| Title | 1: Investigate Disagreements Between Thomson Scattering and ECE Measurements in High Te Discharges | ||
| Name: | White | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | M. E. Austin, R. Prater, P. Bonoli, R. Harvey, S. Scott, B. Bray, C.C. Petty, R. Pinsker | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment is to search for a discrepancy between Thomson scattering (TS) and ECE measurements of Te on DIII-D in discharges with high electron temperature. To carry out the experiment, L-mode discharges are used to attain central electron temperatures of Te(0) = 9 keV with NBI + FW and Te (0) = 15 keV with NBI + ECH (achieved in 2009 in MP 2010-55-01). This year, we will be targeting high power, off-axis ECH discharges to reach Te(0) > 12 keV, while maintaining a Maxwellian f(v) in the core plasma. Electron temperature measurements from TS and the absolutely calibrated Michelson interferometer would be compared to look for any discrepancy between the two diagnostics. With the 2009 data, a gate timing issue with TS was discovered several weeks after the run. This limited the time resolution/available statistics for the experiment (Bray, Friday Science Meeting, May 7th , 2010) and is a main reason the experiment must be repeated. A second motivation for this new experiment this year is that based on using NBI on DIII-D to investigate a specific hypothesis for the cause of the discrepancy (R. Harvey): cold electrons deposited from the NBI may flatten the electron f(v) in the correct vicinity to produce the discrepancy (per modeling by de la Luna and Krivenski) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The hypothesis is that cold electrons deposited from the NBI may flatten the electron f(v) in the correct vicinity to produce the discrepancy (per modeling by de la Luna and Krivenski). Indeed, it is well known that after the NBI is shut off and the influx of cold electrons (cold is born at the ionization energy, 10s of eV) ends, Te increases over several confinement times. This is seen in DIII-D (moderate rises of 100s of eV) and was quite large in TFTR supershots (1-2 keV). At DIII-D, with combination of NBI and ECH, it will be possible to make high Te plasmas with phases where the beams are on and where beams are off. TS and ECE will be compared in both cases. If the cold electrons are responsible for the discrepancy, the discrepancy should only be seen when beams are on.
Reference shot 140715 (beams and FW to high Te(0) > 9 keV). Obtain high quality TS and ECE data for detailed comparisons and modeling. Apply high power ECH near r=0.25 or 0.3 starting when the current ramp is nearly complete, generate plasmas with weak negative shear for good confinement but avoid the very strong eITB that comes with strong negative central shear. Prater??s recipe for using ECH in these experiments: Leave the plasma center free of ECH to maintain a Maxwellian distribution there and use large enough k_parallel that little EC power is deposited via relativistic downshift; split the power equally into positive k_parallel and negative k_parallel to avoid driving excessive local current . In this experiment, we can look with the beams on and then turn the beams off during the highest Te phase of discharge to get the beam off data. | ||
| Background: | One hypothesis from JET was that ICRH-generated fast ions may be related to the cause of the discrepancy, because the discrepancy was seen above 5 keV in NBI+ICRH plasmas but not NBI only plasmas [de la Luna 2008]. However, recent experiments at C-Mod produced high Te plasmas with ICRH (fast ions present) and Mode Conversion heating (no fast ions) and there was no evidence of the discrepancy for Te(0) < 8 keV in either case [White, et al. to be submitted Nuclear Fusion]. This would indicate that fast ions may not responsible for the discrepancy.
In the new experiment at DIII-D we will explore a a different hypothesis related to NBI cold electron influx that may help explain the TFTR discrepancy in particular, where the discrepancy was seen in NBI plasmas [Taylor, 2009] Note that in the core of typical tokamak plasmas with Te(0) < 7 keV the TS and ECE measurements of electron temperature are in very good agreement. Also note that TS and ECE measurements of electron temperature often disagree in high Te(0) discharges that are strongly heated via ECH, but in these cases the disagreement can be explained by a well understood perturbation of the electron energy distribution function caused by the ECH [3]. In contrast to these cases, the cause of the TS/ECE discrepancy in discharges heated with only NBI or NBI + ICRH on JET and TFTR where Te(0) > 5 keV is not known. Such a discrepancy has been observed in NBI discharges and discharges heated with both NBI and Ion Cyclotron Resonance Heating (ICRH) discharges in TFTR and JET [1,2]. In these cases, the central electron temperature Te(0) measured with ECE diagnostics is 10-20% higher than the TS measurement of Te(0) The discrepancy starts at Te(0) ~ 7 keV and increases approximately linearly with electron temperature. Theoretically, a non-Maxwellian electron distribution f(v) with distortion near the thermal velocity may create such a measurement discrepancy between TS and ECE measurements [4], however, no known mechanism can sustain that type of distribution with finite heating power As an ITPA joint experiment, either a positive or a negative result on this topic from DIII-D can significantly impact international efforts to understand the past discrepancies that have been reported on TFTR and JET [1,2]. For example, a positive result, the observation of the discrepancy between TS and ECE on DIII-D, would verify the discrepancy on an additional machine and would therefore motivate a new and detailed investigation of the phenomenon. However, a negative result, the observation of no discrepancy under a variety of conditions with high electron temperature produced with NBI and FWH, would be equally beneficial as it would show that agreement between TS and ECE can be obtained in high temperature tokamak discharges. In both cases, the experimental results and associated modeling will improve the understanding of heating and diagnostic techniques in plasmas relevant for ITER and reactors. | ||
| Resource Requirements: | All gyrotrons
FW heating systems All available NB sources | ||
| Diagnostic Requirements: | Thomson scattering, Michelson interferometer, 40-channel ECE radiometer, ECEI, CER and MSE, fast magnetics, all fast ion diagnostics, all available fluctuation diagnostics. If available, oblique ECE. | ||
| Analysis Requirements: | EFIT, gaprofiles, ECESIM, ONETWO/autoonetwo, GENRAY, TORAY,and CQL3D | ||
| Other Requirements: | (references are here)
[1] E. de la Luna, et al., Rev. Sci. Instrum. 74, 1414 (2003) [2] G. Taylor, PPPL report 4202 (2006) [3] C. C. Petty et al. GA Report A25804 (2007) [4] V. Krivenski et al. 29th EPS Conference on Plasma Phys. and Contr. Fusion Montreux, 17-21 June 2002 ECA Vol. 26B, O-1.03 (2002) | ||
| Title | 2: Probing separation of thermal and particle transport in I-mode plasmas | ||
| Name: | White | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | A. White, D. Whyte, A. Hubbard, M. E. Austin, T. Rhodes, G. Wang, G. McKee, Z. Yan | ITPA Joint Experiment : | No |
| Description: | This experiment supports JRT 2013. <br> <br>This experiment will seek to measure in detail how the density and temperature fluctuations and the nT fluctuation cross-phase angle change in the edge/pedestal across the L-I transition. After the transition, during steady I-modes, density and ECH scans can be used to probe changes in the WCM and I-mode edge turbulence. This experiment is complementary to a second experiment proposed to focus on the changes in core turbulence and transport across the L-I transition. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Following experiment by D Whyte and successful access of I-modes on DIII-D (MP#2011-31-07), we would seek to use variations in density and heating with NBI, ECH and FW to help increase optical depth in edge and pedestal region of I-mode plasmas to allow for clear ECE measurements of fluctuations. Plasmas with well-defined L-I transitions and steady I-modes would be used to collect profiles of n-tilde (BES, DBS, reflectometry), Te-tilde (CECE, 32 channel ECE, ECEI), and the n-T phase angle (CECE/reflectometry) in the edge and pedestal region of L- and I-mode plasmas. Using BES, CECE and CECE-reflectometry, we can probe simultaneously how multi-field turbulence varies across the L-I transition and we can make the first measurements of the nT cross phase angle in the I-mode edge. While the fluctuation diagnostics for this exeriment will be focused in the edge to measure the turbulence, we can also set-up BES with the linear array of channels to monitor reductions in core density fluctuations across the transition. The multifield/multiscale fluctuation measurements, combined with modeling, will be used to explore how heat and particle transport channels are separated in I-mode plasmas. | ||
| Background: | The Weakly Coherent Mode (WCM) is an edge localized, electromagnetic turbulent mode associated with I-mode operation and is believed to play a role in regulating particle transport in I-mode: density profiles in I-mode are ??L-mode like?, temperature profiles are ??H-mode like?. It is unclear exactly how EM drift-wave type instabilities/turbulence can provide this separation of transport channels as the balance of temperature fluctuations, density fluctuations, and their phase angle with respect to electrostatic potential and magnetic field fluctuations is presently not known for I-mode.
At Alcator C Mod, we have shown that the Te-tilde/Te associated with the WCM in I-mode is 1-2%, which is up to an order of magnitude lower than the n-tilde/n fluctuation levels [White Nuclear Fusion (2011)]. In addition to the appearance of the WCM, there is also a reduction of broadband edge turbulence [Hubbard POP 2011] and core turbulence [White, C-Mod/PSFC Turbulence Group Meeting Presentation, 2011] across the L-I transition, all of which may be playing a role in determining the ITER-relevant transport properties on I-mode. While it is not possible to measure the n-T cross phase at C-Mod, this measurement is available at DIII-D and could greatly constrain models for how the WCM can regulate separately the particle and thermal transport in I-mode (e.g. in the absence of n-Phi or T-phi measurements, n-T can provide a constraint on the phases that matter for transport). | ||
| Resource Requirements: | I-mode at DIII-D.
NBI, ECH (maybe FW as well) | ||
| Diagnostic Requirements: | CECE, coupled CECE and reflectometer, BES, DBS, reflectometry, edge profile measurements, 32 channel radiometer, ECEI | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 3: Reduction of core ITG turbulence and destabilization of TEM turbulence in the core of I-mode plasmas | ||
| Name: | White | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | A. White, N. Howard, M. Greenwald, L. Schmitz, C. Holland, D. Whyte, A. Hubbard, M. E. Austin, T. Rhodes, , G. Wang, G. McKee, Z. Yan | ITPA Joint Experiment : | No |
| Description: | This experiment supports JRT 2013. <br> <br>While the edge localized Weakly Coherent Mode (WCM) associated with I-mode operation is believed to play a role in regulating particle transport and giving I-mode it's unique and ITER relevant transport properties, the changes in core turbulence and transport are also an important part of I-mode physics and may hold part of the answer to why I-mode and H-mode particle transport are so different, yet thermal transport similar. One possibility is that a change from ITG to TEM core turbulence across the L-I transition is playing a role in the separation of thermal and particle transport will be studied in detail in this dedicated experiment. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | First as piggy back on MP#2011-31-07 and then as a later dedicated run day, this experiment will seek to measure in how the density and temperature fluctuations and the nT fluctuation cross-phase angle change in the core across the L-I transition. Using BES, CECE and CECE-reflectometry, we can probe simultaneously how multi-field core turbulence varies across the L-I transition and we can make the first measurements of the nT cross phase angle in the I-mode core. Using NBI only we will seek to make an L-I transition where the core turbulence remains ITG dominant. Then, using a combination of NBI and ECH, we will make L-I transitions where the core turbulence changes to TEM dominant in I-mode. These plasmas should be fully diagnosed with MSE and CER to allow for nonlinear gyrokinetic simulations of the I-mode core turbulence. | ||
| Background: | At Alcator C Mod, there is evidence that I-mode core plasmas may be TEM dominant due to a reduction in ITG turbulence caused by a decrease in Ti gradient drive. This is interesting, because in NBI QH-mode plasmas (also a naturally ELM-free regime), core turbulence can remain ITG dominant (although reduced in amplitude overall) [Schmitz, PRL 2008], yet in low collisionality H-mode plasmas the role of TEM/ETG turbulence can become important [Schmitz, Nuclear Fusion 2012]. I-mode plasmas, and I-mode core turbulence dynamics, may resemble the low collisionality H-mode cases. It is unclear how robust this ITG-TEM boundary crossing is for the L-I transiiton, but it could be part of the reason why I-modes can exhibit such stark separation of particle and thermal transport, while the two remain tightly coupled in more standard H-modes (ITG dominant). | ||
| Resource Requirements: | I-mode at DIII-D. NBI, ECH. | ||
| Diagnostic Requirements: | CECE, coupled CECE and reflectometer, BES, edge profile measurements, core profiles, MSE, CER, in between shot ONETWO and TGLF | ||
| Analysis Requirements: | ONETWO, TRANSP, TGLF, GYRO, TGYRO | ||
| Other Requirements: | -- | ||
| Title | 4: ITER baseline with low torque and dominant e- heating | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | GL Jackson, EJ Doyle | ITPA Joint Experiment : | No |
| Description: | The goal of the experiment is to investigate the access and performance of the ITER baseline scenario at low torque and dominant electron heating. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Go to low field (~1.3 T) and use 3rd harmonic ECH so as to reach ITER betaN~1.8 with ECH only. Compare confinement and rotation profiles for ECH only case with equivalent balanced NBI case. Vary torque around 0 Nm (both co and counter), both with maximum and 0 ECH, to look at impact of rotation on confinement, and document similarities or differences in behavior dependent on whether in a dominant e-heated regime. Measure the intrinsic rotation with beam blips, and use NBI torque step to measure the intrinsic torque and check for consistency. | ||
| Background: | Previous attempts at investigating this important region of parameter space have been hampered by reduced confinement typically seen with ECH, coupled with a similar effect from counter NBI. The loss of confinement has been such that even with 3 MW of ECH, the beam power was still typically greater. Based on #147092, going to 1.3 T should allow an ECH only ITER baseline discharge. In particular, this reference used 4.5 MW of beams at T~1 Nm for betaN~2 with H89~1.9. This was 1.24 MA and 1.6 T. If we assume only a modest reduction in confinement from here, H89~1.8 for ECH 0 Nm, then at 1.3 T, 1 MA, we get 2.7 MW ECH, which is within our capability. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons, 210 beams | ||
| Diagnostic Requirements: | ECH hardened fluctuation diagnostics :) | ||
| Analysis Requirements: | TRANSP/onetwo modeling of current drive, heating profiles etc, plus intrinsic torque analysis | ||
| Other Requirements: | |||
| Title | 5: Inhibition of carbon co-deposits in hidden areas by ammonia injection | ||
| Name: | Tabares | Affiliation: | CIEMAT |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | Tony Leonhard, Dimitry Rudakov, Clement Wang | ITPA Joint Experiment : | No |
| Description: | : Ammonia (carbon radical scavenger) will be injected through the DiMES plug at a location withdrawn from the divertor floor by a few cms. The formation of carbon deposits will be investigated and compared to a reference case without ammonia injection. The possible diffusion of ammonia into the divertor plasma will be tested by OES. The effect of ammonia seeding on the optical characteristics of mirrors at DiMES will be also monitored. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The DiMES plug needs to be provided with a fueling line, a pumping system and a mass spectrometer. Ammonia will be injected during 8-10 plasma shots, then carbon deposition on the mirror samples will be measured upon DiMES plug extraction by several techniques. Simultaneous recording of N emission lines in the divertor plasma nearby will be made in order to quantify the degree of plasma perturbation by the injected species, if any. Reference experiment without ammonia injection may be required to quantify the level of carbon deposition under the selected plasma scenario if the existing reference is not close enough. | ||
| Background: | The scavenger technique for the inhibition of tritium trapping in co-deposits at remote locations has been tested in real divertors only in JET and AUG so far. Nitrogen was used then as a scavenger. Very recent results from linear plasma divertor simulators (PILOT PSI, PSI-2) indicate that ammonia is a better scavenger of film precursors (carbon radicals) than nitrogen, and can prevent the formation of deposits even when injected outside the divertor plasma, through ammonia-radical direct reactions. A 4 nm/min film deposition rate was inhibited by 1Pa.m3 flow of ammonia in PILOT PSI, downstream the plasma. However, the possible perturbation of the plasma by the injection of scavengers remains to be proved. The active, real time suppression of tritium retention in remote areas is a pre-requisite for the use of carbon facing components as an alternative to the present, tungsten-based design | ||
| Resource Requirements: | 2 ½ day experiments | ||
| Diagnostic Requirements: | Mass spectrometry, DiMES analysis, divertor spectroscopy, div probes | ||
| Analysis Requirements: | Plasma parameters near DiMES, radiated power | ||
| Other Requirements: | Gas inlet at DiMES plug, ammonia injection | ||
| Title | 6: RMP ELM Suppression q95 windows at low BT, Single and Double Row | ||
| Name: | deGrassie | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | E'quipe 3D | ITPA Joint Experiment : | No |
| Description: | * It is hypothesized that there are more q95 windows for RMP ELM suppression than we have found in DIII-D, because we have been limited in our ability to increase the I-coil current enough. This is especially compelling for the single row versus the double row because the single row does not need to match the field line pitch on the outboard side as does the double row, to strengthen a resonant perturbation. <br>* To exhaust our search we will go to low BT, 1.2 ?? 1.4 T, as low as can be and establish a robust standard RMP suppression condition (q95 ~ 3.5, n=3, even parity), and then search q95 (3.0 ?? 5.0) going up to I-coil current of 7KA (or more if possible and allowed). We want to be able to drive the coils until we achieve either ELM suppression, trigger the loss of H-mode, or encounter a LM disruption.<br>* In the absence of one of these outcomes then this region of q95 space remains an unexplored region of the q95 spectrum. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | *Establish a low BT standard suppression condition.
*Work out a matrix to scan I-coil amplitude (ramp) and/or q95, double and single row, so that the region 3 < q95 < 5 and I-coil current up to the maximum allowed is covered. | ||
| Background: | -- | ||
| Resource Requirements: | *Standard RMP ELM suppression experimental set-up.
*RELIABLE SPAS and C-supplies. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 7: Heat pulse propagation in a perturbed magnetic topology | ||
| Name: | Evans | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | 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), et al., | 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: | -- | ||
| Other Requirements: | Scheduling of this experiment needs to take into consideration the travel arrangements of international participants. | ||
| Title | 8: RMP ELM Suppression at q95 ~ 7+, odd parity n=3: A Paradigm Sieve | ||
| Name: | deGrassie | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | E'quipe 3D | ITPA Joint Experiment : | No |
| Description: | * Revisit n=3 odd-parity RMP ELM suppression at q95 just above 7 (128464).
( q95 = 7.15, betaN = 1.9, nebar = 2.4, neped = 1.9, Teped = 1.2, nustar_e = 0.1) * The reigning paradigm for RMP ELM suppression is that a resonant island opens near the top of the pedestal, limiting further expansion of the pedestal, and thereby suppressing the ELM instability. * In this high q95 condition the mode at the top of the pedestal needs to have m ~ 21-24. * The goals are a) Establish whether or not this is of the same genus of ELM suppression we get at q95 = 3.5, or is it a somewhat dramatic mitigation? b) If it is deemed to be the same, obtain the necessary data for modeling to establish whether or not one could expect an island with such a large m to pose a relevant blockage at the top of the pedestal. Or, are multiple island chains involved here? c) If it is deemed to be different, what is it? d) Test the sensitivity of the RMP ELM suppression here to the intrinsic error field, since we would expect any sensitivity to be greatly reduced from the q95 = 3.5 condition. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | * Reestablish this RMP ELM suppression (128464).
* Fully characterize the domain (q95 window) and amplitude (I-coil) for suppression. * Measure any residual ELM energy dumps, and heat load pulses. * Obtain high quality data set for modeling. * Test the effect of the intrinsic error field by modifying the correction, in such a manner as to be able to compare with q95 = 3.5 conditions. (The point is to know what modes count.) | ||
| Background: | |||
| Resource Requirements: | Standard RMP ELM suppression fare.
RELIABLE SPAs and C-Supplies | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 9: Exploring physics of RMP ELM suppression at low torque | ||
| Name: | Moyer | Affiliation: | UCSD |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Explore the physics of plasma response and RMP ELM control in ITER-relevant low torque/low rotation ELMing H-modes using n = 3 I-coil RMP at low coil current (possibly only a few 100 amps). Demonstration of RMP ELM suppression at low torque would radically 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. If rotational screening models are correct (one or two fluid), we should be able to access an ELM suppressed regime in ELMIng H-modes with low net toroidal rotation (balanced or near-balanced NBI) using very little I-coil current, possibly only a few hundred amps.Previous attempts have always approached RMP ELM suppression at low torque/rotation by starting first with a highly co-rotating ELMing H-mode. Once even a small amount of counter-NBI is applied, the toroidal rotation decayed until the plasma bifurcates to a locked solution, leading to plasma disruption. I propose instead that we start with an ISS low torque, ELMing H-mode, and apply a modest n = 3 even parity I-coil RMP of as low as 100 Amps to access ELM suppression without locking the plasma. With low toroidal rotation, the screening should be significantly lower and far less RMP field should be needed. Even if this approach doesn't suppress ELMs, we should learn a good deal about rotational dependence of plasma response for low net toroidal rotation. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Establish a low torque/near balanced NBI ELMing H-mode in ISS shape. This will establish a target ELMing H-mode plasma with a a zero-crossing in the electron perpendicular velocity but deeper in the core than desired for ELM suppression based on emerging understanding. Apply an n =3 even parity RMP; this should initially brake the edge, leading to RMP penetration and formation of stochastic field lines. the resulting j_r x B torque should spin up the plasma toroidal rotation (as typically seen in RMP ELM-suppressed discharges) and shift the zero-crossing out to the top of the pedestal. Scan the I-coil current from 100 to 1000 A (or until locking is reached). Optimize error field correction and use ECCD feedback on NTMs to widen operating window away from locked modes. Document plasma response, transport, and stability using traveling (rotating) I-coil field from Techrons if needed current is low enough. | ||
| Background: | Previous attempts have always approached RMP ELM suppression at low torque/rotation by starting first with a highly co-rotating ELMing H-mode. Once even a small amount of counter-NBI is applied, the rotation drops until it bifurcates to a locked state; in essence, we keep trying to run these discharges in a controlled manner through the "prohibited" part of the bifurcation curve. | ||
| Resource Requirements: | tokamak in clear state; both 210 NB sources; 30 and 330 L and R sources needed for profiles. 150L for BES fluctuation measurements. I-coil with reliable low current operation. Techrons if we can establish interesting results at sufficiently low current. | ||
| Diagnostic Requirements: | CER across plasma; Full fluctuation and pedestal profile diagnostics. | ||
| Analysis Requirements: | Significant profile, kinetic EFIT, Er, and fluctuation analysis. ELITE P-B stability, M3D-C1, XGC0, and other modeling. | ||
| Other Requirements: | -- | ||
| Title | 10: Compatibility of NRMF-assisted QH-mode with low torque startup | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | KH Burrell, AM Garofalo | ITPA Joint Experiment : | No |
| Description: | The primary aim of the experiment is to investigate whether NRMF driven torques can be used to access QH-mode without an initial large counter NBI phase. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Attempt to produce a clean transition from L- to QH-mode using NRMF applied before the L-H transition and with low (preferably slightly co) NBI torque, and measure the level of NRMF torque achieved. Compare the performance and access of QH-modes formed with significant ctr-NBI. If the NRMF torque is insufficient, determine how much ctr-NBI is required to obtain robust QH-mode operation. If QH-mode can be accessed with NRMF at low torque, go to lower toroidal field and attempt the same with ECH only. This will be helpful in clarifying the role of fast ions in generating the NRMF 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, I-coil (n=3), 210 beams, 6 gyrotorons desirable | ||
| Diagnostic Requirements: | Standard profile diagnostics | ||
| Analysis Requirements: | Analysis/modeling of NRMF torque (IPEC and experimental) | ||
| Other Requirements: | |||
| Title | 11: Investigation of reduced central impurity accumulation during ELM pacing by pellet injection | ||
| Name: | Brooks | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, N. Commaux | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to confirm the reduction of central impurity accumulation observed during ELM pacing and to understand the mechanism producing it. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In the same cryopumped discharges previously employed in ELM pacing experiments, inject trace levels of argon from the outer wall. The combination of continuous injection and continuous cryopumping of this recycling impurity will produce a constant edge source. Observation of the beam-modulated, argon charge exchange lines from H-like, He-like and Li-like argon with the XUV SPRED instrument provides a means to monitor the core concentration of argon. Compare discharges with and without ELM pacing. | ||
| Background: | A reduction in central metal concentration has been reported (L.R. Baylorâ??s IAEA abstract) in discharges in which shallow penetrating pellets have been employed to generate rapidly paced ELMs. The roles of impurity source, impurity screening and ELM impurity purging in producing this effect is difficult to untangle in the case of the intrinsic metallic impurities. Previous use of argon as a probe impurity in puffâ??nâ??pump experiments on DIII-D has demonstrated the value of such gas injection in understanding impurity accumulation. | ||
| Resource Requirements: | beams, LFS rapid pellet injector | ||
| Diagnostic Requirements: | modulation of 30 beam, X-ray camera and fast camera imaging also useful | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 12: 3-D Fields and ECH Density Pumpout | ||
| Name: | deGrassie | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | Andrea Garofalo | ITPA Joint Experiment : | No |
| Description: | The purpose of the experiment is to see if applied 3-D fields, or error fields in non-specific 3-D field experiments, play a role in the density pumpout when ECH is applied to a NBI target H-mode discharge. ECH density pumpout will be parameterized as a function of target discharge toroidal rotation at varying levels of applied 3-D field, or error field if there is an effect for small levels of perturbation. | 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 | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 13: Measure Intrinsic Rotation Size scaling in DIII-D alone -II | ||
| Name: | deGrassie | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | Wayne Solomon, Keith Burrell, John Rice (MIT) | ITPA Joint Experiment : | No |
| Description: | *Size Scaling NEEDED to confidently extrapolate to ITER.
*Continue experiment started with 2009-51-01 *There, steady (enough) conditions were not obtained in the small and large extremes in major radius, presumably because of lack of operational time with new shapes. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | *Measure the Rice scaling slope, Rs, for three similar shapes at different size. Here, V = Rs*W/Ip.
*Focus on ECH H-modes + NBI blips to get unpolluted intrinsic rotation. *The sizes listed below have a variation in R^2 of 1.39, which we should be able to measure in the slope. *An R^2 scaling is indicated by direct comparison of the slopes between C-Mod and DIII-D, and fits with one dimensionless fit to the international database, that of MA ~ BetaN, where MA is the so-called 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 | 14: Scan of RMP coil current below suppression threshold | ||
| Name: | maingi | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans | ITPA Joint Experiment : | No |
| Description: | The idea is to do a scan of RMP coil current from a reference ISS ELMy discharge to just above the suppression threshold, and obtain 4-5 I-coil levels in between, to see if the RMP affects the pedestal continuously despite the fact that ELM suppression is a threshold. A reference discharge #126443 and scan are described in Evans NF 2008 and Fenstermacher PoP 2008. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Scan of RMP I-coil current: vary the RMP coil current from 0 kAt to 4.8 kAt, starting from #126443 as a reference. There are existing data at 0, 4.0, and 4.8 kAt, the last one being above the min. threshold needed for ELM suppression. The idea is to get 4-5 data points between the reference and ELM-suppressed states to see if the pedestal and ELM characteristics change continuously with current, as indicated by the existing 3 point scan. | ||
| Background: | Analysis of the I-coil current scan has shown that the magnitude of the density and pressure gradient drop scales continuously with the current, but there is very little data to determine if the dependence is continuous just below the ELM threshold. | ||
| Resource Requirements: | About 10 good discharges, starting with an ISS ELMy H-mode, and determining the threshold current needed for suppression. Take 2 discharges per current level below the threshold, if possible. | ||
| Diagnostic Requirements: | High resolution edge Thomson, CER | ||
| Analysis Requirements: | Profile analysis with Osborne's tools; stability analysis with ELITE. | ||
| Other Requirements: | -- | ||
| Title | 15: Direct Measurement of E_rad Corrugation at Rational Surfaces | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | M. E. Austin | ITPA Joint Experiment : | No |
| Description: | Use the combination of co and counter MSE views to directly measure the corrugation in the radial electric field at rational q surfaces that is responsible for transport barriers. The target plasmas are balanced-NBI L-modes with early heating so that q>2. The analysis will focus on MSE channels that view the radius where a rational surface, such as the q=2 surface, first enters the plasma. In the absence of E_rad effects, the co and counter viewing MSE channels will measure the same magnetic field pitch angle. Thus, a separation between the co/counter MSE signals at the time a rational q surface enters the plasma is a direct measurement of the E_rad corrugation effect. There is the option to add the off-axis beam to slow the current profile evolution, which should make the corrugation easier to observe. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Establish L-mode plasma with early beam heating to slow the evolution of the current profile. (2) Use 30LT and 210RT beams without modulation to collect continuous MSE and CER data. (3) May need to move the plasma location around to make sure the MSE channels are looking exactly at the location where the rational q surface (especially q=2) first enters the plasma. (4) Try adding the off-axis beam (at maximum downward angle) to slow the evolution of q_min. | ||
| Background: | Previous experiments by Max Austin found corregations in the electron temperature profile when a rational q surface entered the L-mode plasma. These corrugations were observed for both co-NBI and balanced-NBI (although only the co-NBI cases resulted in long lasting transport barriers). The GYRO turbulence simulation code predicted the existence of these corrugations by means of a equilibrium ExB shear flow driven by the zonal flows. This experimental proposal will look for direct evidence of this ExB shear flow by means of the E_rad sensitivity of the MSE diagnostic. | ||
| Resource Requirements: | NBI: 30LT and 210RT essential. 150 beamline tilted downwards is desired (but need to consider impact on BES). | ||
| Diagnostic Requirements: | MSE is critical. Fluctuation diagnostics are desirable. | ||
| Analysis Requirements: | Need GYRO simulations. | ||
| Other Requirements: | |||
| Title | 16: Measurement of Inductive Poloidal Current | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the poloidal current density profile induced by ramping the toroidal field coil. Compare with the poloidal current expected from the parallel Ohm's law to determine if the perpendicular conductivity is large enough to give a significant contribution. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Study H-mode plasmas with beta_pol near unity so that the "natural" poloidal current is negligible. Compare discharges with positive and negative ramps of the toroidal field to cases with no BT ramping. Keep the plasma current, density, and temperature constant during these ramps. Study two cases, a low electron temperature plasma with NBI heating only, and a high electron temperature plasma using ECH in addition to the diagnostic beams. | ||
| Background: | The magnitude of the perpendicular conductivity has not been measured to my knowledge in tokamaks. In this experiment, ramping the toroidal field will induce a poloidal electric field that can be exactly computed using Faraday's law. Multiplying this E_pol by the parallel conductivity gives the parallel contribution to Ohm's law, while multiplying E_pol by the perpendicular conductivity gives the perpendicular Ohmic current density. Using the MSE data (although not necessarily equilibrium reconstruction), both the poloidal current density and the parallel current density can be measured. By comparing plasmas with and without a BT ramp, it will be possible to determine if the measured change in the parallel current density is enough to explain the total measured poloidal current (i.e., the perpendicular conductivity is negligible). | ||
| Resource Requirements: | NBI: Co and counter MSE beams, and at least 2 other sources.
EC: Minimum 6 gyrotrons. | ||
| Diagnostic Requirements: | MSE is critical. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 17: Modulation of Bootstrap Current | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | 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 | 18: RMP ELM Suppression at the NTV Offset Rotation | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Establish RMP ELM suppression in a plasma with mild counter rotation. Allow the rotation to "lock" to the offset rotation given by NTV, using additional NTV torque from the n=3 C-coil. Evaluate the confinement and stability properties of this discharge. Compare even and odd parity to vary the relative contributions of resonance and nonresonant effects. The NTV offset rotation frequency should be made as large as possible by operating at low Ip (i.e. low Bp) and low density (i.e. high Grad_Ti). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Use reverse Ip configuration so that most of the neutral beams are injecting in the counter direction. (2) Establish ELMy H-mode plasmas with Ip=1.0 MA and q_95=3.6. Lower Ip may be used if the beam ion confinement is good enough. (3) Start with even parity of I-coil. Apply RMP to suppress ELMs, with the n=3 C-coil added for additional (counter) NTV torque. Allow the density to pump out to a low level to obtain a high gradient in the ion temperature. (4) Determine the sensitivity of the toroidal rotation rate during RMP application with the amount of counter-torque injection. If the effect of nonresonant braking is large, then the toroidal rotation should be a stronger function of the NTV offset velocity than of the NBI torque. (5) Compare even and odd parity of I-coil, ideally in same discharge if SPAs are used. | ||
| Background: | For co-rotation discharges, applying the RMP to suppress ELMs results in a reduction of the toroidal rotation. This reduces the confinement time, and also can lead to locking of the plasma if the resonant braking effect becomes large. It is predicted that the nonresonance braking effects of an RMP coil on ITER may dominate over the co-torque injection from neutral beams, in which case the toroidal rotation on ITER should "lock" to the NTM offset value. This experiment proposes to study the consequences of this effect by starting with a counter rotation frequency close to the NTM offset value. | ||
| Resource Requirements: | Reverse plasma current configuration.
RMP I-coil configuration. Use SPAs so that even and odd parity can be compared in same discharge. C-coil in n=3 configuration. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 19: Effect of Islands on ECCD | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R. Prater | ITPA Joint Experiment : | No |
| Description: | ECCD is an important tool to control MHD, such as tearing modes. While DIII-D has done detailed studies of ECCD, these have been for an axisymmetric plasma. The helical perturbations from tearing modes may significantly change the ECCD profile, which in term could affect its application to MHD control. This experiment will examine two facets of the effect of islands on ECCD. First, the flux-surface-average parallel current density will be compared for deposition at the island O-point or X-point. Second, the ECCD profile will be decomposed into separate toroidal and helical components. This second case requires a slowly rotating island, which can be achieved using entrainment with the I-coil. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Part I: Effect of islands on flux-surface-average parallel EC current density. (1) Target plasma is to be taken from successful modulated ECCD experiment to stabilize the 2/1 NTM. Probably a mixture of co/counter NBI will be used to slow the island rotation frequency to <5 kHz. (2) During the ECCD measurement phase, the 30LT and 210RT beams should be on continuously for MSE data acquisition, (3) With EC deposited at the island O-point, compare co/radial/counter ECCD injection. For the co-ECCD case, the power should be limited so that the island is NOT stabilized. (4) Repeat last step for ECCD deposition at the island X-point. (5) Repeat last step with continuous ECCD (i.e. not modulated).
Part II: Helical current from ECCD (1) The target plasma should be taken from a successful entrainment experiment where the I-coil is used to force a 2/1 tearing mode to rotate at a frequency <1 kHz. (2) Apply co/counter/radial ECCD at the q=2 location continuously (i.e. not modulated). (3) Compare co/radial/counter ECCD injection. (4) Compare modulated ECCD at island O-point or X-point for co/radial/counter injection. | ||
| Background: | Experiments on DIII-D over the last 10 years have made detailed comparisons between ECCD theory and experiments on the local level. However, these experiments specifically avoided MHD such as sawteeth and tearing modes. Thus, the ECCD studies were done in a axisymmetric plasma configuration. The highly localized region of ECCD led to the development of methods for direct analysis of the MSE signals without equilibrium reconstruction. This direct analysis method was able to determine the ECCD profile with spatial resolution limited only by the MSE diagnostic itself. Later, this methodology was extended to include the helical perturbations from tearing modes. The helically perturbed current for a m/n=2/1 "quasi-stationary" mode was successfully determined using MSE data and was reported at the 2006 EPS meeting. | ||
| Resource Requirements: | NBI: Both co and counter beams are required.
EC: 6 gyrotrons are required. I-coil: Entrainment of rotating 2/1 mode required for Part II of this experiment. | ||
| Diagnostic Requirements: | MSE is critical, with highest time resolution possible. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 20: New Optimal Plasma Shape for AT Scenario? | ||
| Name: | Petty | 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 | 21: ECE Imaging of ELM-NTM coupling | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The ECE imaging camera being developed by U.C. Davis and collaborators will be used to measure the changes in the electron temperature profile during ELM events in hybrid plasmas. The 2D images will help us understand the physics behind this coupling, and perhaps improve our understanding of magnetic flux pumping in hybrids that maintains the safety factor minimum slightly above unity. Both n=2 and n=3 tearing modes will be studied. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment can piggyback on another hybrid experiment as long as the toroidal magnetic field is high enough (BT~2 T). The target hybrid plasmas should have type-I ELMs with ~40 Hz frequency and q95>4 so that sawteeth are suppressed. We want to image both the usual hybrid case with a 3/2 NTM as well as hybrids with a dominant n=3 NTM (such as 4/3 or 5/3). We will likely not want to use ECCD to stabilize the 3/2 NTM because the required filtering needed to remove the 110 GHz radiation will compromise the ECE images (this needs further study, however). Conditional averaging over many ELMs will be used to improve the SNR of the ECE imaging diagnostic. | ||
| Background: | Using the ECE radiometer array, a modification in the electron temperature profile was observed previously during ELM events near the rational surfaces for 3/2 and 5/3 NTMs (but interestingly, not for 4/3 NTMs). This demonstrated a clear coupling between ELMs and NTMs, but the physical mechanism is not clearly understood. | ||
| Resource Requirements: | NBI: 6 co sources are requested for long pulse lengths. | ||
| Diagnostic Requirements: | ECE imaging and MSE are critical. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 22: Test of Turbulence Spreading Using Turbulence Propagation | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The question of turbulence spreading, that is, whether turbulence is or is not a strictly local phenomenon, can be precisely tested by modulating the turbulence (and plasma profile) at a fixed location and then monitoring the propagation of the turbulence (and plasma profiles) away from this region. If the turbulence propagation speed is much faster than the temperature or density propagation speed, then this can be attributed to turbulence spreading. For this purpose it does not matter much how the turbulence is modulated; it can be a simple amplitude modulation or something more sophisticated such as a modulation of the radial correlation length. The most likely source of modulation is ECH, either as a monopolar change in the electron temperature profile or as a "swing" experiment where the ECH deposition is alternated between two (closely spaced) location. The turbulence diagnostic must be capable of covering a large radial range, so the 32 channel linear array of the BES diagnostic is ideally suited for this experiment. An 8 channel DBS diagnostic would also be useful to monitor the propagation of intermediate k turbulence. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To minimize MHD, this experiment will use an L-mode plasma with 1-2 sources of continuous NBI for diagnostic purposes (BES, CER, MSE) and 6 gyrotrons for turbulence modulation. If the beam power needs to be limited to 1 source, repeat shots can be taken to switch between beams. The ECH modulation rate should be relatively high (~100 Hz) to allow an accurate measurement of the propagation speed. Actually it is preferable to study several different modulation rates, so repeat shots will be taken to cover the range 25-200 Hz. | ||
| Background: | While the ECH "swing" experiment led by Jim DeBoo has similarities to this proposal, in that case the ECH modulation was too slow to obtain the phase delay information that is crucial to this proposal. Also the radial spread of the tubulence modulation was limited in DeBoo's case, perhaps a consequence of the "swing" arrangement. Therefore, a monopolar modulation of the ECH at relatively high frequency is preferred for this proposal. | ||
| Resource Requirements: | Beams: 30LT, 330LT, 150LT
ECH: Six gyrotrons | ||
| Diagnostic Requirements: | BES 32 channel linear array
DBS 8 channel array | ||
| Analysis Requirements: | GYRO simulations will be done after the experiment. | ||
| Other Requirements: | |||
| Title | 23: Bootstrap Current Change During RMP ELM Suppression | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | D.M. Thomas, H. Stoschus | ITPA Joint Experiment : | No |
| Description: | Using the Li Beam and MSE diagnostics, directly measure the change in the H-mode pedestal bootstrap current that is caused by the change in the pedestal gradients when the RMP is turned on. The most accurate measurement would be a RMP/non-RMP comparison. It would be good to also look at cases where q_95 is outside the resonance window to see if there are any differences. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This should be a "piggyback" or background experiment, in that the data should be obtainable in the course of the RMP ELM-suppression experiments in 2011. The only criteria is that the plasma boundary be located in a favorable place for the edge MSE and Li Beam. In addition, long analysis windows are desired to reduce the random errors in the pitch angle measurements. The basic dataset would contain four discharges: (1) RMP ELM-suppressed case inside the q_95 resonance window, (2) repeat without RMP, (3) RMP case outside the q_95 resonance window (ELMs not expected to be suppressed), and (4) repeat without RMP. | ||
| Background: | Our physics picture of RMP ELM-suppression is that the RMP reduces the H-mode pedestal gradients, thus reducing the pedestal bootstrap current. This makes the plasma stable to ballooning-peeling modes according to ELITE calculations. We would like to test this experimentally by using the Li Beam and edge MSE diagnostics to directly measure the changes in the magnetic field pitch angles. The edge current density is proportional to the channel-to-channel derivative of the pitch angles. The pitch angle data can either be used in a equilibrium reconstruction, or analyzed directly. | ||
| Resource Requirements: | I-coil in RMP configuration. | ||
| Diagnostic Requirements: | Li Beam and MSE are critical. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 24: Comparision DIII-D and AUG high collisionality ELM response to 3D magnetic perturbations | ||
| Name: | Evans | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | W. Suttrop, et al., | 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: | This proposal and proposal #187 by W. Suttrop are intended to be combined and as noted in #187 we would prefer two half day experiments in a single week. Scheduling of this experiment needs to take into consideration the possibility of international participants with specific travel constraints. | ||
| Title | 25: Removed by Author's Request | ||
| Name: | Leblanc | Affiliation: | GA |
| Research Area: | NA | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | Removed by Author's Request | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Removed by Author's Request | ||
| Background: | Removed by Author's Request | ||
| Resource Requirements: | Removed by Author's Request | ||
| Diagnostic Requirements: | Removed by Author's Request | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 26: MGI Mitigation of REs in Vertically Stable Discharges | ||
| Name: | Wesley | Affiliation: | GA |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Evaluate Massive Gas Injection (MGI) for 'early' mitigation of vertically stable RE plateaus. Compare effect of various low-Z and high-Z gases under control conditions where onset of vertical instability does not develop. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use 2.7-mm argon pellet injection or equivalent to make ~300 kA initial RE currents with 'Type B' vertically stable control (eg 145840); assess mitigation effect of massive quantities (~1000 Torr-l) of 'early' (within 20-30 ms of pellet) gas injection for D2, He, Ne, Ar, Xe. | ||
| Background: | 2011 MGI RE mitigation experiments with 'early' gas timing (gas at ~20 ms after pellet) demonstrated significant incremental mitigation (increase in RE current dissipation rate), but were limited by onset of vertical instability within 30 ms of gas arrival. Type A high-elongation control scheme was used. Repeat here with 'Type B' low-elongation scheme to improve degree and rate of mitigation, do a more systematic comparison among candidate gases. | ||
| Resource Requirements: | 2.7-mm Ar pellet injector or equivalent RE generation capability; standard 1.3:1 low-elongation EC heated 'target' plasma (4+ gyros); 'Type B' low-elongation control scheme (eg 145840) | ||
| Diagnostic Requirements: | Standard target diagnostics, standard RE diagnostics including HXR array and synch and impurity light fast camera and spectro. | ||
| Analysis Requirements: | Dynamic model for gas effect | ||
| Other Requirements: | |||
| Title | 27: Gas mitigation in Stationary RE Plateau Discharges | ||
| Name: | Wesley | Affiliation: | GA |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Inject modest quantities of low-Z and high-Z gases into ohmically sustained 'stationary' RE plateau discharges | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Prepare 'stationary' OH sustained RE plateau (~300 k, 4 V loop voltage); inject modest (~ 100 Torr-l) of low-Z and high-Z gases at t = 300 ms; evaluate new stationary equilibrium at same I_RE | ||
| Background: | Low-quantity 100 Torr-l) He injection in #145521 resulted in a new, sustained stationary equilibrium discharge with reduced density-normalized loop voltage and dissipation (relative to no-injection sustained baseline examples). Repeat with various modest quantities of D2, He, Ne, Ar, Xe will allow effect of mGI (modest gas injection) on stationary phase dissipation of be systematically studied with high-reproducible 'mature' stationary plateau 'targets'. | ||
| Resource Requirements: | 2.7-mm Ar pellet (or equivalent RE initiation means), 'type B' vertically-stable RE discharge control (eg 145840), 4+ gyrotrons for pre-pellet target prep | ||
| Diagnostic Requirements: | 'standard' RE diagnostics, including HXR array + fast camera for synch light and/or impurity light imaging; possible post-mGI plasma composition measurements | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 28: Removed by Author's Request | ||
| Name: | Wesley | Affiliation: | GA |
| Research Area: | NA | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Removed by Author's Request | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Removed by Author's Request | ||
| Background: | |||
| Resource Requirements: | Removed by Author's Request | ||
| Diagnostic Requirements: | Removed by Author's Request | ||
| Analysis Requirements: | Removed by Author's Request | ||
| Other Requirements: | |||
| Title | 29: Removed by Author's Request | ||
| Name: | Wesley | Affiliation: | GA |
| Research Area: | NA | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Removed by Author's Request | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Removed by Author's Request | ||
| Background: | |||
| Resource Requirements: | Removed by Author's Request | ||
| Diagnostic Requirements: | Removed by Author's Request | ||
| Analysis Requirements: | Removed by Author's Request | ||
| Other Requirements: | |||
| Title | 30: Real-time TM/Disruption Avoidance Demo | ||
| Name: | Wesley | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Use real-time steerable EC + PCS 'autonomous' detection and control to demonstrate avoidance, repair and/or frequency reduction of pre-disruptive ('dud') or disruptive plasma outcomes. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Couple empirical or ITER-scalable dud or disruption predictors with PCS-enabled real-time steerable ECH or ECCD to effect 'autonomous' dud/disruption avoidance. After development, test/demonstrate during TM or similar 'disruption-prone' operations; quantify benefit. | ||
| Background: | |||
| Resource Requirements: | 2-4 'on-demand' gyrotrons, real-time mirror steering under PCS control, dedicated test shots for development; symbiotic test in 'ITER-like' baseline scenario(s). Need reactive PCS software qualified to command EC response | ||
| Diagnostic Requirements: | mode location and pending 'dud' and/or disruption detection; standard diagnostics to evaluate mode response and mitigation | ||
| Analysis Requirements: | Validation of predictors and/or EC response; preparation and qualification of PCS algorithms | ||
| Other Requirements: | |||
| Title | 31: Rupture Disk Injection for Improved Mass Delivery | ||
| Name: | Wesley | Affiliation: | GA |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Evaluate ability of a close-coupled high-flow Rupture Disk Injector (RDI) to provide improved low-Z and high-Z mass delivery to 'ITER-like' DIII-D target plasmas | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Conduct systematic studies with a close-coupled, high-flow RDI system. Use 'standard' ITER-like beam-heated target plasma to allow comparison with past or new MGI and SPI results for same species. | ||
| Background: | Past DIII-D experiments with the single-valve high-intensity (long tube) and MEDUSA-I and MEDUSA-II baffled MGI valve arrays demonstrate improved gas assimilation and in-plasma gas/mass uptake as peak flow and/or direct coupling of the injector stream improve. ASDEX-U experiments with in-vessel valve(s) demonstrate improved neon deliver realtive to similar ex-vessel valves. A close-coupled RDI system in DIII-D would allow study of characteristics and assimilation of 'sharp' gas jet impingement in ITER-like elongated target plasmas with 1+ MJ thermal energy. | ||
| Resource Requirements: | Multi-RDI array, ideally with electric 'synchronous' initiation. Midplane location with exit < ~10 cm to plasma surface desirable. Alternate with asynchronous 'fill pressure rise' triggering may be workable. Need 10+ shot capability for initial campaign. | ||
| Diagnostic Requirements: | Tri-color CO2 and/or ability to follow very rapid/high dn/dt; camera and spectro imaging of jet interaction region. Multi-azimuth bolo, spectro and SXR useful. | ||
| Analysis Requirements: | |||
| Other Requirements: | Could also be used to investigate 'Putvinski' scheme for RE groweth mitigation. Requires electric triggering plus coordination with reliable early RE generation (very challenging). | ||
| Title | 32: Understand mechanisms by which ECH influences NTM stablility | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | 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: | 1 day expt, 6 gyrotorons with NTM control, 210 beams | ||
| Diagnostic Requirements: | MSE, Thomson | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 33: Error field correction at low rotation | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | Stability and 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, for example, using a single I-coil nearest to that error source (IL30). 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 | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 34: Investigation of Sheath Power Transmission at the DIII-D Divertor | ||
| Name: | Donovan | Affiliation: | U of Tennessee, Knoxville |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | Dean Buchenauer, Jon Watkins, Dmitry Rudakov, Charlie Lasnier, Josh Whaley, Peter Stangeby | ITPA Joint Experiment : | No |
| Description: | The sheath is the ultimate regulator for the interaction of plasmas and materials, and of particular importance is the rate at which power flows through the sheath to the surface, since this sets the thermal loading on PFCs. However, experimental observations often find that the standard sheath theory to predict heat loading is inadequate, often by factors of 2-10, which is completely unacceptable for predictive capabilities of heat loading. Measurements of electron temperature, plasma density, and heat flux are required under a variety of conditions to better understand their impact on the conflicting sheath power transmission factor values. Previously collected evidence from embedded thermocouples suggest that a dedicated experiment to compare infrared measurements with modeling of near surface thermocouples are warranted. DiMES serves as a unique platform to perform these critical experiments and play a role in resolving the sheath power transmission conundrum. Simultaneous data collection on the DiMES probe, the Langmuir probe divertor array, the IR camera, and the embedded thermocouples during an L-mode shot is requested. Additionally, a comparison between Ohmic, ECH, and NBI heating would be very informative as to determining the potential cause of the unusual sheath power transmission factor values. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In order to investigate the flux of ions reaching the probes, this study will compare the long-used domed design with a similar domed design elevated above the magnetic sheath of the divertor tiles and a flat surface probe whose normal is aligned along the local magnetic field (surface normal probe). As these probes have surfaces normal to the divertor heat flux, they will require that the outer strike point be swept across the DIMES location to retain thermal integrity. Possible studies have included
1. determination of equilibration time during strike point sweep across DIMES probe 2. power scaling 3. variation with magnetic field angle 4. density and/or divertor neutral pressure scaling 5. toroidal field scaling For this proposal, we expect that we can reasonable cover studies 1 and 2 using the second half of a shot. If the run goes well, we expect that study 4 could also be accommodated by the main experiment. The measurements would be made between 3.5 to 5.0 seconds of flattop current, with operation well within core plasma stability boundaries to minimize damage to the probe tips and L-mode plasmas to avoid transient effects at the divertor (ELMs). Power levels would be limited to 3-4 sources based on thermal calculations and previous measurements of divertor heat and plasma parameters. We will require that the magnetic configuration be optimized to provide a radial x-point sweep for which the outer strike point moves inward from R=151.5 cm to R=143.0 cm (approximately) with minimal change in the x-point height (nominally 12-15 cm above the divertor floor). A reference shot (80136) was used to provide a similar sweep for a DIMES exposure on October 21, 1993, however some development will likely be required for the present divertor geometry. The rate of the sweep would vary during study 1 and be fixed during study 2. | ||
| Background: | Since its installation in 1992, the Divertor Materials Evaluation System (DiMES) on the DIII-D tokamak has provided a unique platform for the study of plasma surface interactions. Early experiments performed many first-of-kind observations at a divertor surface: quantification of the net erosion rate of carbon, demonstration of reduced erosion during plasma detachment, elucidation of the role of chemical sputtering, quantification of deuterium retention in carbon and metallic coatings, and identification of a critical issue of MHD interaction between liquid lithium and a divertor plasma. These passive measurements have provided data for the benchmarking of PSI codes and helped to improve the operation of DIII-D.
Less well known perhaps is that DiMES can also be a platform for the development of plasma diagnostics. Early design issues have now been improved to provide 12 electrical feedthroughs (+ one pair for a thermocouple) for active measurements (microsecond time response). Sandia California designed the first active DiMES head and has tested Langmuir probes and H-microsensors using the platform. With the improved cabling, we propose to utilize the DiMES platform to address the sheath power trasmission conundrum. Experiments from DIII-D and other tokamaks have demonstrated that the power transmission through the sheath, as determined by divertor Langmuir probes and infrared camera images, is still a mystery. Profiles of the sheath power transmission factor (ratio of heat flux to product of ion saturation current and electron temperature) show this ratio varying across the outer divertor strike point, to values much less than the nominally expected value of ~7 near the peak heat and ion flux (more recent data shows a similar trend). Since this determination relies on the interpretation of probe characteristics and the thermal response of the tiles to the heat flux, the use of DiMES to better understand the plasma measurements and heat load would be beneficial. This DiMES head was refurbished in early 2010 and inserted into the DIII-D divertor for discharges in March and April. The probe design performed well at the high parallel heat loads (7 divertor strike point sweeps over the probe with 25 MW/m^2). These ohmically heated discharges had a modest amount of ECH power, but resulted in x-point sweeps that did not provide for strike points to reach the radius of DiMES (OSP was near 130 cm while the center of DiMES is at 148.5 cm). The 3-probe DiMES head was inserted for a total of 12 shots during the 2011 campaign. Data was collected that has offered insight regarding the possible impact of the effective collection area of the Langmuir probe array on the calculation of the Sheath Power Transmission Factor by comparing the current collected on the planar and domed probes on DiMES. However, many of these shots still contained a significant amount of ELMs and arcs. A dedicated sweep over the DiMES probe during an L-mode shot with simultaneous data capture from the Langmuir probe shelf array and the IR camera would be preferred for optimal data collection conditions. | ||
| Resource Requirements: | The hardware for the DiMES probe is available, along with instrumentation provided by the divertor Langmuir probe array. The experiment would require an IRTV at the 165 degree toroidal location (we are currently exploring alternatives to the TEXTOR camera), in additional to the present divertor camera view. Run time of approximately ½ day would also be required, including NBI availability (no cryo-pumping needed or desired). | ||
| Diagnostic Requirements: | A desirable element of the experiment would be to use longer focal length optics for the IR camera at 165 degrees (we also would require operation in line scan mode to improve time resolution during the x-point sweeps). This would provide better resolution of the heat flux to the domed probes.
Divertor Langmuir probes Probes mounted on the DiMES system, using instrumentation from the divertor probe racks 165 degree lower divertor camera with long focal length optics and line scan mode Other lower divertor camera in customary video mode Divertor spectroscopy (Boron / Nitrogen) Magnetics for EFIT determination of field angles Zeff C02 interferometer Thomson scattering (burst mode for main experiment, but regular mode for piggyback) Fast filterscope channels viewing the lower divertor Other useful diagnostics Tile current array Fast tile thermocouple array Bolometers Edge CER for ion 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. More detailed evaluation of other desirable diagnostics would be welcome to begin to understand the relationship between the sheath power transmission factor and collisional effects in the divertor plasma. | ||
| Other Requirements: | |||
| Title | 35: Comparison of ELM mitigation on DIII-D and MAST just above the L-H transition power | ||
| Name: | Kirk | Affiliation: | CCFE |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | Todd Evans | ITPA Joint Experiment : | Yes |
| Description: | Explore the physics of ELM control with input power near to the L-H transition power.
Compare the effect of the RMP on the L-H power threshold, the ELM-free period and the first ELM for various shapes. Document the changes in plasma parameters, especially the radial electric field, turbulence and flows. In particular try to gain an understanding of why ELM suppression has not been achieved in a DN discharge. These joint MAST/DIII-D experiments would have direct relevance to ITPA PEP-32 and PEP-23 |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start with the usual DIII-D DN shape with input power just sufficient to reach a type I ELMing regime. Scan GAPIN from shot-to-shot at different I-coil currents applied in one shot before the L-H transition and in the next shot after the L-H transition. The goal is to maintain a connected double null magnetic configuration (dRsep = 0) while increasing GAPIN and the alignment of the divertor legs with the upper and lower cryopumps. Investigate the effect of the RMPs on the ELMs including any ELM free periods and on the radial electric field. Compare the results with similar experiments pefroemd on MAST. Compare the pedestal evolution from the L-H transition to full performance H-mode with and without RMPs. Using the GAPIN corresponding to the lowest L-H power threshold from step 1, scan q95 between 3.2 and 4.5 by varying BT with constant Ip from shot-to-shot. | ||
| Background: | ITER will operate at powers close to the threshold power.
