working on this problem for several years now. A preliminary first-principle control-oriented model of current profile evolution in response to auxiliary H&CD systems (NBI, EC) and electric field due to induction was developed for L-mode discharges [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 1/2day experiment showed qualitative agreement between model prediction and experiment, and corroborated that the actuators constraints were correctly taken into account during the control synthesis. A reduced-order first-principles model was 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]. Extensions of these controllers were tested in L-mode discharges in DIII-D in 2011 [6, 7, 8], which represents the first time ever model- based, first-principles-driven, full-magnetic-profile controllers were successfully implemented and tested in a fusion device.
[1] Y. Ou, T.C. Luce, E. Schuster et al., Towards Model-based Current Profile Control at DIII-D, Fusion Engineering and Design 82 (2007) 11531160. [2] Y. Ou, C. Xu, E. Schuster et al., Design and Simulation of Extremum-Seeking Open-Loop Optimal Control of Current Profile in the DIII-D Tokamak, Plasma Physics and Controlled Fusion, 50 (2008) 115001. [3] C. Xu, Y. Ou, J. Dalessio, E. Schuster et al., Ramp-Up-Phase Current-Profile Control of Tokamak Plasmas via Nonlinear Programming, IEEE Trans. on Plasma Science, vol.38, no.2, pp.163-173, 2010. [4] Y. Ou, C. Xu and E. Schuster, Robust Control Design for the Poloidal Magnetic Flux Profile Evolution in the Presence of Model Uncertainties, IEEE Trans. on Plasma Science, vol.38, no.3, pp.375-382, 2010. [5] Y. Ou, C. Xu, E. Schuster et al., Optimal Tracking Control of Current Profile in Tokamaks, IEEE Transactions on Control Systems Technology 19 (2), 432-441 (2011). [6] J. Barton, M.D. Boyer, W. Shi, E. Schuster et al., Toroidal Current Profile Control During Low Confinement Mode Plasma Discharges in DIII-D via First-Principles-Driven Model-based Robust Control Synthesis, Nuclear Fusion 52 (2012) 123018 (24pp). [7] M.D. Boyer, J. Barton, E. Schuster et al., First-Principles-Driven Model-Based Current Profile Control for the DIII-D Tokamak via LQI Optimal Control, Plasma Physics and Controlled Fusion, under review. [8] M.D. Boyer, J. Barton, E. Schuster et al., Backstepping Control of the Toroidal Plasma Current Profile in the DIII-D Tokamak, IEEE Transactions on Control Systems Technology, under review. | ||
| Resource Requirements: | Machine time: Two 2-hour evening sessions + at least 1/2 day experiment.
Actuators: All NB and EC H&CD systems at full power. | ||
| 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) and beta_N 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: | Matlab. MDSPLUS. | ||
| Other Requirements: | -- | ||
| Title | 370: Particle pinch in pedestal | ||
| Name: | Groebner groebner@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | J. Boedo, T. Leonard, T. Osborne | ITPA Joint Experiment : | No |
| Description: | Look for evidence of an inward particle pinch in the pedestal by using reciprocating Langmuir probe to look for an inward fluctuation-driven particle flux. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop H-mode plasma which is benign as possible for the reciprocating Langmuir probe. This means no beams (thus ECH heating or Ohmic H-mode), low density (thus low current) and low temperature (thus as low as possible Bt). Develop an H-mode with as long an ELM-free phase as possible (which will probably not be long under these circumstances). Insert probe as far as possible into pedestal to measure profile of turbulent-driven particle flux. We want to see if this flux is inwards, and if so, how far into the pedestal the flux is inwards. | ||
| Background: | There a number of reasons why it is important to obtain an understanding of the physics processes which govern the density profile. For instance,