There is evidence from MAST and JET that applying RMPs in H-modes with Ploss close to PLH leads to a behaviour of the H-mode normally associated with even lower Ploss (i.e. similar to a decrease in power). Another aspect related to the above is to understand what is the influence of the application of RMPs on the pedestal evolution of the H-mode after the transition up to H=1 and from this back to the H-L transition. This is interesting/important for ITER for two reasons: a) the evolution from L-mode to high performance H-mode evolution in ITER is strongly influenced by density behaviour because this determines the alpha heating evolution which is a major contribution to the plasma heating at high Q; b) exit from H-mode in ITER is likely to lead to large changes of the plasma energy and cause large power fluxes and fast inwards shifts of the plasma and possible impact on the inner wall. Experience with RMP fields in DIII-D has shown a particularly strong density response in DN plasmas. By changing the alignment with the cryopumps during the GAPIN scan we should be able to affect the L-H power threshold, the evolution of the density following the H-mode transition and the onset of the first ELM. | ||
| Resource Requirements: | tokamak in clear state; both 210 NB sources; 30 and 330 L and R sources needed for profiles. 150L for BES fluctuation measurements. I-coil with reliable low current operation. Techrons if we can establish interesting results at sufficiently low current. | ||
| Diagnostic Requirements: | CER across plasma; Full fluctuation and pedestal profile diagnostics. | ||
| Analysis Requirements: | Significant profile, kinetic EFIT, Er, and fluctuation analysis. | ||
| Other Requirements: | |||
| Title | 36: Profile documentation in low-delta discharges with RMP ELM Suppression | ||
| Name: | maingi | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans, T. Osborne | ITPA Joint Experiment : | No |
| Description: | The main goal is to document the profiles changes with RMP in the low delta shapes that showed minimal pedestal degradation, to understand why they differ from the ISS shape. A complete RMP amplitude scan is requested. Maintaining recycling control could be challenging. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Make the ISS shape with the OSP on top of the baffle; will have to rely on cryopumping through the PFR for recycling and density control, which could be challenging. Do an RMP amplitude scan with 3-4 points below ELM suppression threshold, and 3-4 above to see if effect on edge profiles is continuous. | ||
| Background: | The ISS discharges always show a substantial pedestal pressure reduction with application of the RMP, because the neped drops by > 50%, while Teped does not change. The low delta discharges [Evans NF 2008] however show a strong increase in Te and even more so in Ti, leading to only a modest Pped drop. That lost pressure is made up by slightly improved confinement in the core. The idea here is to reproduce those shapes and document the pressure changes with the improved high resolution Thomson, and to see if the effect of the RMP on the edge pressure is continuous. | ||
| Resource Requirements: | 10-15 discharges (~ 1/2 day), cryopumps, RMP | ||
| Diagnostic Requirements: | Thomson, CER | ||
| Analysis Requirements: | Profile analysis with Osborne's tools; stability analysis with ELITE, GS2 and GYRO possibly | ||
| Other Requirements: | -- | ||
| Title | 37: ELM suppression and pedestal characterization as a function of drsep | ||
| Name: | maingi | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | E. Lazarus, T. Evans | ITPA Joint Experiment : | Yes |
| Description: | This proposal is to re-visit ELM suppression with RMP as a function of drsep. ELM suppression in DN has been unsuccessful, partly due to technical reasons and possibly due to underlying physics. In previous expts, suppression (mitigation) was achieved at drsep=-4cm (-2 cm) in the ISS discharges, while no effect was observed at drsep=0. In this expt., start from the successful case at drsep=-4 cm and move away from that in small drsep steps, looking for the threshold between suppression and mitigation, which should provide additional insight into the physics. Finally finish at drsep=0 cm, and use new techniques to eliminate previous technical difficulties. This expt. will also provide the data needed for ITPA PEP-6, discharge and pedestal performance and ELM characteristics as a function of drsep. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Leave the discharge in LSN during rampup, to obtain the favorable evolution of the equilbrium and stability developed over the pas few years.
2. Change drsep at ~ 1.5sec, slowly ramping from -4 cm to the target value; a drsep scan of -3 cm, -2 cm, -1 cm, and 0 cm is desired. 3. Turn on the I-coil after the shape is established, i.e. at ~ t=2 sec 4. If the beta is too high, do a pre-programmed NBI step down to avoid the beta limit; consider turning on beta feedback also. 5. If the coupling to the pumps is too strong with I-coil on, move the major radius of the plasma to reduce the coupling to the pumps, reducing exhaust. 6. Do RMP amplitude variations to determine the ELM suppression requirement at each drsep value. | ||
| Background: | ELM suppression is reproducible in the ISS discharges, but it has been elusive as drsep is changed from -4cm to 0cm toward double-null operation. Operationally the energy confinement increases as DN is established, leading to high beta locked modes. Turning on the I-coil early to induce density pumpout and pressure reduction (ie. reduce beta) has proven to be too effective, i.e. leading to L-mode like conditions. The ELMs in DN also appear to be more complex than the LSN, possibly due to differences in the 3-D topology. Also, ELM mitigation only has been achieved with drsep=-2cm. Thus, a drsep scan is proposed to find the limits of where ELM suppression can be achieved. Examination of the threshold cases will yield insight into the critical physics. Finally, a few new ideas to eliminate the operational problems in DN can be tried. | ||
| Resource Requirements: | ~ 1 day operation time, I-coils, cryopumps | ||
| Diagnostic Requirements: | Thomson, CER | ||
| Analysis Requirements: | profile analysis with Osborne's tools, ELITE, TRIP3D | ||
| Other Requirements: | -- | ||
| Title | 38: Investigation of Sheath Power Transmission at the DIII-D Divertor using Surface Thermocouples | ||
| Name: | Donovan | Affiliation: | U of Tennessee, Knoxville |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | Dean Buchenauer, Jon Watkins, Dmitry Rudakov, Charlie Lasnier, Josh Whaley, David Donovan | ITPA Joint Experiment : | No |
| Description: | An accurate assessment of the heat flux reaching the divertor is essential for making a proper selection of first wall materials for a fusion reactor. The heat flux can be measured directly by IR cameras, calorimeters, or thermocouples, or it can be calculated from the electron temperature and particle flux measured by Langmuir probes. The correlation between heat flux measurements and Langmuir probe measurements is known as the sheath power transmission factor (SPTF). For at least two decades, the experimental measurements of the SPTF on DIII-D have disagreed with theoretical values. Previous heat flux measurements have relied primarily upon IR cameras for comparison with the Lagmuir probes. A more direct comparison could be made between a surface thermocouple positioned immediately adjacent to a Langmuir probe. Surface thermocouples (STC) have been successfully implemented on other tokamaks (C-MOD and NSTX), and offer a more direct heat flux measurement than embedded thermocouples or calorimeter probes. The DiMES removable platform is the ideal means with which to test STCs in DIII-D. By placing a Langmuir probe and a STC on the same DiMES head, side-by-side measurements of heat flux can be made for a much more direct calculation of the SPTF, which could lead to a solution to the discrepancies between theoretical and experimental values. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A DiMES head will be used that contains both a standard domed Langmuir probe and a surface thermocouple. The STC will be Type E with a graphite body and ribbons several microns thick made of chromel and constantan running through the center. The STC is capable of being directly exposed to the plasma due to the unique eroding junction utilized. As the surface of the thermocouple is burned away by the plasma, the thermocouple junction is maintained at the surface by overlapping micro-connections between the ribbons. This technology has been successfully utilized under similar conditions in the C-MOD tokamak.
Preliminary testing of the STC can be done without the need for direct contact with the outer strike point. DiMES can be inserted during nearly any shot and used to measure the surface temperature of the divertor. The STC measurements can then be compared to those from the IR camera and the embedded thermocouples to determine if the temperatures are reasonable for the given conditions. Once the reliability of the STC has been confirmed, a sweep of the outer strike point across the DiMES head would be needed to obtain measurements of electron temperature and particle flux from the Langmuir probe. The SPTF can then be calculated by the measurements of the STC and the Langmuir probe. 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. | ||
| Background: | Since its installation in 1992, the Divertor Materials Evaluation System (DiMES) on the DIII-D tokamak has provided a unique platform for the study of plasma surface interactions. Early experiments performed many first-of-kind observations at a divertor surface: quantification of the net erosion rate of carbon, demonstration of reduced erosion during plasma detachment, elucidation of the role of chemical sputtering, quantification of deuterium retention in carbon and metallic coatings, and identification of a critical issue of MHD interaction between liquid lithium and a divertor plasma. These passive measurements have provided data for the benchmarking of PSI codes and helped to improve the operation of DIII-D.
Less well known perhaps is that DiMES can also be a platform for the development of plasma diagnostics. Early design issues have now been improved to provide 12 electrical feedthroughs (+ one pair for a thermocouple) for active measurements (microsecond time response). Sandia California designed the first active DiMES head and has tested Langmuir probes and H-microsensors using the platform. With the improved cabling, we propose to utilize the DiMES platform to address the sheath power trasmission conundrum. Experiments from DIII-D and other tokamaks have demonstrated that the power transmission through the sheath, as determined by divertor Langmuir probes and infrared camera images, is still a mystery. Profiles of the sheath power transmission factor (ratio of heat flux to product of ion saturation current and electron temperature) show this ratio varying across the outer divertor strike point, to values much less than the nominally expected value of ~7 near the peak heat and ion flux (more recent data shows a similar trend). Since this determination relies on the interpretation of probe characteristics and the thermal response of the tiles to the heat flux, the use of DiMES to better understand the plasma measurements and heat load would be beneficial. | ||
| Resource Requirements: | The hardware for the DiMES probe is available, along with instrumentation provided by the divertor Langmuir probe array. The experiment would require an IRTV at the 165 degree toroidal location (we are currently exploring alternatives to the TEXTOR camera), in additional to the present divertor camera view. Run time of approximately ½ day would also be required, including NBI availability (no cryo-pumping needed or desired). | ||
| Diagnostic Requirements: | A desirable element of the experiment would be to use longer focal length optics for the IR camera at 165 degrees (we also would require operation in line scan mode to improve time resolution during the x-point sweeps). This would provide better resolution of the heat flux to the domed probes
Divertor Langmuir probes Probes mounted on the DiMES system, using instrumentation from the divertor probe racks 165 degree lower divertor camera with long focal length optics and line scan mode Other lower divertor camera in customary video mode Divertor spectroscopy (Boron / Nitrogen) Magnetics for EFIT determination of field angles Zeff C02 interferometer Thomson scattering (burst mode for main experiment, but regular mode for piggyback) Fast filterscope channels viewing the lower divertor Other useful diagnostics Tile current array Fast tile thermocouple array Bolometers Edge CER for ion 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 | 39: ELM suppression in VH-mode | ||
| Name: | maingi | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans, T. Osborne, T. Taylor, J. Canik | ITPA Joint Experiment : | No |
| Description: | The goal is to re-visit the use of RMPs in VH-mode to achieve a steady VH phase with minimal impurity buildup and no terminating X-event. The previous attempts at this were done many years ago; much has been learned practically about how RMP should be optimized, i.e. timing, amplitude etc. There is an important consideration: should the standard high delta DN VH-mode recipe be used (DN has proven elusive in terms of RMP ELM suppression), or should the old reversed-Bt VH-mode in LSN shape to match JET from the mid-90's (Greenfield) be used? | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Establish VH-mode discharges without RMP; consider use of beta feedback to reduce dW/dt
2. Add RMP at a time after VH-mode is established at low amplitude 3. Do a short q95 ramp to look for resonance condition for ELM suppression 4. Use recent experience with RMP optimization to guide how best to apply RMP | ||
| Background: | Recent work by the 3-D ELM task force has suggested that the resonance condition for RMP might be explained by islands at the top of the pedestal that prevent the pedestal width from growing to the point that peeling-ballooning modes are destabilized (Snyder, APS 2011). The stable width in VH-mode is bigger, but VH-modes still suffer the terminating X-event, consistent with peeling-ballooning physics. The basic idea is to use RMPs to limit the width in VH-mode to prevent the X-event; if successful, the pedestal width should be at least 100% larger than the present ELM-suppressed width in ISS discharges, i.e. 6-8% in normalized poloidal flux. The timing of RMP application and VH-mode triggering is pretty important, but the experience with RMP application over the last few years should increase change of success wrt previous attempts. | ||
| Resource Requirements: | ~ 1 day machine time, RMP, cryopumps | ||
| Diagnostic Requirements: | Thomson, CER | ||
| Analysis Requirements: | Profile analysis with Python tools, stability analysis with ELITE, transport analysis with ONETWO and/or TRANSP | ||
| Other Requirements: | -- | ||
| Title | 40: Effectiveness of TBM error field correction in slowly rotating H-mode plasmas | ||
| Name: | Reimerdes | Affiliation: | CRPP-EPFL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J.M Hanson, R.J. La Haye, N. Oyama, J.A. Snipes, T. Tala | ITPA Joint Experiment : | No |
| Description: | This proposal seeks to investigate the effectiveness of TBM error field correction with low n saddle coils in slowly rotating H-mode plasmas. This investigation should reconcile the seemingly contradicting results in low density L-mode plasmas and in fast rotating H-mode plasmas (see background) and test the applicability of the 2011 H-mode results into a more ITER relevant regime. <br>It is hypothesized that the ineffectiveness of the I-coil to recover the TBM induced rotation decrease in fast rotating H-modes is primarily caused by a dominant role of the n>1 components of the TBM field and the related plasma response in braking the rotation. Since minimizing the n=1 plasma response does not coincide with maximizing the rotation higher order n=1 modes must also be important. Assuming that these components are dominantly non-resonant they should contribute less to the rotation braking at low rotation than the resonant components, which a mainly related to the principal n=1 plasma response. In addition they should only indirectly (via a decrease of the effective momentum confinement time [Reimerdes2009]) contribute to the mode locking, which is believed to be driven by penetration of resonant components of the total perturbed field. As a result n=1 error field correction should be efficient in restoring the tolerance to n=1 field penetration at low rotation, similar to the previous L-mode results [Schaffer2011]. In addition a maximum error field tolerance should coincide with a minimum of the n=1 plasma response. <br>This hypothesis can be tested by optimizing the n=1 EFC in a slowly rotating H-mode plasma with respect to its tolerance against error field penetration. In plasmas with the same value of net NBI torque and density the tolerance measured by the maximum tolerable additional n=1 I-coil field should not differ with or without TBM field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiments should be carried out in an ELMy H-mode discharge with ITER similar shape and low net NBI torque. Plasma parameters are chosen similar to ITER baseline parameters. The parameters can be adjusted in order to improve the reproducibility of the discharge. EFC is applied with the I-coil. The currents are optimized by maximizing the tolerance to field penetration. This is achieved by ramping the amplitude of an n=1 field with different toroidal phase angles until field penetration occurs. It is expected that the critical n=1 currents lie on a circle in the toroidal plane with its center indicating optimum correction. The procedure is repeated without the TBM field. The radii of the circles with and without the TBM field yield a measurement of the tolerance against field penetration. The optimum EFC is tested in both cases in order to evaluate the relation with respect to the n=1 plasma response. | ||
| Background: | The 2011 TBM experiment showed that n=1 error field correction (EFC) using the I-coil can only recover a small fraction of the TBM induced decrease of the plasma rotation. This result contrasts the 2009 low density locked mode experiment where n=1 EFC of the TBM could recover the low locking density of a plasma without TBM field [M.J. Schaffer2011]. The 2011 experiment also revealed that minimizing the n=1 plasma response is not synonymous with maximizing the plasma rotation. | ||
| Resource Requirements: | TBM mock-up coil, simultaneous co and counter-NBI. | ||
| Diagnostic Requirements: | Locked mode (RWM) sensors, rotation measurements, interferometer. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | References
[Schaffer2011] M.J. Schaffer, et al., Nucl. Fusion 51 (2011) 103028. [Reimerdes2009] H. Reimerdes, et al., Nucl. Fusion 49 (2009) 115001. | ||
| Title | 41: Effect of lithium wall conditioning on recycling and QH/VH-mode access | ||
| Name: | maingi | Affiliation: | PPPL |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | D. Mansfield, G. Jackson, T. Osborne, (A. Garafalo) | ITPA Joint Experiment : | No |
| Description: | The idea is to use lithium delivery systems from Mansfield to test lithium wall conditioning in DIII-D. The ultimate target is to reduce recycling to ease access to VH-mode and QH-mode. After the experiment, a boronization can be used to cover the lithium and return to normal wall conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use lithium dropper at different rates during the discharge to see if recycling can be controlled. Start with low dropper rate and increase systematically. Use a LSN discharge with large enough drsep that the second X-point is not is above the PFCs, i.e. 'out of the box'. If successful at reducing recycling, attempt to use this to ease access to QH-mode and VH-mode. | ||
| Background: | Lithium wall conditioning in NSTX has enabled remarkable control of the recycling and edge profiles, more so than can be achieved by cryopumping alone. Basically lithium works as a large area pump; the local divertor recycling coefficient is reduced down to 0.85-0.9; with cryopumping, the local recycling coefficient is reduced down typically to ~ 0.95. This has enabled NSTX to reduce the recycling source by > 50%, and have a density profile gradient in H-mode from psi_n=0.7-1. The altered density and temperature profiles are more stable to kink/peeling modes, even up to betan ~ 6, where RWMs are seen. In DIII-D, the first test would be with smaller lithium amounts than in NSTX, to see if some level of recycling control can be achieved. | ||
| Resource Requirements: | 2-3 days to test lithium conditioning, if implemented in DIII-D. | ||
| Diagnostic Requirements: | Thomson, CER | ||
| Analysis Requirements: | Edge profile and stability analysis; core transport analysis | ||
| Other Requirements: | |||
| Title | 42: Density Control and Active Impurity Removal from Double-null H-mode Plasmas | ||
| Name: | Petrie | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | 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 â?¤ 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 toroidal field is â??forwardâ??, i.e., the ion gradB drift is toward the lower divertor. Argon impurities are 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 = +0.5 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 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? | ||
| 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: | This is a proof-of-principle experiment. Machine time: 0.25 (forward Bt), only the upper outer divertor and lower outer cryo-pumps are at liquid helium temperature, minimum 6 co-beam sources. | ||
| Diagnostic Requirements: | Asdex gauges inside both outer baffles, Asdex gauge in the upper PFR, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, divertor IR cameras, and CER. | ||
| Analysis Requirements: | SOLPS/UEDGE, ONETWO | ||
| Other Requirements: | |||
| Title | 43: Establishing an experimental basis for plasma effects on magnetic islands | ||
| Name: | Evans | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | 3D Task Force Group Members | ITPA Joint Experiment : | Yes |
| 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 : | Yes |
| 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 | 44: The Compatibility of Radiative Divertor with AT Plasma Operation | ||
| Name: | Petrie | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This study will combine ALL the essential elements for making the first real test of a radiating divertor concept in an AT/hybrid near-DN plasma, using â??realisticâ?? high triangular shape and a particle exhaust configuration compatible with high performance tokamaks. PRESENTLY, ONLY DIII-D HAS THE CAPABILITY TO MAKE THIS TEST. Argon is injected into the private flux region near the upper outer divertor separatrix target. Enhanced deuterium plasma flow toward the divertor in the low field SOL is enhanced by a combination of deuterium gas injected upstream of both outer divertor targets and active cryo-pumping from both outer divertor locations; â??advancedâ?? tokamaks will likely not use an inner pump for a variety of reasons that will not be discussed here. Previous experiments have shown that setting the ion gradB drift direction toward the lower divertor and taking dRsep = +0.5 cm will yield the best chances of optimizing the benefits of (near) double-null shape with maintaining a high performance relatively clean of impurity accumulation. An additional benefit of having the ion gradB drift direction out of the divertor is that we are running at a significantly lower core density than if the ion gradB drift direction reversed, so RF can be included in our â??advancedâ?? scenarios. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The plasmas are near-DN high performance H-modes that can be reliably maintained for at least 2 seconds. Both outer divertor cryo-pumps are at liquid helium temperature. The gradB-ion drift direction is downward. dRsep = +0.5 cm. RF heating can be used. This experiment is probably best done in as follows:
* First establish the sensitivity of AT plasmas to deuterium gas injection. Scan the deuterium gas puff rate, i.e., establish operational limit to how much D2 gas injection the AT plasma can accommodate before there is appreciable degradation in AT plasma properties. * Scan of the argon injection rate at the most â??reasonableâ?? D2 injection rate established above. Understanding the sensitivity of high performance DIII-D plasmas embedded in a â??radiating divertorâ?? environment is paramount. So, important measurables from this experiment are, of course, the changes in energy confinement and current profile. Other important measurables include changes in the (poloidal) radiated power distribution and heat flux, changes in the density and electron temperature at the divertor targets, and the accumulation of argon in the core and divertor plasmas. In the present DIII-D machine, the methodology discussed above provides the best chance to successfully couple a high performance discharge to a radiating divertor, and, from my experience, the only chance of doing so. | ||
| Background: | High performance â??ATâ?? in the DN and near-DN configurations are attractive for future power reactor operation due to their high toroidal beta and energy confinement properties. However, for futuristic AT-machines (like ARIES-AT), there can be severe divertor power loading problems. One possible way of reducing excessive power loading at the divertor target(s) is to radiate significant power outside the main plasma, mainly in the divertor (and SOL to some extent). But the resulting divertor cooling may also lead to a cooling of the upstream (core) plasma, which, in turn, may result in a marked degradation in AT-edge properties. The expected increase in the argon presence in the pedestal can also be expected to affect the AT-pedestal adversely.
Previous work with radiating divertor H-mode DN plasmas has shown that the â??balancedâ?? DN results in overly rapid accumulation of the seeded impurity (argon) in the core plasma. Two important reasons for this are (1) the relatively easy penetration of an impurity specie from the high field side into the core plasma of the DN and (2) the particle drifts in the scrape-off layer plasma in one of the divertors that always assist in the escape of injected impurities from the divertor region to the vulnerable high field side SOL. On the other hand, the radiating divertor was shown to be effective in magnetically unbalanced DNs (dRsep=+0.5 cm with gradB drift down) for reducing divertor heat flux while still maintaining good H-mode properties. This configuration has also been show to produce the lowest density we can achieve in DIII-D for parameters generally used in AT experiments, and this should facilitate the use of RF heating and current density control. | ||
| Resource Requirements: | Because coupling AT plasmas to radiating divertor conditions has never been attempted, elements of this experiment could be run in piggyback, e.g., assess the sensitivity of an AT plasma to specific gas puff rates. If the results are favorable, then a dedicated follow-up experiment can be planned. Estimated time to execute the experiment, as described above: 0.5 â?? 1.0 day. Both outer baffle cryo-pumps â??coldâ?? should be cold and there should be minimum of six 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, IR cameras monitoring the divertor, and CER. | ||
| Analysis Requirements: | SOLPS/UEDGE, ONETWO | ||
| Other Requirements: | |||
| Title | 45: Comparison of pedestal profiles and confinement in ECRH and NBI heated H-mode plasmas | ||
| Name: | Lore | Affiliation: | ORNL |
| Research Area: | Pedestal | Presentation time: | Requested |
| Co-Author(s): | R.J. Groebner, T. Osborne, R. Prater | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to determine the effect of the heating mixture (ECRH and NBI) on the pedestal structure and confinement properties. Pure ECRH and pure NBI heated H-modes with similar line averaged densities and input power will be compared for 3-4 power levels and 4-5 ECRH deposition radii. Time permitting, a small density scan and/or triangularity scan will be performed.
Results from this experiment will augment a comparison of a limited number of pure ECRH and pure NBI cases made using shots from non-dedicated experiments [J.D. Lore, et al., Nucl. Fusion 2012 (submitted)]. In this case it was found that confinement was similar for ECRH and NBI cases with on-axis deposition, however far-off axis ECRH deposition resulted in a reduced Te pedestal height. As these results are from a limited set of discharges, the robustness of these results should be tested in other shapes, power levels, etc. Determining the effect of dominant electron heating with low torque input has also been identified as an urgent research task for ITER. Temperature, density, and rotation profiles will be compared with a focus on the pedestal region. The effect of heating method on the ELM frequency and magnitude, and global confinement properties will also be investigated. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start with reference NBI plasma, constant density, with 3-4 power levels in the range that can be reproduced with ECRH. Density should be low enough to ensure good ECRH absorption. Power should be held constant long enough to get good profiles. Reproduce shot with NBI power replaced with ECRH power. Repeat ECRH discharge with 4-5 different resonance locations in minor radius.
Time permitting, repeat above set of shots at a higher reference density. Operation at higher collisionality will allow direct comparison to ASDEX results from 2011. These results showed that confinement was independent of heating mixture, however only on-axis deposition and high collisionality cases were tested. Time permitting, test robustness of results versus shape (triangularity scan proposed). | ||
| Background: | A search of the DIII-D database for suitable pure ECRH and NBI H-mode shots was performed in 2011 and presented at the H-mode. A limited number of good comparisons were found, however the results indicate that for the same line-averaged density and input power, ECRH discharges with far off-axis resonance (rho > 0.5) have reduced thermal confinement times and a lower Te pedestal height as compared to NBI plasmas. With on-axis resonance, however, the pedestal structure and confinement is similar in both cases. In mixed (ECRH+NBI) discharges, adding ECRH power decreases the density pedestal height and the global stored energy with far off-axis deposition. These effects are reduced or reversed as the resonance is moved towards the axis. | ||
| Resource Requirements: | All gyrotrons. NB sources at level to match maximum ECRH power. | ||
| Diagnostic Requirements: | Thomson and CER systems. D-alpha arrays. Fluctuation diagnostics. | ||
| Analysis Requirements: | EFIT, profile analysis tools, ONETWO, TORAY, EPED. | ||
| Other Requirements: | -- | ||
| Title | 46: BetaN=5 at min(q)>2 | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Achieving beta_N ~ 5 is considered a necessary condition of steady-state tokamak power plant operation. Many modeling studies have pointed to high min(q) as a suitable solution due to (allegedly) high bootstrap current to lower the recirculating power and high ideal-wall n=1 stability limits due to wall stabilization. Finding this regime in DIII-D for a very energy confinement times would provide confidence in the modeling basis and allow improvement of the assumptions made. It would also inform the decision to install a second vertically steerable NB on DIII-D. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start from existing min(q)>2 scenarios and lower B and I, switching to 3rd harmonic ECH when necessary. | ||
| Background: | Endless modeling studies. | ||
| Resource Requirements: | 20 MW NB (can use ctr-NB if needed for power, but priority to co-NB). 150 NB steered full off-axis. EC desirable, but not required. | ||
| Diagnostic Requirements: | Standard set for kinetic EFITs for stability analysis. Fast diagnostics for instability identification. MSE EFITs for CD analysis. | ||
| Analysis Requirements: | Yes. | ||
| Other Requirements: | |||
| Title | 47: Coupling of Correction Coils to Internal Kink | ||
| Name: | Lazarus | Affiliation: | ORNL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Recent 3D equilibrium calculations (VMEC) show an equilibrium state best-described as a saturated internal kink. The proposal is to study this with a goal of controlling sawtooth amplitude. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Calculations show that for kappa=2 elliptical shapes at A~3 an edge pertrubation delta_R/R0~1e-4 will produce the internal mode. | ||
| Background: | The model equilibria have high shear at q=1, thus differ from the work of W.A. Cooper, et. al., proposed for last year's TJA experiment. | ||
| Resource Requirements: | DIII-D, NB, ECH | ||
| Diagnostic Requirements: | ECE, CER, magnetics | ||
| Analysis Requirements: | VMEC, Terpsichore | ||
| Other Requirements: | |||
| Title | 48: Runaway electron characteristics and loss dynamics with fast visible imaging | ||
| Name: | Moyer | Affiliation: | UCSD |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | J. Yu, E. Hollmann | ITPA Joint Experiment : | No |
| Description: | Investigation of runaway electron generation, acceleration, transport and loss dynamics using fast 2D imaging of synchrotron emission. Fast visible imaging has been used successfully in DIII-D {J. Yu et al, in preparation] to study the physics of RE generation and loss. For runaway electrons (RE) in the tail of the distribution function (energies up to 60 MeV) that can be imaged in the visible, the typical pattern of synchrontron emission is a field-aligned forward cone of emission whose projection on a 2D plane as imaged by the camera is an oval containing information on the location, energy and pitch angle of the RE population vs. time. However, in many cases (particularly in CY11), this oval distribution developes into a significant crescent shape reminiscent of a (1,1) island; this pattern is often lost following rapid events with significant RE loss (indicated by gamma ray bursts from REs hitting the walls). Another anomalous behavior is the formation of "satellite" RE beams of much smaller spatial extent to the main beam. These satellites display a tendency to move in the opposite direction to the main RE beam, a result which is quite puzzling. The goal of this experiment will be to document these RE behaviors, and to characterize the parameters which affect these dynamics, using primarily in-situ fast high spatial resolution visible imaging of the synchrotron emission with the fast framing cameras.
- development of crescent shaped emission over more common ovals - events with lead to RE loss and re-symmeterization of the emission profile - formation and evolution of small satellite RE emission profiles - test models for dynamical evolution of RE populations |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use the fast visible cameras with forward and backward views to image the evolution of synchrotron emission from RE in the visible range to study the creation, acceleration, transport, and loss of REs in a post-disruption RE plateau. Analysis of the visible emission provides high temporal and 2D spatially resolved information on the RE energy distribution (tail), on location in the vessel, and on dynamics of acceleration and loss during the plateau. In most cases, the RE "beam" appears as an oval of emission due to the relativistic forward peaking and the finite pitch angle of the magnetic field lines. In many cases in CY11, however, the RE emission developed into a crescent shape. This pattern is re-symmeterized in fast events (MHD instabilities?) that produce a fast loss of REs to the wall and that leave the RE emission distribution once again oval. Several models have been proposed for the origin of this non-symmetric emission pattern. One involves the development of an annular outer region of high parallel electric field as the outermost REs are scraped off on the vessel structures, producing a secondary RE acceleration in an annular region that should lead to a shell of emission. Alternative models include island-like MHD. The goal of this experiment will be to characterize this generation, acceleration, transport and loss, and to investigate the impact of "soft-landing" or RE beam dissipation/diffusion techniques for controlling the RE beam and it's interaction with the plasma-facing components. | ||
| Background: | Massive Gas Injection has been shown to be very effective in reducing both thermal and mechanical loads associated with disruptions. However, damage to the plasma-facing components due to runaway electrons produced in both the disruptions and the rapid shutdowns remain a significant threat to ITER. DIII-D is one of the few tokamaks with the ability to study runaway electron physics in a systematic way. This experiment will provide important information on the RE generation, acceleration, and loss dynamics needed for the upcoming US design of a rapid shutdown system for ITER. D3D has demonstrated reliable production of RE plateaus following "killer" pellet injection, and the ability to control the position of the resulting RE beam for long periods in the vacuum vessel (moving the beam to larger and smaller Rmajor while maintaining vertical stability). These capabilities make D3D ideally suited to study this critical ITER physics. | ||
| Resource Requirements: | 2.7 mm "Killer" Argon pellets to trigger RE plateaus. RE and disruption diagnostics (DISRAD, SXR arrays, BGO scintillators, fast cameras, visible survey spectrometers, etc.). Fast magnetics AFTER the thermal quench into the RE plateau. | ||
| Diagnostic Requirements: | Fast camera with forward and backward views of synchrontron emission in the visible portion of the spectrum. Survey visible spectrometers for both view to validate what emission is being measured by the cameras. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 49: Double Null Elm Suppression | ||
| Name: | Lazarus | Affiliation: | ORNL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | T. Evans, M. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | Obtain ELM suppression in Double Null discharges. Study the ELM in high-symmetry cases. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Similar to the discharges of last year, but increase I-coil current, scan safety factor, use C-coil to optimize RMP spectrum | ||
| Background: | Last year we concluded that we could not achieve ELM suppression in DN. With hindsight, we realized that we were very close. Although the ELM structure is altered, the ELM perturbation was at the low level that is commonly seen when the safety factor is just outside a suppression window. | ||
| Resource Requirements: | Usual complement for RMP ELM experiments.
Will want to repeat some parts of the experiment with reversed grad-B drift. | ||
| Diagnostic Requirements: | Usual for RMP ELM experiemts | ||
| Analysis Requirements: | |||
| Other Requirements: | Improved method for selecting windows for kinetic EFIT. Probably based on sumultaneously meeting D_alpha criterion at an upper and a lower photodiode. | ||
| Title | 50: Checkout of real-time steerable mirror NTM control | ||
| Name: | La Haye | Affiliation: | Retired from GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Lohr, B. Penaflor, D. Humphreys, R. Prater, A. Welander | ITPA Joint Experiment : | Yes |
| Description: | The new capability for real-time steering of ECCD launch mirrors with control of "poloidal" angle by the PCS makes routine use possible; no need to move plasma (Rsurf or Zsurf) or change BT for alignment. A preliminary slow sweep in April 2010 was successful. PCS control of pre-programmed sweeps, search and suppress, and active tracking all need to be developed and checked out using the mirrors. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a steady m/n=3/2 NTM as in 142650 for example.
For all gyrotrons, one at a time, do preprogrammed slow sweeps across q=3/2 and back to check biggest dip and mirror locations. Repeat with 100 Hz modulation of gyrotron power for ECE measure of where absorption occurs. Repeat (no mod) with pre-programmed single "instantaneous" steps of each mirror (0.5 degrees) close to biggest dip to check PCS/Ethernet latency and mirror response. Multiple gyrotrons powered simultaneously to study the power and alignment requirements for complete stabilization, i.e. map out rate of decrease of n=2 MIRNOV Btheta with misalignment and/or Peccd; condition for complete stabilization. | ||
| Background: | ITER will use real-time steerable mirrors, not plasma shifts or changes in BT for alignment as we have had to do up to now. Our successful control techniques of "search and suppress", "target lock" and "active tracking" can be modified for mirror steering. | ||
| Resource Requirements: | All gyrotrons (6 in 2011) one at a time to start. Fast mirror upgrade by PPPL was installed in 2011. Reduction of PCS latency needed as planned. Safety interlocks on hardware and PCS to avoid mirror failures. | ||
| Diagnostic Requirements: | ECE in particular. All kinetics (MSE, CER, Thomson) for TORAY-GA. | ||
| Analysis Requirements: | Auto analysis codes for surveying OK. | ||
| Other Requirements: | -- | ||
| Title | 51: BetaN=5 without wall stabilization | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Beta_N~5 is considered a necessary condition for steady-state tokamak power plant operation. Demonstration of stable profiles at this level for several energy confinement times without the need for wall stabilization would provide a critical benchmark for modeling and projections. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Build on previous results using new or untried tools available: possibility of slightly off-axis NBI, more EC power, non-axisymmetric magnetic fields and to limit the pedestal | ||
| Background: | |||
| Resource Requirements: | 20 MW NBI (ctr-NB ok if needed) | ||
| Diagnostic Requirements: | Standard diagnostics for kinetic EFITs for stability analysis and for MSE EFITs for CD analysis. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 52: Performance extension of advanced inductive with no applied torque | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Experiments were limited in achievable betaN and duration by power and energy available in the 210 NBs. Going to 1.25 T would allow exploration of higher beta_N at fixed power, and longer duration at lower beta_N. The role of ECH in stability can be probed (there will be no ECCD) by altering the heating location somewhat (first-pass absorption will not be complete, but dominant). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Translate existing scenarios to lower B. | ||
| Background: | |||
| Resource Requirements: | 210 NBs must work. Longer pulse operation highly desirable. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 53: ITER low-activation scenario simulation | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Run H beams into He plasmas. Look at fueling, L-H threshold, flux consumption, and ELM mitigation by RMP issues. Likely will need to run a few different B fields and different q95. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | He plasmas, conversion of at least 2 beamlines to H, probably all 4. Must think hard about whether to argon frost cryopumps or allow the preferential pumping of H to mitigate the fueling ratio difference to ITER. | ||
| Diagnostic Requirements: | Would like to measure H and He ion densities. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 54: Particle transport in H mode | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use stationary and perturbative techniques to explore particle transport in H mode. Compare EC (or EC+FW) plasmas to NB plasmas (central fueling vs. none). Test model of density peaking vs. collisionality with and without central fueling from NB. Compare transport with co-NB and balanced-NB and ctr-NB. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | Need to optimize for reflectometer data. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 55: Effect of applied torque on energy and momentum confinement | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Execute detailed engineering (torque) and dimensionless parameter (Mach number) scans in L mode and H mode to clarify the role of rotation in energy confinement. Momentum confinement comes for "free". Compare, if possible, to EC or EC+FW plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | 210 NBs must be functioning. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 56: PCS Detection and Management of ECCD for Prompt 2/1 NTM Suppression | ||
| Name: | La Haye | Affiliation: | Retired from GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Lohr, N. Eidietis, D. Humphreys, B. Penaflor, R. Prater, A. Welander | ITPA Joint Experiment : | Yes |
| Description: | Show that the PCS can control the ECCD mirror aiming fast and accurately enough to prevent the onset of an m/n=2/1 NTM or lock onto it and suppress it before it grows to large amplitude. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Needs successful checkout of the real-time steerable mirror control of gyrotron aiming, and implementation into the PCS of mirror versions of active tracking (without a mode) and search and suppress or target lock (with a mode). Show that an otherwise unstable mode can be raised to higher beta stably with active ECCD tracking by mirror alignment. Show that with ECCD initially off, the PCS can detect the growing mode (capability already exists), can turn on the gyrotrons (capability already exists), use a logic of both search and suppress (target lock) andactive tracking to stop the mode growth and then completely suppress it. | ||
| Background: | ITER relies on 2/1 NTM control by ECCD using mirror steering. DIII-D has pioneered techniques of search and suppress and active tracking but only by promptly changing the plasma position for alignment (not possible in ITER) or by less promptly changing BT (also not possible in ITER). We now have a real-time mirror steering capability that needs to be exploited, particularly to deliver a mode controller to DIII-D AT plasmas. The logic for ITER control described in the above exp approach/plan is described in Figure 9 of La Haye et al. Nucl. Fusion 2009 but has never been demonstrated on any device with any alignment technique. | ||
| Resource Requirements: | Authorization for real-time steering which will allow basic checkout (see 50) at a minimum to be done.
6 gyrotrons. Previous thorough checkout of PCS mirror control. Implementation of PCS mirror versions of active tracking and search and suppress (target lock too). Faster mirror steering by new fast mirrors (installed by PPPL) and work on ETHERNET/PCS link to reduce latency (concept in hand). | ||
| Diagnostic Requirements: | Standard. ECE and ECEI for island location. | ||
| Analysis Requirements: | Standard, TORAY-GA etc. | ||
| Other Requirements: | -- | ||
| Title | 57: Simultaneous perturbation analysis of energy, momentum, and particle transport | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | For over 20 years, it has been recognized that the electron temperature, ion temperature, ion rotation, and particle transport are linked. Two attempts have been made to look at the coupling (JET-early 90s, DIII-D-late 90s). We have a unique capability with the fast CER to do all channels together. Propose starting with a pure source (EC) and looking at the direct and cross responses in sawtooth-free L mode (to avoid the "noise" of MHD). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat 2011 experiments varying the EC location, but optimized for CER and reflectometer measurements of perturbations. | ||
| Background: | |||
| Resource Requirements: | 5 gyrotrons min | ||
| Diagnostic Requirements: | Fast CER (and requisite NBs), reflectometers | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 58: Scaling of L-H threshold with density | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER operation takes credit for the strong density dependence of the L-H threshold. This should be remeasured in an ITER-shaped plasma, with variations in the x-point height and the dome. Also, there is a minimum threshold power as density is reduced--the scaling of this with engineering and dimensionless parameters needs to be quantified. This should be done with no applied torque if possible, to eliminate that as a source of systematic error. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | 210 NBs must be working. 5 gyrotrons. Reversed B (with possible extension to normal B). | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 59: Test of ITER divertor model | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Present ITER modeling claims that the divertor is in a different regime from present-day experiments. The modeling also neglects drifts. We should push DIII-D as far as possible in the direction of ITER by running high density (meaning high current) with gas puffing and compare the divertor behavior in quasi-SN plasmas (|dRsep| > 2 cm) with both upper and lower null to see the effects of drifts. In the lower case, use the more extensive diagnostic set to benchmark modeling of the divertor and look for approach to the predicted ITER regime. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | Need new IR periscope functional. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 60: ITER baseline scenario access to SOF | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | ITER (mistakenly) believes they can save volt-seconds in the inductive ramp by ramping faster. This experiment will first explore the limits on current ramp (slow and fast) and document the flux consumption. Full-bore ITER scenario with EC startup assist should be used. Then the impact on flux consumption of auxiliary heating during the ramp should be documented. IR camera measurements of the inside-wall limiter phase would be highly desirable. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | |||
| Background: | Part of IOS-1.1 joint experiment. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 61: Identify RMP transport modification mechanisms | ||
| Name: | Boedo | Affiliation: | UCSD |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov | ITPA Joint Experiment : | No |
| Description: | The way the RMP affects the plasma and transport is still up for debate. We propose an experiment where high-diagnostic-density probes will be inserted in RMP perturbed plasma with n=1, n=2 and n=3 perturbations in order to clarify already observed mechanisms involving electric fields.
We will use OH, L-mode and H-mode plasmas at various power levels to also test the effect of rotation and collisionality on the perturbation evolution and penetration. Use of ECH will be extensive to comare discharges with and without torque. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create low power L-mode and H-mode discharges with naturally low rotation. Use inner-limited geometry to allow larger islands. Create n=1 perturbations first in L-mode and add modulated beams (L or R source) to vary torque and see RMP penetration and then increase power to shift into H-mode. Move on to n=2 and n=3 perturbations. | ||
| Background: | |||
| Resource Requirements: | NBI, ECH | ||
| Diagnostic Requirements: | scanning probes, fast cameras. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 62: Mapping of access to stationary ITER baseline scenario | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Experiments in 2011 found stable stationary operating conditions for one set of initial conditions, and access problems for lower betaN and same current ramp rate. The map of access conditions vs. initial li and SOB torque need to be completed. For discharges reaching stationary conditions, it is important to know if there is a unique final state or if it depends on the path. After reaching a stationary state, exploration of nearby states with higher or lower pressure is important for setting physics requirements on burn control. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | Part of IOS-1.1 joint experiment | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 63: Active means for access to ITER baseline scenario | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Experiments in DIII-D have found relatively narrow conditions for access to stable operation near the ITER baseline scenario parameters. This experiment would investigate whether active means to control the 2/1 tearing mode stability with direct ECCD allows a wider range of access conditions to stable operation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | Part of IOS-1.1 joint experiment. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 64: Identify basic edge/SOL flow mechanisms | ||
| Name: | Boedo | Affiliation: | UCSD |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov | ITPA Joint Experiment : | No |
| Description: | Various mechanisms have been identified as affecting flows in the edge/SOL. A recent one is the presence of a large momentum source straddling the LCFS. There is also a Pfirsch-Sluter mechanism and obviously the classical divertor pressure asymmetries and drfits. More work is needed to separate and quantify the relative importance of these mechanisms | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | USN and LSN, ELM-free, low power H-mode and L-mode discharges with varying divertor conditions and drift direction. Divertor conditions will be varied by alternating pumping (in USN) or varying OSP position in LSN.The diamagnetic terms will be varied by the change from L to H-mode. LCFS rotation will be modified by using ECH or NBI (co-counter) | ||
| Background: | |||
| Resource Requirements: | NBI, ECH | ||
| Diagnostic Requirements: | probes Xpt and midplane | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 65: Divertor detachment transport | ||
| Name: | Boedo | Affiliation: | UCSD |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov | ITPA Joint Experiment : | No |
| Description: | We propose to identify and quantify the source of the effect of divertor detachment on downstream and upstream turbulent particle and heat transport.
We can also quantify the fraction of heat and particle flux in the divertor corresponding to bulk flow. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create LSN discharges at low and medium power and detach gradually while sampling upstream SOL and divertor area in 2D by shifting geometry. | ||
| Background: | It is known divertor detachment is accompanied by an increase in parameter fluctuations in the divertor and upstream, but there is no clear quantification of this effect, its origin and if it actually affects transport in the divertor and upstream/downstream SOL. | ||
| Resource Requirements: | NBI, ICH | ||
| Diagnostic Requirements: | divertor TS, scanning probes (Xpt and midplane) , floor probes, etc | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 66: Prompt torque and zonal flow damping | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | Transport and 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 on offsetting radial current, which then imposes a torque on the background plasma. However, this offsetting radial current grows up on the ion-ion collision time. During this time, the poloidal rotation and the radial electric field both evolve. If we use out of phase modulation of the counter and co beams, we can periodically reverse this torque, creating a square wave modulation. If the modulation period is fast compared to the fast ion slowing down, we only need to consider the prompt torque. For a plasma with 15 keV central temperature and 5 x 10^19 m^-3 density, the fast ion slowing down time is greater than 100 ms even for the 1/3 energy component. The damping of the overall plasma poloidal rotation is the same as the damping time of the plasma electric field. Accordingly, CER measurements of any impurity ion can be used to determine the overall poloidal rotation damping. More importantly, this damping time of the plasma electric field is the zonal flow damping time, which is crucial to turbulence behavior. Theory predicts that this damping time is of order the ion-ion collision time which is around 20 ms in our candidate plasmas. | ||
| Resource Requirements: | Reverse Ip. 7 NBI sources. 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 | 67: Investigate angular momentum diffusion and pinch using off-axis torque | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | Transport and 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. he 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 | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 68: Private Region Physics | ||
| Name: | Boedo | Affiliation: | UCSD |
| Research Area: | General B&PP | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov | ITPA Joint Experiment : | No |
| Description: | Identify and quantify the sources of transport into, and out of, the private region. There is a lot of promise in learning how to control the transport in this region particularly if it leads to an increased divertor heat footprint and concomitant reduction in heat flux. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create LSN plasmas and scan the magnetic geometry to sample in 2D with divertor diagnostics. Scan density and Ip to change radial transport and approach detachment. carefully inventory particles and their sources and sinks. | ||
| Background: | The private region has been virtually ignored for many years despite being of key importance for its proximity to the core (via Xpoint) and of its relevance for rector operation by being in close contact with divertor domes. Depsite having a strong particle and heat sink in its immediacy (the floor), this region contains a surprising amount of particles and heat. There is a lot of promise in learning how to control the transport in this region particularly if it leads to an increased divertor heat footprint | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | scaning probes, divertor TS, floorprobes, gages | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 69: Comparison Mach probe and CER measurements of intrinsic rotation profile of main ions in helium plasmas | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | J.S. DeGrassie, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure the edge main ion toroidal rotation profile in helium plasmas with the Mach probe with the CER system and then compare them. The goal is to verify that we see the same edge rotation structure on both diagnostics, including the localized peak seen previously (see background). The data will be used to compare with XGC0 calculations to see if we can achieve a physics understanding of this work. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize low power ECH H-mode plasmas like 140420-437 as the basic target. Measure edge main ion profile with Mach probe and CER at various times both before and after the L to H transition. Since the Mach probe plunges at one or two times and since the beam blip for CER seriously affects the rotation, multiple shot will be required to obtain a complete time history. Obtain complete edge profile data needed for the XGC0 modelling. | ||
| Background: | Experimental measurements of the edge main ion rotation profile in ECH H-modes (shot range 141444-141487) using the reciprocating Mach probe showed a localized peak in the deuteron rotation profile with the top of the peak on the separatrix. Data mining of helium H-mode plasmas from several years ago (shots 140420-437) demonstrated a similar structure in the main ion rotation. From the standpoint of angular momentum transport, this localized peak is quite surprising, since it is inconsistent with simple transport models. The structure may be a consequence of ion orbits crossing the separatrix; such effects have been predicted by the XGC0 code, although the spatial structure doesnot exactly match the experiment. | ||
| Resource Requirements: | Helium plasmas with ECH | ||
| Diagnostic Requirements: | Mach probe. Standard profile diagnostics including CER. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 70: Test of Neoclassical Toroidal Viscosity theory using modulated I-coil currents | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | Use modulated I-coil currents to investigate the theory of braking of plasma toroidal rotation by non-resonant error fields | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate the I-coil currents to modulate the non-resonant drag on the plasma. Investigate the effects as a function of modulation frequency, background plasma rotation, collisionality and I-coil parity. | ||
| Background: | This experiment was given 1/2 day in 2008. Unfortunately, there were issues of machine cleanliness since it was run after an experiment with significant gas puffing. Accordingly, the QH-modes were poor. Attempts to perform this experiment in ELMing H-mode lead to locking of the ELM frequency to the I-coil modulation. Although this locking was a significant discovery, the ELM effects on the rotation masked the direct I-coil effects. We need to perform this experiment in high quality QH-mode plasmas, since this avoids the ELM problem while still allowing us to probe H-mode plasmas. | ||
| Resource Requirements: | I-coil system connected to 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 | 71: Dependence of momentum transport on Te/Ti | ||
| Name: | Yoshida | Affiliation: | QST |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | W. Solomon, P. Gohil | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the dependence of momentum transport diffusivity and momentum pinch velocity on Te/Ti in order to increase an understanding of the momentum transport in ITER.
Recently the electron and ion transport dependence on Te/Ti has been clarified in DIII-D [L. Schmitz, et. al., NF2012]. Regarding the momentum transport and the toroidal rotation velocity, the intrinsic torque with ECH has been observed in [W. Solomon, et. al., NF2011]. Momentum transport on the density scale length, collisinality and gyro radius were investigated in [T. Tala, et. al., NF2011] and [M. Yoshida, et. al., H-mode workshop 2011]. However the dependence of momentum transport on Te/Ti still remains an open issue DIII-D has advantages in studying momentum transport in ITER-like plasmas: electron heating with low collisionality, low external torque input using co & counter NBs and EC, and ITER-like plasma shape. Using this advantage the dependence of momentum transport on Te/Ti will be obtained. The data taken in DIII-D will be compared to the data in JT-60U where the ion heating was dominant or Te/Ti<1. By comparing the measurements in DIII-D to those in JT-60U, we will make a progress in understanding of the properties of momentum transport in the plasmas dominated by electron heating and ion heatings. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Momentum transport diffusivity and momentum pinch velocity will be evaluated using NB modulation or magnetic perturbative experiments.
Dependency of the momentum transport diffusivity and momentum pinch velocity on Te/Ti will be investigated through a scan of the power ratio of BAL-NB to EC at a totally constant absorbed power. The target plasma is as follows. - type-I ELMy H-mode plasmas - ITER-like plasma shape - low external torque input using a combination of the co- and counter-NBs - low density to avoid the coupling between electrons and ions due to collisional equipartition Relation between the changes of momentum transport and of turbulence properties will be investigated using turbulence measurements including long and short wavelength. | ||
| Background: | In the momentum database for ITPA Transport and Confinement group, we have recently found the ratio of momentum diffusivity to ion heat diffusivity tends to be smaller with a large Te/Ti, and the ratio of pinch velocity to momentum diffusivity tends to increase with decreasing Te/Ti.
However these trends depend on JT-60U and JET data. In addition, data at Te/Ti<1 dominate the database. | ||
| Resource Requirements: | 1 day experiment. CO & Counter NBs, ECH (~3 MW, ~3 sec) | ||
| Diagnostic Requirements: | Standard diagnostics, especially fast CER for toroidal and poloidal rotation and ion temperature, Thomson for electron density and temperature, ECE for sawtooth and electron temperature, high k FIR scattering, BES. | ||
| Analysis Requirements: | Transient momentum transport analysis using modulated NBs or magnetic perturbations | ||
| Other Requirements: | |||
| Title | 72: Maintain low-rotation QH-mode with ECH only | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | A. Garofalo, W.M. Solomon | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to demonstrate that low-rotation QH-mode 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 2009-2011 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 | 73: Investigate in-out density asymmetry at large toroidal rotation | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | Transport and 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 measurements inside magnetic axis. | ||
| Diagnostic Requirements: | CER chords viewing points inside magnetic axis. All profile diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 74: Effect of off-axis beams on Alfven eigenmode stability and mode structure | ||
| Name: | Heidbrink | Affiliation: | UC, Irvine |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Energetic Particle group | ITPA Joint Experiment : | Yes |
| Description: | Study the effect of variations in the fast-ion gradient on the stability and mode structure of RSAEs, TAEs, and BAAEs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Basically, this is a repeat of the successful 2011 experiment but with the toroidal field reversed to maximize the change in the fast-ion gradient associated with switching between on-axis and off-axis injection. Experimental conditions will match the best shots from the 2011 experiment (early beam injection into L-mode). | ||
| Background: | In 2011, we conducted a very successful 1/2 day experiment on this topic. For that experiment, we used normal (-) toroidal field so the FILD diagnostics would work. As a result, the change in the fast-ion gradient was not as great as it could be with +BT off-axis beam injection &, in preliminary analysis, it is not certain that the gradient changed sign in the core. Cleaner data is desirable for my IAEA paper. | ||
| Resource Requirements: | Nearly every beam source is required. | ||
| Diagnostic Requirements: | Good FIDA and ECE-I data are essential. | ||
| Analysis Requirements: | The usual: EFIT, TRANSP, FIDASIM, NOVA. | ||
| Other Requirements: | |||
| Title | 75: Alfven eigenmodes during the current ramp in H-mode plasmas | ||
| Name: | Heidbrink | Affiliation: | UC, Irvine |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | C. Petty & the EP group | ITPA Joint Experiment : | Yes |
| Description: | For many years, we have studied Alfven eigenmode activity during the current ramp in L-mode plasmas. The purpose of this experiment is to see whether the observations of stability thresholds, mode structure, and fast-ion transport are similar in H-mode plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Induce an early H-mode transition. Use a shape with good pumping to control the density. Inject beams early to produce Alfven eigenmode activity. Attempt to match q profiles of L-mode plasmas but with H-mode density profiles. Vary beam power to find stability thresholds for the various Alfven eigenmodes. | ||
| Background: | Since 2005, for diagnostic reasons, our studies of Alfven eigenmodes in reversed shear plasmas have been almost entirely in L-mode plasmas. It is important to confirm that our observations in those experiments are also applicable in more reactor-relevant plasmas. | ||
| Resource Requirements: | Nearly every neutral beam source. ECH | ||
| Diagnostic Requirements: | FIDA, FILD, ECE-I, BES, ... | ||
| Analysis Requirements: | The same arsenal of tools used in L-mode plasmas | ||
| Other Requirements: | |||
| Title | 76: Diagnostic spatial cross calibration using edge sweeps in QH-mode | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | Pedestal | 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: | n 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 machine. 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. 7 NBI sources. | ||
| Diagnostic Requirements: | CER, MSE, BES are essential. Standard profile diagnostics are also needed. ECE-I for EHO studies. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 77: Compatibility of radiative divertor operation with ITER baseline scenario | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | ITER expects to use impurity seeding to mitigate the steady-state heat load in the divertor. Little information exists on the impact on the core performance. Various gases and values of B should be tested to see the effects of optimal and non-optimal matching of the radiative states to the edge conditions. Tests with the ion grad-B drift into and out of the divertor would be helpful to understand the role of drifts. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | Part of IOS-1.2 | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Strongly desirable is the new IR periscope | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 78: Collisionality scaling of energy confinement in advanced inductive plasmas | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Analysis of the ITPA joint experiment database on advanced inductive plasmas showed a strong dependence of the H98 factor with decreasing collisionality As ITER is far in that direction, understanding whether this is a true scaling or an artifact of correlations in the database would be valuable to accurately projecting to ITER. The collisionality scan is quite simple on DIII-D. It entails maintaining constant density while varying I and B together by about 40% for a factor of 4 scan. This is fairly straightforward with the pumping capability in DIII-D. If possible, it will be done in conditions similar to the JET/DIII-D rho* scan so that the results can be interpreted together. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | IOS-4.3 joint experiment. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 79: Compare edge particle transport in ELMing H-mode and QH-mode with and without NRMF | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure the edge impurity particle transport in ELMing H-mode plasmas and contrast it with the edge particle transport in QH-mode both with and without n=3 nonresonant magnetic fields (NRMF). This will allow us to determine separately determine the net edge loss due to ELMs, EHO and EHO plus NRMF. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create a QH-mode plasma with moderately low toroidal rotation which can be run both with and without NRMF. Inject LiF pellets into QH-mode phases both with and without NRMF. Turn off NRMF and raise density until ELMs return; perform same measurements in ELMing H-mode. These measurements can be used as part of other QH-mode parameter scans to map out particle transport as a function of those parameters. | ||
| Background: | QH-mode is an extremely attractive operating mode for future devices, since it exhibits H-mode confinement combined with steady-state operation without ELMs. Work in the 2009 and 2010 campaigns on D III-D has demonstrated QH-mode operation with zero-net NBI torque by replacing the NBI torque with torque from neoclassical toroidal viscosity produced using nonresonant n=3 magnetic fields. A key part of developing QH-mode with NRMF as an operating scenario for future devices is developing a predictive understanding of the edge particle transport. To develop a predictive understanding, we need to be able to measure the edge particle transport. Previous experiments in QH-mode without NRMF have used injection of pellets doped with LiF [K.H. Burrell et al, Phys. Plasmas 12, 056121 (2005)]. The essential features of LiF are 1) lithium and fluorine do not recycle, thus allowing direct measurement of the edge loss rate from the decrease of core density and 2) LiF contains a low and a moderate Z element, allowing determination of the loss rate as a function of Z. | ||
| Resource Requirements: | Lithium fluoride injection using shell pellet injector or lithium pellet injector. | ||
| Diagnostic Requirements: | All standard profile and fluctation diagnostics, especially edge BES and ECE-I for EHO studies.CER tuned to Li or F lines for impurity transport study | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 80: Access conditions for ITER advanced inductive operation | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Analysis of an ITPA joint database of JET and DIII-D advanced inductive discharges is now underway. Certain operational conditions have been identified in the DIII-D dataset and the JET dataset for access to advanced inductive operation. Conversion to physics conditions is being attempted now. This proposal is a placeholder to indicate interest in testing the physics hypotheses formulated from the on-going analysis in the next few months. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | IOS-4.2 joint experiment. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 81: Develop low-torque, high normalized fusion performance QH-mode for ITER/FNSF | ||
| Name: | Garofalo | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | K. Burrell, W. Solomon | ITPA Joint Experiment : | No |
| Description: | Demonstrate QH-mode operation at high values of βN and low values of q95 (for high β), and with reactor relevant values of the NBI net torque. We plan to use the magnetic counter-Ip torque mainly 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 raising betan, lowering q95, and increasing the co-Ip NBI. | ||
| 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. | ||
| Resource Requirements: | In order to maximize the NBCD from the tilted beamline, forward IP with reversed BT will be used. In order to maximize the NRMF counter-IP torque, 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 odd-parity n=3 operation. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | This experiment should be scheduled immediately after a boronization. | ||
| Title | 82: RMP Response Above No-Wall Limit | ||
| Name: | Wade | Affiliation: | ORNL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The Ideal MHD plasma response is expected to increase non-linearly with plasma pressure above the no-wall limit. Measurements of how the plasma responds to applied n=3 RMPs as the plasma pressure is increased in this regime should provide clues as to ideal vs vacuum response. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize the 3 s stationary phase of betaN = 3.5 discharges developed in the Steady-State Core Integration Working Group and apply toroidal phase modulated n=3 RMPs to assess the plasma response. BetaN can be scanned from 2.5 to 3.5 to determine dependence. | ||
| Background: | |||
| Resource Requirements: | 6 Co-NBI Source
6 Gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 83: Assess steady-state conditions with far-off-axis NBI | ||
| Name: | Wade | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | Utilize small, upper shifted plasma used for off-axis NBI measurements in 2008 and develop scenario with 4 NBI sources driving current at rho = 0.5 and two far off-axis NBI source driving current as far out as rho =0.75. Compare with discharge that is centered in the vessel with nominally all NBI aimed on-axis. Shape should be possible with higher plasma current if needed for reduced NB orbit losses and for sufficient duration to test steady-state capabilities. Main issue is whether sufficiently high beta can be achieved in such a shape. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 84: RMP Response with Counter NBI | ||
| Name: | Wade | Affiliation: | ORNL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | Utilize toroidal phase modulations of applied n=3 RMP to assess variation in plasma response as rotation is systematically varied from co-rotating to counter-rotating. Because electron perpendicular velocity cannot be zer in counter-rotating cases, this should provide a test of whether island-like structures can be observed in these cases. Do not insist on ELM suppression as this has shown to be problematic at low rotation. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 85: Turbulence and Transport Variation with Mach number and Shear in Hybrid discharges | ||
| Name: | McKee | Affiliation: | U of Wisconsin |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | C. Holland, C. Petty, T. Rhodes, L. Schmitz, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Study 2D turbulence structure, magnitude, spectra, and its likely suppression as a function of rotationally varied ExB shear in long-duration, moderate to high beta hybrid H-mode plasmas via co- and near-balanced NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | An experiment to systematically study turbulent eddy structure as a function of Mach number in hybrid discharges will help address these issues by directly measuring eddy structure, 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. The long-duration, stationary hybrid discharges make these an excellent platform in which to study the turbulence characteristics. The low-amplitude of fluctuations in the core of hybrid plasmas makes their study more challenging, but the stationary qualities (several seconds) allow for ensemble-averaging of the 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.
In terms of the experiment, discharges similar to those already developed by C. Petty et al. would be used, with the exception that the neutral beams used for beta feedback will need to be changed to allow for the BES measurements (which require the 150 left (preferably, 150 R if necessary) beam on steady). We will run relatively high q95 (~5.5) to increase turbulence magnitude. Several repeated discharges would be performed for full radial measurements. | ||
| Background: | 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-Nuclear Fusion, 2008). This has been shown to be consistent with the a reduction in ExB shearing at lower Mach number from GLF23 modeling. Previous measurements of turbulence characteristics in hybrid discharges (McKee, APS-2005) with the upgraded BES diagnostic, showed that turbulent eddies exhibit a strongly tilted structure in co-injected hybrid discharges. This is in sharp contrast to the more radially-poloidally symmetric eddy structure typically observed in the core of L-mode discharges. The direction of this tilted eddy structure is consistent with the ExB shear flow in these plasmas, although it was questionable as to whether the shear magnitude could bring about such a strong eddy tilt. | ||
| Resource Requirements: | All NBI sources, flat injection | ||
| Diagnostic Requirements: | BES, DBS, PCI | ||
| Analysis Requirements: | Lots...TGLF, GYRO | ||
| Other Requirements: | |||
| Title | 86: Confinement dependence of impurity-seeding in H-mode plasmas | ||
| Name: | McKee | Affiliation: | U of Wisconsin |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | G. McKee, 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: | |||
| Diagnostic Requirements: | BES (8x8), UF-CHERS, DBS, CECE, FIR, PCI, etc. | ||
| Analysis Requirements: | TGLF, GYRO | ||
| Other Requirements: | Neon, Argon and (possibly) N2 gas injection | ||
| Title | 87: Beta limit of ITER shape steady-state scenario | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Previous experiments on the steady-state scenario in the ITER shape gave some surprising results in the overall beta achieved and the height of the pedestal achieved. With the availability of the off-axis NB and slightly more EC power, it would be good to revisit these results and extend them to map the beta limits as a function of q95, min(q), changes in bulk rotation, and wall-plasma separation (changing the scale factor). This does not need to be done at fully non-inductive conditions so the ctr-NB can be used and the restriction on pumping can be relaxed. Data should inform directly modeling studies of ITER, specifically those trying to explore the needs for H&CD system upgrades. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | IOS-3.1 joint experiment. | ||
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| Title | 88: Access to steady-state scenario conditions in ITER | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | Steady-state scenarios presently envisioned for ITER require elevated min(q) at start of flattop (SOF) and start of burn (SOB). This experiment would explore closed-loop control in the current rise (ARTAEMIS and Lehigh methods) to reach different target q profiles at SOB, then explore the beta limits and potential for fully non-inductive operation of those profiles in the ITER shape. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | IOS-3.2 joint experiment. | ||
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| Title | 89: Coupling of FW to ELM suppressed H mode plasmas | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | At present, the baseline scenario for ITER is a low-q95 H mode with RMP ELM suppression/mitigation. ICRH is a day-1 heating scheme, but there is concern that the 15 cm outer gap will require active techniques to ensure adequate coupling. The standard active technique is gas puffing, either globally or in the vicinity of the FW antenna. This experiment proposes to operate a standard H mode near the ITER parameters in the standard RMP ELM suppression q95 window for DIII-D, then vary the outer gap and toroidal phase of the RMP to look for axisymmetric and non-axisymmetric variations in the SOL near the antenna and measure the loading. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | IOS-5.2 joint experiment | ||
| Resource Requirements: | |||
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| Title | 90: Demonstration of burn control of the ITER baseline scenario | ||
| Name: | Luce | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | Yes |
| Description: | Entry into burn (SOB in ITER language) is a point where excursions in stored energy could lead to tearing modes or giant sawteeth that prevent the burn from proceeding as desired to meet the physics objectives. The ITER heating systems need to have defined physics requirements for power control to see what mix of systems are suitable for this critical stage of the plasma evolution. Exit from burn (EOB) is also critical. This experiment will focus on developing physics requirements that can be translated to ITER from DIII-D experiments. If time permits, this experiment should address the issues of when to turn on and off the RMP for ELM mitigation and integrating control of the radiative divertor operation at SOB and EOB. But the first priority is avoiding core instabilities. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | IOS-6.3 joint experiment. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 91: Optimization of castellation for a new W-divertor of ITER: studies of shaped castellation in DIII-D | ||
| Name: | Litnovsky | Affiliation: | Juelich |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov (UCSD), 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 for collect the deposited material at the bottom of castellation. The gaps and instrumented plates will be analyzed in the American and European laboratories. | ||
| 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: | Machine Time: One week piggyback exposure using DiMES manipulator system, preferably LSN operation, NBI-heated ELMy H-mode. | ||
| Diagnostic Requirements: | DiMES TV, floor Langmuir probes, in particular the probe at the DiMES radial location, MDS chord looking at DiMES. | ||
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| Other Requirements: | |||
| Title | 92: Far off-axis NBCD | ||
| Name: | Park | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Evaluate far off-axis NBCD physics | 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 | 93: E-GAM Physics On DIII-D | ||
| Name: | Chen | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Requested |
| Co-Author(s): | R.Nazikian, W.W. Heidbrink, G.J. Kramer, D.C. Pace, R.K.Fisher, M.A. VanZeeland and C.M. Muscatello | ITPA Joint Experiment : | No |
| Description: | E-GAM Physics on DIII-D:<br> 1) In fusion reactor it may be possible to produce E-GAMs in the core of high q-min plasmas, relevant to proposed steady state plasma regimes.<br> 2) E-GAM is a remarkable instability for detailed understanding of nonlinear dynamics of wave-particle interactions.<br> --- Large scale, large amplitude, low frequency, large loss<br> 3) E-GAM observed at high qmin on DIII-D:<br> --- Strongly driven by counter-injection<br> --- Drive and loss dominated by counter going beam ions near loss cone<br> --- Strong coherent losses observed at FILD, as well as incoherent losses<br><br>Key questions:<br> 1) To what extent does the coherent loss represent first orbit loss?<br> --- Toroidal source dependent coherent AE losses are observed<br> 2) Does E-GAM lead to anomalous heating of ions due to ion Landau damping on thermal ions?<br> --- Mode driven to large amplitude, <~n/n>~10%<br> 3) SPIRAL predicts that co injection damping of mode is strongly dependent on amplitude of mode due to fractional resonances. Can we validate it experimentally? | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) E-GAM is often observed when counter-going beams are injected in the current ramp-up phase of the discharge.
2) Operate in reverse Ip, to modulate 30L, 330L for mode drive and diagnostic data and to keep 150L full on for BES. This will give about 5 MW power for full counter injection E-GAM excitation needed for large mode amplitude. 3) 10ms beam pulses from 30L and 330L will be injected alternatively with constant power during a plasma current ramp. This will allow E-GAM losses toroidal dependence checkout. --- Also allow toroidal dependence checkout of AE coherent losses during counter-injection 4) Use fast CER to determine if anomalous thermal ion heating results through the bursting phase of the E-GAM. Use main ion CER and compare to carbon. 5) Shift the start time of the beam to reduce qmin and reduce E-GAM amplitude with similar target plasma conditions. See if thermal Ti/Te is lower with the lower mode amplitude. 6) Use 210R & 210L (modulate) (now co beams) keeping constant counter injection power to see if damping of E-GAM depends on E-GAM amplitude. ---Strength of fractional resonances depends on mode amplitude, as predicted by SPIRAL. ---Use FIDA to investigate the stochastic fast ion transport | ||
| Background: | -- | ||
| Resource Requirements: | 30L, 150L, 330L, 210L, 210R | ||
| Diagnostic Requirements: | FILD, CER, BES, FIDA, neutrons, core plasma diagnostics | ||
| Analysis Requirements: | EFIT, Reverse Orbit Modeling, TRANSP, SPIRAL, NOVA | ||
| Other Requirements: | One of the FILDs need to be modified (such as rotating the aperture) to allow collecting lost ions in reversed Ip plasma. And favorable plasma shape will be used according to which FILD is modified (i.e. oval for FILD1, circular for FILD2, generally up-down symmetric plasma is preferred for NOVA run). However, experiment can be redesigned if the modification is not provided. | ||
| Title | 94: 2-D Impurity Transport Studies of High-Z Impurities | ||
| Name: | Reinke | Affiliation: | U of York |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | David Pace
Eric Hollmann |
ITPA Joint Experiment : | No |
| Description: | The poloidal variation of the high-Z impurity density is to be studied in the core of Ohmic, electron cyclotron and neutral beam-heated plasmas.
The results of this experiment should help the community better understand neoclassical parallel impurity transport, understanding necessary to accurately predict poloidal impurity rotation. Additionally, recent studies of anomalous radial transport have shown radial impurity peaking to be sensitive to poloidal variation of impurity density. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental approach is relatively straightforward. A high-Z impurity (Kr or Xe) will be injected into the current flattop and the 2-D structure of the steady-state emissivity profile measured with soft x-ray tomography and bolometry.
ECH and NBI will be added and the changes in the poloidal structure documented. | ||
| Background: | Poloidal variation of high-Z impurities has been observed on a number of tokamaks. Centrifugal force is well known to lead to accumulation of impurities on the outboard side of the plasma, but small poloidal electric fields can also lead to large asymmetries. Weak poloidal electric fields can be sustained by cyclotron heating and neutral beam injection. When heated with cyclotron waves, a particles perpendicular energy is increased, enhanced magnetic trapping. This leads to an in/out density asymmetry, requiring a poloidal electric field to maintain quasineutrality. For NBI, the orientation of the beams impact the poloidal electric field, with more radial beams injecting a larger fraction of particles into trapped orbits.
Recent measurements of Mo asymmetries in ICRH plasma on C-Mod have led to extension of neoclassical parallel impurity transport theory, taking into account energetic particle driven poloidal electric fields. The goal of this research is to test this theory against ECH and NBI-driven in/out flux-surface asymmetries of high-Z impurity density on DIII-D. | ||
| Resource Requirements: | 5 plasmas, each with an Ohmic and auxiliary-heated phase.
x3 w/ ECH, scanning deposition from HFS->on-axis-> LFS. (if possible, intra-shot scan would be preferred) x2 w/ NBI, with toroidal NBI and another with (more) radial NBI Specific target plasmas are not sufficiently important and this research could piggyback on other plasmas if high-Z impurity puffing was allowed, perhaps near the end of a discharge. | ||
| Diagnostic Requirements: | Soft x-ray tomography and resistive bolometer tomography to view 2-D radiation patterns.
Charge exchange spectroscopy to find the toroidal rotation and ion-temperature profiles. In ECH and Ohmic plasmas, beam-blips can be used. Electron density and temperature profiles. Available fast-ion diagnostics to determine in/out asymmetry of fast-ion density in NBI plasmas. Zeff, other impurity monitoring | ||
| Analysis Requirements: | The existing tomography tools should be sufficient to determine variation of high-Z impurity emission on a flux surface.
The first cut analysis will compare results to predictions based on the centrifugal force acting alone. | ||
| Other Requirements: | Seeding of a high-Z impurity (Kr or Xe gas puffing) which is a non-standard technique on DIII-D. | ||
| Title | 95: Effect of RMP on L-H transition power threshold | ||
| Name: | Yan | Affiliation: | U of Wisconsin |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G. McKee, P. Gohil , P. Diamond | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the physics of the RMP effect on L-H transition power threshold. It will also provide an experimental test of the theoretical prediction that RMP damps zonal flow. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main idea of the experiment is to scan RMP amplitude across L-H transition at different L-mode densities and q95. It will be LSN, low triangularity plasma. Diagnostics like BES, UF-CHERS, DBS, CECE and RS probe wil be used to measure the turbulence spatial structure and other characteristics from the outer core, across the pedestal, and into the SOL. The outcome of the experiment will provide for investigation of the physic of RMP effect on L-H transition power threshold and a test of the theoretical prediction that RMP damps zonal flow. | ||
| Background: | MP has been shown in many experiments as an effective way to mitigate ELMs, and the commissioning of ELM control is planned for ITER. Therefore it is important to investigate the degree to which the MP affects the H-mode threshold power and understand the physics behind. Previous studies have shown an increase of the L-H power threshold with RMP above some level [1]. Theoretical prediction also showed that RMP damps zonal flow [2]. By doing this experiment we can exam how RMP affects turbulence and flow, which will help understand the physics of the effect of RMP on L-H transition power threshold.
[1] P. Gohil, et al., Nuclear Fusion 51, 103020, 2011 [2] M. Leconte, et al., APS 2011 invited talk, 2012 | ||
| Resource Requirements: | 4 neutral beams | ||
| Diagnostic Requirements: | BES, UF-CHERS, CER, TS, DBS, CECE, PCI, Reynolds stress probe | ||
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| Title | 96: Collisionality and q95 scaling of the coupled turbulence/zonal flow during LH transition | ||
| Name: | Yan | Affiliation: | U of Wisconsin |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | George McKee, Jose Boedo, Dimitry Rudakov, Rich Moyer, L. Schmitz | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the collisionality and 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 electron density to vary collisionality at different momentum input , as well as a q95 scan 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: | 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- H transition [1]. It is shown that the zonal flow can be damped linearly through collisions and have strong q95 dependence. L-H transition power threshold also depends strongly on density.
In the 2011 campaign a density scan has been performed, but mainly focused on limit cycle condition. Moreover, the density scan was incomplete. The lowest density condition was not obtained, which will be very interesting since the L-H power threshold increases again at lower density. No q95 scan has been done. By completing these it will help understanding the underlying physics of L-H transition and power threshold scaling, which is a key issue for ITER. [1] K. H. Burrell, Phys. Plasmas 4, 1499 (1997) | ||
| Resource Requirements: | 4 neutral beams | ||
| Diagnostic Requirements: | BES, UF-CHERS, CER, TS, DBS, CECE, PCI, Reynolds stress probe | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 97: Expand the high li, betaN >4 operating regime through instability avoidance and higher heating power | ||
| Name: | Ferron | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make use of ECCD stabilization of 2/1 tearing modes and modifications in the discharge evolution during the betaN ramp up in order to extend the high-performance pulse length and allow operation at lower values of q95. Take advantage of the additional neutral beam made available in 2010 and the sixth gyrotron to push betaN above 5 and test the effect of wall stabilization at high li. Make measurements of the fast ion profile in order to understand anomalous losses. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In order to operate at lower values of q95, it is necessary to avoid the early, fast-growing n = 1 mode. Stability of previous discharges is still being studied in order to understand this mode, but its occurrence is likely coupled to the current profile which has a region of negative flux surface average current just inside the H-mode pedestal which is a result of the negative surface voltage produced by the Ip feedback system because of the current overdrive by the noninductive current. One approach would be to avoid this negative current by holding the surface voltage to more positive values (a technique also used in fNI = 1 discharges in TCV). Holding the surface voltage at 0 would also allow a clear demonstration of noninductive current overdrive. The other possibility is to modify the time evolution of beta and density during the beta ramp up. ECCD would be used to avoid the 2/1 mode that terminates the high performance phase. | ||
| Background: | In 2008, high li discharges with betaN >4.5 were obtained that had fNI = 1.2 and betaN above 4 for 1 s. Bootstrap current fraction was above 80%. In the early portion of the high beta phase when li was near 1.4, even with betaN = 4.5 the discharge was operating below the no wall n = 1 ideal stability limit. BetaN was limited by available heating power. The duration of the high-performance phase was limited by onset of a 2/1 tearing mode. Best performance was obtained with q95 near 7. At lower values of q95, the high beta phase was terminated during the beta ramp up by a fast growing n = 1 instability. Comparisons with ONETWO indicate significant anomalous fast ion loss, possibly a result of semicontinuous 1/1 mode activity. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Would make use of FIDA. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 98: Maintaining high li at high betaN using RMP and near-axis current drive | ||
| Name: | Ferron | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make high li, betaN >4 discharges more stationary by reducing the rate of decrease of li through replacement of ohmic current near the axis with ECCD and by using RMP to reduce the H-mode pedestal density in order to reduce the edge bootstrap current. Results would be used to determine what would be required to produce a true, steady-state, high li, high betaT discharge. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | There are two primary parts to this experiment. First, the deposition profile of the ECCD would be varied in order to determine its effectiveness in replacing the core ohmic current. A accompanying goal would be to determine if q(0) can be raised slightly in order to avoid 1/1 activity. The new off-axis beam injection capability may be useful for tailoring the NBCD profile near the axis. It is unlikely that, even with six gyrotrons, the ohmic current can be completely replaced in the core. However, we can obtain scaling information in order to compare with models and determine how much external current drive would be necessary to make a stationary high li discharge. In the other part of the experiment, the RMP fields would be used to reduce the H-mode pedestal pressure and, particularly, the density. Supposedly there is a window for ELM stabilization with q just above 7 which matches the best high li discharge from 2008. However, ELM stabilization isn't necessarily needed, and the RMP fields are reported to have an effect on the pedestal height even away from the resonance required to stabilize ELMs. The best discharges in 2008 were strongly overdriven with noninductive current which may have contributed to the rate of decrease of li. Some exploration of the balance between total plasma current and the noninductive current would be done in order to test the effect on the li decrease. | ||
| Background: | In the fNI >1 discharges produced in 2008, li decreases slowly as the trapped ohmic current decays in the discharge core and the bootstrap current builds in the H-mode pedestal. In the best performance discharge, because Ip was relatively low (q95 about 7), and the noninductively driven current exceeded the total programmed current, the edge surface voltage was driven negative. | ||
| Resource Requirements: | |||
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| Title | 99: Dependence of confinement and stability on toroidal rotation in high li discharges | ||
| Name: | Ferron | 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 | 100: Fast ion diffusion in high betaN, steady-state scenario discharges | ||
| Name: | Ferron | 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 comparison of predicted and measured fast ion density profiles in the steady-state scenario at high betaN. Optimize the discharges for all of the fast ion diagnostics including FIDA and the fast ion loss detectors. Compare losses from on-axis and off-axis beams. Make use of techniques such as short beam pulses to look at the rise and fall times of the neutron signals. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | See the "Description" paragraph. | ||
| Background: | Analysis of the high betaN steady-state scenario discharges requires calculation of the noninductive current density profiles from models. In order to calculate the neutral beam current density profile, the beam deposition profile must be calculated and this calculated profile must be assumed to be correct. However, typically if the measured thermal pressure and the calculated fast ion pressure profiles are summed, the on-axis pressure and the total stored energy are inconsistent with equilibrium reconstructions using EFIT with magnetics and MSE data. It is necessary to assume an anomalous fast ion diffusion profile in order to obtain agreement. It is highly desirable to evaluate whether this technique of using an anomalous diffusion profile actually produces fast ion density profiles from the model that match the experiment. Otherwise, analysis of the steady-state scenario discharges and calculation of the noninductive current fraction involves significant uncertainty because the total neutral beam driven current is not well known.
The problem is particularly significant at high betaN. | ||
| Resource Requirements: | |||
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| Title | 101: Maximize betaN and fBS with direct tearing mode stabilization in steady-state scenario discharges | ||
| Name: | Ferron | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The primary goal is to produce high betaN, fNI = 1 discharges for long duration. Part of the approach to achieving this goal would be to develop the capability to preemptively stabilize tearing modes in steady-state scenario discharges using ECCD narrowly deposited at the resonant surface. With this capability activated, push the betaN to the maximum attainable in order to study the limit determined by modes other than resistive tearing modes and in order to maximize the bootstrap current fraction. With tearing modes stabilized, maximize the duration of fNI = 1 at high betaN. With some or all of the gyrotrons dedicated to narrowly deposited ECCD, there will be reduced or zero current density in the broadly deposited noninductive current density profile normally provided by ECCD. Use the off-axis neutral beams to replace the broad ECCD profile normally used. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dedicate the first portion of the experiment to development of the techniques for direct tearing mode stabilization using ECCD. Apply the standard methods of search and suppress and active tracking. Preferably, use real-time aiming of the gyrotron steering mirrors rather than plasma motion or toroidal field changes in order to adjust the ECCD deposition profile to lie on the resonant surface. Standard steady-state scenario discharges at high betaN would be used as the targets rather than a custom discharge with a reproducible tearing mode. It is highly likely that tearing modes will appear often in the high betaN discharges. Once there is confidence that tearing modes can be avoided, the primary goal of the experiment would be to produce fNI = 1 discharges for long duration. The approach would be to maximize fBS by increasing betaN. The direct stabilization would be counted on to maintain tearing mode stability both as betaN is increased and as the current profile evolves during the increased duration of the discharge. | ||
| Background: | Tearing modes (2/1 and/or 3/1) impose the betaN stability limit and limit to discharge duration in steady-state scenario discharges. Progress toward developing fNI = 1 discharges has been very slow because most discharges are interrupted by the growth of a tearing mode. It is essential that we do something to eliminate these modes. The general approach has been to either assume that the resonant surfaces will not be present if q_min is maintained at a high enough value or that tearing modes can be avoided using broadly deposited ECCD. Tearing modes have not been reproducibly avoided using these techniques, though. A proven technique for avoiding tearing modes is direct stabilization with ECCD narrowly deposited at the resonant surface. Very precise aiming is required but searching techniques to find the correct aiming have been developed. This technique may be very valuable in steady-state scenario discharges at high betaN, but the work to test this has never been done. The hesitation to use the direct stabilization technique has stemmed from the need to use the available gyrotron power to provide a broad profile of off-axis noninductive current density in order to produce a total noninductive current profile compatible with stationary, fNI = 1 operation. However, now that the off-axis beams are available, it may be possible to divert some or all of the gyrotrons to tearing mode stabilization and still have enough off-axis noninductive current drive. | ||
| Resource Requirements: | Real-time steerable mirrors are highly desirable. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 102: Beta limit and bootstrap current fraction in ITER steady-state scenario discharges | ||
| Name: | Ferron | 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. A portion of this experiment addresses ITPA IOS group high-priority experiment 3.1. Increase the plasma current over what has been used previously to push q95 down to 5 in order to reach conditions that project to Q = 5 in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce the fNI = 1 discharges produced in 2008 and vary q_min, beta and density gradient in order to test the effect on the achieved bootstrap current and beta limit. Use the ECCD to better advantage to avoid 2/1 tearing modes in order to either raise the achievable betaN or establish the maximum betaN value as determined by ideal stability. Do this in a discharge shape that better matches the ITER scaled shape in the outer, lower squareness. Using shape adjustments, modify the density by taking advantage of the divertor cryopump. As conditions are varied, test the effect of the outside gap on the betaN limit. Make use of the off-axis beams to improve the capability to reach elevated values of q_min. | ||
| Background: | During 2008 the first attempts were made at making a fNI = 1 discharge in a scaled ITER shape in DIII-D. FNI = 1 was successfully obtained at relatively low betaN = 3.1 with fBS = 0.7. The beta limiting instability was a 2/1 NTM and the outside gap seemed to have a moderate effect on the achievable beta. This contrasts with the double null shape steady-state scenario discharges which had less density gradient and correspondingly less bootstrap current but which have been operated at betaN = 3.7 without a 2/1 NTM. The discharge shape that was used doesn't quite match the intended ITER scaled shape. | ||
| Resource Requirements: | Real-time steerable mirrors are highly desirable. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 103: Simultaneous achievement of the non-inductive and Q=5 goals of steady-state scenario in ITER shape | ||
| Name: | Park | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate fully non-inductive and Q=5 condition in ITER shape using off-axis beam and higher power ECH | 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 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 | 104: Transport and turbulence validation in high-density, low fast ion fraction QH-modes | ||
| Name: | Holland | Affiliation: | UCSD |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | K. H. Burrell, L. Schmitz, T. Rhodes, G. Wang, G. McKee, R. Prater, S. P. Smith, C. Petty, G. Staebler, J. Kinsey, R. Waltz | ITPA Joint Experiment : | No |
| Description: | Perform an ECH heating scan starting from a baseline high density, low beam ion fraction QH-mode (nominal target: 131920 4200 ms) with T_e/T_i ~ 1, and use to test gyrokinetic & gyrofluid turbulence & transport predictions. Results will contribute to 2012 transport JRT and help better understand issue of L-mode near-edge "shortfall" (e.g. does it occur in H-mode conditions). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -Begin by reproducing conditions in shot 131920 4200 ms, with possible alterations to lower Z_eff if possible.
- Perform series of discharges with increasing near-axis ECH heating applied in additon to beams, identify maximum level which can be run steady-state -Perform series of repeat shots scan fluctuating diagnostic across plasma to obtain radial profiles of low and intermediate-k turbulence measurements | ||
| Background: | A 2010 TMV task force validation experiment focused on validating turbulence and transport predictions against an ECH power scan in "standard" QH-modes. Initial model predictions presented at 2010 APS did not include the significantly stabilizing dilution effect from from the large fast ion population (n_fast/n_e ~ 25% at rho = 0.6), and overpredicted transport by 10x or more. New simulations incorporating this dilution yield significantly lower transport levels closer to experriment. GYRO simualtions of 131920 4200 ms (with n_fast/n_e ~ 5%) yield transport levels consistent with power balance. Therefore propose to repeat 2010 validation experiment using high-ne conditions as baseline.
Additional motivations: -will contribute to 2012 JRT on electron transport -2010 experiments have Ti/Te ~ 2 at rho=0.6 with and without ECH input. Target discharge has Ti/Te ~ 1.1-1.2 due to increased n_e. | ||
| Resource Requirements: | -standard QH-mode setup (reverse Ip etc)
- co and counter NBI - full ECH capability | ||
| Diagnostic Requirements: | -profiles: reflectometer, Thomson, CER, ECE
- turbulence: BES, ECE, DBS | ||
| Analysis Requirements: | CER, kinetic EFITs, Er analysis, characterization of turbulence (spectra, correlation lengths, etc.)
Significant TGLF and GYRO modeling | ||
| Other Requirements: | |||
| Title | 105: Measurement of electron shielding of beam ion current | ||
| Name: | Park | 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. | 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 | 106: Develop low-torque, high normalized fusion performance QH-mode for ITER/FNSF | ||
| Name: | Garofalo | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | K. Burrell, W. Solomon, M. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | Demonstrate QH-mode operation at high values of N and low values of q95 (for high ), and with reactor relevant values of the NBI net torque. We plan to use the magnetic counter-Ip torque mainly 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 : | Yes |
| Experimental Approach/Plan: | The plan is to further improve on the Nov. 2011 experiments [D3DMP No.: 2012-31-01] by exploring higher betan, lower q95, and higher co-Ip 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. | ||
| Resource Requirements: | In 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: | This experiment should be scheduled immediately after a boronization. | ||
| Title | 107: Turbulence and transport model validation in low-rotation plasmas | ||
| Name: | Holland | Affiliation: | UCSD |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | G. R. McKee, T. Rhodes, K. Burrell, R. Waltz | ITPA Joint Experiment : | No |
| Description: | Use combination of balanced injection and ECH "rotation pumpout" to minimize rotation (and rotation shear) in high density L-mode. Goal is to test turbulence & transport predictions in plasmas with strong ion heating but low shear suppression. As much as possible, tailor plasma to maximize gyroBohm flux normalization Q_gB = n*T*c_s*(rho_star)^2 such that normalized fluxes are as low as possible, pushing plasma as close to marginal stability as possible. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Starting from balanced injection cases of 2006 L-mode rotation expt. by McKee, optimize to maximze Q_gB/minimize Q/Q_gB using ECH, density, B_field, etc.
Perform repeat serires to obtain radial profiles of fluctuation measurements. H-mode analog possible? | ||
| Background: | -Most validation studies at DIII-D utilizing beam-heated plasmas have not minimized rotation/shear suppression. Finite shear changes thresholds for stability, particularly low-k ITG. It is therefore hard to validate model predictions of the ITER/reactor-relevant case of near-marginal transport at low rotation. Have studies of electron dominated, ECH-only plasmas with small rotation- should complement with measurements of ion-dominated cases. | ||
| Resource Requirements: | NBI (co- and counter), ECH | ||
| Diagnostic Requirements: | Full profile and fluctuation sets. | ||
| Analysis Requirements: | CER, Er, EFIT postprocessing, GYRO and TGLF after that. | ||
| Other Requirements: | |||
| Title | 108: Compare FW, EC, NB in Advanced Inductive Scenario using power modulation | ||
| Name: | Ryan | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | R. Pinsker, J. Hosea, G. Taylor, D. Rasmussen, M. Murakami, A. Nagy, S. Diem, M. Kaufman, R. Perkins | ITPA Joint Experiment : | No |
| Description: | Ascertain the effectiveness of replacing on-axis NBI with FW H&CD in AI discharges, potentially freeing up NB for future off-axis injection. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Individual comparison of heating systems: Apply incremental (~2 MW), modulated power from each of the three heating systems (FW, EC, and NB) to a baseline AI plasma (e.g., 146571) for comparison. Fast modulation (~50 Hz) for central heating measurements (delta-T, delta-W break in slope) and slow (~2-10 Hz, chosen to avoid ELM frequency) for MSE measurements of current drive.
2. Constant power comparison: sequentially apply ~2 MW of incremental power from each H/CD system into AI plasma; alternate the system order for succeeding shots. 3. Synergistic comparison: overlap FW and EC pulses for constant power (i.e., 2 MW FW, 1 MW FW+1 MW EC, 2 MW EC). | ||
| Background: | This is a continuation of the AI work in 2011 (notably shot 146571) which compared 1.5 MW incremental EC and FW on top of 8 MW NB and 1.5 MW EC. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 109: Central safety factor (q0) control with FW | ||
| Name: | Ryan | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | R. Pinsker, J. Hosea, G. Taylor, D. Rasmussen, M. Murakami, A. Nagy, S. Diem, M. Kaufman, R. Perkins | ITPA Joint Experiment : | No |
| Description: | Use FW to make discernible changes in q0/qmin; determine control parameters for q change. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish the required central driven current required to optimize q(0) by driving one 90 MHz system in co-CD phase, the other in cntr-CD, and varying the power ratio between them. For a constant power scan this requires operation at 50% of full power capability. Assuming the optimum driven current will be determined to be in the counter-current direction, the counter-CD antenna will be then be operated at full power, and the powers and phases of the remaining two antennas will be adjusted to drive this current at the highest possible power. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 110: Show we can increase betaN in the weak shear scenario with one off-axis beam | ||
| Name: | Holcomb | Affiliation: | LLNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Primary goal: Achieve betaN>4 and qmin>1.5 in weak magnetic shear DN plasma using off-axis NBI and ECH. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Reproduce 147634 that had qmin>1.5, betaN=3.5 for 3 s, fNI=0.7, fBS=0.4, peaked johmic profile, G=0.3, q95=5.2, using 10.75 MW co-Ip NBI and Bt=1.65 T.
2. Turn voltage up on all beams to achieve max 14.2 MW observed in 2011. With H89=2.1, this scales to a max betaN=3.9. 3. Turn Bt down to 1.4 T. Turn Ip down to 0.89 MA to keep q95=5.2. With 14.2 MW co-Ip NBI and H89=2.1, this scales to a max betaN=4.5. Reaim ECH for on-axis 3rd harmonic heating. Start high power phase earlier to compensate for anticipated faster qmin evolution. 4. Scan high power phase start time earlier to scan up the qmin being tested, and evaluate achievable max betaN vs. qmin. Evaluate tearing stability & off-axis fishbone stability. Use MHD spectroscopy to measure the plasma response to n=1 (and n=2, if possible). In the high power phase, always use maximum off-axis beam tilt and power allowed to maintain the broadest profiles. At each step, try using ctr-Ip beams at least once. Also evaluate the best turn-on time for ECH: early or at the start of the high power phase. | ||
| Background: | It is critically important to show access to high betaN for the long term success of the steady state approach in a tokamak, both for fusion gain and for maximizing the bootstrap current fraction.
The off-axis NBI was provided chiefly to access broader current and pressure profiles that are predicted to have higher ideal MHD betaN limits. Experiments in 2011 showed we can create broader profiles with predicted ideal-wall n=1 betaN limits over 4. However, we were not able to achieve betaN this high with qmin>2 due to low confinement, and experiments at lower qmin (>1.5) did not have a primary goal of maximizing betaN. We need to show incremental progress in the ability to obtain higher betaN in elevated qmin, weak shear discharges to inform a decision about a 2nd off-axis beam in the future. With the power available in 2012, we will need to turn down the B-field to do this. The reference shot mentioned above has a peaked ohmic current density during the high power phase (nvloop & onetwo analysis), so additional on-axis NBI power should not overdrive the current profile. | ||
| Resource Requirements: | ALL NBI, but chiefly co-Ip beams. Elevated beam voltages on 30rt, 330's, 150's. 3+ MW of ECCD/ECH. | ||
| Diagnostic Requirements: | MSE, CER, TS, MHD spectroscopy, ECE, Reflectometer, BILD, filterscopes, Asdex gauges, C02, | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 111: FW Loading comparison with respect to plasma shape | ||
| Name: | Ryan | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | R. Pinsker, J. Hosea, G. Taylor, E. F. Jaeger, D. Green, D. Rasmussen, M. Murakami, A. Nagy, S. Diem, M. Kaufman, R. Perkins | ITPA Joint Experiment : | No |
| Description: | -- | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Investigate the FW power coupling properties of non-conformally shaped antennas to three general plasma configurations: AI, AT, and DIIID Demo for ITER Baseline. The AORSA code will predict loading for the three configurations as a function of gap spacing. Loading measurements will be taken to compare to these AORSA calculations, with the eventual goal of developing FW compatible plasma configurations. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 112: Minimum pellet size for ELM pacing | ||
| Name: | Baylor | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | N. Commaux, A. Loarte | ITPA Joint Experiment : | Yes |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 113: Validation of plasma response models with rotating n = 2 and fast BES imaging | ||
| Name: | Moyer | Affiliation: | UCSD |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | M. Van Zeeland | ITPA Joint Experiment : | No |
| Description: | Validate models of plasma response to RMPs using rotation n = 2 RMPs and 2D imaging BE spectroscopy. In CY11, the UCSD fast camera was used to document the boundary displacement due to rotating n = 2 RMPs. The modification of the BES profile on the LFS midplane has been shown to be primarily a displacement of the edge. At q95 ~ 3.8 this displacement = 2 cm, which is consistent with the displacement of the steep gradient region of the pedestal density profile as measured on the LFS midplane using profile reflectometry. This displacement is a fundamental prediction of both the vacuum and plasma response models where it is due to displacement of the stable and unstable manifolds (vacuum model) and/or a driven kink response (m3d-c1). New dedicated run time is needed in CY 2012 to apply this capability to study the parametric scalings in these response models since most of the CY2011 discharges did not have 150R for fast camera based 2D imaging of the beam emission profile. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Oscillate the I-coil with an n = 2 RMP at 10 Hz. During the I-coil oscillation, run the 150R beam unmodulated to obtain continuous BES profile measurements. By raising Bt to 2 Tesla, we will also optimize the ECEI of this behavior (previous shots had a lower Bt leading to measurements too far out. ) Primary parameter scans for model validation include:
1) reverse Bt to change the direction of the displacement poloidal rotation; 2) vary q95 which appears to affect the magnitude of the displacement in CY11 data; 3) vary "target" rotation systematically to look at rotational screening response. | ||
| Background: | Understanding the plasma response to externally applied magnetic perturbations (RMPs) is critical for understanding how these magnetic fields affect ELM behavior, leading to a predictive model for ELM suppression in ITER. In addition to the trip3d-mafot vacuum field model, in which the externally applied fields are superimposed on the 2D Grad-Shafranov axisymmetric equilibrium, several models of plasma response are becoming available, including marsf, msd-c1, and xgc0. These models need to be validated against experimental measurements in order to develop a predictive model for RMP ELM suppression and for boundary control with externally applied magnetic perturbations. | ||
| Resource Requirements: | I-coil in n = 2 with 10-20 Hz oscillation.
profile reflectometry 150R unmodulated during the I=coil modulation to allow continuous fast camera based 2D imaging of beam emission. Bt = -2 T to permit ECEI further up the pedestal than in CY11 experiments. 210 beams to allow variation of the torque and toroidal rotation of the target axisymmetric plasma. | ||
| Diagnostic Requirements: | Full pedestal and profile diagnostics.
Phantom v7.1 on 225T0 port Phantom v7.3 on 90T0 port with passive imaging of the HFS. ECEI mm wave based fluctuation diagnostics to monitor turbulence changes. | ||
| Analysis Requirements: | Will require significant trip3d-mafot, m3d-c1, and other response model simulations for comparison to measurements. | ||
| Other Requirements: | |||
| Title | 114: HFS Pellet Fueling with simultaneous pellet pacing ELM control | ||
| Name: | Baylor | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | N. Commaux, A. Loarte | ITPA Joint Experiment : | Yes |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 115: Comparison of n=2 RMP field effects on ELMs and pedestal properties in KSTAR and DIII-D | ||
| Name: | Evans | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | YM Jeon, J. Kim, Y. Oh, J. Kwak, W. Kim, JY Kim. et al., | 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. 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 2011 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). If ELM suppression is not obtained we will use experience acquired during our 2011 n=2 ELM suppression experiments to map out what needs to be done in KSTAR to obtain n=2 ELM suppression. | 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 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 and in KSTAR with n=2 fields have not yet resulted in suppression or significant mitigation. Understanding the mechanisms involved in these diametrically opposed results between the two machines will provide new insight into the key physics mechanisms responsible for RMP 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 | 116: Develop ITER-relevant access to G=0.4 QH-mode | ||
| Name: | Garofalo | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | K. Burrell, W. Solomon, M. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | Demonstrate ITER-compatible access to QH-mode operation at G=0.4 and ITER equivalent values of NBI torque. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Using an ITER-similar plasma shape, reproduce the G=0.4 QH-mode from the Nov. 2011 experiments [D3DMP No.: 2012-31-01].
Then, work on the discharge front-end to replace a Bt ramp down with a constant Bt waveform, and to remove the reliance on net counter-Ip NBI torque by adjusting timing of NRMF application and power ramp up. The main issue to confront is that the counter-torque from NRMFs (needed for QH-mode) is weak at low betan. Therefore, betan should be ramped up already during the L-mode phase, to enable the NRMF torque needed for QH-mode access and avoid ELMs between the L- and QH-mode phases. If access to QH-mode with zero-net NBI torque appears precluded, we could explore using I-mode operation as an intermediate phase between L-mode and QH-mode. | ||
| 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: an upper-biased DND plasma cross section, 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, large counter NBI torque during the L-mode phase for robust access to QH-mode.
For application to ITER, highest priority issues to address are: extend G=0.4 QH-mode to ITER similar shape, develop a zero-torque ELM-free path to the G=0.4 QH-mode state, and remove reliance on Bt ramp during Ip ramp up. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | Very recent boronization. | ||
| Title | 117: Fast ion bootstrap current increased by high harmonic ion cyclotron resonance of off-axis beam ions | ||
| Name: | Park | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Improve off-axis NBCD efficiency by increased fast ion bootstrap current. Deliver RF power to off-axis beam ions to generate high energy tail through high harmonic ion cyclotron resonance, which is a much more efficient way of increasing bootstrap current than any thermal electron/ion heating. The fast ion bootstrap current is inherently off-axis (so good for steady-state scenario) The radial profile of fast ion bootstrap current could be controlled by adjusting the resonance location (even towards further off-axis current drive). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | off-axis beam, FW | ||
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| Title | 118: Establish the incremental confinement of EC power in high betaN steady-state scenario discharges | ||
| Name: | Ferron | 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. | 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. | ||
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| Title | 119: Tandem n=1 EFC by the I and C-coils | ||
| Name: | La Haye | Affiliation: | Retired from GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | R. Buttery, M. Schaffer and E. Strait | ITPA Joint Experiment : | Yes |
| Description: | The best n=1 error field correction with the I-coil and the C-coil simultaneously optimized (in tandem) will be found. The optima by themselves are clearly "incomplete" but together may make for a better match to the intrinsic n=1 field error structure. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | An ohmic discharge in a biased up DND to avoid low density H-mode would be run at q95~3.2 as previous to find the low density locked mode limit. The previous best I240-coil correction (from 4 current sweeps of the quadrants to find the best shifted circle) would then be applied and the lower (~50%) low density limit found. To this I-coil correction, 4 sweeps of the C-coil (at medium constant density) would be added to find the next shifted circle. The optimum C-coil current would be added to the optimum (old) I240-coil and the new lower limit found. Some iteration may be required to get the best tandem correction. If better ohmic n=1 EFC is found, this would be a starting point for higher beta H-mode etc dynamic n=1 EFC elsewhere. | ||
| Background: | The best 4 quadrant sweeps of either the I or C coils separately do not recover a low density limit below 50%, if that. Attempts at pre-calculating the I and C currents together for best correction of the known intrinsic errors have also not worked well. We need to let the plasma determine how best to use the coil sets.
A similar optimization approach was used several years ago to determine that a 240-degree phase difference between the upper and lower I-coils was best. This led to better error correction than with the C-coil. Here we extend the approach to the optimization of two coil sets (I240 and C), and expect an improvement over error correction with the I240-coil alone. Previous work on multi-mode error field correction showed improvement through optimized use of the old n=1 coil and the C-coil [J.T. Scoville and R.J. La Haye, Nucl. Fusion 43, 250 (2003)]. To date, there has been no attempt to optimize the combined use of the I-coil and C-coil for error field correction based on the empirical plasma behavior. Unlike #148 of 2011, this will limit the possible combos to I240 (best U&L I-coil phasing by itself) and the C-coil. | ||
| Resource Requirements: | I240-coils and C-coils with SPAs and standard q95~3.2 UDND ohmic so can be ready early in run and needs limited diagnostics. Note that success at lower locked mode density with tandem at q95~3.2 would then require q95~4.6 at (higher BT) for interpolation of a new I and C algorithm for basic use in the PCS. | ||
| Diagnostic Requirements: | Standard magnetics and ECE. | ||
| Analysis Requirements: | If the plasma tells us how to better correct the intrinsic error field we might then use IPEC etc to see what the optimization is doing, i.e. resonant versus non-resonant correction etc. | ||
| Other Requirements: | None. | ||
| Title | 120: Divertor Heat Flux Variation with Pellet ELM Pacing | ||
| Name: | Baylor | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | C. Lasnier, N. Commaux, A. Loarte | ITPA Joint Experiment : | Yes |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | |||
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| Title | 121: Establish the best technique for dynamic error field correction in steady-state scenario discharges | ||
| Name: | Ferron | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Dedicate a series of steady-state scenario, high betaN discharges to optimizing the parameters of the dynamic error field correction algorithm. The relevant parameters include frequency response (high versus low), gains, audio amps versus SPA on the I coils. Determine the effectiveness of DEFC at minimizing the n = 1 response of the discharge, and the corresponding effect on rotation and confinement. Determine whether there is an effect on RWM and/or fishbone stability. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Dedicate a sufficient number of discharges to tuning of DEFC parameters to establish how beneficial this tool is for steady-state scenario discharges and to establish the best setup to use. This wouldn't necessarily need to be an entire day of shots, but it is necessary to have enough reasonably reproducible shots in a row available in order to scan parameters. It might be necessary to have shots on two different days in order to test audio amp versus SPA. | ||
| Background: | The use of dynamic error field correction seems to improve the performance of steady-state scenario discharges. However, the best way to use DEFC in these discharges has not been established. Usually the operators of the DEFC have only a couple of shots in a day in order to make adjustments. The shots may not be reproducible enough for shot to shot comparison. Also, there is the possibility that fast frequency response with audio amps on the I coils is the best setup, but this is difficult to test because switching between audio amps and SPA on the I coils is a large patch panel change. | ||
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| Title | 122: Plasma Triangularity Effect on Pellet ELM Triggering and Pacing | ||
| Name: | Baylor | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | N. Commaux, | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | Yes | |
| Experimental Approach/Plan: | |||
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| Title | 123: Dynamics of rapid pellet injection for ELM pacing using fast visible imaging spectroscopy | ||
| Name: | Moyer | Affiliation: | UCSD |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Object of this proposal is to use the existing fast framing camera capabilities in DIII-D to characterize the interaction of rapid pellets from both the midplane (using 90T0 port) and lower divertor (using 225T0). In CY11, rapid (up to 60 Hz) pellet injection proved quite effective in reducing the impulsive power loading to the divertor [L. Baylor et al; IAEA 2012]. Initial experiments utilized fast framing of carbon impurity lines from the 225T0 view to document and compare the pedestal and boundary response to pellets from the LFS midplane and the lower divertor. In these experiments, we will simultaneously run the 90T0 camera to investigate the extend of localized versus global (toroidal extended and/or symmetric) responses to both pellet injection locations. Previous results were successful using CIII emission, but additional imaging in Dalpha and in BES of the 150R source might also provide useful information. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use 225T0 camera view to image both the LFS midplane and lower divertor pellet injection and interaction with the plasma pedestal and boundary. Use the 90T0 view to image the "global" extent of the pellet interaction and subsequent ELM-like events. | ||
| Background: | Pellet pacing remains the chosen "fall-back" ELM control approach for ITER. In CY11, D3D made significant progress is scaling existing results to higher pellet injection frequencies (up to 60 Hz achieved). Prior to these experiments, results from many devices suggested that the reduction in heat flux to the divertor target plates dropped less than linearly with increasing pellet pacing frequency, suggesting that the impulsive power loading reduction might saturate well above the required limits in ITER. However, CY11 results, to be reported by L. Baylor et al at IAEA 2012, indicated significant heat flux reduction for pellet injection frequencies up to 60 Hz. Additional experiments are needed to characterize the effectiveness of the pellet pacin, to determine the nature of the physics involved in the pellet-induced ELM-like events, and to optimize the gas load on the vacuum vessel/pumping system. | ||
| Resource Requirements: | R-2 and R0 high frequency pellet injectors | ||
| Diagnostic Requirements: | Fast cameras on 225T0 and 90T0 ports
Full profile, pedestal and boundary diagnsotics, particularly IRTVs. | ||
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| Title | 124: Test of the compatibility between pellet fueling and ELM suppression techniques | ||
| Name: | Commaux | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, T. E. Evans | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | |||
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| Title | 125: Fast pellet mass drift physics | ||
| Name: | Commaux | Affiliation: | ORNL |
| Research Area: | General BPP | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, P. B. Parks, B. Pegourie | ITPA Joint Experiment : | No |
| Description: | -- | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -- | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 126: Isolate tauE & fBS dependence on qmin, off-axis NBI power, & rhoqmin | ||
| Name: | Holcomb | Affiliation: | LLNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | This experiment systematically isolates the roles of high qmin and off-axis NBI in reducing the energy confinement time. It will also address this question: does the bootstrap current fraction still cease to increase with increasing qmin, even with larger rho_qmin provided by the off-axis NBI? (I.e. how does fBS scale with rhoqmin?) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce 6 to 9 good shots, all with betaN=2.7, q95=5, equal density, same EFC, same BT, shape, wall conditions, ECCD, etc. Sequence:
1. qmin=2.1, with max off-axis NBI power at betaN flattop. 2. qmin=1.5 with max off-axis NBI power at betaN flattop. 3. qmin=1.1 with max off-axis NBI power at betaN flattop. - Then, tilt 150's back to on-axis injection, (1-2 hours). Either do this in the middle of the day, or at 5 PM after an afternoon experiment. Then in the PM (or next morning): Steps 4-6: repeat steps 1-3 exactly with the only difference being the beam angle. The q-profile evolution should be somewhat different. Steps 7-9: adjust the discharge evolution as needed to match the qmin obtained in steps 1-3. | ||
| Background: | Attempts to go to high betaN with qmin>2 using off-axis NBI were stymied by low energy confinement time. We clearly showed in a pair of discharges at low qmin that the use of off-axis NBI alone results in lower confinement, i.e. more total beam power was required to reach the same betaN when off-axis NBI was used. But there also still seems to be a comparable or greater reduction in confinement simply by going to qmin>2. A careful, systematic scan of both qmin and off-axis NBI fraction was not done. Such a scan might provide an empirical confinement time scaling that includes qmin and fraction or amount of off-axis NBI power, and this would be useful for predicting performance with 10 MW off-axis NBI.