1) ITER needs to know what the fuelling requirements are to build up the necessary density pedestal; 2) Control of the density profile appears to be part of the physics for why some ELM-suppression regimes work. Thus control of the density pedestal may be a tool to optimize ELM control. 3) Control of the pedestal density is predicted by EPED to be a tool that would enable optimization of the pedestal height. The actual physics that forms the density pedestal structure is not well understood. In the past, some of us have thought that the neutral penetration depth determined the width. However, we now have a number of reasons why we suspect that this is not the full story and that details of particle transport also play a role. In particular, we suspect that there is an inward particle pinch in the pedestal that helps build up the density and also to push the density barrier further into the plasma. We have made attempts to infer the presence of a pinch via different experimental means, particularly by producing plasmas with dense scrape-off layers, which would limit neutral fuelling of the pedestal. However, our studies have not been definitive and it looks to be very difficult to indirectly infer a pinch. It would be much better to directly measure an inward particle flux, and the reciprocating Langmuir probe is capable of making these measurements. In fact, there have been reports in the past of inward-directed particle fluxes observed in H-mode on Langmuir probes. The goal of this experiment would be to fill in gaps from past work and attempt to get a clear measurement of an inward particle flux under plasma conditions which were as benign to the probe as possible. | ||
| Resource Requirements: | DIII-D tokamak, ECH heating, neutral beam blips for diagnostic purposes. | ||
| Diagnostic Requirements: | Reciprocating Langmuir probe, TS, CER | ||
| Analysis Requirements: | Analysis of Langmuir probe data to produce particle fluxes. The results could be compared to particle fluxes computed with a code, such as TGLF. | ||
| Other Requirements: | -- | ||
| Title | 371: Effect of icoil current on energy confinement in RMP ELM-suppressed discharges | ||
| Name: | Groebner groebner@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Determine how the plasma energy confinement varies with icoil current in discharges with ELMs suppressed via the application of RMP fields. Determine if confinement degradation with the application of these fields can be minimized by choice of icoil current. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop a robust ELM-suppressed discharge with RMP fields, applied with the icoil. Shot by shot, vary the icoil current and measure the H-factor and pedestal height for those discharges which remain ELM-suppressed. | ||
| Background: | ITER cannot tolerate large ELMs and the application of 3D fields via internal coils is planned on ITER as one technique to mitigate ELMs. This plan is strongly motivated by success on DIII-D in eliminating ELMs with the application of RMP fields from the icoil. However, ELM-suppressed H-modes typically show a reduction of energy confinement compared to that in the ELMing plasma prior to ELM suppression. This reduction of energy confinement is a concern to ITER scientists. The actual scaling of energy confinement with important control parameters, particularly the icoil current, is not well known. We need to determine the dependence of energy confinement on icoil current in RMP ELM suppressed regimes. It is plausible that for a current just above the threshold required to obtain ELM suppression that any reduction in energy confinement will be very small and acceptable. We need to find out if this is the case so that we can inform ITER. | ||
| Resource Requirements: | DIII-D tokamak, neutral beams, icoil | ||
| Diagnostic Requirements: | TS, CER | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 372: Pedestal optimization via impurity injection | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | T. Osborne, R. Groebner | ITPA Joint Experiment : | No |
| Description: | The EPED model predicts that, in strongly shaped plasmas, the pedestal height depends strongly on collisionality, and can be optimized dynamically by starting at low collisionality, and slowly increasing collisionality with time. In addition to density variation, it is possible to modify the pedestal collisionality by injecting low-Z impurities to increase Zeff.
Here we propose to control edge collisionality by injection of Neon or Nitrogen, to study the physics of this effect, and to optimize the pedestal pressure to allow high performance at a range of densities. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In a very strongly shaped plasma, scan density and Zeff via impurity injection (Ne or N). Make predictions with EPED ahead of time to guide choice of shape, Ip, Bt and optimal range of Zeff and density.
Experiment could be conducted either in ELMing or QH mode discharges. | ||
| Background: | |||
| Resource Requirements: | Ne or N injection at substantial gas rates. Core pellet fuelling also desirable | ||
| Diagnostic Requirements: | Pedestal profile and turbulence diagnostics | ||
| Analysis Requirements: | EPED studies ahead of experiment | ||
| Other Requirements: | |||
| Title | 373: Pedestal optimization in VH mode | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | T. Osborne, R. Groebner, G. Jackson, R. Maingi | ITPA Joint Experiment : | No |
| Description: | Very high pedestals (or possibly, double barriers) and high performance have been achieved in DIII-D VH-mode operation. However, nearly all VH mode operation occurred well before recent pedestal diagnostic upgrades and advances in understanding pedestal physics.
The goal is to re-visit, study, and optimize pedestal structure in VH-mode discharges, employing recently upgraded diagnostics. Primary objectives include: 1) Determine whether the VH-mode ETB is a single broad barrier, or a pair of separate barriers with a narrow gap in between. 2) Study dependence of VH-mode pedestal on density and impurity content (ideally with Ne or N puffing) 3) Make extensive predictions with EPED before expt, and compare to observations to determine whether VH mode observations can be understood in terms of this model, and also to determine the relationship between VH-mode and the "Super H-Mode" regime predicted by EPED 4) Develop stationary VH-mode discharges either as ELMing discharges (eg recovering high Zeff rapidly after each ELM with impurity injection), pellet-paced ELM discharges, or as ELM-free (employing working model of RMP ELM to predict necessary q and other conditions) |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a high performance VH mode in a configuration consistent with good pedestal diagnostic coverage. Jog plasma slowly across edge diagnostics to study ETB structure in detail. Vary density and impurity injection rates. Establish stationary VH mode by employing one of the above 3 techniques. | ||
| Background: | VH mode operation may be relevant to the JRT, particularly to comparisons with NSTX | ||
| Resource Requirements: | at least 1 day | ||
| Diagnostic Requirements: | pedestal profile and turbulence diagnostics | ||
| Analysis Requirements: | EPED study before experiment, revisiting old VH mode discharges | ||
| Other Requirements: | |||
| Title | 374: Pedestal optimization with RMP ELM Control | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | A key question for low collisionality RMP ELM control is whether sufficient pedestal pressure can be obtained in the presence of RMP ELM control to enable high performance.
A developing model for RMP ELM control based on EPED and plasma response calculations suggest possible routes toward optimizing the pedestal in the presence of ELM control. Exploring these routes enables: 1) Detailed tests of existing understanding, and valuable information for improving that understanding 2) Potential for higher performance RMP discharges 3) Data and model validation to support the possibility of high performance operation with RMP ELM control on ITER Primary tools to enable pedestal optimization with RMP are: 1) Impurity injection (Ne or N) to increase collisionality at a given density 2) Density control (model predicts pedestal pressure should increase with density) 3) Shape (optimize "natural" pedestal height and width) 4) Precise q control |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish RMP ELM control in the standard shape with a patch panel enabling significantly stronger shaping. Increase density (ideally via core pellets) in a series of discharges until ELM control is lost. Select and intermediate density, and increase Zeff with impurity puffing until ELM control is lost. Move to stronger shaping, with appropriately adjusted shaping and repeat density and Zeff scans. | ||
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| Diagnostic Requirements: | pedestal profile and turbulence diagnostics. edge current measurements highly desirable | ||
| Analysis Requirements: | Series of EPED runs before expt | ||
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| Title | 375: QH-mode pedestal optimization | ||
| Name: | Snyder snyderpb@ornl.gov | Affiliation: | ORNL |
| Research Area: | Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, T. Osborne, R. Groebner | ITPA Joint Experiment : | No |
| Description: | Previous theory-motivated optimization of the pedestal in QH mode via improvements in shape and increasing density have enabled high pedestal pressure and high performance in QH mode. The EPED model predicts that further optimization, including possible access to "Super H-Mode" should be possible, and enable substantial performance improvements.