The experiment proposed here also expands the range of q-profiles accessed in a 2009 experiment to evaluate the bootstrap current fraction as a function of the target q-profile. One key result of this work was that fBS did not continue to increase with increasing qmin, as simple scaling expressions predict. The proposed experiment would produce pairs of shots with the same qmin but hopefully different rhoqmin (on vs off-axis), and we could evaluate how higher rhoqmin effects fBS. | ||
| Resource Requirements: | All co-Ip NBI and ~3 MW ECCD. Plan to change beam and possibly source tilt in the middle of a day or between a PM and an AM shift. | ||
| Diagnostic Requirements: | All standard diagnostics. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 127: Particle balance and transport at high density | ||
| Name: | Commaux | Affiliation: | ORNL |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, E. A. Unterberg | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | |||
| Background: | |||
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| Title | 128: Effect of 3D-fields on tearing stability: locked QH-mode cases, stable with n=3 field | ||
| Name: | Garofalo | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | K. Burrell, W. Solomon | ITPA Joint Experiment : | No |
| Description: | 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: | |||
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| Title | 129: Influence of pellet injections on the L/H transition threshold | ||
| Name: | Commaux | Affiliation: | ORNL |
| Research Area: | L-H Transition | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, P. Gohil | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
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| Title | 130: Neon shattered pellet injection in a runaway beam | ||
| Name: | Commaux | Affiliation: | ORNL |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, E. M. Hollmann, D. A. Humphreys, J. C. Wesley | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
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| Title | 131: Test of high Z shattered pellet injection for runaway collisional suppression | ||
| Name: | Commaux | Affiliation: | ORNL |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, E. M. Hollmann, D. A. Humphreys, J. C. Wesley, P. B. Parks | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
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| Title | 132: Influence of q95 and shape on D2 shattered pellet assimilation | ||
| Name: | Commaux | Affiliation: | ORNL |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | L. R. Baylor, T. C. Jernigan, E. M Hollmann, D. A. Humphreys, P. B. Parks, J. C. Wesley | ITPA Joint Experiment : | No |
| Description: | ITER IO Urgent Research Task : | No | |
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| Title | 133: Control of toroidal mode number of EHO | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | A. Garofalo | ITPA Joint Experiment : | No |
| Description: | Investigate how the toroidal mode number n of the edge harmonic oscillation<br>(EHO) changes with plasma shape, q, density and n-number and parity of the<br>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. | ||
| 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 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. 7 NBI sources. | ||
| Diagnostic Requirements: | All profile diagnostics. Fluctuation diagnostics for edge measurements of EHO, especially BES and ECE-I | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 134: Test stability and predicted fBS of baseline FDF scenario | ||
| Name: | Holcomb | Affiliation: | LLNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | A. Garofalo | ITPA Joint Experiment : | No |
| Description: | Produce a reversed shear q-profile with minimal on-axis beam current drive to test the stability of the baseline FDF scenario at betaN=3.3-3.7. Determine if this q-profile is consistent with steeper temperature gradients and a high fBS (70%), while still having good alignment between the noninductive and total current profiles. Evaluate energetic particle loss. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Approximate the FDF baseline scenario in DIII-D. (See background). Use a DN shape. Apply all available ECCD at rho=0.4-0.6. Apply high power off-axis NBI, and 4 balanced on-axis beams for heating with net zero current drive.
This beam arrangement will move towards the low rotation expected in an FDF with no NBI. | ||
| Background: | The present FDF baseline scenario (Garofalo, Jan. 20 2012 Fri Sci Meeting) has q0~7, qmin~1.3, rhoqmin~0.5, q95~4, pressure peaking factor~3.1, fBS~0.7, fNI=1, betaN=3.7, H98~1.2, and only ECCD & LHCD (no NBI). Due to our limited elongation, the stability equivalent betaN in DIII-D is 3.3.
In weak shear DIII-D plasmas with betaN near 3.3, fBS has been calculated to be between 0.4 and 0.6, but not 0.7. The latter relies on obtaining higher Te and Ti gradients probably associated with the strong magnetic shear reversal. | ||
| Resource Requirements: | All NBI, ECCD. | ||
| Diagnostic Requirements: | Standard diagnostics. | ||
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| Title | 135: Is there a more favorable q-profile at low rotation for AI operation? | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of this experiment is to attempt to recover the loss in confinement observed in advanced inductive discharges when going to low torque. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce suitable low torque advanced inductive discharge from FY11 campaign and investigate the dependence of confinement on the q-profile (particularly varying the core shear, including reverse shear). Ideally, modeling with TGLF should be conducted beforehand to guide the experiment (underway). During these experiments, the confinement dependence of rotation should be mapped to the counter direction also. | ||
| Background: | Experiments have shown that there is a significant degradation in the confinement in advanced inductive discharges as the torque and rotation are lowered. This appears independent of whether one starts with a high torque, high performance discharge and ramp the torque down, or form the plasma in a low torque state from the beginning. This is likely explained at least in part by the loss of ExB shear associated with the lower rotation. A key question remains, however, whether the same flat q-profile (which is also a critical aspect of the improved confinement) is appropriate for use at both high and low rotation, or if a different optimization is necessary. Addressing the loss of confinement toward H~1 at low rotation is an important issue, since this reduces the attractiveness of the AI scenario, which relies on high beta and high confinement to allow operation at lower plasma current for the same fusion gain. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons with NTM control, 210 beams | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Preferable simulation/modeling before experiment. TGLF etc modeling of actual discharges. | ||
| Other Requirements: | |||
| Title | 136: Document turbulence change with rotation in AI, and compare with theory | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | 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 | ||
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| Title | 137: Compatibility of low torque Advanced Inductive discharges with RMP ELM suppression | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of the experiment is to investigate whether RMP ELM suppression can be applied to a low torque advanced inductive plasma, and if so, to characterize the impact on performance, confinement and stability. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce hybrid discharge #129949. Raise Bt to -1.65 T (at fixed q95) to move the 3rd harmonic EC resonance out of the plasma, allowing 2nd harmonic access to both the q=2 and q=3/2 surfaces. Ramp to low torque before applying the RMP, and utilize ECCD at the q=2 surface to avoid the 2/1 NTM (an alternative approach is to aim at q=3/2 to limit the size of the 3/2 NTM, although low rotation experience suggests that the plasma will still be susceptible to the 2/1 at low rotation anyway). Ideally should be operated at high enough betaN to be consistent with ITER Q=10 (~betaN>2.8). Perform q95 sweeps to isolate the ELM suppression window if it exists at low rotation. Document the stability, confinement and performance and compare with no RMP. | ||
| Background: | RMP ELM suppression has been achieved with mixed success in the advanced inductive regime. A prototype example is #129949. An issue with these early experiments was that the NTM associated with AI operation would tend to slow and lock with the RMP. This situation is likely exasperated by low torque/rotation. | ||
| Resource Requirements: | 1 day expt, 6 gyrotorons with NTM control, 210 beams, I-coils for RMP | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 138: rho* scaling of intrinsic torque between DIII-D and JET | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | 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 expt, 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 | 139: Coupling of Core Electron and Ion Transport in TEM-Dominated Regimes | ||
| Name: | Schmitz | Affiliation: | UC, Los Angeles |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | T.L. Rhodes, C. Holland, J. Hillesheim, K.H. Burrell, G. Wang, W.A. Peebles, G.R McKee, Z. Yan | ITPA Joint Experiment : | No |
| Description: | The objective of this experiment is to explore the scaling of core electron and ion thermal transport at high Te/Ti. We propose to separate the effect of ExB shear (due to core toroidal rotation), collisionality, and Te/Ti via variation of ECH deposition location and power, NBI power, and NBI torque, to understand how electron and ion transport are coupled in electron-heat dominated regimes. Understanding this synergy between ion and electron transport is potentially important for burning plasma with predominant alpha heating. In particular, we try to address three objectives: 1) Understand whether ion transport can be directly driven by TEM modes; 2) Is there excitation of ion-relevant low-k modes in the TEM regime (addressed by core BES measurements); 3) is the ITG critical gradient substantially reduced at high Te/Ti? | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using ECH in relatively low density QH-mode plasmas, conditions with central Te>Ti can be achieved at ITER-like collisionality (central electron temperature > 10 keV and central ion temperature > 7 keV). These plasmas have been found to be TEM-dominated [1] with substantially increased electron temperature fluctuations and linear TEM growth rates far exceeding ITG growth rates. Reverse I_p operation allows independent control of NBI power and torque to offset the reduced toroidal rotation with ECH. Varying the neutral beam torque will then allow to discriminate the effects of core ExB shear on ion transport. Varying the ECH deposition location and power (and the density in a moderate range) will allow Te/Ti (and the collisionality) to be controlled over a reasonably large range. | ||
| Background: | An important synergy between electron and ion core transport has been found at high Te/Ti in a nominally TEM-dominated regime. In particular, an increase in the Gyro-Bohm normalized ion thermal diffusivity has been observed in the TEM regime,despite the fact that R/L_Ti was likely below the ITG critical gradient. Substantially increased electron temperature fluctuations (characteristic of the TEM regime) have been measured, with density fluctuations unchanged across 0.6 < k rho_s < 2. It is not clear what causes the increased ion transport at high Te/Ti
in this (nominally) TEM dominated regime. Increased ion transport with TEM is seen in at least one gyrokinetic simulation [2]. | ||
| Resource Requirements: | 7 beams, all gyrotrons, reversed I_p, wall conditions allowing high quality QH-mode operation | ||
| Diagnostic Requirements: | All fluctuation diagnostics | ||
| Analysis Requirements: | TGLF calculations (pre- and post-experiment),
Selected GYRO runs resolving TEM relevant wavenumbers. | ||
| Other Requirements: | -- | ||
| Title | 140: Long Pulse/Medium Pulse-ITER Baseline Scenario experiments, part Deux | ||
| Name: | Jackson | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | W. Solomon, F. Turco, T. Luce, R. Buttery, A. Hyatt, E. Doyle, J. Ferron, and R. LaHaye, M. Murakaim, J.M. Park | ITPA Joint Experiment : | No |
| Description: | Extend the successful ITER baseline scenario (IBS) experiments from 2011 by: <br>1. Varying Ip ramp to demonstrate startup at lower li which is a concern for ITER. This fills in a missing part of the matrix which was part of MP 2012-35-01. <br>2. Examine stability of IBS plasmas with OANB <br>3. Vary deposition of EC (q=1,q=3/2, q=2 w/ radial and oblique launch) and the effect on 2/1 TM stability <br>4. Establish a turn-off time for EC still with stable plasmas, i.e. does EC have to be used continually? <br>5. Simulate a complete IBS low torque discharge including the startup phase. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Discharges stable to 2/1 TMs reached stationary conditions in last year's experiments. In 2012, better quantify the parameter space for successful operation keeping in mind the ITER design. Start with shorter discharges, ramping down about 7 sec. so normal hardware settings can be used. Use long pulse set-up (ge 10 s) when required to demonstrate stability that can't be done at shorter duration. | ||
| Background: | see Experimental Approach | ||
| Resource Requirements: | At least one day would be long pulse if needed. This was demonstrated up to 10s in 2011. | ||
| Diagnostic Requirements: | Normal diagnostics, but temporal coverage needs to be extended (for long-pulse). | ||
| Analysis Requirements: | Kinetic Efits, ONE-TWO, Toray, Physics based modeling to predict the effect of OANB | ||
| Other Requirements: | -- | ||
| Title | 141: Co-Ip NBI Torque Limits for QH-mode sustained with NRMF from C-coil Only | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | M.E. Fenstermacher, A.M. Garofalo | ITPA Joint Experiment : | No |
| Description: | Using NRMF produced by the C-coil only, establish the limits on the co-Ip NBI <br>torque that still allows QH-mode operation. Determine this limit as a function<br>of q_95 over the range 3.6 to 5.3. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Re-establish a shot like 145066 at q_95 = 5.2 using n=3 NRMF from the C-coil
only. Scan beam torque to establish the co-Ip torque limit. Using the toroidal field rampdown technique from shots like 147351, produce QH-mode at q_95 = 3.6. Scan the NBI torque to find the limits under these conditions as well. | ||
| Background: | The use of NRMF sustained QH-mode in future devices would be much more straightforward if it is possible to create the NRMF using coils that are outside the toroidal coils, as is the case for the C-coil on D III-D. During the 2011 campaign, we were able to create NRMF sustained QH-mode with zero net NBI torque using the C-coil to produce the n=3 field. The I-coil was used for n=1 error field correction. However, we did not have enough experimental time to determine how much co-Ip torque we could use while still sustaining QH-mode. In addition to establishing this under the conditions we used in 2011 (see shot 145066, for example), we should also determine what this limit is at lower q. This can be done using shots like 147351. Determining these limits is extremely important for the QH-mode presentations to be made this fall at APS and IAEA. If the limits are large enough, they would provide an existence
proof for QH-mode operation that is easier to apply in future machines. | ||
| Resource Requirements: | 7 NBI sources. Reverse Ip | ||
| Diagnostic Requirements: | All profile diagnostics. Fluctuation diagnostics optimized for EHO measurements. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 142: ELM mitigation in q95=7 steady state plasmas | ||
| Name: | Holcomb | Affiliation: | LLNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | ITER IO Urgent Research Task : | No | |
| Experimental Approach/Plan: | Develop a q95=7, fNI=1 target plasma, and sustain it for a few resistive times to evaluate stability and stationarity. Then apply I-coil RMP to modify or eliminate the ELMs. Evaluate the impact on the pedestal, the ELMs, and the ability to maintain fully noninductive operation. This would be a first step in determining if steady state operation and RMP ELM control are compatible. | ||
| Background: | A window for ELM suppression exists at q95=7. We have produced net noninductive plasmas with betaN~3.7-3.9 and q95~6.3 for ~0.7 tauR. It should not be too much of a stretch to adjust q95 up to get into the resonance window. | ||
| Resource Requirements: | All co-Ip beams, 3+ MW ECCD, EFC with C-coils and I-coils for RMP ELM suppression. | ||
| Diagnostic Requirements: | All standard diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 143: ITER Demo Discharges with dominant electron heating | ||
| Name: | Jackson | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | E. Doyle, I. Chapman, R. LaHaye, R. Buttery, T. Luce, F. Turco, W. Solomon | ITPA Joint Experiment : | No |
| Description: | Demonstrate discharges with t_flattop/tau_R > 4 with dominant electron heating, as would be the case in ITER | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Evaluate the use of sawtooth mitigation (ECCD just inside q=1)and EC deposited at q=3/2 to also mitigate the 2/1 TM. | ||
| Background: | This experiment was first carried out in 2011, but the dominant electron heating phase was transient due to the onset of 2/1 TM. Subsequent work by I. Chapman, R. LaHaye, et al.(submitted to Nuc. Fus., 2012) showed a beneficial effect on 2/1 TM stability when ECH was deposited just inside q=1, reducing sawtooth amplitude and increasing the frequency. With this successful demonstration, the dominant electron heating experiment should be repeated with sawtooth mitigation. | ||
| Resource Requirements: | 6 (or 7) gyrotrons. Probably 3 at q=1 and 3 or 4 further out. Test both q=3/2 and q=2/1 ECCD deposition and compare to radial launch. | ||
| Diagnostic Requirements: | ECE is a key diagnostic as is mse | ||
| Analysis Requirements: | Toray, good mse EFITs. ONE-TWO and kinetic EFITs after the experiment. | ||
| Other Requirements: | -- | ||
| Title | 144: Interaction between Off-axis fishbone mode and RWM | ||
| Name: | Matsunaga | Affiliation: | QST |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | M. Okabayashi (PPPL), Y. In (FAR-TECH), G. L. Jackson (GA), E. J. Strait (GA), T. C. Luce (GA), J. R. Ferron (GA), W. W. Heidbrink (UC Irvine), M. Takechi (JAEA), A. Isayama (JAEA) | ITPA Joint Experiment : | No |
| Description: | The aim of this experiment is to clarify mechanism that off-axis fish bone mode (OFM) can drive resistive wall mode (RWM). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For OFM-driven RWM, high beta plasma above no-wall limit should be targeted, and finally, RWM onset is required. Under our hypothesis that this phenomenon is related to energetic particle (EP) profile and transport, EP diagnostics such as FIDA and BILD are needed. RWM stability can be proved by active MHD spectroscopy with frequency of hundreds and kilohertz by using I-coil and lock-in amplifier. To be marginal RWM stability, plasma rotation will be declined as low as possible while keeping higher beta. Moreover, to investigate error field effect on OFM behavior, static EFC can be scanned. | ||
| Background: | Recently, RWM induced by EP driven mode are observed in both DIII-D and JT-60U even with enough plasma rotation for RWM stabilization. This is thought to be due to reduction of EP stabilization on RWM. The EP modes, that are called Off-axis Fishbone mode (OFM) on DIII-D and EP driven Wall mode (EWM) on JT-60U, can enhance EP transport and finally lose EPs. According to MARS-K that is kinetic RWM stability code, this interpretation is qualitatively consistent with that results. The focusing phenomenon implies that EP can contribute to stabilize RWM, however, control of EP driven mode is needed to avoid its driven RWM for burning plasmas with high beta. | ||
| Resource Requirements: | 1 day experiment with
NBs to exceed no-wall beta limit and ECCD for NTM suppression | ||
| Diagnostic Requirements: | Standard for RWM and EP experiments | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 145: ELM triggering by EP driven mode | ||
| Name: | Matsunaga | Affiliation: | QST |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | M. Okabayashi (PPPL), Y. In (FAR-TECH), G. L. Jackson (GA), E. J. Strait (GA), T. C. Luce (GA), J. R. Ferron (GA), W. W. Heidbrink (UC Irvine), M. Takechi (JAEA), A. Isayama (JAEA) | ITPA Joint Experiment : | No |
| Description: | The aim of this experiment is to clarify mechanism of ELM triggering by energetic particle (EP) driven mode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For EP driven mode that can trigger ELM, high beta plasma above no-wall limit should be targeted. To investigate EP effect to pedestal, EP profile and transport, EP diagnostics such as FIDA and BILD are needed. Also, edge measurements such as CER, ECE and BES are useful to clarify the causal connection of this phenomenon. | ||
| Background: | Recently, ELM triggering by EP driven mode are observed in both DIII-D and JT-60U. The EP modes are called Off-axis Fishbone mode (OFM) on DIII-D and EP driven Wall mode (EWM) on JT-60U. On JT-60U, the EWM-triggered ELM has higher repetition frequency and smaller energy loss compared with natural ELM. On DIII-D, OFM triggered ELM was observed in high beta operation in SSI experiments. These phenomena are quite important for high beta plasmas with EPs, thus, reactor relevant burning plasmas. | ||
| Resource Requirements: | 1 day experiment with NBs to exceed no-wall beta limit | ||
| Diagnostic Requirements: | Standard for RWM and EP experiments | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 146: Direct measurement of Zonal Flow Damping via Limit Cycles | ||
| Name: | Schmitz | Affiliation: | UC, Los Angeles |
| Research Area: | L-H Transition | Presentation time: | Requested |
| Co-Author(s): | G.R Tynan, G.R. McKee, L. Zeng, Z. Yan, T.L. Rhodes, E.D. Doyle | ITPA Joint Experiment : | No |
| Description: | Zonal Flow damping may have a profound effect on the L-H transition power threshold and its density scaling/safety factor scaling. We propose to exploit the recently investigated limit cycle oscillations to directly measure Zonal Flow damping versus density/ion collisionality, combining local probe measurements of the turbulent energy transfer to the flow (up to ~1 cm inside the separatrix) and direct measurements of the Zonal Flow amplitude evolution by Doppler Backscattering and BES. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The proposed experiment is a density scan from the lowest achievable L-mode density (in an LSN, non-pumping, low triangularity configuration) to densities Each shot in these scans will be repeated to obtain data via DBS, BES and the reciprocating midplane probe at 2-3 radii. The nonlinear energy transfer will be calculated from the probe data. The instantaneous ExB shear is evaluated from DBS data; the equilibrium (diamagnetic) main ion ExB shear is approximated via profile reflectometry and CER ion temperature data; | ||
| Background: | Limit cycle oscillations (LCO) preceding the L-H transition near power threshold have been mapped with high spatial/temporal resolution in a previous experiment, and the role of low frequency Zonal Flows in edge barrier formation has been established. However only limited probe data exist so far mostly at one radius inside the separatrix, and the scaling of the LCO with density has only been documented within a very limited density range. No useful q scan has been
performed so far. Previous measurements have already shown that the limit cycle period increases with plasma density, however quantitative measurements of ZF evolution and energy transfer rate across a larger collisionality range are needed to establish the Zonal flow damping rate and its scaling. | ||
| Resource Requirements: | 30,330,150 Beams, ECH | ||
| Diagnostic Requirements: | DBS, midplane reciprocating probe, BES, V/Q-band profile reflectometer | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 147: Non-inductive and minimal V-sec startup with EC | ||
| Name: | Jackson | Affiliation: | GA |
| Research Area: | Plasma Control and General Issues | Presentation time: | Requested |
| Co-Author(s): | N. Eidietis, A. Hyatt | ITPA Joint Experiment : | No |
| Description: | Non inductive startup, up to 33 kA, was demonstrated with EC only (Jackson, Nuc. Fus.,2011). With the addition of more EC power, and using the outer F-coils to provide flux, this current can be significantly increased. A transition to a diverted discharge with only a modest V-s addition from the solenoid and NBI and ECCD could provide useful information for FNSF and beyond. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use all available ECH, ramp outer F-coils to provide a small V-sec contribution, then divert. Continue ramping with the E-coil, but apply NBI soon after diverting. Minimize the V-sec contribution from the Ecoil | ||
| Background: | Solenoidless startup (Leuer,Nuc. Fus., 2011) and NI EC startup (Jackson,Nuc. Fus., 2011) have been demonstrated to modest current levels. However these experiments tried to achieve 100% solenoidless startup (Leuer) or non-inductive startup (Jackson). Relax this constraint by allowing some V-sec from the Ecoil solenoid, then try to minimize that current. Compare OANB injection with all sources on-axis. | ||
| Resource Requirements: | 6 or 7 gyrotrons, 6 Co-NB sources. | ||
| Diagnostic Requirements: | standard set | ||
| Analysis Requirements: | ONE-TWO, TRANSP, kinectic EFITs, Toray, physics based predictive modeling of OANB CD. | ||
| Other Requirements: | -- | ||
| Title | 148: Operate at increased Bt to produce fNI = 1 discharges using a model-guided approach | ||
| Name: | Ferron | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | 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. This model points to relatively high Bt >1.9 T as the optimum point for operation at betaN near 3.8. Produce discharges both with and without off-axis beam injection for comparison as the model (which has, as yet, only been developed for on-axis injection) predicts q_min near 1.6 even with only on-axis injection. Focus on 1.5 | ITER IO Urgent Research Task : |
No | |
| Experimental Approach/Plan: | Most of the approach is described in the description paragraph. The key is operation at increased Bt compared to previous experiments. One thing to note is that the presence of anomalous fast ion diffusion will significantly alter the results. Also, depending on the exact value of H89 obtained in the discharges (which can't be controlled ahead of time), the necessary beam power could be as high as 17 MW, more than we have available presently. | ||
| 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, at a given betaN and q95, there is a specific value of Bt at which fNI = 1. At lower Bt values, betaN must be larger than we have normally been able to obtain. For Bt = 1.9-2.1 T, the required betaN is near 3.8 while at Bt = 1.75 T, betaN must be about 4.3. In all these cases, fBS is about 0.45, lower than desired. A high fBEAM results in fNI = 1. FBS increases at higher density but in that case increased beam power is required. In the past, the value of Bt for high fNI discharges has been chosen without an understanding of the parameter values in a self consistent set. | ||
| Resource Requirements: | Maximum neutral beam power. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 149: Improve confinement of high qmin steady-state scenario | ||
| Name: | Park | Affiliation: | ORNL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Aim at increasing betaN by confinement improvement up to the predicted stability limit for qmin > 2 scenario with off-axis beams. Focus on optimizing heating and current drive "location" under constraints of the available power. Use off-axis beam to make a broad current and pressure profile with qmin > 2. Divert some (or all) of gyrotrons for central heating (rather than off-axis current drive) to deliver power where the local confinement is better, until we still can keep qmin>2. Try to use central FW heatingif possible. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | In 2011, the new off-axis beam allowed us to explore high qmin (>2) steady-state scenario. The predicted betaN limit is higher ( > 4) but confinement is not good enough to test the betaN limit for this scenario. This is partly because we put most of power into off-axis region where the local confinement is poor (in order to make a broad current profile). TGLF predicts that benefit of a broad current profile (flat q) for better confinement increases rapidly with beta. Increase of betaN and confinement by additional central electron heating could be self-apmplifying in high qmin (>2) scenario. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 150: Investigate NTV torque in QH-mode plasmas sustained by NRMF | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | A.M. Garofalo | ITPA Joint Experiment : | No |
| Description: | Experimentally measure the torque due to neoclassical toroidal viscosity in QH-mode plasmas with nonresonant magnetic fields (NRMF) by scanning the NBI torque both with and without NRMF. Compare measurements with the predictions from theory. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Run a forward Ip QH-mode discharge similar to the ones run in October, 2011.
Scan the NBI torque with and without the NRMF to map out the curve of plasma angular momentum content versus NBI torque. The difference between the curves with and without NRMF provides information on the NTV torque. For proper interpretation of the data, we need to produce QH-modes with both counter-Ip and co-Ip NBI torque, including QH-mode plasma with and without NRMF at strong co-Ip torque. We will vary the amplitude of the NRMF to check the theoretically predicted scaling with the square of the amplitude. In addition, we will do the scans at two different beta values up to the no wall limit to see if there is any contribution due to plasma amplification of the NRMF. | ||
| Background: | The NTV torque from NRMF is a key tool allowing QH-mode operation with co-Ip NBI torque in the range equivalent to the torque planned on ITER. In order to extrapolate to ITER, we need to validate the NTV torque theory. This can be done using torque scans in QH-mode plasmas both with and without NTV torque.This was done during the October, 2011 run. However, we did not get the data on strongly co-rotating QH-mode with NRMF torque needed to complete the data set. In addition, we have not investigated the parametric dependences predicted by the theory, such as the quadratic dependence on NRMF amplitude. | ||
| Resource Requirements: | 7 NBI sources. | ||
| Diagnostic Requirements: | All profile diagnostics. Fluctuation diagnostics optimized to see the EHO in the edge | ||
| Analysis Requirements: | Comparison with IPEC+NTV code | ||
| Other Requirements: | |||
| Title | 151: SOL width in top-limited discharges | ||
| Name: | Rudakov | Affiliation: | UCSD |
| Research Area: | SOL Physics | Presentation time: | 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 | 152: Pedestal current density in a 3D magnetic boundary | ||
| Name: | Stoschus | Affiliation: | Oak Ridge Associated Universities |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | D. Thomas, C. Petty | ITPA Joint Experiment : | No |
| Description: | The Lithium Beam diagnostic is capabable to measure the pitch angle of field lines in the pedestal region with high spatial resolution. This allows to deduce the edge current density, which may differ significantly from the no-RMP value and for different locations within the induced 3D magnetic topology due to magnetic field screening. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Comparison of discharge with robust ELM suppression versus no-RMP case
(2) comparison to non-resonant fields (3) slow, e.g. ~1-2Hz, two point rotation of n=3 fields to measure at two locations within topology, e.g. OP vs. XP of an island Discharges may be adapted for both good LIBEAM and MSE data with long flattop phases and therefore integration time. | ||
| Background: | A working hypothesis is that RMPs reduce the pedestal pressure gradients and therefore the bootstrap current. Thus, the plasma parameters go from a ballooning-peeling unstable to a stable regime. Additional effects like magnetic field screening due to induced currents on the rational flux surfaces potentially add up to the current.
We aim to test this hypothesis by directly measuring the edge current density directly and compare it to the modeled bootstrap current. Therefore, the Libeam and MSE diagnostic are suitable. The measurements will allow to constrain efits in the pedestal region as being input for field line tracer and stability codes. | ||
| Resource Requirements: | n=3 I-coils, NBI | ||
| Diagnostic Requirements: | LIBEAM and MSE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 153: Investigate Neoclassical Theory of Poloidal Rotation | ||
| Name: | Chrystal | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | Keith Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to test neoclassical predictions of poloidal rotation using CER. This will be done by isolating certain plasma parameters for adjustment and comparing their effect on the measurements with the theory. Also, plasmas that should have low agreement with neoclassical theory will be investigated. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure impurity poloidal rotation in several different scenarios: a plasma where rho* is scanned, a plasma where nu* is scanned, and a plasma with an internal transport barrier. | ||
| Background: | Neoclassical theory predicts that poloidal rotation is dependent on rho* and nu* and varying these parameters should allow a range of poloidal rotations to be measured.
Transport barriers are of interest because their sheared electric field is partially determined by the poloidal rotation. Also, at the barrier, the ratio of the gyroradius to the gradient scale length can become significant, and violate assumptions of neoclassical theory. Measurements of poloidal rotation will be done with a set of toroidal CER chords that view inside the magnetic axis. These chords can be used to measure poloidal asymmetries in toroidal rotation that can then be used to calculate the poloidal rotation. The advantage of this method is the ability to avoid any calculated atomic physics corrections. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER, all other profile diagnostics. Special beam modulation for optimum CER data. | ||
| Analysis Requirements: | Comparison with neoclassical codes (NCLASS etc.). | ||
| Other Requirements: | |||
| Title | 154: pitch angle in standard and reversed Bt/Ip configurations | ||
| Name: | Stoschus | Affiliation: | Oak Ridge Associated Universities |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | D. Thomas, C. Petty | ITPA Joint Experiment : | No |
| Description: | The Lithium beam is suitable to measure the pitch angle of field lines in the pedestal. From that the current density can be deduced. The measurement technique and analysis shall be tested in standard and reversed Bt and Ip scenarios. The results will be compared to modelling results from efit. This allows in turn to test bootstrap current models for these scenarios. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Piggyback to experiments with reversed and standard Bt/Ip, ideally with same plasma parameters. All H-mode plasmas with medium density and the pedestal at the position of the Libeam view chords are suitable. Long (>1s) flat top phases and therefore integration time is needed. | ||
| Background: | The pitch angle is supposed to be the same for reversing both Bt and Ip and changes sign for changes of either Bt or Ip. Those scenarios represent an ideal test case to validate the LIBEAM measurements and to compare to bootstrap current models. | ||
| Resource Requirements: | +/- Bt/Ip | ||
| Diagnostic Requirements: | LIBEAM, MSE | ||
| Analysis Requirements: | EFIT | ||
| Other Requirements: | |||
| Title | 155: Studies of arcing on divertor PFC surfaces | ||
| Name: | Rudakov | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | 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 | 156: Dust generation from deposited layers and leading edges | ||
| Name: | Rudakov | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | 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 <br>pre-deposited hydrocarbon films and specially machined leading edges. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | DiMES samples with pre-deposited hydrocarbon films and specially machined leading edges
will be exposed to known particle/heat fluxes at the strike point in LSN configuration. Dust generation will be characterized by available diagnostics (visible cameras, IR TV, MDS) and postmortem analysis of the samples. | ||
| Background: | Dust production and accumulation present potential safety and operational issues for ITER by contributing to tritium inventory rise and leading to radiological and explosion hazards. In
addition, dust penetration of the core plasma can cause undesirably high impurity concentration and degrade performance. Projections of dust production rates based on experience from existing devices are needed. ITER physics work programme for 2009-2011 calleds 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 | 157: Aerogel targets to study velocity, size and composition of dust particles in DIII-D SOL | ||
| Name: | Rudakov | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | S. Ratynskaia (Royal Institute of Technology, Stockholm) | ITPA Joint Experiment : | No |
| Description: | Targets made of silica Aerogel highly porous material composed of clusters of 2-5 nm solid silica spheres with up to 95 % empty space will be used to study velocity, size and composition of dust particles present in the outboard SOL of DIII-D during plasma discharges. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Aerogel target will be installed in Midplane Material Evaluation Station (MiMES) and kept just
outside of the wall tile radius in 240R0 port during plasma discharges for a few days. Then the target will be removed and analyzed for captured dust. | ||
| Background: | Dust penetration of the core plasma in ITER can cause unacceptably high impurity concentration and degrade performance. Therefore, knowledge of the dust transport and dynamics is important. Detection of dust present in the plasma during discharges is non-trivial. The main parameters of interest, apart from the particle material, are dust velocity, size and number density. Due to the uncertainties in the present estimates of the dust parameters it is important that diagnostics cover the maximum possible range of these parameters in order not to overlook some dust populations. A new method for dust collection has recently been proposed and is based on the use of aerogel - a highly porous, very low density material. Aerogel collectors can capture dust grains without destroying them, even in the high velocity range. Analysis of the tracks of captured particles allows to evaluate the dust velocity and the dust composition can be deduced upon particle extraction. | ||
| Resource Requirements: | MiMES with a slot for an Aerogel target. Piggyback experiments, no machine time requested. May be performed during plasma startup. | ||
| Diagnostic Requirements: | Fast USCD camera coupled to 135T0 port is highly desirable. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 158: Measurements with CER chords inside the magnetic axis | ||
| Name: | Chrystal | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | Keith Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to test tangential CER chords whose points of intersection with the neutral beam are inside the magnetic-axis. Combining their measurements of toroidal rotation with the same measurements from existing tangential chords allows the poloidal rotation to be calculated. These calculations do not require any atomic physics corrections, and scans of toroidal field and temperature should illustrate any differences between this method and the standard method (vertical CER views). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure tangential and poloidal rotation of carbon impurities in plasmas with an ion temperature scan and plasmas with a toroidal field scan. The rate the temperature and field can be scanned is increased if ELM's are infrequent (as in a QH mode for instance) since this allows for greater time resolution of quality CER data points. | ||
| Background: | Tangential CER chords that span the magnetic axis can measure poloidal asymmetries in the toroidal rotation. This information can be used to calculate the poloidal rotation without using any cross-section corrections, which are necessary for the standard measurement of poloidal rotation using vertically viewing CER chords. These corrections depend on ion temperature and toroidal field, so scans of these parameters should create the clearest picture of differences between these two methods for measuring poloidal rotation. This information will be useful for determining the amount cross-section corrections influence the standard measurement of poloidal rotation. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER, all other profile diagnostics. Special beam modulation for optimum CER data. | ||
| Analysis Requirements: | Comparison with neoclassical codes (NCLASS etc.). | ||
| Other Requirements: | |||
| Title | 159: Stabilization of NTMs with ECCD | ||
| Name: | Isayama | Affiliation: | QST |
| Research Area: | Stability and 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, (c) Preemptive (or very early) stabilization: identify the minimum EC wave power for preemptive (or very early) stabilization and (d) ECCD width effect: investigate the stabilization effect for different ECCD deposition width by changing the toroidal injection angle. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Experimental condition to obtain an m/n=2/1 NTM is based on the previous experiments (e.g. Volpe et al., Phys. Plasmas 16, 102502 (2009)). After the discharge scenario for obtaining 2/1NTM 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 probes, ECE (also oblique ECE if available), MSE, CER | ||
| Analysis Requirements: | REVIEW, NEWSPEC, TORAY | ||
| Other Requirements: | |||
| Title | 160: ELM Control Using Optimized n=2 Magnetic Fields | ||
| Name: | Lanctot | Affiliation: | Department of Energy |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | Buttery, Hanson, Evans, Orlov | ITPA Joint Experiment : | No |
| Description: | Demonstrate sustained ELM and disruption-free operation at high performance using optimized n=2 I-coil fields | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | 1.Re-establish ELM suppression using the static n=2 I-coil in even parity at 4.3 kA. Find threshold in I-coil current for suppression. Consider options for independent control of IL30 to counteract intrinsic error field from 30 deg bus.
2.Once target is obtained, rotate n=2 field at 2 Hz starting shortly after ELM suppression. Confirm that the n=2 I-coil amplitude and the near resonant poloidal harmonics are constant. 3.Increase (or decrease) I-coil current to maximize plasma performance while maintaining suppression. Attempt to avoid locked modes and/or disruptions by varying input neutral beam torque/power, and using off axis NBI for higher qmin operation. 4.Map out operating space for suppression by varying q95, betan, and Icoil amplitude/phase. 5.Use pedestal breathing technique to document pedestal stability and investigate possible island structures. 6.Patch panel change to 240 deg. I-coil phasing and repeat step 2. | ||
| Background: | Experiments in DIII-D during 2011 demonstrated the complete suppression of type-I ELMs in discharges matching the ITER shape and collisionality by applying a static even parity n=2 I-coil field. Suppression was obtained at the same values of q-95 found in previous experiments using the even parity n=3 I-coil configuration. However, suppression was not maintained when the field was rotated. The high confinement phase of the discharges was often terminated prematurely by the onset of locked modes particularly at low q-95.
SURFMN analysis of the applied vacuum field shows the Fourier amplitudes of the AC field spectrum were changing drastically when the Icoil phase was varied. This is not due to an interaction with the intrinsic error field, but to the relative phase of the I-coil currents. In most cases, suppression occurred when the amplitude was a maximum near 4.3kA. In certain cases, there is a direct correlation between suppression and an increase in the amplitudes of the field line pitch-resonant and kink-resonant field components. | ||
| Resource Requirements: | Fully functioning I/C coil power supplies are critical. | ||
| Diagnostic Requirements: | Full diagnostic set | ||
| Analysis Requirements: | Equilibrium, TRIP3D, ELM stability, plasma response calculations | ||
| Other Requirements: | -- | ||
| Title | 161: ITER Baseline in Reverse-Ip with RMP and NTV | ||
| Name: | Lanctot | Affiliation: | Department of Energy |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | Garofalo, Snyder, Hanson | ITPA Joint Experiment : | No |
| Description: | Sustained ELM Suppression and Rotation Control in the ITER Baseline scenario with Minimal Counter-injected NBI Torque<br><br>This experiment has a scenario development component and a physics goal. In terms of scenario development, the aim is to make progress toward demonstrating RMP ELM suppression in low torque discharges where the plasma is less resilient to error fields and locked modes. The physics goal is to test a proposed mechanism for ELM suppression from Snyder, which attributes the increased transport (that leads to ELM stabilization) to the presence of an island, or strong pitch resonant fields, near the top of the pedestal. The island is thought to persist in the highly conductive plasma where the perpendicular electron velocity (v_perp_e) vanishes. This hypothesis will be tested by attempting ELM suppression in a discharge where v_perp_e is not close to zero near the pedestal. Such a rotation profiles has been created in reverse Ip discharges with low NBI torque following the application of a non-resonant n=3 field, which spins up the plasma due to NTV effects. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | 1.Establish a low torque startup in the ITER baseline scenario in reverse Ip
2.Use odd parity n=3 I-coil and C-coil fields to modify the initial toroidal rotation profile, which is expected to be co-Ip in the core, near zero at mid-radius, and again co-Ip in the edge. (Typically the toroidal rotation is counter-Ip in the edge.) Adjust field to maximize change in the rotation profile. 3.Once maximum rotation is attained at the top of the pedestal, change the I-coil to n=3 even parity to obtain ELM suppression. Optimize the applied fields and adjust plasma parameters to avoid disruptions and improve performance. 4.If ELM suppression is not obtained, modify the rotation profile to create v_perp_e=0 point close to pedestal. 5.Document operational space. 6.Compare results to RMP ELM suppression in the ITER baseline scenario in normal Ip. This would require six or more discharges during an ITER scenario development day. | ||
| Background: | Most RMP ELM experiments are done in rapidly rotating H-modes, which is not ITER relevant. Feasible ELM control techniques for ITER must be able to suppress ELMs in low torque discharges. The requirements for suppression in this case may be significantly different than those found in plasmas with strong rotation. ITER also needs rotation control. This experiment would test the compatibility of these two control methods, which share a common actuator.
The success of RMP ELM suppression in future machines is uncertain because of the lack of a validated theory linking the applied field to the increased transport that leads to suppression. Recent theories point to the importance of the plasma rotation in the physics. The rotation profile found in reverse-Ip discharges following the application of non-resonant fields is drastically different from that found in typical discharges where the only external torque comes co-injected NBI, and would lead to a direct test of the Snyder model. This method has the added benefit that the amount of NTV torque can be adjusted by varying the externally applied field. | ||
| Resource Requirements: | Fully functioning power supplies for simultaneous C and I-coil operation. | ||
| Diagnostic Requirements: | CER rotation measurements. Usual RMP diagnostic set. | ||
| Analysis Requirements: | CERFIT (poloidal rotation profile will likely be important). Usual RMP diagnostic set. | ||
| Other Requirements: | -- | ||
| Title | 162: Core thermal barrier formation as a tool to understand flow self-generation | ||
| Name: | Schmitz | Affiliation: | UC, Los Angeles |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | T.L. Rhodes, G.R. McKee, M. Austin, E.D. Doyle, J. Hillesheim, Z. Yan, W.A. Peebles | ITPA Joint Experiment : | No |
| Description: | Internal transport barrier formation dynamics offers an important tool for understanding turbulence/Zonal-flow interaction and flow self-organization. Ion barriers can form both near the q=2 rational surface and, at sufficiently high beam heating power, outside of or in the absence of a q=2 surface. In particular, core Zonal flow excitation under conditions where collisional damping is very small, may allow investigation of the eddy dynamics just prior to flow generation (conjectured to be mediated by a modulational instability). Many effects that potentially complicate understanding edge barrier formation, are absent in the core (ion orbit loss, transition to the open field line region, and instability drive in the SOL). Detailed radial mapping of the ZF formation with microsecond time resolution has become possible via multi-channel DBS. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Ion ITBs are routinely obtained in DIII-D, at moderate beam power, at the q=2 with early reversed magnetic shear. At higher beam power, ITB formation can occur without the presence of the q=2 surface. We propose to investigate both scenarios to elucidate the barrier trigger and formation dynamics. Two toroidally separated DBS systems (13 channels), and BES (large array) can map ZF evolution preceding barrier formation, using repetitive shots and fine-tuning the plasma density. From the DBS data, the turbulence level, ExB velocity and velocity shear, v_ExB - correlations, the turbulence decorrelation rate, and the radial ZF propagation velocity can be extracted.
Co-injected, moderate density L-mode plasmas with moderate to high NBI power, and low density reverse-I_p counter-injected QH-mode plasmas offer the best diagnostic access. | ||
| Background: | Recent measurements have identified Zonal Flows and detailed Zonal Flow dynamics preceding the L-H transition, in both limit cycle (dithering) and regular L-H transitions, and in electron internal transport barriers at the q=2 rational surface. Measurements with a 5 channel core DBS system indicate a quasi-stationary well structure of the radial electric field that is not related to the ion pressure gradient, and possesses the characteristics of a Zonal Flow with both stationary and broadband fluctuating components. Anti-correlation between the turbulence amplitude (measured in the TEM range of wavenumbers),and the rms density fluctuation level has been observed. Earlier measurements in DIII-D have identified Zonal Flows in an ion ITB at the q=2 rational surface. However, none of these measurements has resolved the early ITB formation stage. | ||
| Resource Requirements: | 7 beams, possibly ECH (all gyrotrons) | ||
| Diagnostic Requirements: | All fluctuations diagnostics, MSE | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 163: Development of advanced LIBEAM pedestal measurements | ||
| Name: | Thomas | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | H. Stoschus, J. Munoz,, T. Osborne, R. Groebner | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to examine the utility of the LIBEAM diagnostic for making simultaneous edge profile measurements of a number of plasma parameters; including plasma density, ion temperature, and Zeff, as well as the current density. The delivery of all of these parameters at the same spatial location without recourse to equilibrium reconstructions or flux mappings would serve as a powerful constraint on pedestal models | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize LIBEAM injection into well-characterized h-mode pedestals, along with the proper atomic physics modeling and proper spectroscopic dispersion on the existing views, or additional views. Compare measurements with other diagnostics.
Vary pedestal parameters through techniques such as gas puffing or edge heating to determine sensitivity of measurements. | ||
| Background: | Lithium beams are used worldwide for a number of edge measurements yet we do not exploit this capability on DIII-D. The beam modeling is sufficiently mature now to pursue these measurements at the same time as we are doing the poloidal field measurements, or with small modifications to the equipment. Density profiles could be acquired continuously with mm resolution with the addition of different fiber bundle at the machine (5mm with existing views). Zeff and T can be obtained through simultaneous views of different wavelengths. Ion temperaure can be obtained through charge exchange spectroscopy on the injected lithium. | ||
| Resource Requirements: | LIBEAM, can run in background mode until technique is demonstrated. Variation of edge collisionality or resistivity to assess effects when system is working. Continued development of beam model with up-to-date cross sections. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 164: Investigation of edge current density systematics in Reversed BT/IP using LIBEAM | ||
| Name: | Thomas | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | H. Stoschus, C. Petty | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to examine any systematics in the LIBEAM measurements when both IP and BT are reversed, which should ceteris paribus result in an identical pitch angle from the system. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize LIBEAM injection into forward and reversed BT/IP discharges where the edges are made to be a s similar as possible. Vary pedestal parameters through techniques such as gas puffing or edge heating to determine sensitivity of measurements. | ||
| Background: | The LIBEAM poloidal field measurements work by precisely determining the pitch angle at a finely spaced array of viewpoints in the outside midplane. These pitch angles should be insensitive to a simultaneous reversal of both IP and BT. | ||
| Resource Requirements: | LIBEAM, Normal and reversed IP/BT discharges. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 165: Test of bootstrap models in DIII-D H-mode pedestals | ||
| Name: | Thomas | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | H. Stoschus, T. Osborne, R. Groebner | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to procure good LIBEAM data over a range of collisionalities and compare the Chang and Sauter models | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize LIBEAM injection into H-mode pedestals where we can vary the collisionality. Compare with predictions of the various models. | ||
| Background: | Previous LIBEAM poloidal field measurements have demonstrated the collisional dependence of the bootstrap current. Through more precise measurements we hope to distinguish between the aforementioned models that have differing collisionality effect. | ||
| Resource Requirements: | LIBEAM, other pedestal diagnostics, comparative analysis of differing models. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 166: Decoupling of pressure gradient and current density in DIII-D ELMing H-mode pedestals | ||
| Name: | Thomas | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | H. Stoschus, T. Osborne, R. Groebner | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to procure good LIBEAM data during ELMing H-modes that will allow us to study the inter-ELM behavior of the pedestal. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize LIBEAM injection into H-mode pedestals with a long ELM period. Use conditional averaging to build up a picture of the temporal evolution of the pedestal pressure and current. Compare with predictions of the various models. | ||
| Background: | The ELM limit cycle most likely occurs in the interplay between the edge pressure gradient and current. Previous LIBEAM poloidal field measurements have given a tantalizing glimpse of a temporal disconnect between these two parameters, where the pressure gradient recovers rapidly yet the current does not. Through more precise measurements we hope to clarify this behavior. | ||
| Resource Requirements: | LIBEAM, other pedestal diagnostics, good H-mode pedestal with long ELM period for best statistics. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 167: Test of NBI species mix as an Actuator on DIII-D | ||
| Name: | Thomas | Affiliation: | GA |
| Research Area: | Torkil Jensen Award | Presentation time: | Not requested |
| Co-Author(s): | B. Grierson, D. Humphreys | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to demonstrate that we can vary the beam energy fractions in a controlled fraction for use in feedback controlled torque/power experiments for the DIII-D program. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | By varying the Neutral Beam source operating conditions, change the relative values of the full, half, and third energy components during series of shots at varying densities on DIII-D. Document the changes in rotation and pressure profiles that result from the differing deposition profiles. Measure the beam profile using the main-ion CER system. Eventually learn how to control the beam component mix using he PCS system. Apply feedback to obtain a desired rotation profile (for example) | ||
| Background: | The bulk of the US fusion program, including DIII-D relies on positive ion sourceâ??based neutral beam injection which results in three primary beam components-full, half, and third energy. Each component has unique torque and power deposition characteristics based on the plasma density and atomics physics of the deposition.
If we can control and vary the generation of these components during a shot, then we add an additional actuator to our control arsenal in the same fashion as off-axis NBI or steerable ECH. The momentum deposition profile could be altered much more easily than tilting the beam. This could help in the development of a number of plasma scenarios if successful. It could also be duplicated on the many other machines relying on positive ion source NBI. The existence of the main-ion charge exchange profile diagnostic on DIII-D is a key element of determining the success, or feasibility of this idea. | ||
| Resource Requirements: | A neutral beam where the source parameters can be altered in a safe and consistent fashion during a shot. Detailed multi-component beam deposition modeling. Eventual integration of PCS control of relevant power supplies | ||
| Diagnostic Requirements: | main ion CER/BES measurements | ||
| Analysis Requirements: | TRANSP/NBEAM | ||
| Other Requirements: | PCS feedback development | ||
| Title | 168: Removed by Author's Request | ||
| Name: | Takechi | Affiliation: | QST |
| Research Area: | NA | Presentation time: | Not requested |
| Co-Author(s): | Removed by Author's Request | ITPA Joint Experiment : | No |
| Description: | Removed by Author's Request | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Removed by Author's Request | ||
| Background: | Removed by Author's Request | ||
| Resource Requirements: | Removed by Author's Request | ||
| Diagnostic Requirements: | Removed by Author's Request | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 169: Destabilization of sawteeth by local ECCD in the presence of energetic ions | ||
| Name: | Chapman | Affiliation: | CCFE |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R La Haye, RJ Buttery, G Jackson | ITPA Joint Experiment : | Yes |
| Description: | ITER will need to deploy ECCD sawtooth destabilisation to avoid large sawteeth triggering the onset of low betan NTMs with potentially large size at mode onset. DIII-D made a key demonstration of this in relevant ITER-like baseline scenarios with significant heating power and fast ion beta in 2011 [Chapman et al, accepted Nucl Fusion, â??Sawtooth control using electron cyclotron current drive in ITER demonstration plasmas in DIII-Dâ??]. However, a demonstration of the technique proposed for ITER (with ECCD to change local magnetic shear using steerable mirrors rather than field/current ramps and preferably in real-time feedback) is required, to ascertain whether other strategies (eg ICRH) will be needed in ITER. This demonstration should extend the 2011 results to tracking the q=1 radius to provide a viable demonstration for ITER and to ascertain impact on NTM threshold. DIII-D can make a major impact in this area, which is of high priority for ITER baseline scenario. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Utilise the ITER demonstration ELMy H-mode low-density scenario with long period sawteeth in the presence of NBI in the core. Apply real-time mirror sweep of co-ECCD from outside q=1 to inside q=1. Repeat scan with counter-ECCD. Heating ramps should be applied to measure 3/2 and 2/1 mode onset thresholds comparing ECCD and no ECCD cases. Ideally, real time systems can be used to keep ECCD tracking q=1 surface. | ||
| Background: | Long period sawteeth have been observed to result in the low-beta triggering of neoclassical tearing modes [Chapman, NF 2010], which can significantly degrade plasma confinement. In ITER this problem is likely to be exacerbated by the strongly stabilising effect of the energetic alpha particles, which are predicted to lead to long sawtooth periods. Consequently it is imperative to develop a robust control actuator which can deliberately destabilise the sawteeth, even in the presence of very energetic ions. Initial results from Tore Supra [Lennholm et al, PRL, 2009], AUG [Igochine, accepted PPCF, 2011] and DIII-D [Chapman et al, NF, 2012] suggest that ECCD can destabilise sawteeth in the presence of core fast ions. However, robust sawtooth control in feedback by tailoring the magnetic shear remains to be demonstrated in high performance plasmas with a significant fast ion beta. See Chapman PPCF 2011 for topical review on this subject. | ||
| Resource Requirements: | A mixture of on-axis co- and counter-NBI to reach betaN>2 and low torque. Maximum ECRH power from gyrotrons with mirrors for sweeping desposition. Real time mode targeting systems.
It is intended that much of this can be done parasitically to ITER H-mode scenario development, with a few dedicated pulses to test the NTM onset beta w and w/o active sawtooth control. | ||
| Diagnostic Requirements: | CER, MSE, Magnetics, ECE, TS, SXR | ||
| Analysis Requirements: | TORAY-GA simulations of ECCD. TRANSP simulations for fast ion population. Drift kinetic simulations to show effect of fast ions on kink stability. | ||
| Other Requirements: | |||
| Title | 170: Mitigated ELMs under the influence of n=3 RMPs | ||
| Name: | Jakubowski | Affiliation: | Max-Planck Institute for Plasma Physics |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | Todd Evans, Charles Lasnier | ITPA Joint Experiment : | No |
| Description: | Investigate power loads in mitigated RMP scenarios. Power loads of mitigated ELMs need more attention as this is also a possible scenario for ITER discharges. DIII-D shows that if the amplitude of an ELM is small enough the power exhaust follows strictly 3D magnetic topology of stochastic boundary. However, the larger events show some deviation from the magnetic topology. One of the questions for ITER would be, what is a threshold in terms of ELM energy, where power loads do not follow exhaust channels defined by the manifolds as theses events could lead to a damages of tungsten plates (this particular request comes from ITER team). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1)Reproduce discharges ##139742-46, where ELM amplitude could be controlled by q95 at ITER-like collisionalities and triangularity.
(2)Minimize the amplitude of mitigated ELMs by adjusting q95 (3)Compare the power loads (measured by fast IRTV) of non-RMP type-I ELMs with mitigated ELMs | ||
| Background: | As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved. The question arises if such scenarios can be reproduced at ITER similar collisionalities and ITER-like plasma shape. | ||
| Resource Requirements: | n=3 I-coils, NBI | ||
| Diagnostic Requirements: | IRTV | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 171: Heat Load associated with halo current during current quench phase of disruptions/VDEs | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | M. Sugihara,E. Hollmann, R. Pitts, 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 indicate that significant fraction of magnetic energy seems to be dissipated convectively and/or conductively. This indication is supported by the observed small radiation energy during this phase, especially for slow current quench discharges. Since plasma directly sits on the divertor dome and halo current flows to the inner divertor baffle during current quench phase of downward VDE and centered disruptions in ITER, more careful design of the wall panel of these components must be done, if significant energy actually deposits during current quench phase. | ||
| 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 | 172: Validation in DIII-D of proposed 15MA ITER scenario with burn extended beyond plasma current flattop | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | Y. Gribov, T. Luce | ITPA Joint Experiment : | No |
| Description: | Study extension of ITER burn by allowign the current to decay naturally in its own L/R time. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | In DIII-D discharge relevant to ITER 15MA inductive scenario (at current flattop: q â?? 3, κ â?? 1.8, βN â?? 1.8, H-mode), starting with plasma initiation in low toroidal electrical field and fast plasma current ramp-up with early divertor formation and auxiliary heating (in L mode), switch off plasma current control at the end of current flattop allowing natural decay of the plasma current in H-mode. Auxiliary heating should be varied keeping Ptot/Pthreshold â?? 2.3. Plasma shape control keeps plasma shape as is was at the end of current flattop. When plasma current slowly reduces to about half of the flattop value, plasma current control should be switched on, H-to-L mode transition should be triggered and plasma termination should be continued fast in divertor magnetic configuration with reduction of plasma elongation till plasma current significantly less than the flattop value.