An important tool that has not been employed is the controlled puffing of low-Z impurities (eg Ne or N) to enable control of Zeff at a given density. The EPED model predicts that this should increase pedestal pressure and possibly enable QH-mode access. It also is predicted to be an avenue for obtaining high performance in ITER without exceeding the Greenwald limit. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Conduct a detailed optimization study beforehand with the EPED model to determine optimal shape and range of density and Zeff.
In a very strongly shaped, but single null, configuration with good wall conditions, obtain QH mode. Ramp down rotation with time to produce a density scan in each discharge. Increase impurity puff rate to vary Zeff across a series of shots. Select optimal conditions for detailed diagnosis and performance extension. | ||
| Background: | |||
| Resource Requirements: | 1+ days | ||
| Diagnostic Requirements: | pedestal structure and turbulence diagnostics. | ||
| Analysis Requirements: | EPED runs before the expt | ||
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| Title | 376: Role of q and magnetic shear of pedestal formation time and turbulent transport | ||
| Name: | Diallo adiallo@pppl.gov | Affiliation: | PPPL |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | R. Groebner, D. Eldon, J. Canik, T. Rhodes, T. Osborne, W. Guttenfelder, and R. Singh | ITPA Joint Experiment : | No |
| Description: | The goal of this proposal is to investigate the underlying physics mechanism during the density and temperature pedestal formations (e.g., time scales) and to identify the instabilities responsible for edge transport after an ELM crash. Dependencies of pedestal formation time and instability characteristics (amplitude, spectra) for various plasma currents will be investigated to elucidate the role of q and magnetic shear, and the scaling with gradient scale lengths on fast time scale. The experiment will be focused on pedestal profiles within few milliseconds after the ELM crash. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiments will target measurements of the pedestal dynamics (over short time scale < 10 ms) with great accuracy to enable measurements of the pedestal formation time and identify the fluctuations and their role in particle and heat transport. To elucidate dependencies in q and magnetic shear, scans in plasma current are to be performed. Specifically, we propose two approaches (both will have plasma current scan components) depending on the availability to ELM triggering capabilities.
Approach A: 1. Rely on intrinsic ELMs. a. Choose ELMy discharges (142XXX) with long inter-ELM periods. b. Vary the bunching the 7 laser pulses to resolve the pedestal formation i. Start with 100 microsec apart ii. Increase to 300 micro sec iii. Increase to 0.5 ms c. Combine the Thomson measurements with profile reflectometry i. The reflectometry will resolve the fast (25 micro sec) time scale of the pedestal formation. Note that Thomson could provide 100 micro sec resolutions. d. Particular attention will be given at obtaining fluctuation data during the pedestal formation. Such measurements include DBS, and BES. Approach B: (if the ELM triggering capability is available) 1. Trigger an ELM (30 â?? 70 Hz) This could be done with either the Deuterium pellet or lithium slapper, whichever is available during the campaign. 2. Synchronize the pellet triggers with laser firings and both profile reflectometry and DBS (if possible) a. Vary as the laser firing as approach A. b. The profile reflectometry may not need to be synchronized. c. Note that the time resolution of the profile reflectometry is 25 microsec and 100 microsec for Thomson. | ||
| Background: | It is largely accepted that the fusion performance is dictated by the pedestal characteristics. More specifically, the pedestal pressure height and width are key parameters for the fusion gain. While the pedestal pressure is key in the edge MHD stability, it is important to independently assess from the edge transport point of view the temperature and density pedestal characteristics during the pedestal formation. An understanding of the pedestal formation is therefore a key issue in developing predictive model and optimizing the pedestal for maximum core fusion. The two recent theories that address the pedestal dynamics are the KBM theory in the EPED model, which postulates the clamping of the pressure pedestal gradient with the onset of KBM instability [SNYDER, PoP 2012]. The onset of the KBM generates a steady-state transport in all channels (particle and heat). One other theory proposes a model on turbulent particle pinch due to ETG in the edge for rapid formation of the pedestal [KAW, IAEA 2012 TH/P4-15]. Clearly to elucidate the physics at play, the next step in experiments is to identify the individual transport mechanisms at play during the pedestal formation to better develop predictive capabilities of particle and heat fluxes in the pedestal. More specifically, several tests will include estimating correlations between the density fluctuations amplitudes as measured from DBS and BES and the density and temperature scale lengths. Variations of these correlations with plasma current (q and magnetic shear) and comparisons with gyrokinetic scaling studies (gradient, q, shear) using the experimental profiles (see example of such studies in Canik et al. IAEA 2012 submitted to Nucl. Fusion 2013) will unambiguously provide information on the instabilities present in the pedestal region during its formation. In summary, we test the pedestal formation before it hits the ballooning limit and characterize the density, temperature, and turbulent fluctuations in the pedestal during the first 10 ms after the ELM crash for three plasma currents. Such characterizations and scaling with plasma current will provide an estimate of the pedestal width to be expected in next-step devices such as ITER and FNSF. | ||
| Resource Requirements: | Fast-sweeping reflectometer sampling the pedestal region. Pellet injection capability for ELM triggering, DBS, BES, Thomson with the ability to arbitrary (> 100 microseconds) space the laser pulses. The experiment will require Er measurements (of time scales faster than CER) with DBS (if possible). | ||
| Diagnostic Requirements: | All profiles diagnostics. Special attention on the firing of the lasers DBS measurements and reflectometry, BES array centered in pedestal region, and midplane calibrated Dalpha signals. | ||
| Analysis Requirements: | Profile analysis, fluctuations analysis. The profile characteristics will provide inputs for GS2 linear and GYRO nonlinear analyses to identify the instabilities, compare with fluctuation data, and identify the transport channels at play during the pedestal formation. | ||
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| Title | 377: NTM locking disruption avoidance by the EM torque with toroidal-phase forward magnetic feed | ||
| Name: | Okabayashi mokabaya@pppl.gov | Affiliation: | PPPL |
| Research Area: | Disruption Mitigation | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, R. LaHaye, D. Shiraki, Ted Strait and Francesco Volpe. | ITPA Joint Experiment : | No |
| Description: | For successful operation of reactor-oriented devices like ITER, the NTM mode-locking is one of potentially-serious MHD events leading to major disruptions. Here, it has been proposed to apply accelerating electromagnetic (EM) torque and to overcome the mode-locking torque due to imperfect 2D magnetic fields (error fields). This overcoming against mode-locking is produced by magnetic feedback system utilizing internal 3D coils. A key element is to introduce the feed-forward toroidal-phase shift between the observed NTM mode and the applied feedback magnetic field. An advantage of feedback approach is that the applied total torque increases quadratically when NTM amplitude increases. Secondly, the presetting feed-forward phase shift is convenient to control the amount of applied toque input. Thirdly, the dynamic error field correction (DEFC) process takes place simultaneously since the feedback parameters are typical DEFC settings. This feedback scheme is expected to find a quasi-steady state NTM rotation condition at very low rotation during the slowing down period toward mode-locking. A simple toroidal-phase control stability model predicts that the direction of mode propagation depends on the direction of toroidal-phase shift setting and that the NTM frequency at a very low rotation steady state is the order of the inverse of the filtering time constant preset in the feedback system. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Preliminary results of its application in DIII-D high beta plasmas are promising. By proper presetting of the feed-forward toroidal phase shift, the NTM propagating initially with ~ 5-6 kHz was slowed down, but was sustained around ~50 Hz as a new low frequency steady-state equilibrium point without leading to locking. The stored plasma energy was kept within 70% range of initial level. The C-VI rotation was initially around 5kHz at q=2 surface and reduced to well below 500 Hz. It is hard to estimate accurately the bulk D-plasma rotation frequency. The final NTM frequency of ~50 Hz corresponds to the inverse of the filtering time constant (~ 1/40 ms). The mode propagation was found to flip its direction depending upon the preset of toroidal-shift direction. During the feedback in the low frequency quasi-steady state, the toroidal-phase difference between the NTM and the feedback field was fluctuated but remained stable, consistent with a simple model prediction. However, non-linearity of this toroidal phase difference in time implies the possibility of the influence due to uncorrected error field.