The plasma current ramp-up rate and the power of auxiliary heating is adjusted for achieving the value of li(3) at the start of plasma current flattop about 0.82 and βp â?? 0.1. The Greenwald ration during the current ramp-up is about 0.35. At the start of current flattop the main heating starts with H-to-L mode transition flowed by increase of plasma density till about 0.82. | ||
| Background: | Such scenarios was recently designed and simulated with the DINA code. In this scenario at the end of current flattop (610s), when current in CS1 module is close to its engineering limit, feedback control of plasma current is switched off allowing natural decay of the current (in L/R time scale). Feedback control of the plasma shape keeps the separatrix shape (plasma â??wall gaps) as it was at the end of current flattop. At the end of current flattop the auxiliary heating is started gradually increased from its value at the end of current flattop, 48MW, to its maximum value, 73MW (33MW NBI + 20MW ICH +20MW ECH). In this scenario plasma current slowly decay to 7.7MA (at 2000s) staying deeply in H-mode (Ptot/Pthreshold > 2.3) with significant fusion power (64MW at 2000s). Plasma density is reduced keeping the same Greenwald ratio as it was at the end of plasma current flattop (0.82). At 2000s 33MW of NBI is switched off ( Feedback control of plasma current is switched on at 2020s. Since that time (7.7MA) plasma current is reduced rapidly (in 64s) to 1.8MA in divertor magnetic configuration. Plasma elongation is reduced consistently with reduction of the plasma current. In the simulation of this scenario plasma vertical stabilization is produced by the outer superconducting coils (VS1 system). All engineering parameters and the plasma-wall gaps are within the allowable limits. Such scenario allows increase of fusion fluence more than by a factor of 2 relative to the scenario with standard plasma current ramp-down (with burn termination at the end of current flattop). Moreover such scenario with the longest plasma current ramp-down require only slow reduction of the plasma density, which could be beneficial in the in the light of rather uncertain capability of density pumping performance during the current ramp-down (behavior of walls at this phase of scenario in ITER is rather uncertain). At present we have DINA data necessary for plotting of waveforms of the main plasma and engineering parameters. By the end of February we should have report on the corresponding DINA contract where this scenario will be described in one of the report section. | ||
| Resource Requirements: | beams for ITER-like plasma in DIII-D. ECRH may be needed for heating and for low field breakdown. | ||
| Diagnostic Requirements: | Standard core and pedestal diagnostics for core and edge ne, Te, Ti, MHD activity and impurity monitoring. | ||
| Analysis Requirements: | Analysis of plasma confinement, current profile and MHD | ||
| Other Requirements: | No | ||
| Title | 173: Study of L-H transition with ITER-like target plasma with opaque SOL | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | L-H Transition | Presentation time: | Requested |
| Co-Author(s): | A.R. Polevoi, M. Fenstermacher, R. Groebner, P. Gohil, L. Baylor, P.Snyder | ITPA Joint Experiment : | No |
| Description: | Study of L-H transition with ITER-like target plasma with opaque SOL | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Take as a reference a scenario in deuterium with heating marginally sufficient for L-H transition (preferably pure RF w/o core fuelling from NBI). Increase the SOL opacity trying to get a transition from L-mode to I-mode like profiles with pedestal for temperature, T, and w/o pedestal for density, n (instead of L-H). (The SOL opacity can be increased by SOL heating or by increase of the distance between the separatrix and PFCs by shaping keeping the same plasma surface area, S as in the reference case). Start core fuelling by injection of small pellets to check whether the pedestal for density will appear and whether it is possible to control pedestal pressure by fuelling to avoid ELMs. The studies with opaque SOL will help to assess the ??natural? ETB height and width (caused by transport, rather than by peeling-ballooning limit or ionization length), to study control of pedestal density by pellet fuelling (= simulation of ITER fuelling), to reduce the effect of recycling on particle transport studies (with pellet fuelling). | ||
| Background: | The SOLPS predicts for ITER the SOL opaque for penetration of the edge neutrals. Thus, even in presence of the ETB after the L-H transition the pedestal appears on temperature profile (similar to I-mode). Pedestal on density appears only with core fuelling by pellets. Thus, the density in the pedestal area is purely controllable (by pellet fuelling) which potentially can help to avoid ELMs keeping the edge below the peeling-ballooning limit in future reactors. | ||
| Resource Requirements: | ECR, (+NBI) heating, fuelling by small pellets. | ||
| Diagnostic Requirements: | High resolution measurements of pedestal and SOL areas. | ||
| Analysis Requirements: | Stability analysis of the peeling-ballooning modes | ||
| Other Requirements: | Keep opaque SOL for the whole dischsrge. | ||
| Title | 174: L-H and H-L Power Thresholds in He Plasma Diluted by Hydrogen | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | L-H Transition | Presentation time: | Requested |
| Co-Author(s): | A.R. Polevoi, P. Gohil | ITPA Joint Experiment : | No |
| Description: | Study of dependence of L-H and H-L power threshold on helium contamination by hydrogen in ITER-like conditions. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Make a scan in the range ne ~ 2-5 1019m-3 with EC heating only and He fuelling by gas puffing. For H-L threshold studies: For each density of this scan create an type-I ELMy H-mode plasma in a pure He. Then increase fraction of hydrogen gradually by LHS hydrogen pellet injection or hydrogen puffing keeping the electron density constant. Determine the critical hydrogen fraction for Type-I-III and for H-L transitions. For L-H threshold studies: For each density of this scan in the L-mode vary hydrogen fraction by LFS pellet injection keeping ne constant. Determine PL-H and P(Type-III-I). | ||
| Background: | In ITER pre-DT phase H-mode operation is more likely in He plasmas with dominant electron heating, low torque input, He fuelling by gas puffing in the range ne ~ 2-5 1019m-3. The LFS hydrogen pellet injection required for ELM pace making will cause He contamination by hydrogen. Core fuelling by hydrogen is also required if particle pinch is not sufficien to keep density above the NBI shine-through limit. Power threshold for the H- mode operation has different dependence on density in low density range for hydrogen and He. Thus, compound of the mix can be critical for H-mode operation. Experiments with H-NBI in He plasmas are not ITER relevant because of central fuelling by hydrogen, uncertain CX loss and impact of torque. | ||
| Resource Requirements: | EC heating, He-puffing and hydrogen puffing (or hydrogen pellets) | ||
| Diagnostic Requirements: | Resolution of He/H species in the mix. | ||
| Analysis Requirements: | |||
| Other Requirements: | No hydrogen NBI. | ||
| Title | 175: Control of TAE modes via external perturbations | ||
| Name: | Podesta | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Requested |
| Co-Author(s): | EP group | ITPA Joint Experiment : | No |
| Description: | Explore the possibility of affecting the TAE dynamics by means of externally applied, low-n perturbations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce L-mode plasmas with TAE modes. 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 rotating perturbations imposed through the Internal Coils and characterize modifications of TAE dynamics. | ||
| 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 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 (~kHz) | ||
| Diagnostic Requirements: | All fast ion diagnostics, ECE-I, possibly reflectometers | ||
| Analysis Requirements: | |||
| Other Requirements: | EFIT, TRANSP, NOVA | ||
| Title | 176: Comparative study of chirping TAEs in DIII-D and NSTX | ||
| Name: | Podesta | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | EP group | ITPA Joint Experiment : | No |
| Description: | Identify critical parameters regulating the dynamics of bursting/chirping TAE/RSAE modes in L-mode DIII-D and NSTX plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Identify DIII-D L-mode scenarios with bursting/chirping TAEs. Establish connection to NSTX scenarios by including discharges at low magnetic field (~0.6 T). Study the dependence of TAE dynamics on fast ion parameters (e.g. fast ion distribution function) and plasma parameters such as q-profile, thermal plasma density and temperature. | ||
| Background: | TAE modes in L-mode NSTX plasma usually present a bursting/chirping character, up to the extreme case when so-called 'avalanches' develop with subsequent enhancement of the fast ion transport and losses. Scenarios on DIII-D do not typically show these features. This study will investigate the conditions leading to the development of strong TAE bursts by comparing DIII-D and NSTX results. | ||
| Resource Requirements: | Scenarios at low magnetic field, approaching the standard values of NSTX (~0.55 T). Variable mix of NB sources (co- vs. counter, different NB injection energies). | ||
| Diagnostic Requirements: | Fast ion diagnostics (FIDA,FILD). internal mode structure measurements (ECE-I, reflectometer). q-profile data. | ||
| Analysis Requirements: | EFIT, TRANSP, NOVA. Possibly FIDASIM. | ||
| Other Requirements: | |||
| Title | 177: Parametric study of poloidal velocity and comparison with neoclassical predictions | ||
| Name: | Podesta | Affiliation: | PPPL |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | B. Grierson | ITPA Joint Experiment : | No |
| Description: | Expand previous studies on the parametric dependence of poloidal velocity on plasma parameters and comparison with neoclassical predictions | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Starting from the best scenario achieved in the 'DIIID/NSTX similarity experiment' (K. Burrell and R. Bell, FY2010), perform systematic scans of (i) magnetic field and current (including sign), (ii) toroidal rotation (through mix of co- and cntr- NB sources) and (iii) plasma profiles (e.g. Te/Ti ratio, etc.). | ||
| Background: | Previous results from NSTX and DIII-D have shown that the measured poloidal velocity is consistent with predictions from neoclassical models in specific conditions, namely at low magnetic field (and, arguably, lower temperatures than what is normally achieved on DIII-D). On the other hand, discrepancies between measured and neoclassical poloidal velocity are sometimes observed for more standard DIII-D operating regimes. The successful comparison between the two devices for NSTX-like conditions leads to exclude that aspect-ratio effects are important (K. Burrell, TTF 2011). This study will complement previous results from both DIII-D and NSTX. The range of parameters for which a comparison with neoclassical theory is performed will be extended to investigate the reason(s) of discrepancy between measurements and theory. | ||
| Resource Requirements: | Scenarios at low magnetic field, approaching the NSTX values (~0.55 T) | ||
| Diagnostic Requirements: | CER, main ion CER | ||
| Analysis Requirements: | EFIT, TRANSP, NCLASS, GTC-NEO | ||
| Other Requirements: | |||
| Title | 178: Model-based control of the current profile and βN for AT scenarios | ||
| Name: | Moreau | Affiliation: | CEA Cadarache |
| Research Area: | Plasma Control and General Issues | Presentation time: | Not requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University) | ITPA Joint Experiment : | Yes |
| Description: | This proposal aims at completing the development of model-based current profile control in AT operation scenarios started in 2011 on DIII-D. The goal is to apply the control as early as possible during the ramp-up phase in order to obtain, in a reproducible manner, various requested target q-profiles and betaN for the high-betaN phase of the advanced scenarios. The magnetic and kinetic profile control algorithms to be tested here have been developed in recent years [1-2] and implemented in 2011 in the DIII-D PCS. The first closed-loop experiments were performed successfully in September 2011 on DIII-D [3] (see background). Now the control method needs to be fully assessed with a broader variation of the target profiles and controller configurations and parameters. The control actuators are (i) co-current NBI power, (ii) counter-current NBI power, (iii) balanced NBI power, (iv) total ECCD power from all gyrotrons in a fixed off-axis current drive configuration, and (v) loop voltage. Off-axis NBI will possibly be used as an additional actuator for better flexibility in the profiles, depending on the execution of the experimental proposal # 182 ("Experimental identification of the plasma response to off-axis NBI"). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | After the successfull experiments in 2011, the full assessment of the proposed current profile control method will require one day in 2012. Prior to these new experiments, some dedicated tests in the PCS should be conducted in a couple of 2-hour preliminary sessions. In the experiment, the control of the poloidal flux profile, psi(x), and iota profile (i =1/q) will be activated with various psi(x), i(x) and betaN targets, starting control at t = 0.5 s for 5 seconds. First we shall reproduce the test made for i(x) control in our last 2011 shot, with the same profile controller parameters as in shot 147709 but with better Vsurf control, as in the previous shots. Then, the following shots will be dedicated to varying the psi(x), i(x) and betaN targets and, if necessary, the controller order and gains/weights. Particular attention will be given to the choice of the targets, actuator weights and anti-windup gain to avoid or limit the duration of actuator saturation. | ||
| Background: | Real-time control of the plasma current profile and betaN 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 [1-3]. 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 betaN) 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 [2], not only during the phase when the system identification was performed (t > 2.5 s) but also during current ramp-up (from t = 0.3 s). This shows the robustness of the data-driven models for control applications. The first experimental tests of this control method performed in September 2011 on DIII-D were successful and the results will be submitted to the 2012 IAEA conference through the ITPA (Integrated Operation Scenarios, Joint Experiment IOS-6.4) [3].
References: [1] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. [2] D. Moreau et al., Nucl. Fus. 51 (2011) 063009. [3] D. Moreau et al., "Integrated Magnetic and Kinetic Control of Advanced Tokamak Scenarios on DIII-D Based on Data-Driven Models", submitted to the IAEA Fusion Energy Conf., San Diego, 2012. | ||
| Resource Requirements: | NBI at full power is needed and with waveforms generated in real-time by the PCS, including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. Full power ECCD from 6 gyrotrons will also be required. Note: If the experiment proposal # 182 has not been executed ("Experimental identification of the plasma response to off-axis NBI"), the 150 beams should aim on-axis and should be able to provide the same input characteristics (geometry, voltage, etc ...) as in November 2009. | ||
| Diagnostic Requirements: | Real-time magnetic measurements, MSE and equilibrium reconstruction including the poloidal flux and the q-profile (RTEFIT2) are essential. Measurements of the density profile, toroidal rotation profile, as well as ion and electron temperature profiles are also required for analysis, not necessarily in real time. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 179: Material migration during ELM suppression | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | Plasma-material Interface | Presentation time: | Requested |
| Co-Author(s): | R. A. Pitts, A. W. Leonard, T. E. Evans, P. C. Stangeby, W. R. Wampler, D. Rudakov, J. A. Boedo, J. G. Watkins | ITPA Joint Experiment : | No |
| Description: | Measure material migration during ELM suppression | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Establish reliable ELM suppressed H-mode discharge. No decision has yet been made but it is likely that the chosen configuration will have to be one in which reliable ELM suppression can be guaranteed over many repeated discharges (required to inject enough 13C for post-mortem detection). Previous migration experiments have used an ELMing H-mode equilibrium (e.g. #123400) with low upper triangularity, high inner-wall clearance and primary divertor strike points on the lower shelf. Ideally, a migration experiment with a perturbed edge would use a similar equilibrium, but this is likely to require significant development to obtain an optimized shape and RMP configuration. In this case, the least time consuming solution would be to inject into an already optimized ELM suppressed discharge (e.g. #126006), possibly tweaking the configuration somewhat to increase the inner wall gap. This plasma has much higher triangularity than the older H-mode migration equilibrium, but has the advantage of being close to the expected ITER baseline inductive shape. A migration experiment with this discharge would then be similar to the earlier study at DIII-D in which injection was performed from the vessel floor into a high triangularity SNU discharge with drsep ~1.5 cm. Most of the 13C was in this case recovered from tiles in the secondary divertor region. In this new experiment, injection would be from the upper, toroidally continuous injection location and tests would be required beforehand to check that methane injection at the levels required for the migration session does not significantly perturb the ELM suppressed regime and perhaps to modify the inner wall gaps. These tests can be done with ordinary methane. Unfortunately, if an LSN equilibrium is used (to make sure of a reliable ELM suppressed discharge), diagnosis of the upper secondary X-point area would be compromised. The new main chamber wide-angle IR should provide some idea of the upper strike power deposition. Based on previous migration experiments on DIII-D in a standard H-modes, it Is prudent to request 2 or at least 1.5 days to provide the required pre-characterization and discharge set-up. | ||
| Background: | Focus to date in the field of material migration has been on understanding migration pathways in the plasma boundary under what might be called more conventional conditions. Experiments worldwide, including many excellent examples on DIII-D itself, have used L-mode or ELMing H-mode plasmas. Tracer gas, typically methane labeled with 13C, is repetitively injected at various different locations (one in each separate experiment) and plasma-facing components are then removed for post-mortem analysis to track the deposited carbon. Whilst these experiments (mostly in C dominated devices thus far), and subsequent modeling, have taught us much about the basic migration processes at work, they have not been performed in an ELM mitigated/suppressed edge, with which it is highly likely ITER will have to work. An ergodized plasma boundary implies different transport dynamics in comparison with the conventional, non-perturbed edge, and this is seen both in experiment and modelling. Simulations being performed for a magnetically perturbed SOL in ITER show the development of complex flow patterns absent in the undisturbed equilibrium. Such perturbations may, or may not strongly affect the migration pathways from main chamber to divertor and this needs to be tested, particularly if operation with magnetic perturbations in ITER does turn out to be the standard operational mode which is adopted. With its previous history in the migration area and the obvious fact that DIII-D has pioneered the use of RMPs for ELM suppression, the facility is well placed to continue these studies to the next level and investigate material movement in what might turn out to be a much more relevant plasma edge for ITER than the conditions which have been investigated to date.
- Resource Requirements: 1 run day of repeated ELM suppressed discharges for sufficient 13CD4 to be injected. At least 0.5 run days beforehand to set up the required target, establish the optimum methane fuelling levels and have all diagnostic in operation. Heating requirements as for normal ELM suppressed discharges. 13CD4 gas source. | ||
| Resource Requirements: | 1 run day of repeated ELM suppressed discharges for sufficient 13CD4 to be injected. At least 0.5 run days beforehand to set up the required target, establish the optimum methane fuelling levels and have all diagnostic in operation. Heating requirements as for normal ELM suppressed discharges. 13CD4 gas source. | ||
| Diagnostic Requirements: | IR cameras for divertor footprints and if possible the new main chamber system to gauge wall interactions (especially if using an ELM suppressed equilibrium with low inner wall clearance). Divertor Langmuir probes and if possible main SOL reciprocating probes for some idea at least of the far-SOL structure. Pedestal profiles. Divertor and midplane manipulators for sample insertion to have some measurement of 13C deposition at two locations which can be accessed faster than post-mortem analysis of DIII-D tiles. | ||
| Analysis Requirements: | will require extensive post-mortem surface analysis of a number of tiles extracted after the experiment (as done previously In DIII-D migration experiments). No decision has yet been taken as to how many, nor which tiles it might be necessary to remove, but it is likely that the number will be larger than has previously been the case in earlier migration tracer experiments. | ||
| Other Requirements: | The actual migration itself must be performed on the last run day of the campaign before in-vessel intervention required. Avoid all disruptions/off-normal events. Pre-characterization could be done earlier in the campaign. | ||
| Title | 180: Kinetic control of the toroidal rotation and temperature profiles together with the current profi | ||
| Name: | Moreau | Affiliation: | CEA Cadarache |
| Research Area: | Plasma Control and General Issues | Presentation time: | Not requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University) | ITPA Joint Experiment : | Yes |
| Description: | This proposal aims at providing a tool to study the physics of low rotation AT discharges on DIII-D. The goal is to control the full radial profile of the plasma rotation and ion temperature (or betaN) together with the current profile in advanced steady state discharges, and to achieve, in a reproducible manner, different current/rotation/temperature profiles for the high-betaN phase of the advanced scenarios.<br>The control actuators are (i) co-current NBI power, (ii) counter-current NBI power, (iii) balanced NBI power, and (iv) total ECCD power that was shown to have a significant effect on rotation [1]. The controller takes into account the coupling between the rotation/temperature profiles and the measured current profile, as well as the response to the various actuators. Off-axis NBI could possibly be included as an additional actuator for better flexibility in the profiles, depending on the execution of the experimental proposal # 182 ("Experimental identification of the plasma response to off-axis NBI"). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The demonstration of adequate control will require one or two experimental days, after some dedicated tests of the algorithm in the PCS have been conducted in a couple of short (2 hours) preliminary sessions.
First we shall reproduce the reference shot from the system identification series (1.8 Tesla, betaN-controlled AT scenario, at a central plasma density, ne0 ?? 5 ? 1019 m-3 and plasma current, Ip = 0.9 MA). In subsequent shots, betaN control will be disabled at t=2.5 s (i.e. after a 1 s current flat top) when rotation or temperature profile control will start for periods of time that will be short in the first tests, and will increase if the closed-loop response of the plasma is favourable, possibly up to the end of the flat-top phase. Then we shall request different target rotation and temperature profiles, including low and, possibly, zero rotation. If the tests are successful, attempts will then be made to control simultaneously the toroidal rotation and ion temperature profiles together with the current profile. | ||
| Background: | Real-time control of the plasma current and toroidal rotation profiles is important to achieve stable and reproducible operation of tokamaks in the advanced steady state regime. A multi-variable profile control approach based on data-driven, semi-empirical plasma models has been proposed in which the controller uses a combination of the available heating and current drive systems in an optimal way to control simultaneously the evolution of magnetic and fluid/kinetic plasma parameters and profiles [2-3]. The two-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.
The model to be used to determine the controller matrices has been obtained from system identification experiments performed on DIII-D in 2009. It was shown to provide excellent fits to the experimental data [1]. The development and experimental tests of such control methods is requested by the ITPA under the IOS group (Joint Experiment IOS-6.4). References: [1] D. Moreau et al., Nucl. Fus. 51 (2011) 063009. [2] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. [3] D. Moreau et al., "Integrated Magnetic and Kinetic Control of Advanced Tokamak Scenarios on DIII-D Based on Data-Driven Models", submitted to the IAEA Fusion Energy Conf., San Diego, 2012. | ||
| Resource Requirements: | NBI at full power is needed and with waveforms generated in real-time by the PCS, including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. Full power ECCD from 6 gyrotrons will also be required. Note: If the experiment proposal # 182 has not been executed ("Experimental identification of the plasma response to off-axis NBI"), the 150 beams should aim on-axis and should be able to provide the same input characteristics (geometry, voltage, etc ...) as in November 2009. | ||
| Diagnostic Requirements: | Real-time magnetic measurements, MSE, equilibrium reconstruction (RTEFIT2) and toroidal rotation profiles (CERQUICK) are essential. Measurements of the density profile as well as ion and electron temperature profiles are also required for analysis, not necessarily in real time. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 181: Measure counter streaming flows in 3D SOL | ||
| Name: | Schmitz | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | J. Howard (ANU), S. Allen, M. Fenstermacher, M. Lanctot, C. Lasnier, T. Weber (LLNL), T. Evans (GA) | ITPA Joint Experiment : | Yes |
| Description: | Recent EMC3-Eirene modeling for DIII-D ELM suppressed plasmas have shown that counter streaming flow pattern in the plasma edge and SOL can evolve. These flow pattern have the potential to alter or completely transform the particle transport due to the 3D boundary and might be in part an explanation for the density pump out. This has in particular a strong impact on the migration of eroded material and hence on the erosion/deposition balance. The aim of this experiment is to measure these 3D flow patterns with the interferometric fast camera setup developed by LLNL&ANU. This represents a unique diagnostic capability to measure this generic transport effect in the edge during application of 3D control fields and in particular during ELm suppression. As the EMC3-Eirene code is currently being used for modeling of the edge transport and divertor target loads with RMP ELM control fields. This experiment presents an important test bed to demonstrate the 3D flow dominated SOL transport characteristic. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | We want to run plasmas in ITER similar, high triangularity shape and apply the typical n=3 spectrum with n=1 EFC as used for ELM suppression.
This experiment will have two sequences: (1) L-mode plasma This is used to establish the method in a shallow profile plasma. Here in using different interferometric patterns, the flow of C ions and potentially D atoms as well can be measured aiming at a radial resolution in the flow pattern. Puffing of hydrocarbons and He is foreseen to enhance the signal. We aim at the following sequence: - no RMP plasma - RMP plasma, I-coil current 4kA, 0 degree phase - RMP plasma, I-coil current 5kA, 0 degre phase - RMP plasma, I-coil current 6kA, 0 degre phase (this scan shall increase the size of the flux tubes and make correlated flows stronger) - RMP plasma, I-coil current 4kA, 60 degree phase - RMP plasma, I-coil current 5kA, 60 degre phase - RMP plasma, I-coil current 6kA, 60 degre phase (the phase flip shall move the flow structure poloidally at the measurement position) - RMP plasma, I-coil current 6kA, 60 degree phasing, gas puff through Dimes capilary, CH, low flow - RMP plasma, I-coil current 6kA, 60 degree phasing, gas puff through Dimes capilary, CH, medium flow - RMP plasma, I-coil current 6kA, 60 degree phasing, gas puff through Dimes capilary, CH, high flow (this flow scan shall change the downstream conditions and alter the flow pattern and velocity) (2) H-mode with full ELM suppression - noRMP setup, get stable H-mode with type-I ELMs - RMP setup, get suppression, reference #132731 - RMP scan, step 1, from suppression current (#132731, i.e. 4.2kA, 60 degree phasing) - RMP scan, step 2, 5kA - RMP scan, step 3, 6 kA (this scan shall again increase flux tube size and make flows stronger) - try suppression at maximum I-coil with 0 degree phasing to flip flux tubes poloidally - gas flow scan, suppression setup, low gas flow - gas flow scan, suppression setup, medium gas flow - gas flow scan, suppression setup, high gas flow (this scan shall change the divertor conditions and hence drive/reduce the flows) -- in all discharges a nozzle from the upper sealing shall be used to puff He for He-1 line ratio measurements, see related ROF proposal --- | ||
| Background: | Imprints of the invariant manifolds of the separatrix where seen in heat and particle fluxes at DIII-D during ELM suppression and also in L-mode RMP experiments. They are a clear signature of formation of a 3D boundary. The separatrix manifolds generate 3D magnetic flux tubes of short connection length. Therefore the axisymmetric unform SOL gets transformed into a mesh of short connection length magnetic flux bundles around the separatix. Modeling with the 3D fluid Monte-Carlo plasma tarnsport code EMC3 coupled to a kinetic neutral modell Eirene have shown that these magnetic structures induce 3D plasma flows. As adjacent magnetic flux tubes connect to the target in a counter direction, the 2D flow pattern exhibits counter streaming flux patterns. This dense mesh of counter streaming flux tubes is likely to induce a very different transport in particular for the fueling hydrogenic particles as well as for the impurities realsed at the walls. Hence, the resulting fueling, exhaust and the resulting erosion/deposition balance might be changed. This has vital impact for the operational space and for the divertor life time at ITER and this experiment represents a tes bed for the recently identified challenges for transport with 3D fields during RMP ELm control at ITER. | ||
| Resource Requirements: | ISS plasma with ITER04 patch pannel
I-coils on maximum current capability, 0 and 60 degree phasing on same day, n=1 EFC all beams available, ECH for mode control | ||
| Diagnostic Requirements: | - LLNL/ANU flow measurement at lower Xpoint and - if setup already - at upper X-point as well
- TanTV cameras with C, D and He-I filters - DimesTV with CII and D_a - IR TVs - midplane FS with He filters - divertor FS (upper and lower divertor) - Xpt soft X-ray - all profile diagnostics, in part. high resolution TS, profile reflectometer - ECE-I - CER | ||
| Analysis Requirements: | - EMC3-Eirene
- kinetic EFITs for H-mode shots - TRIP-3D | ||
| Other Requirements: | |||
| Title | 182: Experimental identification of the plasma response to off-axis NBI. | ||
| Name: | Moreau | Affiliation: | CEA Cadarache |
| Research Area: | Plasma Control and General Issues | Presentation time: | Not requested |
| Co-Author(s): | M.L.Walker (GA), J.R. Ferron (GA), E. Schuster (Lehigh University) | ITPA Joint Experiment : | Yes |
| Description: | The objective of this experiment is to experimentally characterize the plasma response to off-axis NBI on DIII-D. It requires only a small number of shots (4-5) with off-axis NBI and is a simple extension to the experimental model identification that was performed during the last experimental campaign. These 2009 experiments allowed very good control-oriented models to be obtained to describe the response of the poloidal flux and toroidal rotation profiles to 5 actuators, namely co-current NBI, counter-current NBI, balanced NBI, ECCD and loop voltage (see reference [1]).<br><br>This new experiment will be very useful for the evaluation of off-axis NBI physics as they will provide an experimentally measured space-time plasma response that can be compared with theory-based modeling.<br><br>The off-axis current drive capability of the DIII-D NBI system will also provide additional flexibility for controlling the current profile during ramp-up and/or during the high performance phase of AT steady state discharges, and also for the combined control of the current profile, N and/or the toroidal rotation profile. The proposed experiment will provide essential data to complete the 2009 model in order to use off-axis NBI as an extra actuator for model-based control applications. This would therefore allow future real-time control experiments to be done while making use of the best heating and current drive mix available on DIII-D (see other proposals, e. g. # 178 & # 180). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental plan will be the same as in November 2009, but with modulations of the off-axis beams only (4-5 shots).
First we shall reproduce shot #140090, the reference shot without power modulations (2 shots). Then, up to t=2.5 s, all subsequent discharges will be similar to the reference one (1.8 Tesla, betaN-controlled AT scenario, at a central plasma density, ne0 ?? 5 x 1019 m-3 and plasma current, Ip = 0.9 MA). At t=2.5 s (i.e. after a 1 s current flat top), in all discharges, the Vloop control mode (i.e. the use of Vloop as a control actuator) is enabled and the Ip and betaN controls are disabled in order to avoid feedback in the response data. Between t=2.5 s and t=7 s, modulations of the off-axis beams will be applied, first 150L only, then 150R and both. The modulation waveforms are determined in advance and uploaded into "futureshot" files that can be readily used during the experiment. | ||
| Background: | The algorithms that are used to numerically identify the various elements of a semi-empirical plasma response model using experimental data have been developed for model-based control purposes and used successfully on JET, JT-60U and DIII-D [1-3]. The models relate a set of (machine-dependent) input parameters or actuators (e.g. H&CD powers) to measured output profiles for which control will be needed, namely, the current density (or safety factor) profile, which characterizes the magnetic state of the plasma, and one or several fluid/kinetic parameters and profiles (betaN, plasma rotation velocity, ion and/or electron temperature, etc ?).
A model-based controller can then use all the available heating and current drive (H&CD) systems in an optimal way to regulate the evolution of the plasma profiles [2-3]. The development of such integrated control-oriented models is requested by the ITPA under the IOS group (Joint Experiment IOS-6.4). References: [1] D. Moreau et al., Nucl. Fus. 51 (2011) 063009. [2] D. Moreau et al., Nucl. Fus. 48 (2008) 106001. [3] D. Moreau et al., "Integrated Magnetic and Kinetic Control of Advanced Tokamak Scenarios on DIII-D Based on Data-Driven Models", submitted to the IAEA Fusion Energy Conf., San Diego, 2012. | ||
| Resource Requirements: | NBI with various power waveforms is needed including on axis co-current (30 R/L, 330 R/L), off-axis (150 R/L) and counter-current (210 R/L) beams. This experiment requires that commissioning of the off-axis beams has been completed (full power modulations). Other additional heating and current drive systems from 6 gyrotrons will also be required although not essential. The discharges are to be run partly in the loop voltage control mode (PCS). The system identification requires the availability of the profile data from MATLAB. | ||
| Diagnostic Requirements: | Magnetic measurements, MSE, CER, Thomson. Equilibrium reconstruction including the q-profile (RTEFIT2) are essential, and measurements of the density profile, toroidal rotation profile, as well as ion and electron temperature profiles, are required. | ||
| Analysis Requirements: | MATLAB software / data stored in mdsplus. | ||
| Other Requirements: | -- | ||
| Title | 183: Measure counter streaming flows in 3D SOL | ||
| Name: | Schmitz | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | J. Howard (ANU), S. Allen, M. Fenstermacher, M. Lanctot, C. Lasnier, T. Weber (LLNL), T. Evans (GA) | ITPA Joint Experiment : | Yes |
| Description: | see proposal ID181 in 3D ELM control task force
I think this is an overlap topic and adresses transport effects relevant in both groups. |
ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 184: Imaging of 3D SOL flux tubes with He-I line ratio technique | ||
| Name: | Schmitz | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | M.E.Fenstermacher, T. Evans, E.A.Unterberg, J. Munoz | ITPA Joint Experiment : | Yes |
| Description: | We propose to use puffing of small amounts of He (1017 He atoms / s) through capilaries at the crowne of LSN plasmas as well as through the Dimes capilary to image the 3D magnetic flux tubes druing ELM suppression. Using suited spectral filters, electron density and temperature fields can be extracted. This can be done during any ELM controlled plasma given that the small He puffing is accepted. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | -- | ||
| Background: | Imprints of the invariant manifolds of the separatrix where seen in heat and particle
fluxes at DIII-D during ELM suppression and also in L-mode RMP experiments. They are a clear signature of formation of a 3D boundary. The separatrix manifolds generate 3D magnetic flux tubes of short connection length. Therefore the axisymmetric unform SOL gets transformed into a mesh of short connection length magnetic flux bundles around the separatix. Modeling with the 3D fluid Monte-Carlo plasma tarnsport code EMC3 coupled to a kinetic neutral modell Eirene have shown that these magnetic structures induce 3D plasma flows. As adjacent magnetic flux tubes connect to the target in a counter direction, the 2D flow pattern exhibits counter streaming flux patterns. This dense mesh of counter streaming flux tubes is likely to induce a very different transport in particular for the fueling hydrogenic particles as well as for the impurities realsed at the walls. Hence, the resulting fueling, exhaust and the resulting erosion/deposition balance might be changed. This has vital impact for the operational space and for the divertor life time at ITER and this experiment represents a tes bed for the recently identified challenges for transport with 3D fields during RMP ELm control at ITER. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | tanTV cameras equipped with red He-I line filters (667.8nm, 706.5nm, 728.4 nm and preferably 587.2nm as well) | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 185: Plasma rotation effects on Global and Compressional Alfven Eigenmodes | ||
| Name: | Kramer | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Fast-ion group | ITPA Joint Experiment : | No |
| Description: | Compressional and Global Alfven Eigenmodes (CAE/GAE) are routinely seen in NSTX and they have also been observed in DIIID (Heidbrink et al NF 46 (2006) 324). In NSTX those modes are linked to anomalous electron transport (Gorelenkov et al. NF 50 (2010) 084012). NOVA calculations for GAEs in NSTX have shown that an off-axis extremum in the Alfven continuum where the GAEs reside, can be created with a sheared rotation profile and a monotonic q profile. When the rotation is changed from co to counter in the NOVA simulations this extremum disappears and no GAE solutions could be found any more. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Excite the CAE/GAEs in a normal shear discharge as was done before (Heidbrink et al NF 46 (2006) 324) using co-NBI. In the next discharge replace the co_NBI with counter-NBI to reverse plasma rotation and see if the CAE/GAEs are still excited. Create a third discharge in which the plasma rotation is varied form high to low (or vise versa) and observe how the spectrum of CAE/GAEs change with rotation. It would be advantageous to use a toroidal field that is low enough to excite the modes and high enough so that the diagnostics that depend on the magnetic field still work (BES, ECE, ...). | ||
| Background: | In recent NOVA simulations of GAEs for NSTX it was found that the sheared plasma rotation profile played a crucial role in the formation of an off-axis extremum in the Alfven continuum. The GAEs are located at this extremum and the calculated eigenfrequencies agreed well with the observations. When the plasma rotation in the NOVA calculations was reversed this extremum disappeared and no GAE solutions were found indicating that the plasma rotation plays a crucial role in for the existence of the GAEs. | ||
| Resource Requirements: | Co- and counter NBI | ||
| Diagnostic Requirements: | All fast-ion diagnostics that work at low torodal fields | ||
| Analysis Requirements: | |||
| Other Requirements: | Main analysis codes: EFIT, NOVA, SPIRAL | ||
| Title | 186: ICRF absorption on beam ions | ||
| Name: | Kramer | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Fast-ion group, Nicola Bertelli, Rory perkins | ITPA Joint Experiment : | No |
| Description: | Fast waves can couple efficiently to high-energetic beam ions in the plasma and accelerate them to high energies. Those high-energy ions are not well confined in the plasma and lost near the low-field side mid-plane. Those losses reduce the ICRF heating and beam-ion fueling efficiency. In simulations performed for NSTX with the full-orbit following SPIRAL code it was found that as much as 50% of the full-energy beam ions can be expelled with fast waves. DIII-D is the ideal place to measure the ICRF-induced beam ion losses and study the synergy between NBI heating and fast waves. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Two complementary experiments can be done to study the NBI-ICRF synergy: 1. in a steady plasma with constant ICRF at 60 MHz inject pulses of NBI and record the evolution of the fast ions in the plasma with FIDA, neutron cameras, etc and observe the losses with the FILD detectors. Use pulses from the different beam lines to vary the pitch of the beam ions at birth and vary the acceleration voltage of the various beams. In this way the the coupling between the NBI ions at various pitches and energies can be obtained cleanly for comparison with simulations. 2. in a steady plasma with constant NBI inject pulses of ICRF at 60 MHz and record the change in the confined fast-ion population and the losses. In this experiment the coupling between ICRF and the whole beam-ion slowing down distribution is emphasized. 3. repeat the experiments with the ICRF tuned to 75 and 90 MHz to investigate the effects of higher harmonics. | ||
| Background: | In full-orbit simulations with fast waves and neutral beams for NSTX it was found that the fast waves can efficiently accelerate the beam ions from 90 to 200 keV or higher. Those high-energy ions are not well confined and got lost quickly to the LFS wall thereby decreasing the efficiency of the fast-wave heating and beam-ion fueling. | ||
| Resource Requirements: | Co- and counter NBI, ICRF at 60, 75, and 90 MHz and 1-2 MW of power (or higher). | ||
| Diagnostic Requirements: | FILD, FILD2, FIDA, NEUTRON DETECTION, ECI, and all other fast particle diagnostics | ||
| Analysis Requirements: | Main analysis codes: EFIT, TORIC, SPIRAL | ||
| Other Requirements: | |||
| Title | 187: Comparison of High-collisionality ELM mitigation regimes in DIII-D and AUG | ||
| Name: | Suttrop | Affiliation: | Max-Planck Institute for Plasma Physics |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | Todd Evans et al | ITPA Joint Experiment : | Yes |
| Description: | Produce ELM mitigation in comparable plasmas to ASDEX Upgrade ELM mitigation regime (Suttrop, PRL 2011). This concerns the high-collisionality case (Evans et al, NF 2005, 115467, 119690). The purpose is to come up with matching experiments in both machines which document the requirements for the high-collisionality ELM mitigation scenario. <br>The approach is to systematically work from the successful 2005 DIII-D scenario, changing one parameter at a time, so that the critical parameters for a change of ELM behaviour can be identified. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | 1. Begin by reproducing the documented DIII-D high collisionality ELM mitigation regime (115467 or 119690) with n=3 perturbation. If needed, adjust strike point location to change pumping and/or coils currents (increase from 4 kA to 6.3 kA).
2. Then repeat with emulation of n=2 perturbations. This can be made with symmetric or less symmetric coil current (i.e. different n=4 sideband spectral leakage). Both types should be tried to assess importance of perturbation spectra. 3. Adjust plasma shape (stepwise) towards the low triangularity shape used in the AUG experiments. 4. Measure density dependence (in gas puff ramps) at two different q95 values. If a threshold behaviour in density is observed, check pedestal collisionality values. 5. Repeat with "optimum" off-resonance coil setup (i.e. minimising the resonant field amplitude) | ||
| Background: | The decision to build a set of perturbation coils for ITER and the layout of such a system needs to be guided by current experimental results. Hence seemingly different experimental results from the various tokamaks should be sorted out, in order to prepare a common experimental observation base for the ITER decisions and help understanding the physics of ELM mitigation (or ELM suppression).
First AUG results with n=2 perturbations from 2x4 in-vessel coils (Suttrop, PRL 2011) differ from previous DIII-D high collisionality plasmas (Evans, NF2005) in plasma shape and perturbation mode number used (n=3 in DIII-D). A first experiment aiming to reproduce the AUG parameters in DIII-D in September 2011 reproduced the kinetic profiles, plasma shape and n=2 magnetic perturbation (see e.g. shots 146443, 146446) but did not show clear ELM mitigation in the fashion observed in ASDEX Upgrade. First attempts to reproduce the DIII-D 2005 high collisionality regime (119690) have been unscuccessful (146520, 146524) but the limited time then did not allow to work forward step-by-step from the old settings and hence find the reason for the discrepancy. This experiment is carried out as part of the ITPA pedestal group effort to understand ELM mitigation with perturbations from off-midplane coils (PEP-23) | ||
| Resource Requirements: | Detailed resource, diagnostic, analysis and other requirements are listed in D3DMP No.: 2011-01-05 (Todd Evans).
http://fusion.gat.com/pubs-int/MiniP/review/2011-01-05.pdf | ||
| Diagnostic Requirements: | Thomson scattering, CER to assess edge gradients and stability. | ||
| Analysis Requirements: | These plasmas may produce another test case for the EPED model of pedestal width and ELM suppression. | ||
| Other Requirements: | This experiment should be combined with #24 Todd Evans: Comparison of DIII-D and AUG high collisionality ELM response to 3D magnetic perturbations.
I'd appreciate run time allocation such that I can organise a trip to GA and participate, if possible. It may be great to split the experiment into two sessions in one week to have more time to analyse the first results and iterate on the strategy. I am happy to speak 5 min at the ROF but this would have to be by remote videoconferencing. | ||
| Title | 188: The rotation and torque due to the TBM | ||
| Name: | Tala | Affiliation: | VTT Technical Research Centre |
| Research Area: | Transport and 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 torque | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Based on the single shot where the TBM modulation was applied in the recent TBM experiments on DIII-D, this method turned out to be a very good one to understand the behaviour of rotation due to TBM. To be able to state robustly that the effect of the TBM on rotation originates from the edge, a few more pulses with TBM modulation would be needed. Now in the only pulse performed in 2011 experiments, the error field correction (EFC) by the I-coils was the one that we developed to maximize the toroidal rotation of the H-mode plasmas for intrinsic error field alone. Other options for EFC coils should be used, in particular those where the rotation changes due to TBM are maximised are important as they give data with higher signal to noise ratio. One 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. One way obtain more quantitative data is to use different TBM frequencies and compare the phase and amplitude profiles of the modulated rotation. | ||
| 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. However, this analysis is based only on single pulse analysis so far as only one TBM modulation pulse was executed due to the lack of time in 2011. | ||
| 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 | 189: Measure pedestal evolution with minimal core particle fueling | ||
| Name: | Canik | Affiliation: | ORNL |
| Research Area: | Pedestal | Presentation time: | Requested |
| Co-Author(s): | Leonard, Groebner, Osborne | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to try to enter a regime in which the pedestal density buildup between ELMs is clearly too large to be accounted for by the particle source alone, thereby showing the existence of a pinch. This will be accomplished by operating at high plasma current to minimize penetration of neutrals through the pedestal, and reversed Bt to maintain an attached divertor. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The overall approach is to run plasmas with relatively low heating power (compared to the LH threshold) in order to produce regular low-frequency ELMs, while minimizing core particle sources. Reversed Bt will be used to promote an attached inner divertor (similar to the Opaque SOL experiments of 2011), and a current scan will be performed up to 2MA. The input power will have to be tailored to be produce regular ELMs with long inter-ELM periods in order to measure the pedestal evolution. Further, the beam input will be minimized to reduce the core particle input, either by running ECH-only or by taking companion discharges with some NBI replaced with ECH in order to separate the effect of beam fueling. | ||
| Background: | During recent experiments, a plasma current scan was performed at low input power (~2 MW; shots 144981/77/87). This produced long inter-ELM periods so that the pedestal evolution could be precisely measured, and the time dependence of the particle content of the plasma analyzed. SOLPS modeling has been performed to yield the neutral recycling source inside the separatrix. This analysis indicates that as plasma current is increased, the particle source inside psi~0.96 is reduced compared to the rate of increase of the particle content inside the same radius. At the highest plasma current (1.5 MA), the analysis indicates that the density rise was too large to be explained by fueling; this would then require a pinch to explain the observed density increase. However, it is known that 2-D codes do not capture the detachment of the inner divertor correctly and that this can affect inferred fueling rates by a factor of as much as ~3. Therefore, to be more conclusive, this experiment should be repeated but ensuring an attached inner divertor if possible. In addition, the plasma current should be maximized to extend the observed trend of fueling vs. density rise to a more convincing point. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Thomson, CER, filterscopes, floor probes, IR cameras | ||
| Analysis Requirements: | Profile analysis with python tools, SOLPS/UEDGE/OEDGE | ||
| Other Requirements: | |||
| Title | 190: RF-Only H-Modes | ||
| Name: | Taylor | Affiliation: | PPPL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | R. Pinsker, J.C. Hosea, R. J. Perkins, R. V. Budny, P. Ryan, R. W. Harvey, A. Nagy | ITPA Joint Experiment : | No |
| Description: | This proposal aims to generate H-mode discharges using a combination of fast-wave (FW) and electron cyclotron (EC) heating. The experiment includes the measurement of FW generated power flows to the lower divertor and FW modulation to assess the FW coupling efficiency to the bulk plasma. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A plasma diverted on the lower divertor would be used for this experiment in order to study FW power flows in the scrape off layer with the large array of diagnostics available in the lower divertor. The FW and EC systems would be configured for co-Ip current drive.
There are several phase to the experiment: 1.Generate FW H-mode and add EC power later in the H-mode phase to look for synergies. 2.Generate EC H-mode and add FW power later in the H-mode phase to look for synergies. 3.Reduce the FW and EC power to find the L-H transition power 4.Generate FW H-mode with maximum FW power available and modulate FW power to get FW coupling efficiency, ensuring that the FW power does not fall below the L-H transition power | ||
| Background: | For the first 2-3 years of ITER operation there will be no neutral beam injection (NBI) and experiments will rely on FW and EC heating to generate and sustain plasmas in the H-mode regime. High-harmonic FW (HHFW) experiments on NSTX have generated HHFW-only H-modes with only 1.4-3 MW of RF power, albeit with the assistance of lithium conditioning to reduce edge density and improve FW coupling. Recently on NSTX an Ip=300 kA H-mode discharge with 70-100% non-inductive current fraction was generated with only 1.4 MW of HHFW power using current drive phasing. A significant power flow along magnetic field lines in the scrape off layer was observed in NSTX during HHFW-only and HHFW+NBI H-modes that may be an important loss mechanism in FW-heated H-modes at lower FW harmonics in DIII-D and in future in ITER. RF-only H-modes have been generated previously in DIII-D; in the late 1980??s with electron cyclotron heating [J. Lohr, et al. Phys. Rev. Lett. 60, 2630 (1988)] and in 1991 with fast wave heating [C. C. Petty, et al. Phys. Rev. Lett. 69, 289 (1992)], but have not been studied since. In addition FW heating with current drive phasing was never used to generate FW H-modes in DIII-D any synergies between EC and FW heating were never investigated in DIII-D RF-only H-modes. | ||
| Resource Requirements: | Machine Time: 1 day Experiment
Number of gyrotrons: 4-5 Number of neutral beam sources: beam blips for MSE Three FW systems, one at 60 MHz and the others at 90 MHz. | ||
| Diagnostic Requirements: | MSE, ECE, edge and profile reflectometry | ||
| Analysis Requirements: | TORIC, GENRAY | ||
| Other Requirements: | -- | ||
| Title | 191: ECE imaging of core-localized TAEs | ||
| Name: | Kramer | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Fast-ion group, Ben Tobias | ITPA Joint Experiment : | No |
| Description: | Image the core-localized TAEs (C-TAE) that are observed before giant sawtooth crashes and determine if they begin as reversed-shear Alfven eigenmodes (RSAE). At the onset of c-TAEs it is often observed that the mode frequency increases first before it decreases. When the q profile inside the q=1 surface is monotonic no frequency increase should be observed. A possible explanation of the initial frequency increase at the onset is that the q profile at the plasma center is reversed and that the C-TAEs begin their life as reversed-shear AEs. Mirnov coils at the plasma edge only detect the C-TAEs that have sufficient amplitude at the edge. The ECE-I system is well suited to detect the C-TAEs earlier in the core when they are not yet visible on the Mirnov coils and by measuring the mode structure at the high and low-field side one can distinguish between RSAEs and C-TAEs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In an up-down symmetrical sawtoothing discharge inject sufficient ICRF (60 MHz at 2-3 MW?) to stabilize the sawtooth instability. For creating a sufficiently high fast-ion population some NBI might be injected to which the ICRF can couple efficiently to fast ions. When NBI is used try to keep the plasma rotation low (up to ~20 kHz) by using balanced injection to avoid complications due to the rotation in the analysis. Image the plasma center with the ECE-I system at both the low and high-field side to detect the core-localized TAEs and possibly RSAEs. Measure also the q profile with BES, fast-ion contents with FIDA, and fast-ion losses with the FILD detectors. | ||
| Background: | Core-localized TAEs are frequently observed just before giant sawtooth crashes. From modeling it was found those TAEs are born in the plasma center when q on axis drops below the value that those TAEs can exist. When q decreases further the resonant surface expands and therefore the TAE moves away from the plasma center and decreases in frequency. In a large number of experiments, however, it is observed that the modes do not decrease in frequency from the start but they increase in frequency first. An explanation for this behavior can be that the q profile during sawtooth stabilization experiments is reversed in the core and that the core TAEs start of as RSAEs. ECE-I might be able to detect those RSAEs before they become visible in the Mirnov coil signals at the plasma edge. | ||
| Resource Requirements: | ICRF at 60 MHz at a power level that the sawtooth are stabilized, Co- and counter NBI. | ||
| Diagnostic Requirements: | ECE-I, BES, all fast-ion diagnostics. | ||
| Analysis Requirements: | Main analysis codes: EFIT, NOVA | ||
| Other Requirements: | |||
| Title | 192: Optimizing fast-wave power coupled to the DIII-D plasma core | ||
| Name: | Perkins | Affiliation: | PPPL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | J. C. Hosea, G. Taylor, R. Pinsker, C. Lasnier, J. Watkins, P. Ryan, S. Diem, C.K. Phillips, R. Budny, R. Harvey, N. Bertelli | ITPA Joint Experiment : | Yes |
| Description: | This experiment will enhance the core-plasma heating by fast-wave (FW) power as much as possible by improving the plasma-antenna loading while simultaneously minimizing edge losses. These results will be sought by varying the antenna-plasma gap and edge density, using the stored energy and pressure profiles to compute core coupled power and monitoring energy confinement and divertor diagnostics for FW edge losses. Enhanced loading is of course important for reducing the antenna voltage required for coupling high levels of FW power (â?¥ 3 MW), but additional precautions are needed to ensure that the FW power is not coupled straight into an edge-loss mechanism.
Recent experiments on DIII-D have shown that most of the FW power can indeed be coupled to core electrons but only under certain conditions. With an antenna-plasma gap of 4.5 cm (shot 146571), the stored-energy response to the FW pulse indicates that, within the uncertainties, approximately all the FW power was coupled to the core plasma. However, the plasma-antenna loading was not sufficient to sustain fast-wave power above ~ 2 MW. In principle, the loading improves with a smaller gap, but the experiment proposed here will test this hypothesis. Indeed, shots with a reduced gap (down to 3 cm, shots 146578 and 146579) did not result in a larger rise in stored energy, and preliminary analysis suggests a combination of degraded discharge confinement and possibly also edge losses that prevent FW power from reaching the core. Such edge losses were found for NSTX FW-heated H-modes, where a large fraction of the FW power can flow along SOL magnetic field lines in front of the antenna and to the divertor [1,2,3]. Thus, the experiment proposed here, in addition to analyzing the core energy, will look for FW power deposition in the divertor regions using an IR camera and Langmuir probes. Combined with computational analysis (GENRAY, CQL3D, TORIC, AORSA), we expect to determine where the coupled FW power is deposited and how the antenna-plasma gap and edge density can be arranged for optimal core FW heating. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will first reproduce the conditions for which the highest level of FW coupling was achieved with demonstrated efficient heating of the core plasma (shot 146571). However, a lower-diverted plasma will be used so that any FW power coupled to the SOL will be preferentially lost to the lower divertor and can be measured with an IR camera as well as Langmuir probes. A systematic scan of the outer gap size will then be performed to see if core coupling degrades and edge losses increase as the gap is made smaller, approaching the 3 cm gap which has been shown to be too small to support efficient core heating. Secondly, scans of edge density, obtained first through gas puffing and then by increasing the overall density, will then be performed for the smallest gap that gives a coupling efficiency similar to that for shot 146571. This experiment will therefore determine the enhancement of antenna plasma loading that can be obtained while documenting any degradation of core coupling and increase in edge losses. | ||
| Background: | Antenna loading essentially determines the amount of RF power coupled to the plasma. However, a portion of this power may couple to the SOL and subsequently be lost to the divertor, giving rise to a situation of high loading but poor core heating. Such an edge loss of RF power not only degrades the overall efficiency of an RF system but can also lead to sputtering and erosion, which is of serious concern for ITER where the planned 20 MW ICRH system could seriously erode the machine if even a small fraction of this power couples directly to the divertor. FW-heated H-mode experiments on NSTX have shown that a significant fraction of the high-harmonic fast wave power can couple to magnetic field lines in the SOL in front of the antenna and subsequently propagate to the divertor, causing strong RF-induced heating of the divertor plates when the density close to the antenna exceeds the onset density for FW propagation [4]. We believe that this phenomenon is related to surface wave propagation and could be a universal effect for FW systems and that these edge losses must be minimized on DIII-D during the quest for very high power FW operation.
Recent experiments on DIII-D have successfully raised the amount of FW power coupled to the antennae, stressing the need to efficiently couple this power to the core. Preliminary analysis suggests that, for proper gap size and edge density, most of the power successfully couples to the core, but, for small gap sizes, the core heating is not as good and the edge losses are apparently enhanced. Similarly, Langmuir probe data for previous FW experiments (e.g. shot 141517) show that some Langmuir probes respond strongly to the RF pulse, suggesting that a fraction of the FW power is flowing through the SOL and going straight to the divertor. Surface wave propagation was previously proposed as a significant loss mechanism for low current-drive efficiencies on DIII-D [5], but the topic was not pursued due to lack of proper diagnostics. The recent increase in FW power along with the IR and probe diagnostics now in place will make these effects more clear and distinct. Finally, comparison of ECH and FW heating may help further determine the difference between inefficient core heating due to degraded confinement and inefficient heating due to FW power being lost at the edge. | ||
| Resource Requirements: | Machine Time: 1 day Experiment
FW heating systems All gyrotrons All available NB sources (8 MW as for shot 146571) | ||
| Diagnostic Requirements: | All standard diagnostics, IR cameras (including an upcoming periscope camera system with a wide-angle tangential view), Langmuir probes, Thomson scattering, 40-channel ECE radiometer, fast magnetics, all available fluctuation diagnostics. | ||
| Analysis Requirements: | GENRAY, CQL3D, TORIC, AORSA, EFIT | ||
| Other Requirements: | [1] J.C. Hosea et al., AIP Conf. Proceedings 1187 (2009) 105.