This approach can be applied to any phase when the NTM or TM is ecited in the middle of disruption mitigation. | ||
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| Title | 378: Study of the connection between I-mode and low density Type III ELMs and optimization of I-mode | ||
| Name: | Osborne osborne@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | R. Maingi | ITPA Joint Experiment : | No |
| Description: | Both the predator-prey oscillations near the L-H transition (Schmitz, Boedo) and the recent tentative I-mode discharge on DIII-D have characteristics similar to low density Type III ELM regime studies in the mid-1990s. In the proposed experiment build a better understanding of the connection between these regimes and exploit our understanding of the Type III regime and of pedestal optimization to improve the performance of I-mode discharges. ELM like relaxation oscillations are observed on D-alpha signals in all these regimes, being very high frequency in the I-mode discharge. Broadband density fluctuations seen on the reflectometers were observed to shutoff at the transition from the low density Type III ELM regime to the ELM free regime. In recent Type III discharges these modes have also been associated with a magnetic perturbation in the 50kHz range with n~-8 (electron drift direction). ECE measurements localize these modes to the region near the top of the pedestal in a zone where the Te gradient is still large but the density gradient is relatively low suggesting etae may be an important parameter. The pedestal pressure gradient in this regime is well below the (EPED1 predicted) KBM limit. The pedestal width is observed to rapidly narrow in the transition from Type III to the Type I ELM regime with the pressure gradient quickly coming into agreement with the EPED1 predicted KBM limit. This fact suggests the possibility of obtaining high pedestal pressure in this regime while avoiding Type I ELMs if sufficient pedestal width were obtained. The power required for transition from Type III to type I scaled as P=Ip**2.4/ne**2 which allowed operation continuously in the Type III regime at heating power well above the L-H transition power (6xPLH was achieved) at high enough plasma current and low enough density and discharges with continuous low density type III ELMs have been obtained at sufficiently low power or density. The threshold conditions for transition from Type III to I regimes in terms of local parameters was shown to be associated with a critical alpha parameter, suggestive of the Rogers-Drake and Pogutse-Igitkhanov models for the scaling of the low density branch of the L-H transition. During the type III studies it was found that the pedestal pressure and H factor increases as the threshold to transition to standard ELM free H-mode (followed by Type I ELMs) is approached (H93H=1 was achieved). If the recent I-mode discharges are indeed related to the Type III discharges, these scalings could act as a guide to obtaining high performance ELM free I-mode discharges. | ITER IO Urgent Research Task : | No |
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| Title | 379: Pedestal optimization of ITER baseline scenario discharges. | ||
| Name: | Osborne osborne@fusion.gat.com | Affiliation: | GA |
| Research Area: | Pedestal Structure | Presentation time: | Not requested |
| Co-Author(s): | P. Snyder, R. Groebner, T. Rhodes, G. McKee, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Previous work (Osborne APS 2012) has demonstrated a variation in pedestal pressure of up to 50 % can be obtained ITER baseline scenario discharges. In this experiment we would build a better understanding of the reasons for this variation. High pedestal pressures were associated with the build up of pedestal density at low ELM frequency with the pedestal temperature saturated early in the inter-ELM period. This enhanced the pedestal collisionality allowing higher pressure at the peeling mode limit, however the dominate effect was an increase in the pedestal width. In the lowest ELM frequency discharges the width substantially exceeded the EPED1 width prediction. This is in qualitative agreement with the expectation that the KBM stability should be reduced at high collisionality allowing higher pressure at the peeling limit. In the new experiments we would obtain better fluctuation data to determine if the high width discharges are indeed still KBM limited or in fact if some other process has limited the pressure gradient below the KBM threshold. In addition the mechanism for the variation in the ELM frequency and pedestal pressure was not clear from the previous experiment. The build up of high Z impurities or the betan feedback control of the heating power may have played a role. | ITER IO Urgent Research Task : | No |