[2] G. Taylor et al., Physics of Plasmas 17 (2010) 056114. [3] R. Perkins et al., Bull. Amer. Phys. Society, 56, 16 (2011) 28 [4] J.C. Hosea et al., Physics of Plasmas 15 (2008) 056104. [5] C. Petty et al., Nucl. Fusion, 39, 10 (1999) 1421 | ||
| Title | 193: Determination of minimum confinement degradation for ELM supression with I coils | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | T. Evans, O. Schmitz. M. Fenstermacher | 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 and coil spectrum (n=2,3) | 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 and a low delta configuration with optimum q95 ~3.5, n=3 and a pre-I coil betaN_1.8. Document the pressure decrease at ELM suppression and the associated plasma energy decrease. Repeat at higher beta (beta_N = 2.2). Repeat the experiment with n=2 | ||
| Background: | ELM suppression can cause a decrease of the pedestal pressure and overall plasma energy but the level of decrease depends on plasma configuration and on the level of current in the I-coils | ||
| 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 analisys | ||
| Other Requirements: | |||
| Title | 194: Best match for correcting the TBM error field | ||
| Name: | La Haye | Affiliation: | Retired from GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | R, Buttery, M. Schaffer, J. Snipes, E. Strait | ITPA Joint Experiment : | Yes |
| Description: | The C-coil will be run non-standard with the two adjacent single window panes at 259 and 319 with opposite currents; the two vertical legs at 289 degrees make a field that is the best match to that of the TBM at 270 degrees. This should be the best we can do with existing coils in the direction of better matching between the TBM (localized toroidal field bump at the plasma closest approach) and the correcting coils we have. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | An ITER-like plasma (H-mode, betaN~2, q95~3 etc) would be run and the TBM field applied starting with the best no-TBM n=1 EFC; it is expected that rotation and confinement drop. The C259-319 current would be varied to best recover the rotation noting that a good starting guess can be made by minimizing the sum of the TBM and C-coil Btor perturbation at the plasma. | ||
| Background: | The first set of TBM n=1 EFC experiments in ohmic low beta plasmas showed virtually complete recovery of the low density locked mode limit by redetermining the I240 correction currents. However in discharges closer to ITER-like H-mode at betaN~2, little recovery could be made. Of course the I240 coil makes n=1 and the TBM makes many n-components etc. A scheme for better correction of the deleterious effect of the TBM on rotation needs to be found for ITER. | ||
| Resource Requirements: | TBM back on. Rerun ITER-like H-mode plasmas (although ohmic ITER-like also of interest). I240 coil standard. C-coil reconfigured (on patch panel) for same opposite currents in C259 and C319; but a higher order effect would be to keep the capability of imbalancing C259 and C319? | ||
| Diagnostic Requirements: | Standard, particularly CER for rotation. | ||
| Analysis Requirements: | Standard. | ||
| Other Requirements: | -- | ||
| Title | 195: Compatibility of ELM suppression with pellet injection | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, T. Evans, O. Schmitz, M. Fenstermacher, N. Commaux | ITPA Joint Experiment : | No |
| Description: | This experiment is aimed at demonstrating whether pellet injection in a suppresses ELM regime will unavoidably lead to ELMs being triggered by the pellets or if this can be avoided. In particular the effects of the pellets themselves in the possible triggering ELMs versus the appearance of ELMs by the increasing collisionality associated with pellet fuelling will be addressed. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiment would start from a high delta q95 ~3.5 with bN_2.0 and ELM suppression with n=3. In these conditions pellets of the smallest size and velocity (similar to those used for pacing) will be injected aiming at not changing the plasma density. If no ELMs are triggered the size, velocity and the frequency of the pellets will be increased until ELMs appear. The energy losses and ELM power fluxes for these ELMs will be determined | ||
| Background: | The injection of shallow pellet can lead to the triggering of ELMs and this remains an open compatibility issue of ELM suppression with pellet injection for plasma fuelling in ITER. | ||
| Resource Requirements: | NBI heating, I-coils in n=3 configuration and pellet injection from various locations and with different frequencies and velocities. | ||
| 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 | 196: Direct measure of bootstrap current and compare to models | ||
| Name: | Grierson | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | G. Kagan | ITPA Joint Experiment : | No |
| Description: | Measure the bootstrap current near the electric field maximum in a low collisionality pedestal, as well as the electric field itself. By looking at different spatial locations/shots find this current dependence on the radial electric field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Lithium beam diagnostics similar to the one described in D. Thomas et al Phys. Rev. Lett. 93, 065003 (2004). By looking at different spatial locations/shots find the bootstrap current dependence on the radial electric field. | ||
| Background: | The strong radial electric field, inherent to a subsonic tokamak pedestal, cannot modify drift orbits of electrons as it does for ions, because the poloidal gyroradius of the former is much less than that of the latter. Indirectly, however, electrons do feel the electric field through their friction with ions, whose net flow is substantially modified by this field. A revised expression for the bootstrap current including the effect of the electric field predicts that in a banana regime pedestal this current is larger than it is given by conventional neoclassical formulae. Due to indubitable practical importance of the bootstrap current direct observation of the described effect could strongly impact major tokamak experiments. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Lithium beam diagnostics similar to the one described in D. Thomas et al Phys. Rev. Lett. 93, 065003 (2004). | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 197: Direct measure of main-ion temperature in pedestal | ||
| Name: | Grierson | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | G. Kagan | ITPA Joint Experiment : | No |
| Description: | Perform direct measurements of the background ion temperature profiles for a wide range of the pedestal width to the poloidal ion gyroradius ratios. Determine if the discrepancy between the ion temperature and plasma density scales grows as this ratio goes from ~5 to ~0.5. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use edge CER system tuned to D-alpha and optimize plasma location and ELM characteristics for main-ion CER measurement. | ||
| Background: | A first-principle based analysis finds that in a banana regime pedestal the main ion temperature profile must be much wider than the drift ion orbit; i.e. its characteristic scale must be noticeably greater than the poloidal ion gyroradius. Plasma density does not have such a limitation and, in fact, is found to have a scale comparable to the poloidal ion gyroradius in many experiments. Hence, when the pedestal width to rho_pol ratio is small the ion temperature profile must be much wider than that of the plasma density, whereas once this ratio becomes larger the two profiles are allowed to have similar scales. Direct measurements of the main ion temperature by deGrassie supports this point in the pedestal as wide as (1/2)rho_pol, but comparing the two profiles in the series of shots with pedestal width to rho_pol ratio ranging from ~5 to 0.5 would provide a more solid evidence for the mechanism underlying temperature equilibration in a banana regime pedestal. Clarification of this mechanism is necessary for adequate theoretical description of pedestals. Currently the ion temperature profile is often taken to be as narrow as the pedestal itself when modeling H-Mode, which is justified by impurity ion temperature measurements. However, impurity ion species is more collisional than the main one, making physics behind establishing its temperature profile quite different. It is therefore crucial to measure the temperature of main ions directly rather than to deduce it from that of impurities to elucidate the issue. | ||
| Resource Requirements: | Dedicated use of edge CER at wavelength other than carbon | ||
| Diagnostic Requirements: | Edge CER at non-standard wavelength | ||
| Analysis Requirements: | Pedestal emission analysis from charge exchange between beam and fuel deuterium ions. | ||
| Other Requirements: | |||
| Title | 198: Fast Electron Driven Alfven Eigenmodes | ||
| Name: | Pace | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Requested |
| Co-Author(s): | R. Prater, J. Lohr, R. Granetz, G. Wallace, and the Energetic Particles Group | ITPA Joint Experiment : | No |
| Description: | Alfvenic modes can be excited by both energetic ions and electrons. Electron cyclotron resonant heating in DIII-D can be applied to the purposeful creation of a fast electron population that excites MHD modes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A survey of discharges that are amenable to maximum ECRH power will be identified. Through collaboration with the ECH group, we will purposely drive a large anisotropy in the electron population by way of oblique injection. The ECRH deposition location is adjustable, which will be applied to produce an off-axis fast electron distribution, analogous to the off-axis NBI experiments of 2011.
A major component of this plan is the addition of a companion experiment at Alcator C-Mod. The lower-hybrid (LH) current drive system at C-Mod will be similarly applied to drive an off-axis fast electron population. The combined experiment will therefore cover a parameter space from 1-8 T in magnetic field, and 10^18 - 10^20 m^-3 in electron density. | ||
| Background: | Recent work at DIII-D has quantified the transport of beam ions [1,2] in experiments that utilize early neutral beam injection during the current ramp to excite AEs. A natural progression of this work requires the ability to study wave-particle interactions in steady state plasmas that better approximate reactor conditions in which super-Alfvenic fusion alphas drive the modes. This also enables DIII-D to study fundamental features of AEs, including damping rates that are investigated elsewhere through the use of active MHD antenna systems [3]. An advantage of the fast electron excitation method compared to active MHD antennas is that it is theoretically capable of driving large amplitude and core localized modes.
Fast electron driven fishbones were observed during early experiments employing ECCD on DIII-D [4], suggesting that the increased power available from the present system is capable of extending this behavior to other modes. More recently, fast electron driven TAEs have been observed on Alcator C-Mod during lower-hybrid current drive experiments [5]. The successful generation of TAEs (modes for which there is a fundamental understanding [6]) through ECRH fast electrons will provide additional constraints on the physics of EC wave damping and serve as an additional constraint to models and codes that attempt to calculate the resulting electron distribution. [1] M.A. Van Zeeland, et al., Phys. Plasmas 18, 056114 (2011) [2] W.W. Heidbrink, et al., Phys. Rev. Lett. 99, 245002 (2007) [3] A. Fasoli, et al., Plasma Phys. Control. Fusion 52, 075015 (2010) [4] K.L. Wong, et al., Phys. Rev. Lett. 85, 996 (2000) [5] J.A. Snipes, et al., Nucl. Fusion 48, 072001 (2008) [6] W.W. Heidbrink, Phys. Plasmas 15, 055501 (2008) | ||
| Resource Requirements: | - gyrotrons at high power
- in-shot steering of ECRH | ||
| Diagnostic Requirements: | - ECE (considering non-thermal effects)
- Thomson (secondary temperature profiles) - hard X-ray detectors | ||
| Analysis Requirements: | - mode stability for the measured plasma profiles (NOVA-K or similar)
- electron distribution from CQL3D | ||
| Other Requirements: | -- | ||
| Title | 199: Compatibility of ELM suppression with n=3 and low torque input | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | O. Schmitz, T. Evans, M. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | Demonstrate that ELM suppression with n=3 is compatible with a low torque input heating such as expected in ITER | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Start from a high delta beta_N =2.0 q95 ~3.5 discharge with supressed ELMs with n=3 I-coils with co-injection NBI heating and the lowest possible current in the I-coil. Optimize the error field correction for this configuration. Then carry out a input torque scan by replacing in steps the co-NBI bu counter-NBI sources at constant input power until the lowest possible torque input and evaluate what changes are required in q95 or coil current to keep ELM suppression. Repeat scan replacing co-NBI with counter NBI at constant beta_N. | ||
| Background: | ELM suppression in ITER must be obtained with heating methods that provide very low momentum input. This has not been demonstrated so far | ||
| Resource Requirements: | Co and counter NBI and I coils | ||
| Diagnostic Requirements: | Core and pedestal plasma measurements, IR divertor power fluxes, MHD activity | ||
| Analysis Requirements: | Analysis of previous experiments and evaluation of conditions to perform the experiment | ||
| Other Requirements: | |||
| Title | 200: Fast Ion Transport by 3D Fields | ||
| Name: | Van Zeeland | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | EP Working Group | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate fast ion redistribution/losses from I-coil imposed field perturbations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will begin with discharge 146121 which utilized a 10Hz rotating n=2 I-coil field and small circular plasma. The discharge shape, current, beam mix, and field will be varied as well as I-coil modulation frequency and amplitude. Fast ion losses and transport will be measured using several diagnostic techniques including fast ion d-alpha (FIDA), pitch-angle and energy resolving scintillator detectors, foil detectors, 3.5 MeV neutron counters, and newly installed thermal infrared imaging of the vessel wall. The impact of the fields on the plasma will be measured with BES imaging and reflectometry. | ||
| 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 several exploratory discharges to investigate I-coil induced losses were also carried out. The preliminary I-coil discharges utilized a rotating n=2 perturbation which was observed to cause measureable loss signals. | ||
| Resource Requirements: | 150R, 30L, 330L/R, 210L/R, I-coils | ||
| Diagnostic Requirements: | Reflectometry, Fast Camera, FILD, Periscope IR View, FIDA | ||
| Analysis Requirements: | Same as TBM analysis | ||
| Other Requirements: | |||
| Title | 201: Energetic Particle Driven Chirping Instability in AI Startup | ||
| Name: | Van Zeeland | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | EP Group | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate the Alfven eigenmode frequency range chirping instability observed in recent DIII-D advanced inductive startups. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This half-day experiment will begin with the discharge in which chirping modes are observed (144889) and in a series of scans, the beam voltage and toroidal field will be varied to map out conditions for chirping instability to set in. These scans will also help identify the nature of the mode itself through measurements of the mode structure, frequency and impact on the fast ion profile. Mode structure will be measured with BES and ECEI while the impact on the fast ion profile will be measured with neutrons, ICE, FIDA and FILD diagnostics. | ||
| Background: | Energetic particle driven instabilities in the Alfven eigenmode frequency range that exhibit bursting and/or frequency chirping are not commonly observed in DIII-D plasmas. However, in recent advanced inductive experiments (144889) utilizing a hybrid-like startup with slightly lower field and higher beam energy than the reference discharge (142400), instabilities fitting this description were created repeatedly whereas only typical AEs were observed in the reference discharge. | ||
| Resource Requirements: | All beams except 210L | ||
| Diagnostic Requirements: | FIDA, FILD, BES, ECE, ECEI | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 202: Island Imaging with the SXR X-point Camera | ||
| Name: | Shafer | Affiliation: | ORNL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | Z. Unterberg, T. Evans | ITPA Joint Experiment : | No |
| Description: | This experiment is designed to optimize the plasma for internal plasma response measurements with the tangential-viewing X-point SXR camera by using: 1.) High-density, in-frequent ELMing plasmas (opaque SOL) with n=1 RMPs and 2.) Argon-puffing with n=3 RMP ELM-suppressed plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | High density provides the most effective signal enhancements for the SXR imaging system. Image integration times required are typically >10 ms, which require long inter-ELM periods for best image interpretation. This can be achieved with the so-called opaque SOL plasmas that used high-current and low-input power relative to the LH threshold (e.g. 146198). Inter-ELM periods were ~30-50 ms and pedestal densities were ~ 10^20 m^-3. With the goal to image islands (no requirement on ELM-suppression), large n=1 RMPs are ideal targeting the 3/1 near the X-point. Vacuum predictions put this island widths ~10 cm given the flux expansion in this region. Continuing with SXR-enhanced plasmas, we seek to inject Argon to enhance the signals during n=3 ELM-suppressed cases (this has not yet be done for this imaging system). This requires several dedicated shots to tune the puffing. | ||
| Background: | A tangential-viewing Soft X-Ray (SXR) camera was installed during the last run campaign to examine 3D field effects. Currently, the best results are measurements of the perturbed boundary with an enhanced-EUV energy filter. Energies measured extended down to the EUV range to enhance signal levels and identify the boundary. However, to examine the internal response, true SXR energy filters are used to focus at and with the pedestal. This reduces the light flux enough to require long integrations to get significant signals. Maximizing the density, tailoring the ELM characteristics, and using large n=1 perturbations provide the best conditions for internal measurements. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | Thomson, CER, Reflectometer, SXR arrays, SXRI | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 203: Sustain current profile of hybrid scenario by external off-axis current drive | ||
| Name: | Park | Affiliation: | ORNL |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate hybrid scenario without 3/2 NTM by sustaining the same q profile (q>1) as with 3/2 NTM by external off-axis current drive (off-axis NBCD + ECCD). Broaden pressure profile by off-axis beams and document change of stability limit. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | The hybrid scenario is an attractive baseline inductive operation scenario for ITER. The 3/2 NTM plays a key role in such hybrid discharges by broadening the plasma current profile, resulting in q>1 that stabilizes sawteeth and thereby prevents a trigger for the 2/1 NTM . Anomalous current diffusion in the hybrid scenario has never been fully understood, limiting a predictive capability for theory-based projection from the present-day experiments to large fusion devices such as ITER. There does not exist a quantitative theoretical basis on which anomalous plasma current broadening works for ITER to maintain q>1. The new capabilities of off-axis beams and high power ECCD will allow us to sustain q > 1 everywhere without anomalous current diffusion caused by the 3/2 NTM. The expected broader pressure profile by off-axis beams will also allow us to extend operation parameter space for the hybrid scenario. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 204: Assessment of effects of pellet pacing on average pedestal pressure and plasma confinement | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, N. Commaux, M. Fenstermacher | ITPA Joint Experiment : | No |
| Description: | The effects of pellet pacing on the pedestal pressure and the plasma confinement are characterized for various levels of ELM frequency enhancement and plasma conditions | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Starting from an ITER-like plasma with fELM = 5 Hz repeat ELM pacing scan by pellets with 20 Hz, 40Hz and 60 Hz. Adjust the level of input power down (if possible) and up so that the natural ELM frequency is changed by - 50% and +100% and repeat pacing scan at the same frequencies. Starting from the initial conditions increase gas fuelling so that the natural ELM frequency increases by 100% and repeat pacing scan. Finally decrease current/field by 50% and adjust power and repeat pacing scan. | ||
| Background: | Pellet pacing has demonstrated succesfully mitigation of ELMs by a factor of 10 and no confinement deterioration at DIII-D. It is necessary to investigate if this very good result in view of its application to ITER is valid over a larger range of pedestal conditions and natural ELM frequencies | ||
| Resource Requirements: | NBI, gas fuelling and pellet pacing | ||
| Diagnostic Requirements: | Core an pedestal plasma ne,Te,Ti, pellet diagnostics, divertor IR for ELM fluxes | ||
| Analysis Requirements: | Analyze existing pellet pacing results | ||
| Other Requirements: | |||
| Title | 205: Main-ion poloidal flow in pedestal | ||
| Name: | Grierson | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | G. Kagan | ITPA Joint Experiment : | No |
| Description: | Measure the main ion species poloidal flow near the electric field maximum in a low collisionality pedestal, as well as the electric field itself. By looking at different spatial locations/shots find the net poloidal velocity dependence on the radial electric field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Of course, it is highly desirable to measure the net poloidal velocity of background ions directly. However, even if it is only the toroidal component of the main ion flow that can be measured directly, but at the same time both toroidal and poloidal flow components can be measured for impurities, the main ion poloidal velocity can be recovered quite robustly through the pressure balance equation. | ||
| Background: | Poloidal flow of background ions is neoclassical in nature. In other words, it is due to the drift motion that ion gyrocenters undergo in the tokamak magnetic field line geometry. In a subsonic pedestal of a width comparable to the poloidal ion gyroradius a strong radial electric field arises to maintain pressure balance, making the corresponding gyrocenter orbits substantially different from their core counterparts. As a result, in banana and plateau regime pedestals, neoclassical phenomena become dependent upon the electric field. Most interestingly, a recent first principle study predicts that in the banana regime pedestal the poloidal flow of background ions is reduced in magnitude, or even reversed, compared to what is seen in the core. This prediction was indirectly confirmed by comparing the net poloidal velocity of boron impurities observed in the C-Mod pedestal with the first-principle based expression accounting for the electric field. Either measuring the main ion poloidal flow directly or deducing it through the pressure balance as described in the previous section should provide a more accurate knowledge of this flow. Hence, the proposed experiment would allow further verifying of the electric field effect on neoclassical flows in the pedestal. | ||
| Resource Requirements: | Edge CER tuned to D-alpha. | ||
| Diagnostic Requirements: | Edge CER at non-standard wavelength. Assessment of D-alpha fiducial for zero of rotation. | ||
| Analysis Requirements: | Modeling of direct charge-exchange and halo emission in the pedestal. May require edge neutral density calculation. | ||
| Other Requirements: | |||
| Title | 206: Edge rotation shear threshold for QH-mode access | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | A.M. Garofalo | ITPA Joint Experiment : | No |
| Description: | See background for details. | 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. 7 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 | 207: Investigate Poloidal Variation of Electrostatic Potential | ||
| Name: | Chrystal | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | Keith Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure the poloidal variation of the electrostatic potential within a flux surface for high rotation plasmas. This will be accomplished with CER measurements of impurity ions inside and outside the magnetic axis. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Create a plasma that has high carbon toroidal rotation and use CER tangential views to measure parameters of the carbon impurities. | ||
| Background: | Neoclassical theory predicts that when ion toroidal rotation is comparable to the ion thermal speed, density will no longer be a constant on flux surfaces. In this scenario, comparing ion density at different positions on a flux surface can be used to determine the change in electrostatic potential between the different positions.
CER chords that view inside the magnetic axis will be used to find the impurity density on two sides of a flux surface. Combining these measurements with knowledge of other plasma parameters allows the change in electrostatic potential around a flux surface to be calculated. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER, all other profile diagnostics. Special beam modulation for optimum CER data. | ||
| Analysis Requirements: | Comparison with neoclassical codes (NCLASS etc.). | ||
| Other Requirements: | |||
| Title | 208: Determination of minimum pellet/size penetration for ELM triggering by geometry optimization | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | L. Baylor, N. Commaux, S. Milora, M. Fenstermacher | 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. | 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 scheme included in ITER. The optimization of pellet pacing relies on the injection of the smallest possible pellets (to reduced througput) but that provide a sufficiently large perturbation in the pedestal. | ||
| 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 | 209: Time-dependent evaluation of plasma response spectrum | ||
| Name: | Hanson | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The proposed experiment would test a novel technique to identify the frequency dependence of the n=1 plasma response using an applied multi-frequency perturbation. In previous work, the spectrum has been identified by combining data from multiple discharges probed with single-frequency perturbations. The proposed method would attempt to obtain the same information using a perturbation with several superposed traveling waves, allowing the time-evolution of the spectrum fit parameters to be quantified. 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: | The experiment would ideally be done using low-triangularity, lower single null discharges for comparison with an existing body of single-frequency spectroscopy data. At fixed rotation and normalized beta, a spectroscopic waveform containing 3â??5 frequency harmonics would be applied. The performance of discharges with single-frequency spectroscopy waveforms and no spectroscopy would then be compared. As a second priority, the beta-dependence of the multi-frequency spectroscopy results would be investigated and compared with existing data. | ||
| 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 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 a fixed coupling parameter. However, the coupling parameter may vary with plasma equilibrium parameters such as shape and outer-gap. | ||
| Resource Requirements: | I-coils with SPA power suppies | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 210: Investigate n=2 plasma response | ||
| Name: | Hanson | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | M. Lanctot, S. Haskey | ITPA Joint Experiment : | No |
| Description: | Detailed investigations of the plasma response to slowly rotating n=2 perturbations are proposed. While preliminary n=2 plasma response data obtained in 2011 in a narrow range of normalized beta and q95 values indicates qualitative agreement with ideal MHD theory, dependencies on the applied frequency, I-coil phasing, and plasma parameters have yet to be investigated. This experiment would generate new data for comparison with theory, with a high degree of practical relevance for high-qmin advanced tokamak scenarios that seek to operate near the n=2 ideal wall limit and experiments exploring ELM suppression using applied n=2 perturbation fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment would ideally be done using low-triangularity lower single null discharges for comparison with a rich existing body of n=1 plasma response data. The following dependencies would be investigated and compared with ideal MHD theory: applied frequency, phasing of applied I-coil waveform, normalized beta, q95, and triangularity. These results would then help inform follow-up piggyback experiments in advanced tokamak and ELM suppression scenarios. | ||
| 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.
Although RMP-ELM suppression experiments will likely not approach the n=2 RWM stability limit, an understanding of the magnetic plasma response to the applied perturbation will likely be valuable for predicting, optimizing, and extending ELM suppression in scenarios with varied plasma parameters. | ||
| Resource Requirements: | I-coils configured for n=2, with SPA supplies | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 211: Further development of QH-mode with strong co-Ip NBI torque | ||
| Name: | Burrell | Affiliation: | GA |
| Research Area: | ELM Control | 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. | 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 2008 experiments and to 2) broaden the QH-mode operating space.
Optimization of existing conditions: 1) Find minimum possible target density by lowering gas injection rate early in the shot and moving beam start time as early as possible. 2) Extend QH-mode duration by operating at higher input power and torque Expand parameter space: 1) Scan Drsep and upper triangularity. 2) Vary safety factor by changing current and toroidal field. 3) Vary outer gap to see the effect on the EHO. | ||
| Background: | QH-mode with all co-injection was discovered during serendipitously during the 2008 campaign and a dedicated experiment was performed for one day. We have just barely begun the investigation of the 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 Ip operation. 7 NBI sources. | ||
| Diagnostic Requirements: | All profile and edge fluctuation diagnostics, especially edge BES and ECE-I for EHO studies | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 212: Investigate n=2 plasma response (Dup 210) | ||
| Name: | Hanson | Affiliation: | Columbia U |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | M. Lanctot, S. Haskey | ITPA Joint Experiment : | No |
| Description: | Detailed investigations of the plasma response to slowly rotating n=2 perturbations are proposed. While preliminary n=2 plasma response data obtained in 2011 in a narrow range of normalized beta and q95 values indicates qualitative agreement with ideal MHD theory, dependencies on the applied frequency, I-coil phasing, and plasma parameters have yet to be investigated. This experiment would generate new data for comparison with theory, with a high degree of practical relevance for high-qmin advanced tokamak scenarios that seek to operate near the n=2 ideal wall limit and experiments exploring ELM suppression using applied n=2 perturbation fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment would ideally be done using low-triangularity lower single null discharges for comparison with a rich existing body of n=1 plasma response data. The following dependencies would be investigated and compared with ideal MHD theory: applied frequency, phasing of applied I-coil waveform, normalized beta, q95, and triangularity. These results would then help inform follow-up piggyback experiments in advanced tokamak and ELM suppression scenarios. | ||
| 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.
Although RMP-ELM suppression experiments will likely not approach the n=2 RWM stability limit, an understanding of the magnetic plasma response to the applied perturbation will likely be valuable for predicting, optimizing, and extending ELM suppression in scenarios with varied | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 213: RWM stability boundary control using NBI feedback | ||
| Name: | Hanson | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to extend preliminary results in controlling the plasma response to an applied magnetic perturbation using feedback modulation of the neutral beams injected power. In a previous experiment in 2010, the control of the plasma response was demonstrated over a range of normalized beta, between 1.1 and 1.9, below the no-wall limit. Once it is demonstrated and optimized above the no-wall limit, this control technique could become a useful tool for avoiding disruptions in high performance discharges and for validating kinetic resistive wall mode (RWM) stability models. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Slowly rotating n=1 I-coil waveforms, in the range of 10-20 Hz, will be used to obtain plasma response measurements. Neutral beam power feedback using the toroidal amplitude of the plasma response as a feedback value will be used. A linear ramp in the plasma response target will be used to map out the range of achievable plasma response and beta values with the available beam power. Step function plasma response target waveforms will then be used to aid in optimizing gain settings and validating models for the feedback dynamics. | ||
| Background: | The response of a stable plasma to applied, low-n magnetic perturbations has been shown to be a sensitive indicator of RWM stability. This measurement technique is feasible above the ideal MHD no-wall limit as long as the plasma remains stable. Equilibrium control algorithms that use the plasma response as an input parameter may be able to facilitate high performance operation on the cusp of marginal stability. Control of the plasma response using NBI power feedback was demonstrated on DIII-D in 2010, but normalized beta values above the no-wall limit were not reached. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 214: Complete sawtooth stabilization | ||
| Name: | Kramer | Affiliation: | PPPL |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Energetic particles | ITPA Joint Experiment : | No |
| Description: | Long sawtooth periods can be obtained in sawtooth stabilization experiments with ICRF but almost always this period is terminated with a giant sawtooth crash. In the period before the giant sawtooth core-localized TAEs (C-TAE) are usually observed and from the evolution of the C-TAE activity it was deduced that the central q is decreasing slowly and resonant surfaces for C-TAEs with low toroidal mode numbers (n) become available. Those low-n C-TAEs have a large radial extend and can induce fast-ion transport from inside to outside the q=1 surface. The fast-ion pressure inside the q=1 surface is an important factor for the stabilization of sawteeth. By stopping the current diffusion to the plasma center by applying off-axis ECCD, the central q can be held to a value above the threshold for exciting low-n C_TAEs and the sawtooth-free period might be extended as long as the discharge last. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In an up-down symmetrical sawtoothing discharge inject sufficient ICRF (60 MHz at 2-3 MW?) to stabilize the sawtooth instability. For creating a sufficiently high fast-ion population some NBI might be injected to which the ICRF can couple efficiently. When the giant sawtooth regime is established inject ECCD near or outside the q=1 surface to stop the inward current diffusion. Use different diagnostics, Mirnov coils, ECE-I, BES, etc. monitor the C-TAEs activity and the q profile evolution in the plasma center. | ||
| Background: | Core-localized TAEs are frequently observed just before giant sawtooth crashes. From modeling it was found those TAEs are born in the plasma center when q on axis drops below the value that those TAEs can exist. When q decreases further low-n C-TAEs with a radial extend outside the q=1 surface develop. The strongly excited low-n C-TAEs are efficient in transporting fast ions from inside to outside the q=1 surface. The fast-ion pressure is a major contributor to sawtooth stabilization and when it drops below a certain value a sawtooth is triggered. By stopping the inward current diffusion low-n C-TAEs are avoided, the fast-ion pressure is kept high, and sawtooth instability is avoided. | ||
| Resource Requirements: | NBI, ECCD, ICRF at 60 MHz at a power level that the sawtooth are stabilized | ||
| Diagnostic Requirements: | ECE-I, BES, FIDA, FILD, all fast-ion diagnostics. | ||
| Analysis Requirements: | Main analysis codes: EFIT, NOVA, SPIRAL | ||
| Other Requirements: | |||
| Title | 215: Evaluation of advanced resistive wall mode control algorithms | ||
| Name: | Hanson | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J. Bialek, Y. In, G. Navratil, M. Okabayashi | ITPA Joint Experiment : | No |
| Description: | The proposed experiment seeks to compare the performance of three resistive wall mode (RWM) feedback algorithms: â??classicalâ?? feedback using proportional and derivative gain, a Kalman filter that simultaneously identifies n=1 and n=3 eigenmodes, and a state-space controller that incorporates a 3D electromagnetic model for the wall, sensors, coils and plasma. A comparative study of this nature will help inform the finalization of power supply requirements for ITERâ??s internal and external non-axisymmetric coil sets and provide useful data for comparison with RWM feedback simulation codes, such as VALEN [J. Bialek, et al., Phys Plasmas 8, 2170 (2001)]. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | If a pressure driven RWM target discharge is not available, controllers will be evaluated based on their ability to enhance the decay of the damped, n=1 plasma response that follows ELM and/or fishbone events, or the plasma response driven by a static n=1 perturbation. After obtaining a suitable no-feedback reference discharge, RWM feedback with the three algorithms will be tested in separate discharges. Feedback with the I-coils and C-coils will also be compared. | ||
| Background: | The control or avoidance of long-wavelength MHD instabilities that arise at high pressures in tokamak plasmas will likely be important for the success of steady-state, high fusion gain scenarios in ITER and for future tokamak devices that seek to maximize fusion output. One such instability, the n=1 RWM, has been successfully controlled using feedback with magnetic coils. In high-beta, RWM-stable plasmas, applying RWM feedback has been shown to enhance the decay of the n=1 plasma response driven by fast MHD events [H. Reimerdes, et al., Plasma Phys. and Cont. Fusion 49 (2007) B349].
VALEN eigenvalue calculations suggest that using a model-based, state-space control algorithm can greatly enhance the performance of external coil sets. In DIII-D, using the state-space algorithm is expected to enable RWM feedback with the C-coils for plasmas approaching the n=1 ideal-wall beta-limit, well above the beta value where proportional gain C-coil feedback fails. The present VALEN DIII-D model has roughly 1300 elements and includes the resistive vacuum vessel wall with port holes and flanges, poloidal and radial magnetic field sensors, and three-dimensional representations of the I and C-coils. In addition to DIII-D, VALEN has been used to design and simulate RWM feedback in the NSTX, HBT-EP, KSTAR, and ITER experiments. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 216: Probe kinetic stabilization of resistive wall modes | ||
| Name: | Hanson | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | M. Lanctot, J. Berkery, S. Sabbagh, I. Chapman, G. Navratil, F. Turco | ITPA Joint Experiment : | No |
| Description: | Test the kinetic resistive wall mode (RWM) stability model by systematic variation of off-axis NBI power and plasma rotation and density. Preliminary experiments in 2011 indicated increased damping of the RWM due to off-axis NBI in a regime where a second source of damping, a resonance between the plasma rotation and the bounce frequency of trapped ions is thought to also be present. The proposed experiment would help solidify understanding of RWM damping due to off-axis NBI by continuing to rotation values below the bounce-frequency resonance. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Feedback control of the NBI will be used to maintain normalized plasma beta near 2.3 and constant toroidal rotation. We will reduce the trapped ion fraction using the off-axis neutral beam to heat discharges where the toroidal field is in the same direction as the plasma current (positive B_T), and the fast ion beta will be reduced by maximizing the plasma current and thermal particle density. Throughout the experiment, the growth rate of the RWM will be measured using slowly rotating n=1 I-coil perturbations. The effect of magnetic braking on the stability of the mode will be documented using dominantly non-resonant (i.e. odd parity) n=3 magnetic perturbations. The level of off-axis NBI power, plasma density, and ramp rate of the plasma current will be varied to change the effect of fast ions on the mode. | ||
| Background: | The validation of RWM stability models with quantitative predictive capabilities is a critical issue for future burning plasma experiments, which will likely need to operate above the no-wall beta limit in order to meet fusion performance goals. Previous DIII-D and NSTX experiments have shown that the observed RWM stability above the no-wall limit is consistent with the kinetic stability model in the MISK code, which identifies the trapped thermal and energetic ions as being responsible for stabilizing the RWM above the no-wall limit [Berkery Phys. Plasmas 17 (2010), 082504; Reimerdes Phys Rev Lett 106, 215002 (2011); Sabbagh IAEA 2010 paper EXS/5-5]. One main difference between the two experiments is that NSTX can access the unstable RWM regime at "intermediate" values of rotation while, in DIII-D, the RWM is stable over the entire range of plasma rotation profiles. However, analysis of DIII-D experiments suggests that the RWM is only marginally stable in weakly-shaped LSN H-mode discharges when beta normalized is 2.3, the plasma rotation is 40 km/s (or 0.9 percent of the inverse Alfven time) at CER chord T6, and beta_fast is approximately 20 percent. The RWM is predicted to be unstable in this parameter range when the stabilizing effect of fast ions is excluded. The off-axis neutral beam is expected to lead to a decrease in the trapped ion fraction when the toroidal field is in the positive toroidal direction (i.e. co-Ip). Also, the ratio of the energies stored in hot ions and in the thermal plasma is inversely proportional to the plasma current and the plasma density. This identifies the optimal conditions for this experiment: a LSN H-mode discharge with zero upper triangularity (to minimize the stabilizing effect of poloidal shaping), positive B_T, Ip consistent with B_T for q95<4.0, pellet fueling and gas puffing to maximize the density. | ||
| Resource Requirements: | Positive B_T (co-Ip). 8 NB Sources are required. Pellet fueling. I-coil error field correction and testing waveforms. Experiment should follow boronization. | ||
| Diagnostic Requirements: | Equilibrium diagnostics including kinetic profiles (Thomson, CER, MSE) and fast ion diagnostics. | ||
| Analysis Requirements: | TRANSP/ONETWO calculations of fast ion pressure, Kinetic EFIT, MISK, MARS-K, and HAGIS. Analysis will be ongoing as more information on the performance of the off-axis neutral beam becomes available. | ||
| Other Requirements: | |||
| Title | 217: Impurity concentration in pedestal | ||
| Name: | Grierson | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell | ITPA Joint Experiment : | No |
| Description: | Standard CER on DIII-D measures the fully stripped carbon +6 temperature, velocity and density. However, in the plasma edge and SOL there can be other ionization states of carbon, making the inference of Zeff from C+6 only too low. A coronal model is implemented in the GAProfiles code and indicates that the C+5 and C+4 contributions can be non-neglibigle in some cases, and this model is a lower limit on the concentration of these other states of carbon. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform a survey of the available spectral lines for charge-exchange (for active spectroscopy) and electron impact excited line emission which fall within the available visible range of the CER spectrometers. Pursue active CX emission first, and if necessary evaluate the feasibility of the truly line-integrated impact excited lines. | ||
| Background: | The composition of impurity ions in the pedestal is of general interest as it enters into the calculation of bootstrap current and comparisions with neoclassical models of poloidal rotation. | ||
| Resource Requirements: | Survey of diagnostic feasibility and use of at least one of the available edge CER systems. | ||
| Diagnostic Requirements: | Edge CER. High resolution TS for ne and te profiles. | ||
| Analysis Requirements: | CX spectroscopy for active measurement. Line-integrated modeling for impact excitation. | ||
| Other Requirements: | |||
| Title | 218: Determination of NBI source effects on H-mode plasma density profile shape at low collisionalities | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This experimental proposal aims at the determination of the role of the particle source by NBI in the observed density peaking at low collisionalities in present experiments | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | The experiment consists on the achievement of two low density (no gas puffing) discharges in H-mode with on-axis and off-axis NBI and q~ 1 (without sawteeth if possible) and to determine the peaking of the electron and ion density. Following this additional ECRH or fast wave heating will be added in the central region to determine the effect of electron heating on the observed peaking with on and off-axis NBI heating. | ||
| Background: | Density peaking is usually observed in H-mode discharges at low densities/low collisionalities. However, these discharges are NBI dominated and with a significant core source from the beams. It is important to determine if the observed peaking is associated to the low collisionality or to the central fuelling with a view to extrapolating these density profiles in ITER. | ||
| Resource Requirements: | On and off-axis NBI, ECRH and fast wave heating. | ||
| Diagnostic Requirements: | Core and pedestal measurements of plasma ne,ni,Te, Ti and fluctuation measurements | ||
| Analysis Requirements: | Analyse existing off-axis NBI discharges with relevant plasma conditions | ||
| Other Requirements: | |||
| Title | 219: Effect of plasma shaping on edge rotation shear in L-mode plasmas and L-H power threshold | ||
| Name: | Fedorczak | Affiliation: | UCSD |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G.R. Tynan, G.R. McKee, Z. Yan, L. Schmitz, J. Boedo | ITPA Joint Experiment : | No |
| Description: | We investigate the influence of plasma shaping on the LH power threshold, and more precisely on the shear amplitude of (turbulence-driven) zonal flows at the edge of the plasma, prior to L-H transition. A model has been recently developed that consider the drive of zonal flows by poloidal asymmetric ballooning modes. This drive is highly sensitive to the edge plasma geometry (LSN, USN, DN, triangularity ...). The experiment aims at elucidating if the model supports the trends observed for the edge electric field profile function of the plasma geometry. It should bring critical information about the physics of the L-H power threshold dependence with plasma geometry. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental approach consists of scanning the plasma geometry while keeping other plasma variables constant. First set of comparison is: LSN plasma, its USN symmetric, and the DN (perfeclty symmetric). In each configuration, the power heating (no torque) is slightly increased until the H mode is triggered. The electric field profile and its evolution are monitored using DBS, BES and the midplane reciprocating probe. Then, we test a DN plasma for which the UN X-point is scanned poloidally. Repeat the scan for the LN X-point. It is important to keep a stationnary control of global plasma quantities to relate the edge evolution to the plasma geometry only. | ||
| Background: | Influence of plasma geometry on L-H power threshold in DIII-D has been investigated in the past (McKee), for LSN and USN, but there is currently no physical mechanisms for the trends that are observed on different tokamaks. It is an important topic to investigate what are the optimal plasma conditions to enter the H- mode in ITER. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Doppler Back scattering, midplane reciprocating probe for Reynolds stress measurements, BES | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 220: TBM mock-up effects on confinement at high β | ||
| Name: | Loarte-Prieto | Affiliation: | ITER Organization |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | J.A. Snipes, Y. In, N. Oyama, M. Schaffer, T. Strait | 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 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). Several methods to optimize the n=1 error field correction were attempted and some 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 may 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 will be devoted to DEFC to compare these two techniques at high βN. 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. Then, DEFC will be applied to re-optimize n=1 error field correction in the presence of the TBM mock-up fields at the same βN. These two methods to optimize error field correction will be compared and several discharges will be repeated with the optimum DEFC to check reproducibility and ECH may again be applied to reduce NTMs, if required. If time permits, low NB torque plasmas will be investigated at the same value of βN, possibly with additional ECH to reduce NTMs or reach sufficient βN and one or both of the error field correction techniques will 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 is required to optimize this technique. | ||
| 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. | ||
| Diagnostic Requirements: | Locked mode (RWM) sensors. Rotation measurements. Interferometer. Thomson scattering and ECE measurements. | ||
| Analysis Requirements: | |||
| Other Requirements: | References
[1] J. A. Snipes, et al, Proc. 37th EPS Conf. on Plasma Physics (Dublin, Ireland, 2010) 34A (ECA) P1.1093, http://ocs.ciemat.es/EPS2010PAP/html/author.html [2] M. J. Schaffer, et al., Nucl. Fusion 51 (2011) 103028. | ||
| Title | 221: Improvement and Characterization of ITER Steady-State Scenario | ||
| Name: | Doyle | Affiliation: | UC, Los Angeles |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | C. Holcomb, J. Ferron | ITPA Joint Experiment : | Yes |
| Description: | Address outstanding key performance and physics issues for the success of the steady-state mission on ITER. Specifically:
1) Increase performance, so as to *simultaneously* meet ITER Q and bootstrap fraction targets 2) Investigate and optimize beta limit 3) Investigate physics of good electron transport in these discharges 4) Characterize edge pedestal |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Improve fusion performance while maintaining high bootstrap fraction - utilize additional off-axis ECCD and off-axis NBI installed since 2008.
2) Investigate and optimize beta limit - vary current profile, beta time trajectory, etc, to increase beta limit. 3) Obtain edge improve edge pedestal characterization 4) Transport/turbulence study of electron transport characteristics | ||
| Background: | ITER steady-state scenario discharges were operated on DIII-D in 2008 using a scaled version of the ITER plasma-cross section, and also matching the ITER aspect ratio and approximate q_95 (Doyle NF 2010). In the four years since then, additional DIII-D capabilities have been added (off-axis NBI, additional gyrotrons), enabling us to now simultaneously meet ITER fusion gain and bootstrap fraction goals for the steady-state scenario. | ||
| Resource Requirements: | All 8 NBI sources, all gyrotrons | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 222: Multi-harmonic Dynamic Error Field Correction | ||
| Name: | Buttery | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | usual crew | ITPA Joint Experiment : | Yes |
| Description: | DIII-D has reached the limits of fixed harmonic mix error field correction. We have seen that such correction is imperfect, still leaving residual fields with significant effects. But DIII-D has the most flexible 3D coil system in the world, with many power supplies, sensors and a controller - it could do better. Also we have a range of dynamic approaches possible, such as magnetic response and rotation optimization, that don't always need dedicated disrupting shots to pursue.
Therefore I propose to use the 18 3D coils on DIII-D more independently to explore how better error field correction can be achieved. We should look to optimize various metrics, such as magnetic response or plasma rotation in H modes - noting that TBM studies tell us these may lead to different optimizations of correction. In such studies an error field correction parameter should be ramped to identify the optimal |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Establish H mode with standard error correction. Execute beta ramp to generate magnetic response to error field, and then vary correction to minimize response at multiple sensors. Iterate shot to shot to improve further. If possible perform under dynamic feedback (multiple independent arrays). Measure benefits of correction with further beta ramps or torque ramp-down
2. Establish stationery H mode, ramp 3D fields in various ways to improve rotation. 3. Ohmic plasmas (needs dedicated shots). Perform phase scan optimization of one array, after optimization of another array. Iterate. | ||
| Background: | Error fields are a critical issue for ITER, which will likely need to multi-harmonic correction to reach its goals. Error fields becoming an increasing challenge to DIII-D operation as low rotation and high beta are pursued. | ||
| Resource Requirements: | Piggy back time in discharges, some dedicated shots. Some preparation of control system. | ||
| Diagnostic Requirements: | Multiple array magnetic locked mode sensors. Rotation measurements. | ||
| Analysis Requirements: | Results should be self-evident. Need to consider PS needs carefully. | ||
| Other Requirements: | Someone to run it | ||
| Title | 223: Test current profile influence in NTM in ITER baseline | ||
| Name: | Buttery | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | Turco, Jackson, Solomon, La Haye | ITPA Joint Experiment : | Yes |
| Description: | Explore current profile influence on TM stability by utilizing transient techniques to strike up different profiles as beta is raised.
2011 long pulse ITER studies revealed a marginal stability to performance terminating 2/1 NTMs. But they could not resolve the governing parameters of this stability, although it has long been hypothesized that current profile plays a crucial role. However the long pulse discharges did not achieve much variation in current profile parameters such as li, and did not reach expected ITER values much. This study is crucial in resolving underlying stability influence and so control requirements for ITEr. It also adds insight to stability behavior for advanced steady state regimes. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Change startup and ramp up of discharges (heating timing & levels, gas, Ip ramp rates) to preform different q profiles. Then execute beta ramps or shot-shot beta variations to map out stability parameter space and understand how details in J profile might change stability. | ||
| Background: | Crucial to approach for ITER baseline stability | ||
| Resource Requirements: | Beams. Maybe ECH | ||
| Diagnostic Requirements: | Excellent MSE. MHD diagnostics. | ||
| Analysis Requirements: | Developing smoothed measurements as a proxy for deltaprime | ||
| Other Requirements: | |||
| Title | 224: Transport control via realtime NTM width modulation | ||
| Name: | Eidietis | Affiliation: | GA |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | Anders Welander | ITPA Joint Experiment : | No |
| Description: | We propose to control plasma beta by realtime modulation of the width of 3/2 or 2/1 NTMs using ECCD. The islands will be grown or suppressed in order to obtain and maintain a beta target by a reduction or improvement in confinement. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will begin with a well-established NTM control scenario in order to establish a nominal beta for the scenario. The target beta will then be lowered and raised in steps from that nominal value in order to exercise the full spectrum of NTM growth and reduction control. That sequence will be followed by a series of over-heated shots in which the beams are turned on at successively higher powers without feedback (a very crude approximation of a nuclear burn) in order to explore the limits of the beta control and to examine the differences (if any) between 3/2 and 2/1 modulation. | ||
| Background: | In present-day tokamaks, beta control is primarily accomplished by modulating the auxiliary heating sources, particularly NBI. However, in burning plasmas the effect of auxiliary heating will be merely a perturbation relative to the much greater nuclear heating. Moreover, the burn of an ignited plasma is liekly to be thermally unstable, requiring active feedback. Beta control will have to migrate from the present "turning the hose on and off" approach to more subtle methods. Some candidates include fuel mixture control and total density control. The present experiment proposes the alternative (or complementary) method of deliberately modifying the confinement properties of the plasma via NTM width control, effectively placing a well-controlled pressure "release valve" on the plasma.
The suppression of NTMs has been well established in DIII-D and other devices. This experiment expands NTM control from a binary control problem (suppression or no suppression) to the continuous problem of matching a target width. This control capability has been available for a while but has remained untested and un-utilized. | ||
| Resource Requirements: | All NBI & Gyros, PCS algorithm development, testing of NTM width control algorithm
TIME: 0.5 + 1.0 day (initial test of new algorithms) +(full experiment) | ||
| Diagnostic Requirements: | Magnetic (fast and slow),CER on 30L and 330L,MSE, Thomson,CO2 interferometers, ECE radiometer | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 225: Real time EC steering tearing control | ||
| Name: | Buttery | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | La Haye et al | ITPA Joint Experiment : | Yes |
| Description: | Tearing modes account for the principle performance limit and cause for disruptions in DIII-D and likely ITER. We have to develop more routine control of tearing modes, and understand how to do this on ITER. We have the world's best systems here at DIII-D to pursue this goal. And it is an urgent DIII-D program requirement, as discharges pushing to low rotation and high beta are increasingly limited by tearing modes. THIS IS THE YEAR TO COMMISSION DIII-Ds REAL TIME STEERING EC TEARING CONTROL. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Project to develop control, interface to PCS and then pursue in experiments. Explore targeting and tracking of resonant q surfaces with ECCD. Test various search algorithms with modes present. Compare with pre-emptive approaches. Test modulation benefits (esp for ITER). Port to other scenario programs as needed. | ||
| Background: | Tearing modes are the most fundamental stability issues for tokamaks, and are expected to get substantially more challenging in future lower rotation devices. This is essential to prep for ITER, and to enable DIII-D scenario campaigns for ITER, FNSF and power plant. | ||
| Resource Requirements: | Support by ECH group to implement real time steering. Provision/agreement to PCS control of EC | ||
| Diagnostic Requirements: | Good MSE q profiles and ECE for locating islands and EC. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 226: ELM-like behavior excited by off-axis-fishbones causing massive carbon influx | ||
| Name: | Okabayashi | Affiliation: | PPPL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | Go Matsunaga, J. R. Ferron, J. Hanson, W. W. Heidbrink, C. Holcomb, Y. In , G. L. Jackson, T. C. Luce, E. J. Strait | ITPA Joint Experiment : | No |
| Description: | The main objective of this experiment is to identify the relation of the off-axis-fishbone mode (OFM) time evolution and the massive carbon influx as was observed in SSI. The main diagnostics are the energetic particle(EP) / edge diagnostics and fast Active MHD spectroscopy. This study is intended to be useful to assess the possibility of minimizing the carbon influx. In addition, the relation of the EP to overall plasma performance is extremely relevant to the futureâ??s burning plasmas. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Since the process is so complex. The first step is to improve our understanding of each process. We will document the time evolution with various diagnostics of EP and profiles in a coordinated manner. This includes, fast sampling rotation profile, Er buildup, and MHD magnetic with fast sampling rate, and active MHD Spectroscopy up to 10 kHz. The second step is parametric sensitivity. The active approach includes the application of to n=3 non-axi-symmetric field to assess the non-axi-symmetric filed influence on the ELM-like development. | ||
| Background: | In the high betan exploration with q_min >1 exploration in SSI discharges, the off-axis-fishbone is often excited. Once the OFM is excited, it evolves into a non-linear-stage showing the â??density-snake-likeâ?? behavior in a very reproducible manner. Then, if the â??ELM-likeâ?? behavior was excited, this leads to the massive carbon influx, which was a major obstacle for the SSI scenario development. This event can be interpreted as the consequence of various fundamental processes: EP-mode formation, its non-linear evolution, EP transport toward the plasma edge, the edge profile evolution and finally the carbon influx due to EP losses.
Run time: two ½ days | ||
| Resource Requirements: | fast sampling of EP diagnostics | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 227: Separating RE formation and prompt loss | ||
| Name: | Hollmann | Affiliation: | UCSD |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux | ITPA Joint Experiment : | No |
| Description: | Try different methods of RE seed formation to try to separate out RE seed and RE prompt loss terms in rapid shutdown experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Consists of two parts: first it to use D = 3 mm Ar pellets injected from the planned ORNL pipe gun at different velocities and size to see if the impurity deposition velocity and amount affects the RE formation rate. The second is to use the Ar pellets to shut down shots with large startup RE populations. | ||
| Background: | Presently, it is not known how large RE currents will be in ITER or how frequently they will occur. The uncertainty is in both the RE seed term and the TQ prompt loss term. In present experiments, it has been impossible to separate these two because no diagnostic can presently measure the RE seed term in the plasma. One proposal is that only RE seeds formed near the magnetic axis survive the TQ MHD. Experiments are proposed to test this hypothesis. First, rapid shutdowns on ITER-like target plasmas are done with variable Ar pellet velocity to vary the depth of the RE seed formation to see if this affects the final RE current. Second, rapid shutdowns are done very early in the shot, where a fast electron seed still exists in the center of the plasma to see if the shot-shot CQ RE population is more reliable. Taken together, these experiments should indicate whether a central RE seed is essential for a CQ RE population. | ||
| Resource Requirements: | 1 run day. 4 gyrotrons | ||
| Diagnostic Requirements: | Fast camera, SPRED, SXR, interferometers, BGO scintillators, fplastic. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | new ORNL Ar pellet pipe gun with variable velocity and size pellets. | ||
| Title | 228: Experiments toward measuring anomalous RE loss term in RE plateau | ||
| Name: | Hollmann | Affiliation: | UCSD |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux, N. Eidietis, D. Humphreys, D. Rudakov | ITPA Joint Experiment : | No |
| Description: | Attempt to determine source of anomalous RE loss term during RE plateau by using various tricks and new diagnostics to try to measure each term. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | First, create large >100 kA RE plateau by using IWL low density target plasma shut down with small Ar pellet. Then, move RE beam with constant radius up along center post. With new CdTe array, expect to be able to see footprint of RE diffusion into center post, to quantify diffusive loss to wall. With new HXR spectrometer, measure energy distribution function to look for anomalous collisional dissipation. Inject various amounts of additional argon gas to intentionally increase collisional dissipation and look for changes in RE energy distribution. | ||
| Background: | Reliable, large > 100 kA RE beams are now created routinely in DIII-D using IWL low density targets shut down with small Ar pellets. The current in this RE plateau appears to dissipate much faster than expected from collisional drag due to electron-electron collisions. Possible alternate loss terms are shrinking of the current profile, drift orbit losses to the outer wall, diffusion losses to the inner wall, or anomalous collisional drag on the fast electrons. This experiment aims at investigating the last two possibilities. | ||
| Resource Requirements: | 1 run day. 4 gyrotrons | ||
| Diagnostic Requirements: | SXR, CdTe array, BGO scintillators, HXR spectrometer, fplastic | ||
| Analysis Requirements: | none | ||
| Other Requirements: | new ORNL Ar pellet pipe gun | ||
| Title | 229: Mode non-rigidity under DEFC and DF | ||
| Name: | Okabayashi | Affiliation: | PPPL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | Jeremy Hanson, Yongkyoon In, R. LaHaye, Lidia Piron, T.Strait | ITPA Joint Experiment : | No |
| Description: | Direct feedback operation sometimes implied that the difficulty of performance improvement is related to the mode non-rigidity. For instance, when the series of ELM-driven RWM were excited and the feedback was able to suppress these RWMs. However, the requested slow coil current component (corresponding to the DEFC component) before / after the event did not remain same magnitude, implying the 3D perturbed equilibrium was drifted before / after each event. Also, the combination of I- and C-coils seems working better, thus, the combined mode pattern could be preferable [NF, 2009]. Possible reason is that more than more than one eigen patterns are responsible for these process leading to the mode non rigidity as was discussed by IPEC. The main emphasis is to observe the sensitivity of m-component in the fine tuning of DEFC. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The SSI target includes the rapid change of plasma rotation and plasma betan for observing the RFA sensitivity.
DEFC process - First, we carry out active MHD spectroscopy independently with three sets of coils. Again the emphasis is to observe the m-component sensitivity - DEFC will be performed with series of sub time domain. Each sub domain is set with different gains, time constant and different choice of coil sets. The shift of coil set will allow the weighting shift of m component. The value of gain and time constant will be chosen based on the results of active MHD spectroscopy. - The central plasma rotation is to be monitored to observe the non-resonant effect. run time : For systematic study, we need one day minimum | ||
| Background: | Three sets of mode structures are prepared with upper-I-coil, lower-I-coil and c-coil as three separated operation units. The purpose is to provide m-component flexibility. The sensitivity of these three groups to plasma was calculated with MARS by Yueqiang Liu. For instance, the calculation was done with additional error field like F7A coil radial shift.
To pursue this, a few technical improvement of PCS is requested. In the iteration process, we want to specify the amplitude and phase in the PCS logic with the toroidal phase shift explicitly, in the preprogrammed waveform (new F-matrix to be added), rather than specifying three coil currents. Thus, we minimize the n=3 components, reducing the possibility of non-resonant influence at core. The second request is the minimization of AC sensor contamination. This can be achieved by applying simultaneously the AC compensation developed by RFX group. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 230: Assessment of simultaneous operation of DEFC and direct feedback (DF) | ||
| Name: | Okabayashi | Affiliation: | PPPL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Jeremy Hanson, Yongkyoon In, R. LaHaye, T.Strait | ITPA Joint Experiment : | No |
| Description: | We prefer to use one set of feedback sytem for dynamic error field correction (DEFC) and direct feedback (DF), since the application of double sets is more cumbersome.
However, the current-driven RWM feedback showed some practical difficulty of one coil system functioning for DEFC and direct feedback(DF), when the growth mode becomes faster. When the unstable mode grows, the direct feedback unwinds the slowly-established-error field correction (equivalent to the DEFC process). If the mode amplitude is not reduced significantly in short time period, the feedback completely unwinds the error field correction. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This observation showed the preference of the functional separation between DEFC and DF. The fast response with low gain is set for the DF.
This experiment is intended to document simultaneous operation of C-coil and I -coil DEFC and DF operation and compare the previous results. This approach can be tested to the RWM driven by the off-axis-fishbones in the SSI target. For the better understanding, we develop a model, which includes the mode allowed to drift in the toroidal direction relative to the error field. | ||
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| Title | 231: Rapid shutdown with large shell pellet | ||
| Name: | Hollmann | Affiliation: | UCSD |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | P. Parks | ITPA Joint Experiment : | No |
| Description: | Assess usefullness of large shell pellet concept for rapid shutdown | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Shut down stable, high-energy ITER-like shape (LSN) target plasmas with injection of a single large (D = 1 cm), boron powder-filled shell pellet. Increase target plasma thermal energy as necessary to ensure pellet shell breakup in plasma before thermal quench. | ||
| Background: | Rapid shutdown with collisional RE suppression is highly desired to avoid RE formation during disruptions in ITER. To date, only 20% of the required large density ncrit has been achieved (using multi-valve He MGI or using D2 SPI). Even if 100% ncrit could be achieved in ITER with MGI, the resulting gas reprocessing time of several days is deemed unacceptable. Shell pellets (thin sphere filled with powder) offer a possible alternative because the injected material (Be in the case of ITER) will coat the walls and not clog the pumping system. Proof-of-principle small (OD = 2 mm) shell pellet experiments were performed on DIII-D in 2008 and first large (OD = 1 cm) shell pellet experiments were performed in 2009. The large shell pellets did not burn through because the ablation rate of the 0.4 mm polystyrene shell was about 4x slower than expected. Now, the GA ICF group (N. Alexander) has made shell pellets with 0.1 mm polystyrene shells, which we expect to burn through. | ||
| Resource Requirements: | 1 run day. 4 gyrotrons, 6 beams | ||
| Diagnostic Requirements: | Fast camera, SPRED, SXR, interferometers. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | Large shell pellet launcher at 45V+2 | ||
| Title | 232: Low rotation ITER scenarios | ||
| Name: | Buttery | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | many | ITPA Joint Experiment : | Yes |
| Description: | The principle limit to low rotation scenario access comes from MHD. Therefore we should seek to integrate better MHD control into ITER baselines, as torque is ramped down. This should pick up on preparatory development work in stability area. Particularly, here we should look to test whether multi-harmonic error correction, and real time steering ECCD can reduce 2/1 modes. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop ITER like H modes with a torque ramp-down easier to start at high rotation. Deploy real time gyrotrons to maintain stability. For error fields execute variations in 3D field mix to optimize metrics like rotation (at fixed torque) and test benefits in terms of low rotation access in the torque ramp-down phase. | ||
| Background: | Crucial to developing ITER regime access | ||
| Resource Requirements: | integration of stability team and EC, RMP. | ||
| Diagnostic Requirements: | MSE, MHD diagnostics | ||
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| Title | 233: Experiments on CQ particle transport using SPI | ||
| Name: | Hollmann | Affiliation: | UCSD |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | N. Commaux | ITPA Joint Experiment : | No |
| Description: | Use the very large localized particle deposition achievable with the shattered pellet injector (SPI) to study CQ particle transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Shut down stable, high-energy ITER-like shape (LSN) target plasmas with injection of a single large shattered D2 pellet. Repeat for different target plasma thermal energies. | ||
| Background: | Rapid shutdown experiments have long attempted to reach sufficient mid-CQ particle density to collisionally suppress runaway electrons (REs). This possibility could be severely curtailed if there is significant loss of particles during the CQ. Previous SPI experiments seem to indicate that there is a large loss of particles during the CQ, although it is not clear where the particles are going. This experiment aims at understanding this apparent sink of particle with carefully diagnosed experiments. | ||
| Resource Requirements: | 1 run day. 4 gyrotrons, 6 beams. | ||
| Diagnostic Requirements: | Fast camera, SPRED, SXR, interferometers, CER spectrometers, ASDEX pressure gauges. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | Large shattered pellet injector with D2 pellets | ||
| Title | 234: Characterization of disruption main-chamber heat loads | ||
| Name: | Hollmann | Affiliation: | UCSD |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | R. Pitts, C. Lasnier, J. Watkins, N. Eidietis, D. Humphreys, C. Wong, D. Rudakov, R. Doerner | ITPA Joint Experiment : | Yes |
| Description: | Measure heat deposition time and footprint for disruptions. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Set up fast diagnostics, especially IR cameras and wall probes, for accurate heat load measurements. Create intentional downward hot VDEs and repeat to get shot-shot repeatibility. Between shots, move around main chamber IR camera view to eventually reconstruct fast data on entire main chamber heat loads (averaged over several shots). Compare hot VDEs with different initial thermal energy. Then compare these with heat loads during mitigated (neon MGI) fast shutdowns. Finally, attempt to reconstruct main chamber heat load data for beta limit, density limit, and current limit disruptions, again by repeating several times and moving the main-chamber IR camera view around. Additionally, as a piggyback, take DiMES data on sample response to large impulsive heat loads from VDE to lower divertor. | ||
| Background: | Understanding of disruptions heat loads is still largely empirical and cross-machine comparisons are few, except in the case of TQ and CQ time, where extensive comparisons have been done. It is desired to create a cross-machine database of main chamber heat loads and footprint from different types of disruptions and fast shutdowns to help make first wall material decisions in future devices. This has been made into a high-priority ITPA task for 2011 (DSOL-24). In 2011, a successful ½ day of data on VDE heat loads was obtained. The proposed run day for 2012 improves on this by adding main chamber heat load data for VDEs, as well as attempting to gather some main chamber heat load data for mitigated disruptions and other types of disruptions. Downward hot VDEs are also the best way to get large impulsive heat loads to DiMES to simulate plasma wall interactions during ELMs and disruptions in ITER. Several piggybacks are proposed, including the effect of disruptions on a small sample of tungsten fuzz and the effect of disruptions on thin sacrificial coatings such as silicon to radiatively protect underlying tile material. | ||
| 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. | ||
| Analysis Requirements: | none | ||
| Other Requirements: | none | ||
| Title | 235: RWM characteristics: comparison of SSI target and 2.5li | ||
| Name: | Okabayashi | Affiliation: | PPPL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Jeremy Hanson, Yongkyoon In, R. LaHaye, T.Strait | ITPA Joint Experiment : | No |
| Description: | The main objective of this study is to assess the universality of feedback performance and kinetic stabilization effects by comparing with these SSI and 2.5li configurations. This study also includes the exploration of the lower rotation configuration by utilizing the counter NBI
An operational path of SSI target has been developed, achieving the high betan to excite routinely the OFM-driven RWM. The preliminary analysis shows that the OFM-driven RWM in SSI target seems to have characteristics similar to these in 2.5li configuration, which is often used for RWM exploration. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will modify the SSI target toward lower rotation with superimposing the counter NBI. The low rotational plasma condition should provide the opportunity to assess the closeness of the onset condition of OFM-driven RWM relative to the natural RWM onset.
This experiment will also verify how the active MHD spectroscopy approach functions to assess the closeness to the natural RWM onset condition. We will survey how the RFA behaves near the marginal condition. Active MHD spectroscopy will be performed to untangle the relation between the off-axis-fishbone branch and external kink branch. | ||
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| Title | 236: Delaying TQ onset during disruption mitigation using ECH | ||
| Name: | Eidietis | Affiliation: | GA |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Use ECH heating of the q=2 surface during pellet injection to delay the onset of the thermal quench and enable increased impurity deposition in the core. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Standard disruption LSN target. Turn on ECH at q=2 surface shortly before injection of Argon killer pellet. Measure the time from the pellet launch to TQ with and without ECH. Vary applied ECH power and timing relative to pellet launch. | ||
| Background: | A basic problem for almost all massive impurity injection techniques is that the TQ often occurs before the impurity payload can be delivered to the plasma core (< q=2). This results in greater heat conduction to the divertor and inhibits collisional suppression of runaway electrons. Delaying the TQ onset time even 1-2 ms would greatly enhance the ability of the various mitigation methods to penetrate to the core. ECH on the q=2 has been shown to delay/avoid numerous types of disruptions on numerous tokamaks.
Obviously, ECH will not be effective if the density at the ECH deposition location spikes. However, the injected impurities have finite toroidal migration time. The ECH at 270 may have a few ms to work before cutoff is reached from impurity injection at the 15(MGI) or 135(pellet) degree ports. | ||
| Resource Requirements: | 6x ECH, Ar killer pellets, 30L (MSE), 0.5 DAY
TIME = 0.5 day | ||
| Diagnostic Requirements: | Magnetic (fast and slow),MSE, Thomson, SXR, scintillators, fast camera | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 237: Sustainment of near-zero rotation of tearing mode by phase-shifted feedback and locking avoidance | ||
| Name: | Okabayashi | Affiliation: | PPPL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | Jeremy Hanson, Yongkyoon In, R. LaHaye, T.Strait | ITPA Joint Experiment : | No |
| Description: | The avoidance of mode locking of NTM or tearing mode is one of important issue for disruption avoidance.
The usage of feedback has two advantages compared with simple application of the rotation field at fixed frequency. With feedback process, the mode and the applied non-axi-symmetric field can be synchronized at any frequency. More importantly, when the uncorrected error field starts to increase the mode amplitude ( locked-mode growth) at nearly zero frequency range, the feedback process, by its nature, starts to work as the dynamic error field correction and to reduce the uncorrected error. Thus, the mode rotation is easier to be sustained even near zero frequency. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | - NTMs onset and mode locking is anticipated to occur at various circumstances from low plasma density to beta-collapse in the ITER operational scenarios. The avoidance of tearing mode locking is a critical issue in every step of discharge developments.
- Here, it is proposed to demonstrate a unique usage of internal coils for avoiding mode locking, if successful, this would be important as an application of ITER internal coils. - As we reported previously, the application of feedback can robustly synchronized tearing mode with a rotating external field and the rotation at 40 Hz and 10 Hz was achieved and the mode rotation was well sustained. | ||
| Background: | - Here, it is intended to achieve the sustainment at the rotation futher lower thna 3 Hz to demonstrate the usefulness of this approach for the ITER internal coil application.
- We also plan to document how the feedback will reduce the uncorrected error field automatically near-zero rotation, which should make the slow rotation possible | ||
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| Title | 238: TBM Error Field correction by DEFC | ||
| Name: | Okabayashi | Affiliation: | PPPL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Jeremy Hanson, Yongkyoon In, R. LaHaye, T.Strait | ITPA Joint Experiment : | No |
| Description: | The DEFC is the attractive approach for minimizing the error field since this can be efficient. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The localized error field by TBM includes variety of toroidal and poloidal harmonics. Thus, the process of DEFC needs to prepare various coil sets. This will be done by operating separately Upper / Lower I-coils and C-coils
It is needed to prepare the sets of coil combinations of Upper / Lower separated I-coil and C-coil for producing the n=1, 2 and 3. As the first step, for n=1 component compensation, we will document independently the DEFC with three sets of upper/lower and C-coils. This will provide the poloidal coil set preference without presetting blindly 240 degree connection. Then, upper and lower will be combined with weighting factor determined by the individual operation for DEFC. For the n=3 component, it is to be tested with C-coil. The compassion of performance with the n=1 and n=3 will provide the estimate of the n=2 compensation. Target is to be sufficiently high betan with lower rotation to observe clear RFA. | ||
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| Title | 239: Measurements of net vs gross erosion of high-Z using DiMES | ||
| Name: | Stangeby | Affiliation: | U of Toronto |
| Research Area: | Plasma-material Interface | Presentation time: | Requested |
| Co-Author(s): | P.C. Stangeby, D.L. Rudakov, N.H. Brooks, W.R. Wampler, J.N. Brooks,
D.A. Buchenauer, J.D. Elder, M.E. Fenstermacher, C.J. Lasnier, A.W. Leonard, A.G. McLean, R.A. Moyer, A. Okamoto, J.G. Watkins, C.P.C. Wong |
ITPA Joint Experiment : | No |
| Description: | The net erosion rate at the divertor strike points of future high power devices such as ITER will probably not be acceptable unless net erosion is reduced substantially relative to gross erosion. Fortunately, just such a reduction is theoretically expected to occur due to prompt local deposition of sputtered particles: once it is ionized, the gyroscopic motion should return the ion promptly to the surface; additionally, the strong E-field in the magnetic pre-sheath should also force the ion quickly back to the surface. These processes are expected to be particularly strong for high-Z, e.g. W and Mo. Experimentally, however, the evidence for prompt local deposition is inconsistent, see Background.
In 2011 the DIII-D DiMES system with a Mo sample was used to undertake a more definitive test of prompt deposition than has been possible previously. It was partly successful in that uniquely detailed measurements were made of net erosion and highly detailed interpretive modeling was carried out. Measurements of the gross erosion rate, however, had large error bars. See Background. The basic objective will be to repeat the 2011 experiment as exactly as possible with improved diagnosis of the gross erosion rate. Further objectives: (a) developing and using experimental conditions to increase the gross/net erosion rate of Mo, (b) extending the study to W. A definitive test requires that both net and gross erosion rates be measured directly in the same experiment. Gross erosion rates can be measured from the intensity of neutral WI (400.8, 429.4, 498.2, and 505.3 nm) and MoI lines (379.8, 386.4 and 390.3; 550.6, 553.3, 557.0 nm). This requires reliable knowledge of inverse photon efficiencies, S/XB(Te), in order to convert spectroscopic intensities into atom influx densities. Recently, S/XB values have been directly measured on PISCES for WI [Nishijima, Phys Plasmas 16 (2009) 122503] and MoI [Nishijima, J Phys Mol: At Mol Phys, 43 (2010) 22570], increasing the reliability of neutral influx measurements for these high-Z elements. The new PISCES SX/B values for WI differ by an order of magnitude from previous measurements, made 1997-2002, which is attributed in part to a ne-dependence of S/XB that the PISCES experiments uncovered. This situation, however, makes it more difficult to do a definitive high-Z test using W. Fortunately, the new PISCES SX/B values for the resonance uv MoI lines at ~ 380nm agree with those calculated from first principles, ADAS, although the more convenient visible MoI lines at ~ 550 nm do not. This discrepancy may soon be rectified with new ADAS calculations, now underway. In the meanwhile the 380 nm MoI lines provide the most solid basis for a definitive high-Z test. Net erosion rates are measured by using very thin samples, 10-100 nm, whose thickness is measured ex situ before and after exposure. The sample is prepared with a depth marker and thickness is measured using ion beam analysis [Wampler, JNM 233-237 (1995) 791]. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. We will repeat as exactly as possible the 2011 Mo DiMES experimental conditions where we used low density SAPP (Simple as Possible Plasma) attached divertor conditions, see Background.
2. We will again use repeat shots of a single plasma condition. Since the C-mod measurements were campaign-integrated, it is hard to rule out that the net erosion could have occurred in off-normal conditions. 3. We apply Brooks' advanced computational modeling code analysis for this experimental condition, where the â??plasma backgroundâ?? needed as input for the Brooksâ?? codes has been generated using the Toronto OEDGE code. Brooksâ?? analysis found that gross/net erosion should be about 2 for these conditions, which is much smaller than indicated by the measurements made in the 2011 experiment. We presently attribute this to uncertainties in the spectroscopic measurements of the gross rate, see Background. 4. In order to reduce the uncertainty in the measurement of gross erosion we will: (i) Use a 1 nm pass-band filter at 380 nm with the CCD camera, rather than the 10 nm filter used in 2011, which passed a number of non-MoI lines, (ii) Measure the gross erosion rate by a non-spectroscopic method by incorporating a 2nd, very small, ~ 1 mm diameter, Mo sample in the DiMES head, just upstream of the primary, 1 cm Mo sample; for such a small sample, the net erosion rate â?? which can be measured very reliably by ex situ analysis (RBS) - should be close to the gross rate. (iii) Measure the gross rate using the high resolution MDS spectrometer, which is already absolutely calibrated for the MoI lines at 550 nm; the new ADAS analysis, in progress, for MoI is expected to resolve the discrepancy relative to PISCES measurements for the S/XB values for these convenient lines. (iv) A uv lab sphere will be acquired to absolutely calibrate MDS for the 380 nm lines, for which there are no known discrepancies for the S/XB values. In further experiments we will again use Mo but make a number of changes aimed at increasing the gross/net erosion rate: (i) Increase the Mo principal sample size to 25 mm from 10mm. (ii) Increase ne at the OSP, while keeping Te at ~ 30 eV, by using more beam power while operating with reversed field to avoid H-mode (ELMs). (iii) Increase the magnetic pitch angle so that the Mo sample is subtended by a wider flux tube. (iv) Decrease B for larger gyro-radius, making the magnetic pre-sheath thicker, thus its E-field more effective, and making the Mo+ gyro-radius larger. In further experiments the foregoing will be repeated using W, which is directly ITER-relevant but where the S/XB values are less certain. If we are able to fully understand the gross and net erosion of Mo, we will be in a position to infer high quality values of S/XB for the WI lines from our DiMES experiments, for actual divertor plasma conditions, which will be of immediate use for experiments in W-PFC tokamaks such as JET and AUG and subsequently in ITER. | ||
| Background: | In 2011 we performed the first experiment in a tokamak where post-exposure net erosion of Mo was measured for a sample exposed to well-controlled plasma conditions, and in-situ
gross erosion was estimated as well. Net erosion of high-Z plasma-facing surfaces in a tokamak is expected to be reduced by local redeposition due to sputtered atom collisions with the impinging plasma. Reduction of net compared to gross erosion has been demonstrated for W in AUG [Krieger JNM 266-269 (1999) 207]. However, in C-mod the measured campaign-integrated peak net erosion of Mo divertor tiles was found to be ~10X higher than that computed using the REDEP/WBC code, while the gross erosion predicted by the code was a reasonable match to MoI influx data [Brooks JNM 415 (2011) S112]. The 2011 experiment was aimed at resolving this discrepancy by measuring both net and gross erosion of Mo under stable well-diagnosed plasma conditions allowing more accurate comparison with the modeling. A 1 cm diameter Si disk coated with 24 nm of Mo was mounted in a graphite DiMES holder and exposed in a series of 7 reproducible lower single null L-mode discharges. The exposure was performed near the attached OSP for a total flattop time of ~28 s. The plasma density and temperature near the strike point were measured by the divertor Langmuir probes, ne ~ 1.5e19 /m3, Te ~ 30 eV. The gross erosion rate of Mo was measured spectroscopically, using an absolutely calibrated CCD camera with MoI filter at 390 nm, bandwidth 10 nm. An upper bound estimate of an average gross erosion rate of 3.75 nm/s was inferred from the data. The 10 nm filter passed a number of lines other than MoI; the relative contribution of MoI light was estimated using high-resolution divertor spectrometer data. Net erosion of Mo was measured by comparing the Mo layer thickness measured by RBS before and after the exposure. The reduction of Mo thickness was 11±1 nm on the average, corresponding to an average net erosion rate of 0.40±0.04 nm/s, i.e. significantly smaller than the gross erosion rate measured by the camera. The Mo erosion from the DiMES sample is being modeled using the REDEP/WBC code package. Plasma conditions at the DiMES sample are calculated by the OEDGE code using the process of empirical plasma reconstruction. Langmuir probe data across the target are used as input to OEDGE. Divertor Thomson data and spectroscopic measurements of hydrogenic emissions from DiMES are used to further constrain the plasma solution. This plasma solution is then used as input to REDEP/WBC which calculates the erosion, transport, ionization and eventual deposition of the eroded Mo from/to the DiMES sample. | ||
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| Title | 240: Low-impurity RE production | ||
| Name: | Eidietis | Affiliation: | GA |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Produce fairly "clean" runaway electron beams using small argon gas puff during rampup phase of inner-wall-limited plasma | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1. Produce inner-wall-limited, low-elongation ECH -heated target plasma.
2. Fire argon puff during rampup & observe if RE plateau or HXR signals form 3. Vary timing and quantity of argon puff. | ||
| Background: | Historically, D3D can only reliably produce runaway electron beams using limited, low-elongation targets and 2mm argon "killer" pellets.
As of now, the large argon pellet injector is no longer available. Thus, it would be helpful to have a alternate method of runaway production for RE studies. Tore Supra can reliably produce runaway plateaus in a circular limited target plasma using only a small gas puff during the rampup phase. This technique may yield similar results on D3D, allowing us to continue RE studies without the pellet hardware being absolutely necessary. | ||
| Resource Requirements: | ECH, medusa valve
TIME = 0.5 day | ||
| Diagnostic Requirements: | BGO, neutron counters, SXR | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 241: Investigation of n=1 RMP effects on ELMs with pitch-aligned configuration of I-coils | ||
| Name: | Park | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | T. E. Evans, R. M. Nazikian, Y. M. Jeon | ITPA Joint Experiment : | No |
| Description: | KSTAR has shown that n=1 RMP can be successful for ELM mitigation and suppression when n=1 is optimized for pitch-alignment. The goal of this experiment is to test if n=1 RMP can also alter ELMs in DIII-D, when n=1 configuration of I-coils is adjusted for better pitch-alignment from the nominal configuration, which is optimized for error field correction by coupling to the Kink mode. For this goal, 120-180 poloidal phasing in n=1 I-coil configuration will be used rather than the typical 240-300 poloidal phasing, and will be applied to ELMying plasmas with q95=3.5-6. The q95 variation is particularly important since resonant surfaces are fewer in n=1 and so the location of each surface or the number of resonant surfaces in the pedestal can 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 : | No |
| Experimental Approach/Plan: | Assess the optimal poloidal phasing (expected as either 120 or 180) and currents of n=1 I-coils depending on target plasmas to meet pitch-alignment condition and Chirikov overlap > 15%, but to avoid locking. Consider C-coil for n=1 error field correction instead. Start with the standard target plasma for n=3 RMP ELM suppression, with low collisionality and q95=3.5, and apply n=1 I-coil fields. If successful, investigate the threshold and the resonant window nearby each q95. Repeat experiments with higher q95=4-6 in 2-3 steps. Acquire high resolution profiles with and without n=1 RMP for analysis. | ||
| Background: | In KSTAR, n=1 RMPs have been successfully used for ELM mitigation and suppression with an optimized configuration, 90 phasing, for pitch-alignment. However, another configuration optimized to the Kink mode, 180 phasing, locked or caused H-L back transition as typically expected by n=1 fields. Results perhaps 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 made 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, similarly to 180 phasing in KSTAR. Therefore, it is important 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 q95. KSTAR had higher q95 when ELMs were suppressed, and the high q95 may also be important in n=1 applications in order to have relevant 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 | 242: Modulated ECCD for 2/1 NTM suppression | ||
| Name: | Welander | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Rob LaHaye, Ron Prater, Ted Strait, John Lohr, Ben Penaflor | ITPA Joint Experiment : | No |
| Description: | The purpose is to evaluate benefit of modulated ECCD for suppression of the 2/1 NTM and to scan how suppression depends on the phase between the ECCD and the island O-point. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The shots will use a combination of co- and counter beams to produce a 2/1 NTM with a frequency of about 5 kHz. A plan for how to trigger NTMs should be tested before this experiment.
Modulation of gyrotrons will be used. This requires: (1) at least two gyrotrons on the same launcher, (2) aim at q = 2/1, (3) launch angles for δeccd about three times as wide as usual, (4) spread launcher angles so they straddle q = 2/1 by dÏ? = 0.6δeccd for further total width, where δeccd is FWHM. The goal is to suppress a 2/1 NTM using modulated ECCD with the best guess of phase for the modulation. In the next step suppression will be attempted with cw-ECCD for comparison. After that modulation in the X point of the island will be tested for comparison with modulation in the O-point. After that a shot with a continuous sweep of the phase between O and X will be done. | ||
| Background: | Continuous wave ECCD has proven effective in completely suppressing NTMs. It can also prevent NTMs, when preemptively applied in the correct radial location. The cw-ECCD requires good alignment and a narrow deposition region with respect to the island width. In ITER the ECCD will have a relatively wide deposition region which will make cw-ECCD less effective since the destabilizing effect from current driven in the X-point will nearly cancel the stabilizing effect from the current driven in the O-point. This problem can be solved by switching the gyrotrons on when the O-point passes by their respective line-of-sight and off when the X-point passes by. Predictions made by F.W. Perkins suggest that a modulation scheme using a square pulse train with a 50% duty cycle will give close to maximum feasible suppression rate. Previous work on ASDEX-UG has demonstrated the efficacy of modulated ECCD. The present experiment seeks to study the suppression of the 2/1 NTM with modulated ECCD, and demonstrate the use of modulated ECCD with real-time frequency/phase detection along with sustained synchronization with the mode. The system to be demonstrated constitutes a general solution for mode frequency/phase detection that can be easily extended to simultaneous suppression of multiple modes using launcher steering in future DIII-D upgrades. | ||
| Resource Requirements: | Plasma current: 1.06 MA (standard direction) Toroidal field: -1.61 to -1.69 T (standard direction) Shape: Lower single null, as in 129330
Cryo-pumps: Lower pump (upper pumps desirable), He cooled Error Field Coils: I240-Coil for n = 1 error field correction. Neutral Beams All sources required including 30L and 330L for diagnostics (MSE, CER) and both 210 sources for counter injection and to control rotation. I coils in I240 configuration, powered by the SPAs. ECRH Two gyrotrons on the same launcher with modulation capability are the minimum required. Injected power > 1 MW for > 2s is needed. Four on two launchers (two on each desired). | ||
| Diagnostic Requirements: | Magnetic (fast and slow) CER on 30L and 330L (for V4,5,6) alternating with 10 ms off/10 ms on from 0.9 to 5.0 s except for continuous 0.274 ms resolution for 160 ms around 4.5 s for a total of 994 pulses. MSE Thomson CO2 interferometers ECE radiometer
oblique ECE SXR diodes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 243: Control/Operations Thursday Evening Development Time | ||
| Name: | Humphreys | Affiliation: | GA |
| Research Area: | Plasma Control and General Issues | Presentation time: | Requested |
| Co-Author(s): | Mike Walker, Al Hyatt, Nick Eidietis | ITPA Joint Experiment : | No |
| Description: | Establish regular 2 hour experimental slots on Thursday 5p-7p for initial testing of experimental use of new control/operational capabilities. Intended uses for 2012 include development and testing of: steerable mirror capability, powered VFI for experimental use, Super X divertor control, burn control algorithm, simserver validation, as well as various control tests needed for upcoming experiments during the campaign. Flexibility of Thursday test periods enables rapid response to presently-unforeseen urgent needs arising during the campaign (e.g. control debugging done in 2011). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use same decision process used in 2011 (specific proposals vetted through Control group, now to be approved by Buttery and DIII-D Director). We recommend that priorities be based on demonstration of readiness to proceed, need for capability in upcoming physics experiments, relevance to goals of DIII-D program, and available manpower and budget constraints (an option should be available to not conduct a particular 2 hour experimental session if constraints are too severe). | ||
| Background: | Experiments that use new or modified capabilities often must take on substantial risk associated with the lack of experimental testing of that capability. In the past, this has sometimes resulted in significant portions of an experimental day or even the whole experimental day being consumed by difficulties associated with a lack of readiness of the new capability. In addition, the time it takes for new systems to come on-line can be increased significantly when scheduled experiments are not willing to take on that risk. In recent years, the use of regularly scheduled 2 hour experimental sessions for operational development has significantly accelerated the development of new capability and reduced the risk associated with experimental use of new capabilities. | ||
| Resource Requirements: | Machine Time: 2 hours experimental time each week.
Other requirements will vary with the tests being performed. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 244: PID (proportional integral derivative)control of Error Field | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Stability and 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. We usually avoid PID control due to the possible numerical noise in obtaining the derivative and feeling that integral term is not of the greatest importance. There are many control methods to overcome the issues related to derivative term. As for the integral term being not that important, it is the combination of P, I and D terms that gives the optimal control effort. <br> <br>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. | ||
| Resource Requirements: | 1/2 - 1 day. Depending on type of control development: static/dynamic/real-time automatic self-tuning. Relay-feedback auto-tune algorithm has to be added to PCS. This has been done at NSTX PCS, addition to DIII-D PCS so should be a relatively easy task. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 245: Symmetric DN discharges with ELM suppression for 3D equilibrium reconstruction | ||
| Name: | Sontag | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | T. Evans, L. Lao, E. Lazarus, E. Strait | ITPA Joint Experiment : | No |
| Description: | The V3FIT code will be used to reconstruct the non-axisymmetric equilibrium for stellarator symmetric DN discharges with ELM suppression or mitigation via RMP and compared to references discharges without RMP fields. V3FIT relies on the VMEC 3D equilibrium solver, which converges to a solution much more quickly for up-down symmetric tokamak discharges. 3D reconstruction of a symmetric DN discharge is much more feasible than for a single null discharge. Success of this analysis would provide valuable insight into the 3D effects of the RMP and how it leads to ELM suppression and provides a good starting point for further analysis with 3D transport tools and equilibrium solvers that allow the existence of magnetic islands and stochastic regions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will piggyback on and provide analysis support for other experiments proposed to suppress ELMs with RMPs in symmetric DN discharges e.g., ROF proposal #35 and #49. | ||
| Background: | The application of RMPs to otherwise ELMy discharges has been shown to modify the pedestal profiles, resulting in ELM suppression. The exact nature of how the applied fields penetrate and alter the magnetic field structure and affect transport is unknown. 3D equilibrium reconstruction with V3FIT provides a good starting point for further analysis of the magnetic field structure with non-ideal 3D equilibrium codes such as SIESTA and HINT2, as well as 3D transport analysis with stellarator transport codes. | ||
| Resource Requirements: | Symmetric DN discharges with RMP ELM suppression or mitigation. | ||
| Diagnostic Requirements: | Magnetics, Thomson scattering, SXR diagnostics, CER, MSE | ||
| Analysis Requirements: | Kinetic EFIT, V3FIT | ||
| Other Requirements: | |||
| Title | 246: Enhancement of Runaway Electron Production Reliability | ||
| Name: | Humphreys | Affiliation: | GA |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | V. Izzo, N. Commaux, N. Eidietis, E. Hollmann, P. Parks, J. Wesley | ITPA Joint Experiment : | No |
| Description: | The goals of this 1-day experiment are to study the role played by the initial equilibrium and resulting MHD activity during the thermal quench in producing large post-TQ runaway beams, and to thereby identify scenarios for producing either large beams for study, or suppressed beams. This should result in higher reliability of RE production in ITER-like targets to enhance study of ITER disruption mitigation scenarios. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using ITER-like LSN target plasmas (and possibly also limited low kappa plasmas) the nature of unstable MHD eigenmodes (and associated seed deconfinement) will be varied by applying co- or ctr-ECCD, OANBI, varying edge q95, beta, and/or rotation. Ar pellets will be injected to trigger disruption and observe the impact on conversion of thermal current to RE. Specific choices may be guided by GATO and/or NIMROD studies predicting varying effectiveness of RE seed (de)confinement, making this experiment an important illustration of scenario development through theoretical prediction. | ||
| Background: | Recent analyses with NIMROD (Izzo) have suggested that the nature of the MHD instabilities triggered during a disruption may determine the size of the RE seed population that survives the thermal quench. Modes with large amplitude penetrating to the core (consistent with deconfinement of seeds even at the core) are correlated with low post-thermal quench RE current. Modes with zero amplitude at the core are correlated with high post-thermal quench RE current. Thus, the amplitude of MHD activity and the spatial form of the unstable eigenmodes may provide effective knobs to enhance the effectiveness of RE current mitigation. | ||
| Resource Requirements: | Cryogenic Ar killer pellet injection to generate RE, OANB, ECH/ECCD | ||
| Diagnostic Requirements: | Usual disruption diagnostics; 5 kHz magnetics, MSE, Thomson, new IR camera strongly desirable, fast cameras, UCSD BGO for RE detection, FPLASTIC, | ||
| Analysis Requirements: | NIMROD, GATO, JFIT, TokSys disruption/RE simulation codes; KPRAD, etcâ?¦ | ||
| Other Requirements: | |||
| Title | 247: Determination of MSE Requirements to Resolve Delta-Prime | ||
| Name: | Humphreys | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | F. Turco, R. Buttery, E. Kolemen, C. Holcomb, R. La Haye, A. Welander, N. Eidietis | ITPA Joint Experiment : | No |
| Description: | The goals of this 1-day experiment is to study the ability of present MSE to measure delta-prime or other profile parameters that correlate with TM stability, and thereby to determine requirements on an upgraded MSE system to enable such measurements. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using a standard 3/2 TM-unstable target, vary beta and/or rotation to vary the degree of (meta)stability, confirming by onset of mode and growth of island. Scan plasma +/- 2 cm in major radial direction under different stability conditions to produce high resolution MSE measurements as. Perform scans with several different fixed beta values, and with slowly ramping beta. Apply ECCD to suppress island and perform scans with actively suppressed islands. | ||
| Background: | In order to use active control to produce and maintain plasma targets that are robustly stable to tearing modes yet operate at high performance, it is essential to be able to determine the the degree of TM stability or proximity to the (meta)stability boundary in real time. From previous studies of plasmas near resistive instability (Turco) it appears that the present MSE system cannot reliably resolve delta-prime or other profile characteristics that are predictive for stability. This experiment seeks to confirm this result and quantify the capability of the present MSE system to produce these measurements in real time, and thereby to quantify requirements on a potentially upgraded MSE system. | ||
| Resource Requirements: | 4-6 beams (co/ctr), 4-6 gyrotrons ECCD, possible modification to NTM algorithm to enable Rp-scanning with active tracking (steerable mirrors strongly desirable?) | ||
| Diagnostic Requirements: | MSE, 2-5 kHz magnetics sampling, Thomson | ||
| Analysis Requirements: | Kinetic EFITs, GATO, NIMROD, Corsica (or equivalent nonlinear model/simulation) | ||
| Other Requirements: | -- | ||
| Title | 248: Snowflake Divertor: Implementation, Study of its Effect | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | General B&PP | Presentation time: | Requested |
| Co-Author(s): | Vlad Soukhanovskii | ITPA Joint Experiment : | No |
| Description: | Reliable heat handling will be necessary for fusion energy to be an alternative source of energy. ITER already has a divertor design but FNSF and DEMO still need tested and proven divertor configurations.
We propose to develop PCS control of the snowflake divertor configuration, demonstrate steady-state snowflake configurations, and study the impact of the snowflake divertor configurations on core confinement, pedestal MHD stability, divertor heat flux (steady-state and transient), impurity production, and compatibility of the snowflake divertor configurations with cryopumping (and density control) and extrinsic impurity seeding for enhanced divertor radiation. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | We have developed Snowflake identification software for NSTX and LLNL has developed one for DIII-D. There has been a control development based on the identification software at NSTX. It was to be tested at the NSTX 2011 run year. We can use the similar methods at DIII-D for Snowflake scenario development and control. | ||
| Resource Requirements: | We need to develop new control algorithms and scenarios for the Snowflake Divertor. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 249: Field scan to validate the 288 GHz R0 polarimeter as internal magnetic field diagnostic | ||
| Name: | Zhang | Affiliation: | UC, Los Angeles |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | Tony Peebles, Max Austin, Troy Carter, Neal Crocker, Edward Doyle, Walter Guttenfelder, Chris Holcomb, George McKee, Terry Rhodes, Guiding Wang, Mike Van Zeeland, Lei Zeng | ITPA Joint Experiment : | No |
| Description: | The purpose of this experiment is to validate the capability of the 288 GHz polarimeter, which was installed on DIII-D at 60 degree R0 in October, 2011, for measuring the internal magnetic field equilibrium and fluctuations. Previous calculations for low BT (~ 0.5 T) and low density (ne ï?£ï? 4E19 m-3) plasmas indicate that in such cases the polarimeter is primarily sensitive to Faraday rotation effect (FR, polarization rotation caused by magnetic field parallel to propagation), while the Cotton-Mouton effect (CM, elliptization due to magnetic field perpendicular to propagation) is generally weak and does not strongly interact with FR. However, in higher BT plasmas, CM becomes more important and the interpretation for the measurements is more challenging as well. A synthetic diagnostic code developed for the polarimeter has been preliminarily verified by showing reasonable agreement in the equilibrium measurements for only a few previous full BT discharges, where the refraction of polarimeter probing beam is less severe.
We propose to validate the diagnostic with following tests: 1. A stepped BT scan. As explained above, this will vary the ratio of contributions from FR and CM. 2. A vertical scan to test the spatial dependence of equilibrium profile measurements. 3. The magnetic fluctuation diagnostic capability can be validated by comparing the experimental measurements with the theoretical predictions for coherent modes across a range of BT. There has been a significant success in the study of Alfven Eigenmodes (AEs) comparing experiment and theory for electron density and temperature fluctuations, making them good candidates for this task. This comparison, if successful, can provide validation for both the AE theory and polarimeter capability. Neoclassical Tearing Modes (NTMs) may also be good candidates because of their large amplitude. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | To accomplish (1) and (2), low electron density (ne ï?£ 4E19 m-3), oval/circular shape, quiet L-mode discharges are required. We are planning a 3-step BT scan, 0.5T, 1.2T, 2.0T, for documentation. After reaching the flat-top, the plasma position will be vertically scanned over about 30 cm (or larger, until polarimeter probing beam is totally refracted) to vary the height of polarimeter probing chord relative to the plasma center.
To accomplish (3), low electron density (ne ï?£ 6E19 m-3), upper single null discharges are preferred. We are planning a stepped BT scan, 0.5T, 2.0T, 1.2T, 1.6T, 0.75T, mostly L-mode, and one H-mode discharge for each BT value. After reaching the flat-top, the plasma position will be vertically scanned over 10 cm. For (1) and (2), the synthetic diagnostic code takes the equilibrium magnetic profile from EFITs and density profile from Thomson scattering (normalized by CO2 interferometers)/profile reflectometer, respectively, and calculates the expected polarimeter outputs, which can be directly compared with the experimental measurements. If necessary, kinetic-EFITs will be run to improve the precision of reconstructed magnetic profile. For (3), NOVA, in combination with the electron density and temperature fluctuation measurements, can predict the magnetic fluctuation profile associated with AEs. With the density and magnetic fluctuation profile inputs, the synthetic diagnostic code can calculate 4 different expected polarimeter fluctuation outputs, with either fluctuation separately switched on/off. The different outputs can be compared with the actual measurements to assess the polarimeter capability for measuring magnetic fluctuations. The feature comparison among the spectra of interferometers, magnetic probes and polarimetry phase can also be a useful indicator for the validation. Similar approach can also be applied to using NTMs for validation. The polarimetry measurements can be compared with expectation from a simple heuristic and/or NIMROD modeling. Reference low-BT shot: #140986 (BT = 0.56 T, Ip = 0.61 MA, ne ï?£ 4E19 m-3, SNB) in â??Joint NSTX/DIII-D poloidal rotation experimentâ?? by K. H. Burrell on 01/11/2010 | ||
| Background: | The interaction between Faraday rotation (ï?µ ï?²nB||dl) and Cotton-Mouton effects (ï?µ ï?²nBï??2dl) can significantly complicate the interpretation of polarimetry measurements. CM is dominated by BT, for the diagnostic geometry used on DIII-D, employing retroreflection along a horizontal major radius. Previous polarimetry calculations also indicate the density fluctuations begin to contribute significantly in higher Bï?? cases.
NOVA modeling agrees very well with the measured electron density and temperature fluctuations associated with AEs in DIII-D tokamak. [M. A. Van Zeeland etc, Phys. Rev. Lett. 97, 135001 (2006)] This experiment, if successful, will become a major data set for my PhD thesis. | ||
| Resource Requirements: | NBI to trigger AEs, and 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 (viewing at the mode region) ECE (ï?³ï? ï?±ï?®ï?¶ï? ï??) DBS CECE fast magnetic coils | ||
| Analysis Requirements: | EFITs, NOVA, possibly M3D, NIMROD (for NTMs theory) | ||
| Other Requirements: | |||
| Title | 250: Real-time NTM Stabilization via Profile Control | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys | ITPA Joint Experiment : | No |
| Description: | In order to operate DIII-D and future machines such as ITER close to but below stability limits, we need to control the plasma profiles (especially q profile). I would like to note that this is not recovery from a disruption but control during the "normal" phase of the shot. The main focus of this control would be to identify NTMs (and/or NTM stability proxies) and control the profile from unstable configuration to a more stable one. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Identify how NTM stability proxies change with profile shape. We will focus on ECCD/ECH and off-axis beam as actuators. We will sweep the plasma past MSE cords at various profile configurations, with ECCD, ECH, off-axis beam power inputs. Identify the variation of Deltaprime or other proxies under these circumstances. An input-output model will be developed. From the model control algorithms for stabilization will be obtained and tested (possibly using Thurday evening periods). | ||
| Background: | -- | ||
| Resource Requirements: | 1 day. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 251: Oblique ECE for radial alignment during NTM suppression | ||
| Name: | Welander | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Dinh Truong, Francesco Volpe, Max Austin | ITPA Joint Experiment : | No |
| Description: | Implement PCS code for use of oblique ECE for swift correct radial alignment of ECCD during NTM suppression and for phase adjustment during modulation of ECCD. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | First step is to write the PCS software and test the real-time analysis without affecting the plasma in a calculation-only mode.
Second step is to radially align the plasma in a piggy back experiment with a rotating NTM, no ECH is needed. Third step is to use the new methods during an NTM suppression experiment. | ||
| Background: | The oblique ECE is a diagnostic that looks at radiation coming out from the plasma from the same direction that the ECH beam is injected and at frequencies both above and below that of the ECH. This allows detection of the temperature fluctuations both inside and outside the surface where ECH is absorbed.
If a rotating NTM is present in the plasma the fluctuations on the oblique ECE channels will be out of phase when the ECH is centered on the NTM. If both channels are viewing one side of the island then a comparison with Mirnow signals can reveal if they are viewing inside or outside the island. With this information the alignment can quickly be corrected. An efficient real-time analysis of the phase between the fluctuations together with an appropriate algorithm holds the promise of making radial alignment of NTMs to ECCD an automatic standard procedure. The first goal of the experiment is to develop and implement this technique. The second goal is to implement a PCS code that will use a combination of oblique ECE and Mirnov data to determine the correct phase and frequency for the modulation of ECCD. The advantage of the oblique ECE for phase detection is that the phase mapping from the diagnostic to the deposition point is simply a slight difference in toroidal angle. The advantages of including Mirnov data are that the correct phase can still be found when all ECE channels are on one side of the island. In previous years the oblique ECE had a large noise but during this year upgrades are planned to produce less noise and increase the channels to 16-20. Some of these new channels could be digitized on ACQ216 and used in the experiment. | ||
| Resource Requirements: | For the third step of the experiment all the usual resources for NTM suppression experiments are required.
Plasma current: 1.06 MA (standard direction) Toroidal field: -1.61 to -1.69 T (standard direction) Shape: Lower single null, as in 129330 Cryo-pumps: Lower pump (upper pumps desirable), He cooled Error Field Coils: I240-Coil for n = 1 error field correction. Neutral Beams All sources required including 30L and 330L for diagnostics (MSE, CER) and both 210 sources for counter injection and to control rotation. I coils in I240 configuration, powered by the SPAs. ECRH Two gyrotrons on the same launcher with modulation capability are the minimum required. Injected power > 1 MW for > 2s is needed. Four on two launchers (two on each desired). | ||
| Diagnostic Requirements: | It is desirable to improve the signal-to-noise-ratio of the oblique ECE diagnostic and this work is under way. Some oblique ECE channels need to be plugged into ACQ 216 for the experiment.
Magnetic (fast and slow) CER on 30L and 330L (for V4,5,6) alternating with 10 ms off/10 ms on from 0.9 to 5.0 s except for continuous 0.274 ms resolution for 160 ms around 4.5 s for a total of 994 pulses. MSE Thomson CO2 interferometers ECE radiometer oblique ECE SXR diodes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 252: Discharges for benchmarking XGC0 simulations of pedestal buildup after the L-H transition | ||
| Name: | Battaglia | Affiliation: | PPPL |
| Research Area: | Pedestal | Presentation time: | Requested |
| Co-Author(s): | R. Groebner, C.S. Chang | ITPA Joint Experiment : | No |
| Description: | The X-transport model of the tokamak pedestal predicts that the edge radial electric field must be negative enough to suppress the loss of thermal ions on neoclassical orbits through the X-point. The impact of X-transport on Er (and thus, the pedestal structure) scales strongly with the edge ion temperature, X-point geometry and the grad-B drift direction. The timescale of the pedestal growth is influenced by the ion orbit loss period, which is much longer in the unfavorable grad-B direction. This experiment aims to illustrate the importance of including the kinetic neoclassical X-transport effects when modeling the structure and formation of the pedestal following the L-H transition. The goal is to maintain a matched equilibrium shape, while characterizing the pedestal formation following the LH transition in the different grad-B drift directions for a range of collisionalities. It would be beneficial to also characterize the pedestal for two different shapes with extremes of X-point radius. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment would aim to characterize the pedestal around the LH transition in a LSN, low-triangularity shape with reversed Bt (unfavorable grad-B drift) over a range of collisionallities. The transition would need to occur during the current flattop. The heating power would be kept near the power threshold through the transition to maintain a slow (at least several ms) pedestal growth time. Matched discharges in the favorable grad-B drift direction would be developed in the second part of the experiment. Care would be taken to get the highest temporal resolution of profile diagnostics around the LH transition. Main-ion measurements would be very helpful for benchmarking the simulations. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | All available profile and divertor diagnostics. | ||
| Analysis Requirements: | XGC0, Kinetic EFIT | ||
| Other Requirements: | |||
| Title | 253: RMP ELM suppression in the unfavorable grad-B drift direction | ||
| Name: | Battaglia | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | T. Evans, R. Nazikian | ITPA Joint Experiment : | No |
| Description: | The experiment will aim to characterize the requirements for ELM suppression in a LSN plasma with the toroidal field reversed (unfavorable grad-B drift). 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 use LSN in order to take advantage of the lower divertor diagnostics and cryo-pump, and to connect to previous ELM suppression experiments in the LSN favorable grad-B drift direction. Also, if possible, larger heating power would be used to maintain H-mode while applying the RMP field. The results of this experiment would provide a valuable test for ELM suppression and 3D transport models. For example, XGC0 predicts that the grad-B drift direction will have a significant impact on the parallel and cross-field flows, which will alter the pedestal structure and RMP field penetration. Preliminary expectations is that for a given magnetic equilibrium and pedestal structure, the magnitude of the ExB flows are smaller in the unfavorable direction, which may lead to a total perpendicular electron flow that does not have a zero-crossing in the pedestal region. Consequently, islands may remain shielded and ELM suppression may not be possible or require larger RMP fields in the unfavorable grad-B configuration. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The primary goal is to achieve ELM suppression in the unfavorable grad-B drift direction and to characterize the pedestal structure and divertor conditions during suppression, especially Er, as best as possible. A baseline ELM suppressed, low-collisionallity discharge in favorable grad-B direction is needed prior to this experiment, preferably with matched diagnostics. The goal would be to maintain the established plasma same shape for the entire experiment. With Bt reversed, ramped Ip (q95) discharges will be used to see if any ELM suppression windows exist. If any suppression windows exist (with preference to any that match the favorable grad-B discharge), the pedestal before and during suppression will be characterized using plasma breathing. If successful at observing ELM suppression, the experiment could be repeated at a different beta-N. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | All available profile and divertor diagnostics. | ||
| Analysis Requirements: | XGC0, Kinetic EFIT, Varyped | ||
| Other Requirements: | |||
| Title | 254: Multi-field turbulence measurements in high beta H-mode discharges | ||
| Name: | Guttenfelder | Affiliation: | PPPL |
| Research Area: | Transport and Rotation | 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 (polarimetry) 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 "NSTX-matched" 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 spherical tokamaks (NSTX, MAST), conventional aspect ratio tokamaks (AUG, DIII-D) and RFPs (RFX). There is a clear motivation to measure turbulence characteristics associated with microtearing modes, in any machine.
Non-linear simulations for both NSTX [Guttenfelder, PRL 2011, PoP 2012] and AUG [Doerk, PRL 2011, PoP 2012] 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 are locally narrow around rational surfaces with spacing Drat=1/k_theta*s_hat (~2-3 rho_s, compared to 5-7 rho_s correlation lengths for ITG/TEM). 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]. 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). If successful, these measurements would 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 | 255: Cancelling the low-m non-resonant RMP fields during ELM suppresion using the C-coils | ||
| Name: | Battaglia | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Requested |
| Co-Author(s): | R. Nazikian | ITPA Joint Experiment : | No |
| Description: | Combining n>1 C-coil fields with n>1 I-coil fields will lead to unique vacuum RMP field spectrum for testing ELM suppression physics. With the appropriate phasing of the I- and C-coils, it should be possible to increase the amplitude of higher-m components of the RMP field, while reducing the amplitude of the lower-m modes. Thus, the RMP fields necessary for ELM suppression can exist near the plasma boundary while reducing the amplitude of the non-resonant low-m fields that lead to rotational drag in the core. The experiment would use two extreme cases, where the n=3 C- and I-coil fields are phased constructively and destructively to compare to models. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The I-coils would be run in the n=3 configuration with two power supplies (one upper, one lower). The C-coils would be run with three power supplies to generate n= 3 with n=1 error field correction. First, a reference discharge with RMP ELM suppression using even parity n=3 I-coil fields would be produced with the C-coil providing only the n=1 error field correction. Then the n=3 component of the C-coil that is in-phase with the I-coil will be added. Presuming this leads to ELM suppression, the I-coil current will be decreased until ELM suppression is no longer achieved. Next, the C-coil current will be increased until ELM suppression is regained. This incremental change of the C-coil and I-coil would be continued until a power supply limit is achieved. Finally, the experiment would be repeated with the C-coil out-of-phase, which is expected to enhance the low-m spectral components and amplify rotational drag. It may also be interesting to explore the combination of the n=3 C-coil with an odd parity I-coil field, which will provide additional unique poloidal spectra to benchmark models for field penetration. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | All available magnetics and profile diagnostics. | ||
| Analysis Requirements: | TRIP-3D, Surfmn, IPEC | ||
| Other Requirements: | |||
| Title | 256: SOL width and divertor peak heat flux in QH-mode plasmas with/without n=3 fields | ||
| Name: | Garofalo | Affiliation: | GA |
| Research Area: | ELM Control | 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 : | Yes |
| Experimental Approach/Plan: | Using the same tokamak set-up of Nov. 1, 2011, we can generate both counter-rotating QH-modes and co-rotating ELMing H-modes, for direct comparisons.
By operating right 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 right after boronization. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | Very recent boronization | ||
| Title | 257: Is the Cause of Density (Greenwald) Limit in Tokamaks Radiation Driven Islands? | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | D. Gates | 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"]. We propose to study and test this theory at DIII-D. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Vary the Ohmic/ECCD heating and non-inductive current fraction compare data to the prediction from theory. Measure the radiated power from the island to see if the power balance criteria is met. | ||
| Background: | One prediction of this work is that direct heating of the rational surfaces that participate in the radiation driven island phenomena should suppress these islands, since this would avoid the shielding process described above. Additionally, radiation driven islands should be exacerbated in plasmas with high non-inductive current fractions, since only the ohmic current participates in heating the interior of the island. This may explain the common practice of using "preventative ECRH" to avoid the onset of neoclassical tearing modes. In fact this phenomenon may partially explain the difficulty in finding a reliable predictor for the onset of neoclassical tearing modes because the radiation driven terms are not considered in neoclassical island threshold analysis. | ||
| Resource Requirements: | 1 day experimental time. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 258: Independent Li Control via Offaxis Beam | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Plasma Control and General Issues | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys | ITPA Joint Experiment : | No |
| Description: | New off-axis beam should enable some degree of independent Li control independent of dIp/dt. We want to implement an independent Li control and make it part of the PCS system. This will enable scientists to specify Li in experiments as they do BetaN today. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use relay-feedback or other other tuning algorithms to get PID control for Li. There needs to be a study of the coupling of this control to other controls. Combine with Beta control to enable independent control of Li and Beta. | ||
| Background: | The main driver for this proposal is that vertical stability is very much dependent on the Li. For AT scenario development scenarios and many other different regimes vertical controllability is the limiting factor. An independent Li control would over come these difficulties. Also, for many experiments Li is a very important parameter to keep constant which would benefit enormously from this development. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 259: Error Field Correction Improvement: Adding BetaN dependence | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Add BetaN dependent term to EFC. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Tune and test the improved EFC algorithm. | ||
| Background: | 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, we propose to add a BetaN dependent term to EFC algorithm. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 260: Higher Beta ELM-Suppressed Hybrids | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Continue RMP ELM control experiments in the ITER shape by controlling the 3/2 NTM amplitude to allow higher normalized beta to be achieved without rotational slowing and mode locking. The 3/2 NTM amplitude can be decreased by either (1) optimizing the error field correction to obtain higher rotation rates, (2) using ECCD at the q=1.5 surface, or (3) trying higher q95 (>4) if a RMP resonant window exists. The goal is to obtain RMP ELM-suppressed hybrids with beta_N~3 (close to the ideal no-wall limit). If naturally dominant 4/3 NTM hybrids can be produced, then the ECCD suppression of the 3/2 NTM is not necessary. The higher q95 cases should work because the coupling between the 3/2 NTM and the wall is weakened as the resonant layer is moved closer to the plasma center. This ELM suppression experiment should first use only co-NBI, but once optimized results are obtained then lower rotation plasmas should be studied using balanced NBI. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Repeat previous best ELM-suppressed hybrid case 129958. (2) Optimize error field correction to obtain highest rotation rates. (3) Use ECCD at q=1.5 surface to reduce and/or eliminate the 3/2 NTM. (4) Determine new beta limit during RMP ELM suppression. (5) Add counter NBI to slow rotation rate such that M<0.1. (6) Steer gyrotrons not needed for NTM control to core deposition to obtain Te=Ti. | ||
| Background: | Experiments in August 2007 used the I-coil to completely suppress ELMs in high beta hybrids for q95=3.7. If a dominant 4/3 NTM was present, then beta_N up to at least 2.5 could be achieved during the I-coil phase (actual beta limit not known). However, if a 3/2 NTM was present, then for beta_N>2.2 the plasma rotation slowed down rapidly during the I-coil phase and a locked mode terminated the hybrid discharge. While reducing or suppressing the 3/2 mode is one method to overcome this problem, improved error field correction may also help to reduce the amount of drag and help sustain the plasma rotation, especially in the pedestal region. | ||
| Resource Requirements: | NBI: All 8 sources are requested for long pulse operation.