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| Title | 380: ELM Suppression at Lower Collisionality with Non-Resonant n=3 I-coil | ||
| Name: | Osborne osborne@fusion.gat.com | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This experiment would attempt to obtain suppression of Type I ELMs by using no-resonant I-coil fields to enhance Type II ELM activity in discharges which naturally exhibit strong Type II activity at lower collisionality. ELM suppression without the strong pedestal pressure reduction observed in RMP low collisionality cases might be obtained with this approach. In this experiment we would obtain discharges with strong Type II activity, perhaps close to double null conditions at lower collisionality and apply odd parity n=3 fields with the I-coil. Scan q95. There is some evidence that the suppression of type I ELMs in higher collisionality odd Icoil parity discharges was through enhancement of the Type II ELM activity. In contrast to RMP suppression of ELMs at low collisionality, these discharges showed little change in pedestal pressure in the absence of Type I ELMs. Although Type II activity is generally enhanced with increasing collisionality it is also a function of plasma shape and other parameters. It is possible that the effect of the non-resonant fields on Type II activity is not a function of collisionality and therefore might still function in a lower collisionality discharge with strong enough Type II activity. | ITER IO Urgent Research Task : | No |
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| Title | 381: Search for turbulence at ion-cyclotron frequencies | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Turbulence & Transport | Presentation time: | Not requested |
| Co-Author(s): | R. Waltz, T. Rhodes | ITPA Joint Experiment : | No |
| Description: | GYRO simulates "low" frequency turbulence from ITG/TEM/ETG modes. However, GYRO does not include electrostatic turbulence at ion-cyclotron frequencies. Interaction between ion cyclotron and low frequency drift turbulence is a potential important physics mechanism which could lead to GYRO underestimating turbulence and and transport in L-mode plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Operate stationary L- and H-mode plasma phases without RF heating. Characterize presence and level of turbulence at ion cyclotron frequency range - measurements in L-mode are key. | ||
| Background: | See Waltz and Dominguez, Phys. Fluids 24 1575 (1981) for theory of ion-cyclotron modes. | ||
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| Diagnostic Requirements: | Extend frequency range of DBS system to cover iion cyclotron frequencies (10s of MHz). | ||
| Analysis Requirements: | Estimation of whether observed turbulence levels are significant for GYRO. | ||
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| Title | 382: Transport and turbulence characterization experiment | ||
| Name: | Doyle doyle@fusion.gat.com | Affiliation: | UC, Los Angeles |
| Research Area: | Steady State Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Rhodes, McKee, Schmitz | ITPA Joint Experiment : | No |
| Description: | Transport and turbulence teams have not been fully engaged with steady-state experiments. Best place to start is to obtain transport reference and turbulence characterization discharges for leading DIII-D scenarios. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Obtain turbulence and transport documentation for the four leading steady-state scenarios on DIII-D, viz q_min~1.5, q_min > 2, ITER steady-state scenario, and off axis-NBI scenario. This requires repeat shots of discharges from each scenario to vary beam mix (BES versus CER), DBS wavenumbers, and CECE location. | ||
| Background: | Transport is the limiting factor in steady-state experiment. Confinement varies between campaigns/years and between scenarios - i.e. ITER steady-state scenario had unexpectedly good confinement, while confinement in other regimes has deteriorated. This argues for obtaining transport and turbulence documentation/reference shots for our key steady-state scenarios. | ||
| Resource Requirements: | 8 beams, all gyrotrons | ||
| Diagnostic Requirements: | BES, DBS, CECE | ||
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