EC: Minimum of 3 gyrotrons, prefer to have 6 gyrotrons. I-coil: Required with n=3 setup. Optimal error field correction is also needed. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 261: Collisionality scaling of particle transport | ||
| Name: | Doyle | Affiliation: | UC, Los Angeles |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | C. Petty, L. Zeng, S. Mordijck, T. Rhodes, W.A. Peebles, R. Pinsker | ITPA Joint Experiment : | No |
| Description: | Test GYRO predictions for collisionality scaling of electron particle transport over extended, range. A specific goal is to extend the data set at low collisionality. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize similarity scaling technique to vary collisionality. in particular, goal is to go to lower collsionaility than was achieved in 2011 (v*~0.01). This will be achieved by:
1) Operating at lower density 2) Utilizing increased electron heating from ECH - additional gyrotrons plus re-aiming. 3) Utilize FW system for electron heating. | ||
| Background: | Experiment in 2011 successfully tested GYRO predictions for collisionality scaling of particle transport for v*>0.01. However, GYRO also predicts a dramatic change in particle transport for v*<0.01, which was not tested. The proposal here is to extend the experimental collsionality range to encompass this second, low collisionality region. | ||
| Resource Requirements: | All gyrotrons
FW system | ||
| Diagnostic Requirements: | Profile reflectometer system
Turbulence diagnostic set | ||
| Analysis Requirements: | TGLF and GYRO modeling | ||
| Other Requirements: | -- | ||
| Title | 262: Increased FW Current Drive Efficiency at High Te | ||
| Name: | Pinsker | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | A. Nagy, P.M. Ryan, J.C. Hosea, R. Perkins, G. Taylor, R.H. Goulding, C.C. Petty, S. Diem, M. Kaufman | ITPA Joint Experiment : | No |
| Description: | Combine 6 or 7 gyrotrons-worth of central 110 GHz ECH (all launchers aimed at or near the center of the discharge without driving toroidal current) with the combined 60 MHz and 90 MHz fast wave power. We would use the minimum neutral power necessary to create the sawtooth-free discharge in which the driven currents can be accurately measured, and to make the (MSE) measurement. Both L-mode and H-mode target discharges would be tried, although at full rf power, one would expect difficulty in keeping the discharge in L-mode. The basic scan would be a density scan, at each case obtaining at least a matching pair of discharges with co- and counter-current FW phasing. The object of the exercise would be to extend the range of central electron beta values, and hence of single-pass absorption of the FW power, considerably beyond what was possible without high power ECH. If time permits, comparison of the current drive efficiency of the 60 MHz and 90 MHz systems could be performed - as the single-pass absorption increases, we expect at some point to observe more efficient current drive at higher launched parallel phase velocity (the higher frequency case). The experiment would seem to be the logical precursor to full utilization of the combined rf systems for AT work involving tailoring of the current profile. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | See description. | ||
| Background: | This experiment was tried on two days in the 2004-2005 campaign. However, technical problems prevented any useful data to be obtained. The experiment was also tried for a fraction of a day in 2010, but again technical problems precluded any advances from being made. This experiment is a natural one to split time with the TS/ECE discrepancy experiment (see A. White's contribution to the ROF), as the setup is almost identical. The DIII-D FWCD system was designed to be most efficient in a plasma with central electron temperature of about 10 keV, but the maximum electron temperature at which we have measured the FWCD efficiency to date is about 6 keV. The theoretical prediction is that the FWCD efficiency scales roughly linearly with central electron temperature, and all experiments to date have conformed with this prediction. | ||
| Resource Requirements: | Machine time: 1 day experiment. 5 gyrotrons minimum,4 NB sources, all three FWCD systems | ||
| Diagnostic Requirements: | All usual profile diagnostics, with MSE being especially important. | ||
| Analysis Requirements: | Analysis of current drive with MSE tools, NVLOOP. | ||
| Other Requirements: | -- | ||
| Title | 263: High Beta Hybrids and Pressure Profile Broadening | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Use 5 MW of off-axis beam injection to broaden the total pressure profile compared to on-axis injection. Most of this will be due to a change in the fast ion pressure profile, but some broadening of the thermal pressure profile may also occur depending upon how stiff the transport dependence is. Determine whether the broader pressure profile allows a high beta_N to be obtained in steady-state hybrid plasmas, with the goal being beta_N=4. Will also calculate whether the ideal wall limit changes significantly with the broader pressure profile for these low q_min discharges. The off-axis beam will probably not effect the current drive profile much since the off-axis NBCD efficiency remains high, and the poloidal magnetic flux pumping inherent in hybrids tends to keep the total current profile constant regardless of the driven current profile. Since the noninductive current drive is not crucial to this experiment, it could be done with either positive or negative B_T values.
For steady-state considerations, the co-ECCD should be deposited inside the q=1.5 surface for this experiment. However, we could broaden the scope of this experiment by re-directing some of the ECH power (probably 4 gyrotrons) to deposit at the q=2 surface to see if we can suppress the 2/1 mode. This would be deem a success if the beta limit comes from a RWM rather than a 2/1 mode (the latter is the current situation). |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) 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); the beta_N will likely be 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 beta_N=3.8 at high density, which is well above the ideal no-wall limit. At lower densities and with central co-ECCD, high beta hybrid plasmas have been created with beta_N=3.4 and nearly zero loop voltage (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 | 264: RF-only H-mode studies | ||
| Name: | Pinsker | Affiliation: | GA |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | G. Taylor, J.C. Hosea, R. Perkins, R.H. Goulding, P.M. Ryan, M. Porkolab, S. Diem, M. Kaufman | ITPA Joint Experiment : | No |
| Description: | ITER may not have enough auxiliary heating power to exceed the L/H transition power in the Day 1 configuration (hydrogen ops). For this reason, several machines have remeasured the L/H transition power in hydrogen with hydrogen or helium beams and compared those results with deuterium. Furthermore, recent work has shown that the L/H transition power has a dependence on plasma toroidal rotation speed, with lower rotation speeds being associated with lower L/H transition power levels. Even 20 percent-level effects may be important in this context. With this in mind, the fact that the Fast-Wave-only H-mode observed in 1991 had a distinctly lower power threshold than with NBI heating in the same discharge may be of importance. The fact that H-mode transitions were observed with fast wave (FW) power as the only auxiliary heating source, under conditions of rather low single-pass absorption was an important piece of evidence that multiple-pass absorption of the FW power can be efficient. By expanding the range of FW frequencies, densities (and hence target electron temperatures), and using 3rd harmonic ECH, we can get a more quantitative measurement of the edge losses by determining the L-H transition threshold power under varying single-pass absorption conditions. This is important to ITER, both from the point of view of improving knowledge of access to H-mode in plasmas with only intrinsic rotation (no torque) and also to improve understanding of FW edge losses under varying edge conditions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experiment consists of scans of target density, rf power (at two different frequencies: 60 MHz and 90 MHz), toroidal field, and whether 3rd harmonic ECH is added (at the appropriate toroidal field), and comparison of co-, counter-current, and push-pull phasing. A beam is used for comparison, later in the shot. Minimal beam blips are used for CER, MSE diagnostics. At each condition, the power threshold for L-H transition is observed for FW, for the comparison beam, and for ECH (at the appropriate fields). | ||
| Background: | H-modes with fast wave heating by direct electron absorption as the only form of auxiliary heating were discovered at DIII-D in July 1991, and have not been studied since. In particular, the fast waves in that experiment were launched with the shortest available parallel wavelength ("Pi phasing") at 60 MHz at around 1 T, and we have never studied H-modes with current drive phasing, either co- or counter-current, or at higher frequency than 60 MHz. Furthermore, the great interest in the dependence of L/H transition power levels on rotation and/or applied torque in recent years has provided a new motivation for this experiment, as mentioned in the description section above. Finally, insofar as this study provides further data on FW edge losses, the emphasis on this area on NSTX in the past several years has increased the need to obtain data at higher toroidal fields than can be run on NSTX to see how these effects scale with BT, to provide data both for possible future STs and for ITER FW heating.
In a piggyback experiment in 2010, it was found that H-mode transitions could be rather easily obtained with a very low NBI power plus 2-3 MW of FW power in directional (counter-current) phasing. In another piggyback experiment in 2011, FW+ECH-only H-modes at 2 T were produced rather easily (no NBI at all), with counter-current phasing of the FW. These results reduce the uncertainty that FW-only H-modes can be obtained in the lower-k-parallel 90 deg phasing. | ||
| Resource Requirements: | Machine Time: 1 day Experiment
Number of gyrotrons: 4 Number of neutral beam sources: 4 Three FW systems, one at 60 MHz and the others at 90 MHz. | ||
| Diagnostic Requirements: | Edge reflectometry with the antennas adjacent to the 285-300 FW antenna, along with the UCLA profile reflectometers, would be a very helpful addition to the usual diagnostic set for this experiment. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 265: First-principles Model-based Current Profile and Beta_N Control in DIII-D | ||
| Name: | Schuster | Affiliation: | Lehigh U |
| Research Area: | Plasma Control and General Issues | Presentation time: | Requested |
| Co-Author(s): | John Ferron, Tim Luce, Mike Walker, Dave Humphreys (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.
The objective of this experiment is to implement first-principles model-based controllers developed for the regulation of the q profile evolution during inductive discharges in DIII-D. Unique characteristics of the control approach are (i) the use of first-principles 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. This experiment intends to continue the successful tests performed during the last experimental campaign in L-mode. However, by adding ECH as an actuator, simultaneous current profile control and beta_N control will be sought in this case to precisely control L-H transitions. This will prevent uncontrolled L-H transitions observed in some of the recent experiments and will provide a mechanism for controlled confinement mode switching. The results of the experiments will also provide: i- some of the insight that is necessary to combine controllers separately designed for induction-dominated and bootstrap-dominated phases of the discharge (which require different dynamic models) and ii- an initial assessment of the performance of first-principles model-based current profile controllers in H-mode. 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 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 and beta_N. 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 either avoid L-H transition while regulating the current profile or to intentionally produce L-H transitions at some instant of the discharge. Different initial and target profiles will be considered. The first-principles model-based current profile control experiment will require half a day to possibly one day. | ||
| Background: | The control group at Lehigh University (LU) headed by Prof. Eugenio Schuster has been working on this problem for more than five years. A preliminary first-principle control-oriented model of current profile evolution in response to auxiliary heating and current drive systems (NBI, EC) and electric field due to induction was developed for the inductive 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 1/2â??day experiment showed qualitative agreement with the q profile evolution predicted by the simplified model 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]. These controllers were tested on 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. The controllers developed from first-principles models used three actuators - plasma current, beam total power and line-averaged density. By adding ECH as an actuator, we intend to experimentally test controllers currently under development for simultaneous current profile and beta_N regulation.
[1] Y. Ou et al., â??Towards Model-based Current Profile Control at DIII-D,â?? Fusion Engineering and Design 82 (2007) 1153â??1160. [2] Y. Ou et al., â??Extremum-Seeking Open-Loop Optimal Control of Plasma Current Profile at the DIII-D Tokamak,â?? Plasma Physics and Controlled Fusion, 50 (2008) 115001. [3] C. Xu et al., â??Ramp-Up Phase Current Profile Control of Tokamak Plasmas via Nonlinear Programming,â?? IEEE Trans. on Plasma Science, vol.38, no.2, p.163, 2010. [4] Y. Ou et al., â??Optimal Tracking Control of Current Profile in Tokamaks,â?? IEEE Transactions on Control Systems Technology 19 (2), 432-441 (2011). [5] Y. Ou et al., â??Robust Control Design for the Poloidal Magnetic Flux Profile Evolution in the Presence of Model Uncertainties,â?? IEEE Trans. on Plasma Science, vol.38, no.3, p.375, 2010. [6] M.D. Boyer et al., â??Backstepping Control of the Plasma Current Profile in the DIII-D Tokamak,â?? 2012 American Control Conference. [7] J. Barton et al., â??Robust Control of the Current Profile Evolution During the Ramp-up and Early Flat-top Phases of the Plasma Discharge in the DIII-D Tokamak,â?? 2012 American Control Conference. [8] J. Barton et al., â??First-Principles Current Profile Control for L-Mode Plasma Discharges in DIII-D via Robust Control Synthesis,â?? to be submitted to Nuclear Fusion. | ||
| Resource Requirements: | Machine time: 1/2 day experiment | ||
| Diagnostic Requirements: | Core and tangential Thomson, CER, CO2, magnetics, MSE, ECH diagnostics, a reasonable set of fast ion instability diagnostics (UF interferometers, FIR scattering, ECE at 500 kHz, fast magnetics with fast delay set in the current ramp), FIDA. For closed-loop experiment real-time magnetic measurements and equilibrium reconstruction including the q-profile (EFIT and RTEFIT with MSE) are essential, and real-time measurements of the ion and electron temperature profiles, as well as line-averaged plasma density or density profile are required. | ||
| Analysis Requirements: | Matlab software. | ||
| Other Requirements: | |||
| Title | 266: ITER with Reduced Solenoid Current: Test Feasibility of the New Scenario | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys, W. Walker | ITPA Joint Experiment : | No |
| Description: | The reduced solenoid current of ITER due to manufacturing difficulties brings many new challanges to the new scenario development in the feedforward and feedback control. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will need to develop a Multi-Input-Multi-Output shape control and improve upon the ITER scenarios developed in the previous years. Near-term experiments would make use of the new powered VFI capacity to enable this approach. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 267: High Beta, Steady State Hybrids | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This experiment will integrate a high beta hybrid plasma with the reactor relevance of Te~Ti and full noninductive current drive. In 2012, the possible addition of a seventh gyrotron and optimization of the 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 essential, but prefer 7 gyrotrons. 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 | 268: Vertical control of RE from ITER-like LSN targets | ||
| Name: | Eidietis | Affiliation: | GA |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Requested |
| Co-Author(s): | D. Humphreys | ITPA Joint Experiment : | No |
| Description: | Demonstrate robust vertical control of RE beams formed from vertically unstable elongated targets, particularly an ITER-like LSN. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with ITER-like LSN target. Kill the plasma using argon killer pellet. If this results in RE beam production, attempt vertical control with the D3D Z vertical control system modified to mimic the ITER system and the other shape control frozen to mimic ITER's long wall time and slow PF coils. If that is not found to be successful, reduce the artificial restrictions until a successful catch-and-hold is achieved.
If LSN targets are not producing RE beams at an acceptable rae, we can attempt to alter the current profile using off-axis beams and ECCD to increase the likelihood of RE beams. | ||
| Background: | The US is tasked with designing the ITER disruption mitigation system by 2013. One large question is whether it is possible to catch and hold a runaway beam in ITER for gradual dissipation, or if every beam is doomed to die in an inevitable rapid VDE after the thermal quench caused by a disruption or mitigation scheme.
Existing ITER modeling of major disruptions (Sugihara 2004 IAEA & 2007 Nucl Fusion) indicates that the plasma shape survives the TQ with the only major shape change being an increase in the outer gap and major radius (not inner gap). The subsequent VDE does not occur until well into the current quench. Given that those studies DO NOT include internal vertical control coils and that the formation of a robust runaway beam would significantly alter the current quench properties, there is the possibility that control of large runuaway currents can be reestablished. | ||
| Resource Requirements: | ECCD, 2mm argon pellets, off-axis NBI,
TIME = 2 days | ||
| Diagnostic Requirements: | BGO, FPLASTIC FZNS, fast magnetics. | ||
| Analysis Requirements: | These cases will serve as excellent bases for NIMROD modeling of RE confinement in elongated plasmas (Izzo). | ||
| Other Requirements: | |||
| Title | 269: Flux savings with ECH during the current ramp in ITER similar discharges | ||
| Name: | Jackson | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | A. Hyatt, J. Lohr, T. Luce, D. Humphreys, R. Prater | ITPA Joint Experiment : | No |
| Description: | Quantify potential flux savings using ECH and ECCD during the current ramp phase in ITER similar discharges. Compare to both OH startup and NBI. Note: although this may not be formally identified as an IO Urgent ITER Research Task, methods for flux savings are currently an area of interest for ITER | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | With an EC power scan, up to 7 gyrotrons, determine flux savings. Use both radial and oblique launch. Compare to OANB and OH startup. | ||
| Background: | ITER will have limited flux to achieve its goals. This has been exacerbated by derating the current in one of the solenoid PF coils. Experiments in DIII-D (Jackson, Phys.Plasmas, 2010) showed up to 20% flux savings during current ramp with EC at the same flattop value of li. This is somewhat higher than predicted and more careful experiments are needed to accurately determine reduction in flux consumption as a function of EC power and/or current drive. | ||
| Resource Requirements: | 6 or 7 gyrotrons, NBI (including OANB) | ||
| Diagnostic Requirements: | standard. mse is a key diagnostic for current profile analysis. | ||
| Analysis Requirements: | Toray, TRANSP, ONE-TWO | ||
| Other Requirements: | -- | ||
| Title | 270: High Power FW Coupling to RMP-ELM-stabilized discharges | ||
| Name: | Pinsker | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | T.E. Evans, G. Taylor, C.C. Petty, J.C. Hosea, R. Perkins, S. Diem, M. Kaufman, P.M. Ryan, R.H. Goulding, M. Porkolab | ITPA Joint Experiment : | No |
| Description: | It is arguable that since uncontrolled large ELMs are probably not acceptable for ITER and beyond and therefore ELM control is an absolute requirement, the most relevant regime for FW coupling is one in which ELMs have been suppressed with Resonant Magnetic Perturbations (RMPs). In this experiment, we would continue the study of high-power FW coupling and central electron heating in ELM-stabilized discharges with RMPs that we began in FY07. In a single day's exploratory experiment, we showed that ~2 MW of FW power could be coupled to an ELM-stabilized discharge, and obtained the first signs of central electron heating due to the FW. Much of the day was occupied with establishing an ELM-suppressed case with much smaller outer gap than had previously been used. Subsequent experiments in the ELM-control area have produced more robust ELM-stabilized cases at smaller outer gaps, so that we have a better starting point. Poor B-supply regulation in 2007 led to significant difficulty in staying in the narrow q-resonant window. We did not get a good no-rf comparison shot in either of the two cases which we studied (different outer gaps). No significant power was coupled at 60 MHz from the 285/300 antenna, due to a problem in the antenna which was remedied in the Fall 2007 vent. We need to continue this experiment, which is the world's first such attempt to couple fast wave power to an RMP-stabilized edge, with all of these issues addressed to obtain a publishable result on this important topic. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Continuation of the experiment on this topic from 2007, in which the beam power (=programmed beta) and outer gap are minimized while maintaining the ELM suppression with RMPs and acceptable outer wall heating, the FW power added and documenting the resulting electron heating. Comparison of different antenna phasings (both co-, both counter, push-pull), measurement of electron heating profile with modulation of the FW power. | ||
| Background: | See description. | ||
| Resource Requirements: | Machine time: one day experiment, Number of beam sources: 6, three FW systems, one at 60 MHz and two at ~90 MHz, 7 kA operation of the I-coil in the configuration used for RMP experiments. | ||
| Diagnostic Requirements: | The addition of the edge reflectometer adjacent to the 285-300 FW antenna, along with the UCLA profile reflectometers, would be a significant plus to these studies. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 271: Low-Z material erosion/deposition experiment to benchmark DIVIMP code being used by ITER | ||
| Name: | Chrobak | Affiliation: | Commonwealth Fusion Systems |
| Research Area: | Plasma-material Interface | Presentation time: | Requested |
| Co-Author(s): | P. Stangeby, A. Leonard, D. Rudakov, A. McLean, N. Brooks, W. Wampler, R. Doerner | ITPA Joint Experiment : | No |
| Description: | There is an urgent need to properly benchmark edge impurity transport codes for low-Z impurities with the use by ITER of the DIVIMP code to estimate the net erosion of the Be wall armor and the tritium retention by Be co-deposition [1]. Experiments are being prepared on EAST and JET to benchmark low-Z erosion/deposition (C in EAST, Be in JET), but only for 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. <br>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. It has been determined that Al is a suitable proxy for Be, capable of handling the power loads and able to provide the low-Z benchmarking data by being distinguishable from the surrounding plasma and plasma-facing surfaces. It is proposed here to expose a specially-designed sample with a surface layer of either Aluminum or Beryllium 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. <br>This unique and comprehensive benchmarking would also constitute a major validation for DIVIMP for application to DIII-D itself where we want to establish the relative importance of divertor and wall sources of impurities as regards to core contamination. DIVIMP automatically outputs all aspects of the calculated impurity behavior, including not just the net erosion and deposition patterns, but also the core contamination levels. When any aspect of the code output is confirmed by benchmarking, then confidence is increased in all aspects of the code output.<br>[1] ??Modelling of Beryllium Erosion-Redeposition on ITER First Wall Panels?, S.<br>Carpentier, R. A. Pitts, P. C. Stangeby, J. D. Elder, A. S. Kukushkin, S. Lisgo, W.<br>Fundamenski and D. Moulton, 19th International Conference on Plasma Surface Interactions in Controlled Fusion Devices, San Diego, May 24-28, 2010. Accepted for publication in J Nucl Mater. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Experimental Approach/Plan:
During the 2012 campaign, it is proposed that an initial experiment be performed to confirm that Al is a good proxy for Be, at least so far as the local deposition is concerned. A graphite DiMES head with small, ~1cm diameter Al-coated area would be exposed to a set of 5-6 LSN, L-mode discharges (identical to the ones used for the Mo DiMES experiment performed in the 2011 campaign) and again with the outer strike point placed on DiMES during the steady part of the shots. The head would then be removed for ion beam nuclear reaction analysis by W.R. Wampler of the amount of Al deposited on the surrounding surface of the DiMES head, as well as the amount eroded from the coating area. The gross erosion rate of the Al would be measured spectroscopically using Al I and Al II lines which are presently being identified on PISCES [2]. These data would then be used to benchmark the DIVIMP code for the case of low-Z materials in divertor type plasma contact, regarding the local deposition aspect of the transport. These results would also be compared to the 1996 results for Be [3]. If the Al behaves similarly to Be regarding the local deposition on the graphite surface of the DiMES head, then a second experiment would be carried out right at the end of the campaign to measure the short range transport of the Al. In this case, the entire head of DiMES would be covered with Al, in order to better approximate the toroidally symmetric plasma contact of a divertor ?? this requires that the mfp for ionization of the sputtered Al atoms be short compared with the Al sample size. During the vessel opening a number of tiles in the vicinity of DiMES would be removed and analyzed by Wampler for the distribution of Aluminum. These data would then be used to benchmark the DIVIMP code for the short range transport of low-Z materials in divertor plasma contact. These local and short range benchmarking experiments for divertor-type plasma contact, combined with the benchmarking tests on EAST and JET for the limiter-type plasma contact, would complete the qualification of DIVIMP for its use by ITER to estimate the erosion-wear of the low-Z Be wall and tritium co-deposition by net eroded Be. [2] R. Doerner, private communication, August 2011 [3] W.R. Wampler et al, Journal of Nuclear Materials 233-237 (1996) 791-797. | ||
| 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 from a 2011 experiment 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].
[4] Memo, N. H. Brooks, A. McLean ,?Camera Measurement of Gross Molybdenum Erosion,? 9/6/2011 [5] W.R. Wampler, Email communication 10/5/2011 [6] ??Measurements of carbon, deuterium, and boron deposition in DIII-D?, W.R. Wampler, et al, Journal of Nuclear Materials, 337-339 (2005) 134-138 [7] ??Experimental observations and modeling of carbon transport in the inner divertor of JET?, A. Kirschner, et al, Journal of Nuclear Materials, 337-339 (2005) 17-24. Proceedings of the 16th PSI conference, Portland, ME, 2004 | ||
| Resource Requirements: | ADAS calculations for selected Al emission lines
DiMES TV camera calibration for Al emission | ||
| Diagnostic Requirements: | DiMES TV camera view with 10nm FWHM spectral filter centered at 877nm or 670nm for Al-I
MDS spectrometer views 1) on DiMES and 2) just off DiMES Filterscope views with Al-I and Al-III (450nm) narrow band filters Tangential TV camera views of D-alpha emission and Al-III emission Langmuir probes Divertor-Thomson DiMES | ||
| Analysis Requirements: | Post-mortem NRA of exposed DiMES sample and surrounding tiles | ||
| Other Requirements: | Fabrication of DiMES sample
Identification of emission lines using PISCES-A at UCSD Measurement and modeling of S/XB values for Al emission lines | ||
| Title | 272: Data-driven Model-based Integrated Rotation and Current Profile Control in DIII-D | ||
| Name: | Schuster | Affiliation: | Lehigh U |
| Research Area: | Plasma Control and General Issues | Presentation time: | Requested |
| Co-Author(s): | Didier Moreau (CEA, Cadarache), John Ferron, Tim Luce, Mike Walker, Dave Humphreys (General Atomics), Egemen Kolemen (PPPL) | ITPA Joint Experiment : | No |
| Description: | Establishing suitable plasma profiles has been demonstrated to be a key condition for the achievement of advanced scenarios with improved confinement and possible steady-state operation. The current approach at DIII-D focuses on creating the desired plasma profiles during the plasma current ramp-up and early flattop phases with the aim of maintaining this target profile during the subsequent phases of the discharge. A closed-loop controller is necessary to regulate the current and kinetic profiles around the target values during the flattop.
The objective of this experiment is twofold. First, we intend to develop tools for routine use of closed-loop current, rotation and kinetic profile control in DIII-D by further developing the DIII-D/LU profile control algorithm. Second, we intend to implement and test model-based controllers developed for the combined regulation of the rotation and current profiles during the flattop phase of the discharge, which may also require some level of kinetic profile control (beta_N or electron/ion temperature profile). The model-based control algorithms are synthesized based on data-driven models identified during previous campaigns. The actuators are (i) co-current NBI power, (ii) counter-current NBI power, (iii) balanced NBI power, (iv) total ECCD power from all gyrotrons in a fixed off-axis current drive configuration, and (v) loop voltage or plasma current. The multivariable, model-based controllers developed within this project differ from non-model-based, empirically-tuned, PID (proportional-integral-derivative) controllers, in two distinctive aspects: 1- both static and dynamic knowledge of the system (identified model) is incorporated during the synthesis of the controller, ii- the relationships among all input and output variables are taken into account during the synthesis of the controller. These two distinctive aspects are indeed the reasons for which improved performance is expected from advanced multivariable model-based controllers. Indeed, the strong coupling between the different physical variables involved in the plasma transport phenomenon and the high complexity of its dynamics make unavoidable the use of information of the to-be-controlled system, i.e., dynamic models, during the synthesis of plasma profile controllers. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The profile control capability of the PCS was expanded in 2011 to include the possibility of simultaneous control of one magnetic profile (Ï?(Ï?), ι(Ï?), q(Ï?) or θ(Ï?)=â??Ï?/â??Ï?) and up to two kinetic profiles such as toroidal rotation and ion temperature together with one scalar variable such as beta_N. Magnetic profiles can be controlled at 20 radii (normalized Ï?=0.05-1) and kinetic profiles can be controlled at 10 radii (normalized Ï?=0.1-1). The profile control algorithm uses a variety of actuators that can include loop voltage, plasma current, line-averaged density, total/individual beam powers and total/individual ECH/ECCD powers. The magnetic profiles are obtained in real time from a complete equilibrium reconstruction using data from the Motional Stark Effect (MSE) diagnostic. Kinetic profiles such as toroidal rotation and ion temperature can be measured in real time through Charge Exchange Recombination (CER) spectroscopy. However, this feature of the algorithm was not ready for the 2011 experimental campaign, which prevented any integrated rotation and current profile control experiment. One of the main objectives of this proposal is to overcome this limitation and enhance routine profile control capabilities in DIII-D.
Dynamic response data was used to identify state-space dynamic models for the evolution of q, rotation and temperature profiles using subspace identification techniques. The identified modes were used for the synthesis of reduced-order controllers that exploit the time-scale separation between kinetic and magnetic variables and optimally regulate the profiles around the nominal values. The data-driven model-based integrated rotation and current profile control experiment will require half a day to possibly one day. However, it is absolutely necessary to dedicate at least two short (2 hours) preliminary sessions to debugging and testing of the control algorithms implemented in the PCS. It is important to emphasize at this point that the PCS (plasma control system) at DIII-D does have infrastructure for implementing such advanced controllers as was shown in the last campaign. However, additional work will be carried out before the experiment by the LU group to add to the DIII-D/LU profile control algorithm the capability of reconstructing in real time the rotation and kinetic profiles based on Charge Exchange Recombination (CER) data. | ||
| Background: | The control group at Lehigh University (LU) headed by Prof. Eugenio Schuster has started working during the 2008 experimental campaign on the identification of a dynamic response model for the q profile evolution during the flattop phase [1, 2]. Further experiments were carried out during the 2009 experimental campaign in collaboration with DIII-D (M. Walker, J. Ferron, T. Luce, D. Humphreys) and CEA-Cadarache (D. Moreau) [3, 4]. A reduced-order state-space model obtained from data using subspace identification techniques was combined with Optimal and Robust Control theory to synthesize closed-loop controllers that optimally regulate current, rotation and kinetic profiles [5, 6, 7, 8]. Several controllers integrating current profile and beta_N regulation were successfully tested during the 2011 experimental campaign. However, rotation profile control experiment could not be carried out because of lack of real-time CER data.
[1] C. Xu et al., â??Current Profile Evolution Modeling via Subspace Identification Algorithms,â?? DPP Annual Meeting of the American Physical Society (APS), 2008. [2] C. Xu et al., â??Transport Parameter Estimations of Plasma Transport Dynamics using the Extended Kalman Filter,â?? IEEE Trans. on Plasma Science, vol.38, no.3, p.359, 2010. [3] W. Wehner et al., â??Feedback Tracking Control of Safety Factor and Rotation Profile Evolutions in the DIII-D Tokamak via System Identification,â?? DPP Annual Meeting of the American Physical Society (APS), 2010. [4] D. Moreau et al., â??Plasma Models for Real-Time Control of Advanced Tokamak Scenarios,â?? Nuclear Fusion 51, 063009 (2011). [5] W. Wehner et al., â??Data-driven Modeling and Feedback Tracking Control of the Toroidal Rotation Profile for Advanced Tokamak Scenarios in DIII-D,â?? IEEE Multi-conference on Systems and Control (MSC), Denver, Colorado, September 28-30, 2011. [6] W. Shi et al., â??Multivariable Robust Control of the Plasma Rotational Transform Profile for Advanced Tokamak Scenarios in DIII-D,â?? Proceedings of the 2012 American Control Conference, Montreal, Canada, June 27-29, 2012. [7] W. Wehner et al., â??Optimal Feedback Control of the Poloidal Magnetic Flux Profile in the DIII-D Tokamak based on Identified Plasma Response Models,â?? Proceedings of the 2012 American Control Conference, Montreal, Canada, June 27-29, 2012. [8] W. Shi et al., â??System Identification and Robust Control of the Plasma Rotational Transform Profile and Normalized Beta for Advanced Tokamak Scenarios in DIII-D,â?? to be submitted to Nuclear Fusion. | ||
| Resource Requirements: | Machine time: At least two 2-hour evening sessions + 1/2 day experiment | ||
| Diagnostic Requirements: | Core and tangential Thomson, CER, CO2, magnetics, MSE, ECH diagnostics, a reasonable set of fast ion instability diagnostics (UF interferometers, FIR scattering, ECE at 500 kHz, fast magnetics with fast delay set in the current ramp), FIDA. For closed-loop experiment real-time magnetic measurements and equilibrium reconstruction including the q-profile (EFIT and RTEFIT with MSE) are essential, and real-time measurements of the ion and electron temperature profiles, as well as line-averaged plasma density or density profile are required. | ||
| Analysis Requirements: | Matlab software. | ||
| Other Requirements: | |||
| Title | 273: Test of NBI Burn Simulation algorithm | ||
| Name: | Eidietis | Affiliation: | GA |
| Research Area: | Plasma Control and General Issues | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | A new beam control algorithm has been implemented in the PCS that allows a subset of the beams to be slaved to a fusion burn calculation, allowing them to act like uncontrolled alpha heating sources. This algorithm is designed to support more full-featured burn control simulations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Place 2 beams under burn simulation control.
(2) Vary beta/density in controlled manner and verify that slaved beams are responding with expected power output. | ||
| Background: | Depending upon your transport model, an ignited plasma may not be thermally stable, requiring active feedback to remain at an acceptable level of power output. Techniques for controlling the burn and/or transport will be necessary to actively stabilize the burn. | ||
| Resource Requirements: | new PCS algorithm, NBI
TIME: 0.5 day | ||
| Diagnostic Requirements: | standard magnetics, RT CER | ||
| Analysis Requirements: | RT CER must be working | ||
| Other Requirements: | |||
| Title | 274: Stability of Non-Inductive Plasma at Low Current | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | J. Leuer | ITPA Joint Experiment : | No |
| Description: | Solenoidless startup (Leuer, 2011) and NI EC startup have been demonstrated to modest current levels. Due to power supply constraints of the DIII-D tokamak, the shape and position control were basically not available in these shots. Coupling of the EC needs position control of the plasma and since this control was not available, the EC coupling and maximum current were limited. Due to lack of shape control, diverting the plasma was not possible, either. <br><br>The question remains, if it is were possible to increase current further can we reach a steady state. <br><br>We propose to start at higher current levels by piggybacking in the rampdown at the end of shot and input all possible power sources. Then, see if we can stabilize the plasma current levels or see an increase in the current levels. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Piggyback at the end of the shot. Input all the available power. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 275: Quick measurement of off-axis NBCD Profile | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make a relatively quick (1 good discharge) measurement of the off-axis NBCD profile during a high beta, steady-state scenario experiment to confirm the current drive profile. This will aid our understanding of the current profile evolution as we look for any anomalous effects (such as transport effects). The key to a "quick" measurement is to modulate between the 150LT and 150RT beams. These beams drive different amount of current drive, which we will measure. The advantage of modulating 150LT/150RT compared to modulating between an on-axis and off-axis beam is that the former will be less perturbative to the plasma profiles and equilibrium. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | (1) Desire a target plasma where only one of the 150 beams is enough to produce a stable scenario of high interest. (2) During a stationary period, modulate between the 150LT and 150RT sources at 5-10 Hz. Need to have the 30LT beam on continuously for MSE measurements. | ||
| Background: | As shown in my 2011 APS poster, modulation of the beam sources can be used to directly measure the NBCD profile. A nice measurement of the 150 beam current drive was obtained using 6.7 Hz modulation in a single discharge. | ||
| Resource Requirements: | Need all 6 co sources. | ||
| Diagnostic Requirements: | MSE critical. | ||
| Analysis Requirements: | Will use direct analysis of the MSE signals. | ||
| Other Requirements: | |||
| Title | 276: Study of Incremental Pure Electron Heating in Advanced Inductive Plasmas | ||
| Name: | Pinsker | Affiliation: | GA |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Requested |
| Co-Author(s): | J.C. Hosea, M. Porkolab, R. Perkins, G. Taylor, R. Budny, S. Diem, M. Kaufman, P.M. Ryan, J. Ferron, R.H. Goulding, A. Nagy | ITPA Joint Experiment : | No |
| Description: | In experiments in 2010-11 in which Fast Wave (FW) electron heating was studied in Advanced Inductive (AI) targets, the effect of ~2 MW of FW power was compared with the incremental effect of a like amount of EC power. Important differences between the gross heating efficiency (change in stored energy/MW of incremental power) and central heating effectiveness (appropriately normalized increment in central electron temperature per MW) between the FW and EC heating were observed. Equally importantly, strong sensitivities of the efficiencies were observed (outer gap at less than ~4 cm, ELM frequency and character, DRSEP, etc.) Since ITER will have almost entirely torque-free, fueling-free, pure electron heating (alphas, high energy NBI, EC, FW), it is crucial to understand the effects of such heating sources on confinement of energy, particles and momentum. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with a case like 146571, in which an incremental pure electron heating power of ~1.5 MW of FW was compared with a similar level of EC. Perturb conditions such as centrality of EC (have to slightly raise the toroidal field to get more central deposition), width of EC deposition (by aiming launchers to spread out radial deposition), timing of EC relative to the FW, density, DRSEP, etc. If possible, reduce the NBI power and raise the FW and EC power levels to increase the fraction of electron heating to total heating power. | ||
| Background: | See above. This experiment is an important step towards the study of the DT-reactor-relevant regime of low-torque, non-fueling, pure electron heating. Furthermore, the DIII-D program aims to add as much EC power as possible to move in this direction - we should strive to understand the effects of such electron heating in relevant regimes with the tools that are available already. | ||
| Resource Requirements: | All available co-NB sources (on-axis), 6 gyrotrons, all three FW systems. 1 day of machine time. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 277: ITER Steady State Scenario Demonstration Revisit | ||
| Name: | Murakami | Affiliation: | Retired |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | JM Park, SSI Group | ITPA Joint Experiment : | No |
| Description: | TER 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 used extensively in ITPA-IOS group activities, such as H&CD mix studies, However, it needs new data based on improved hardware and diagnostic capabilities now available in DIII-D. The below is a resubmission of the old ideas in 2009, but the details need to be updated. In particular, NBI (including ITER-like off-axis NBI) and EC (including ITER-like EC launcher configurations) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 1) Reproduce 134372, ITER shape.
2) betaN scan to see how far can be increase 3) Adjust timing of the high power phase at q_min=2 crossing 4) Add broadly distributed ECCD around rho=0.3 â?? 0.6, and move out 5) Error field minimization 6) Document and evaluate f_NI, f_BS and G=betaN*H/q^2 7) Apply/optimize FWCD to control q(0) as well as off-axis ECCD | ||
| Background: | â?¢ In 2008, 2 successful ITER SS Demo shots: equiv. Ip=8.5 MA and Ip=13 MA. Initial Ip scan (with 2-D SS simulations with the scaled DIII-D edge) shows promise at Ipa at 9 MA < 9 MA with Day-1 H&CD.
â?¢ ITER steady state scenario modeling carried out using scaled edge from DIII-D ITER demo discharge (#134372). However, it comes short in simultaneously achieving fNI=100% and QDT=5. This is used in H&CD mixes discussion in ITPA-IOS activities. â?¢IP scan is important for optimization. â?¢ More credible ITER scenario development ,and addressing the H&CD mixes / Upgrade questions. â?¢We need IP scan with the ITER shape) for optimization of fNI and fusion performance G(Q) â?¢ We need good documentation of the edge profiles (Ï? = 0.8 â?? 1.0) including pedestal | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Full core diagnostics, Fast ion diagnostic (UC, Irvine), MSE (LLNL), edge reflectometer (ORNL, UCLA) | ||
| Analysis Requirements: | Scenario modeling with FASTRAN/ONETWO, TRANSP, ONETWO analysis; CURRAY/ONETWO, AORSA | ||
| Other Requirements: | |||
| Title | 278: Electron Critical Gradient and Heat Pinch | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | T.C. Luce | 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 7 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. | ||
| 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 (to within a constant of integration). This allowed an electron critical gradient to be identified that is in agreement with fluctuation measurements and GYRO calculations. Near rho=0.4 there was evidence for a small heat pinch at the lowest electron temperature gradient, but the evidence was weak at rho=0.6. 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 7 gyrotrons, minimum 6 gyrotrons.
NBI: Need 30LT, 330LT, 150LT and 210RT 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 | 279: FW Heating and CD in ITER Demo AI Discharges and Comparison with EC | ||
| Name: | Murakami | Affiliation: | Retired |
| Research Area: | ITER Inductive Scenarios | Presentation time: | Not requested |
| Co-Author(s): | P.M. Ryan, J.M. Park, RF Group, SSI Group | ITPA Joint Experiment : | No |
| Description: | Success FW source power, coupling to Advanced Inductive (AI) discharges should be extended further: (1) to include more FW power coupled to ITER Demo AI plasmas; (2) to optimize FW CD as well as heating (for long pulse); and (3) to include comparison with EC CD in counter-EC as well as co-EC cases. This is related to proposal for FW HCD in AT, but using ITER Demo AI plasmas, aiming at to help ITER H&CD upgrade and research plans. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | â?¢ Success FW source power, coupling to Advanced Inductive (AI) discharges, which includes comparison between FW (counter) and EC (co) at power of<2MW.
â?¢ DIII-D H&CD capabilities match quite well with ITER day-1 heating power capabilities. And DIII-D experiments utilizing these capabilities help significantly ITER research and H&CD upgrade plans. â?¢ Demonstration of substantial FW H&CD in ITER Demo discharges help for research plans for ITER Long pulse operation (by helping volt-saving and current profile control) â?¢ ITER EC system deign includes counter-EC as well as co-EC. If counter- FWCD in counter -EC is demonstrated to be as efficient as counter-EC CD, then EC system can be planned for more co-EC configuration. â?¢ DIII-D leads the ITPA efforts to develop ITER steady state scenario and scenario modeling using theory-based (GLF23) modeling with self-consistent sources and sink based on the day-1 hardware capabilities | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Scenario modeling with FASTRAN/ONETWO, TRANSP, ONETWO analysis; CURRAY/ONETWO, AORSA | ||
| Other Requirements: | fast wave (90 MHz and 60 MHz); >5 gyrotrons | ||
| Title | 280: W-fuzz exposure to weak and strike point plasma | ||
| Name: | Wong | Affiliation: | GA |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | D. Rudakov, R. P. Doerner, E. Hollmann | ITPA Joint Experiment : | No |
| Description: | Presently, W, because of it low physical sputtering rate and favorable thermal properties at high temperature is the preferred PFC surface material. But at high temperature ~1000 C, and high He fluence ~2x10e26 He/m2, nano-size W-fuzz could be generated which could contribute as high-Z impurity to the plasma. This experiment is to study the evolution and transport of W-fuzz, grown in PISCES, when exposed to weak background plasma,which means the strike point is away from DiMES location, and then plasma strike points. Initial exposures are proposed to be performed in piggyback mode. Specific dedicated experiments will be proposed in the future. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose with the use a DiMES button sample to expose a combination of W-fuzz, flat-W, and graphite buttons to the weak background plasma, which could simulate the chamber first wall to study the deposition of eroded material onto the W-fuzz and flat W button surfaces. Separately, we propose to expose similar combination of buttons to plasma strike points to study the evolution and transport of the W-fuzz and to compare it with the flat-W and graphite buttons. To study the possible migration of W, the W-I lines and core Z-effective will be monitored. | ||
| Background: | W-fuzz can be a major concern for the use of W as the surface material for DEMO, when high surface temperature and high He fluence operation is expected. The concerns would be in plasma core contamination the net erosion of the W-surface due to the formation and erosion of the W-fuzz. We propose to study the evolution and transport of the W-fuzz under weak plasma and plasma strike point conditions. | ||
| Resource Requirements: | W-fuzz, W and C buttons, and the DiMES system, and selected plasma discharges. | ||
| Diagnostic Requirements: | Monitoring W-I lines and core Z-effective | ||
| Analysis Requirements: | Examination of the samples before and after the exposure. | ||
| Other Requirements: | |||
| Title | 281: Fast wave heating and current drive in AT plasmas | ||
| Name: | Murakami | Affiliation: | Retired |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | P.M. Ryan, J.M. Park, RF Group, SSI Group | ITPA Joint Experiment : | No |
| Description: | FW operation significantly improved (FW source power, better coupling with local gas puff, etc). More aggressive application to steady state AT discharges are in order.
supplementing central heating (285-deg long-pulse antenna design), and dynamic central CD control (e.g., co-counter CD by push-pill with 0- & 180-deg antennas. Earlier (2005) attempt of RF application to steady state AT had seen some hint of FWCD. Coupling data with gap scans were done, but hampered by the old bumper limiter. With new bumper limiter and local gas puff, the plasma shape is different from the present AI discharges. We would like to: (1) Robust coupling to AT plasmas with ICRF. (2) Characterize beam ion absorption of FW power at 90HZ and 60MHz (3) Validate scenario modeling of FWCD in AT (4) RF edge diagnostic caracterizations such as reflectometers. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | NBI: >= 4 co-beams + 210RT
EC: >=5 gyrotrons FW: 60MHz + 90 MHz, >2.5 MW | ||
| Diagnostic Requirements: | Fast ion diagnostic (UC, Irvine)
edge reflectometer (ORNL, UCLA) | ||
| Analysis Requirements: | CURRAY/ONETWO, TRANSP, SciDAC-RF [AORSA/CQL3D; ORBIT-RF/TORIC, etc]; RANT3D | ||
| Other Requirements: | |||
| Title | 282: Main chamber wall heat flux during and between ELMs | ||
| Name: | Lasnier | Affiliation: | LLNL |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | M.A. Makowski | ITPA Joint Experiment : | No |
| Description: | Determine the scaling of heat flux to the main chamber wall during and between ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | View the main chamber walls using the IR view of the new periscope. Generate ELMing H-mode. Scan density, IP, and q95. | ||
| Background: | Scaling of heat flux to the main wall has not been determined experimentally. | ||
| Resource Requirements: | neutral beams, D2 puffing | ||
| Diagnostic Requirements: | Periscope IRTV, filterscopes, Thomson | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 283: I-mode studies on DIII-D with FW and EC heating | ||
| Name: | Pinsker | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Requested |
| Co-Author(s): | D. Whyte, A. White, A. Nagy, J.S. deGrassie, R.H. Goulding, P.M. Ryan, J.C. Hosea, R. Perkins, G. Taylor, S. Diem, M. Kaufman, M. Porkolab | ITPA Joint Experiment : | No |
| Description: | This experiment is a follow-on to 2011 ROF #59, "Exploration and Characterization of I-mode", where it was proposed to obtain C-Mod's I-mode on DIII-D with neutral beam heating. On C-mod, I-mode has been obtained with ICRF heating as the sole auxiliary heating. The principal motivation for the study of I-mode is an ELM-free confinement mode with H-mode level confinement, and as such I-mode is potentially a game-changer for ITER. However, I-mode also could also solve another problem for ITER, which is coupling high levels of ICRF power without excessive rf electric fields near the antenna. Those rf electric fields cause impurity problems (particularly in high-Z first-wall environments) and also limit the power density that can be obtained without electric breakdown of the antennas. The rf electric field level needed to couple a given power level scales with the antenna load resistance RL as RL**-2, and the L-mode-like edge plasma in I-mode yields antenna loading that is about twice what is observed after an H-mode transition. Hence, at a given fixed maximum antenna voltage, about four times as much ICRF power can be coupled in I-mode (or L-mode) than in an equivalent plasma with a conventional H-mode edge.
Furthermore, studies of HHFW heating on NSTX indicate that ELMs cause a significant degradation of HHFW heating efficiency, compared with ELM-free regimes {Hosea, et al., 2010). This result is consistent with previously published DIII-D FW results (Petty, et al., NF 1999, Vol. 39, 1421), which showed a dependence of FWCD efficiency on ELM frequency and character - at the highest ELM frequencies, the FW edge losses became much more important than in ELM-free regimes such as VH-mode, or with infrequent Type I ELMs. Hence, we expect that FW edge losses could be significantly reduced in I-mode, compared with ELMing H-modes. A crucial aspect of I-mode accessibility for use with ICRF is whether the accessibility has a dependence on the plasma/wall gap, i.e. the outer gap. To maximize the antenna loading, the outer gap is reduced to the minimum compatible with outer wall heating, acceptable rotation/confinement and accessibility to the desired confinement mode. Hence it is important to extend the I-mode accessibility study of ROF #66 to include the affect of the outer gap on I-mode threshold. Another aspect of I-mode accessibility that should be addressed is the possible role of rotation on I-mode access, in that the applied torque in C-Mod is very much lower with ICRF heating only than will be obtained with co-injection NBI. The scans listed in 2011 ROF #59 might be extended by adding NB torque to the possible controlling parameters, scanning from all co-injection to balanced injection; replacing NBI with non-torqueing FW heating in steps is another approach to this important aspect. Along similar lines, the possible effect of the mix of ion and electron heating will be studied by changing the mix of NBI and FW (and/or ECH.) |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The proposal would be to follow the NBI-only I-mode characterization experiment and start with the best (i.e. most robust I-mode access) case from that experiment. The NB power would be replaced with FW power from shot to shot, maintaining the total heating power level at that which maintains I-mode for as long a period as possible. If the BT for optimal I-mode turns out to be high enough to permit reasonably central 110 GHz X2 heating, and the maximum FW power level is reached (should be ~3.5 MW or higher), further replacement of NBI power with EC power is a possible continuation.
If the outer gap dependence of I-mode access has not yet been addressed by the time of this experiment, it should be studied as part of this work. The I-mode accessibility with NBI only could be extended to include a scan of outer gap and the FW loading measured non-perturbatively during that scan. The high-power FW portion of the experiment should be carried out at the minimum outer gap at which satisfactory I-modes are sustained. | ||
| Background: | Experiments on DIII-D with an L-mode edge but high heating power have a long history. One particular example of this kind of work in the 1990s was the fact that to measure the non-inductive current profile with MSE, as in the FWCD work, it was necessary to maintain a longish sawtooth-free period and an L-mode edge, to keep the density low and the power-per-particle as high as possible. When all three FW systems were added to a sufficiently high NB power to keep q0>1 for a long enough period to make the measurement, it was necessary to bias the shape upwards and reduce the inner gap to maximize the L-H transition power in order to maintain L-mode (see 19960311 for a typical example of a session of this kind.) A more recent example of something similar is 140715, from the TS/ECE discrepancy experiment, where an H-factor relative to 89P scaling of about 1.9, or an H-factor relative to H98y2 of about 0.86, is obtained with an L-mode edge. Again, unwanted H-mode transitions limited the beam power in this recent experiment. It is not being claimed that these discharges are I-modes, only that good confinement with an L-mode edge has been obtained with FW plus NB heating in DIII-D in regimes limited by unwanted H-mode transitions, even after having raised the L-H transition power as much as possible. | ||
| Resource Requirements: | 6 sources NBI, 3 FW systems, all available gyrotrons. 1 day experiment. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 284: Rotation Profile Control: 3D Fields and Combined Control with Current Profile | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Plasma Control and General Issues | Presentation time: | Requested |
| Co-Author(s): | M. Walker, D. Humphreys, E. Schuster | ITPA Joint Experiment : | No |
| Description: | Manipulation of the rotation profile, especially increased shear close to the edge, can get rid off micro instabilities with small scale eddies (turbulence) and suppress long wavelength instabilities.<br><br>3D coils are predicted to be possible actuator for edge rotation by the NTV theory. This has not been taken advantage of at DIII-D. We propose to add this actuator to the rotation control algorithm. <br><br>Current profile control was developed last year at DIII-D. The combination with rotation control would enable highly stable configurations against turbulence and NTMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use relay-feedback or other other tuning algorithms to get PID control for NTV rotation control. Study of the coupling of rotation control with current profile control. | ||
| Background: | We already did analytical studies at NSTX to develop a real-time rotation control algorithm that uses this effect in addition to the beams and developed a new control based on simplified momentum transport. It was ready to be implemented on NSTX but due to the breakdown of the machine, it never was implemented. | ||
| Resource Requirements: | 1 day experimental time. PCS algorithm modifications. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 285: Pedestal buildup rate with ELM frequency and impact on turbulence | ||
| Name: | Diallo | Affiliation: | PPPL |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | T. Osborne, R. Groebner, R. Maingi | ITPA Joint Experiment : | No |
| Description: | The goal of this idea is to investigate the pedestal height and width buildup rates during the inter-ELm phase when the ELM frequency is increased. In addition, a characterization of the associated pedestal top turbulence will be assessed as the ELM frequency is systematically varied. Questions that will be answered will be: Is the rate of increase of the pedestal height and width during the inter-ELM phase depending of the ELM frequency? Is the pressure gradient clamped very early in ELM cycle independent of ELM frequency? This experimental idea will also inform us on the associated turbulence characterizations, and will constrain the type of fluctuations (spatial and temporal scales) capable to exist during the inter-ELM phase and correlated with limiting the pedestal gradient. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We propose to start with the well characterized reference discharge from Osborne H-mode WS 2011, and then slowly increase the ELM frequency. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | Thomson, CHERS, BES, DBS, ECE (if can reach the pedestal top) | ||
| Analysis Requirements: | Profile analysis, turbulence analysis | ||
| Other Requirements: | |||
| Title | 286: Cancer Therapy with Fusion Neutrons | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Investigate tokamak as a possible source of fast neutrons for cancer therapy. Expose cells to D-D fusion neutrons. Investigate neutron penetration depth, biological damage, and effectiveness of Li-6 or B-10 in narrowing energy deposition region. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Layers of tissues of different thicknesses should be placed in the torus hall in small sealed plastic containers. After a day of exposure to fast neutrons, these samples will be sent to a biomedical lab. Cell damage should be obvious by microscopic inspection and comparison with control samples which were not exposed to neutrons. Parameters to adjust or scan might include: distance from the torus and/or duration (number of plasma shots) of exposure, thickness of the layer (to study the penetration depth), type of tissue (bone, muscle, adipose), type and diameter of neutron collimator. Grains of fast-neutron-absorbers (Li-6, B-10) can be placed in some tissues. Enhanced effects are expected in proximity of these grains. Initial studies can be carried out on healthy cells. Subsequently, if promising, they can be extended to actual tumor cells. | ||
| Background: | Fast neutrons (E>1MeV) are an emerging tool for cancer therapy. Similar to gamma rays used for radio-therapy, neutrons generate secondary electrons in their interaction with matter (e.g., with tissues). However, gamma's tend to generate few, highly energetic electrons. As a result of their high penetration depth, these secondary electrons only cause few ionizations in each cell they interact with. By contrast, neutrons generate many secondary electrons of less energy, causing more ionizations per cell. In brief, neutrons are more lethal to targeted cells than gamma's and, in fact, any radiation. Therapeutic neutrons are typically generated by accelerators (Be target bombardment with 50-60MeV protons generated by a cyclotron) or, in one case, by a fission reactor (after slow and epithermal neutrons are filtered out). The advantage of D-D fusion is that it directly produces high fluxes of neutrons of sufficient energy (2.45MeV) for therapeutic use in superficial tumors. D-T reactions would produce 14.1MeV of higher penetration depth, thus appropriate for the treatment of deep tumors. | ||
| Resource Requirements: | Several days of piggyback. | ||
| Diagnostic Requirements: | Standard DIII-D neutron diagnostics.
Possibly dedicated fast-neutron dosimeter at actual location of tissue. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 287: "Spiraling field" EFC | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Infer n=1, 2 and 3 error field under various conditions (not limited to low density) from the non-uniform rotation and amplitude modulation of a pre-existing mode (locked mode or resistive wall mode) in response to a 5Hz uniformly rotating n=1 perturbation that slowly grows in time (â??spiraling fieldâ??). If time, repeat without pre-existing mode: thanks to non-linearity of plasma response, measured fields are expected not to rotate and grow as uniformly as applied a.c. fields. Non-uniformities contain information on static error fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Generate a non-disruptive locked mode by ramping the beams and thus beta in a low-rotation (balanced injection) plasma. Tweak beta, q95 and post-locking NBI to ensure no disruption.
If necessary, add ECH or ECCD at q=2 location: the former is expected to keep the mode small (thus, less disruptive) regardless of being deposited in the island O- or X-point; the latter will introduce a modulation in the mode-amplitude and will make the rotation non-uniform in correspondence of O- and X-point deposition. This modulation and non-uniformity allow an even better characterization of the locked mode and, thus, of the EF. Apply growing, rotating (spiraling) magnetic perturbations (MPs). Infer the EF from the non-uniform response of the mode response (amplitude and phase) measured via internal saddle loops. The mode rotation will initially be intermittent, then complete but non-uniform, then more and more uniform. For simplicity the initial experiments can be carried out in the presence of a â??proxyâ??, deliberately applied EF. The measured EF amplitude and orientation can be compared with the applied values. In the following shots the proxy field can be progressively reduced, until measuring the actual intrinsic EF. | ||
| Background: | The basic idea is that a mode locks to the resultant of all error fields (EFs) and applied perturbations, known and unknown, static and time-varying. By applying a known MP and measuring the mode dynamics, we expect to infer the amplitude and phase of the dominant unknown EF, as we recently demonstrated at EXTRAP-T2R. The method can be reiterated: once characterized, the dominant EF can be compensated, and the experiment repeated, searching for the second dominant n, and so on.
A standard EFC method consists in fixing the MP phase phi_MP and ramping the density down until locking. Then phi_MP is scanned shot-by-shot. The EF is inferred from the values of A_MP (or n_e) at locking. Note that ramps are pre-programmed, they do not stop at locking and often terminate with disruptions. Moreover, 3-4 discharges of this kind are needed for every new EFC, i.e. in principle for every new scenario. The approach proposed here consists in scanning both the MP phase phi_MP and, more slowly, its amplitude A_MP. The resulting MP rotates and grows, i.e. it spirals out. The behaviour of a mode locked to and dragged by the EF+MP resultant changes as different regions of the normalized amplitude A_MP/A_EF and normalized phase phi_MP-phi_EF plane are explored. Here A_EF and phi_EF denote the amplitude and phase of the dominant unknown EF. A_EF can be estimated because it coincides with the smallest A_MP for which the mode is successfully dragged for a complete toroidal rotation. Before then, rotation will be incomplete, and phi_EF will be the mid-phase of the former incomplete rotations. All this requires the presence of a non-disruptive mode in the plasma. This mode can either be pre-existing, seeded by EF-penetration, e.g. by an earlier density ramp-down, or, inevitably, it will automatically be generated during the MP â??spiralâ??, as soon as the total amplitude becomes high enough. This method might represent a non-disruptive, ITER-relevant, quicker (as it requires a fraction of a shot, instead of four shots) alternative to the conventional technique of fixing the MP phase and ramping the density down until locking, then repeating for a different MP phase in a shot-to-shot scan. This conventional EFC is restricted to low-density locked modes and thus, inevitably, low beta. The method proposed, instead, also works at high density, high beta and low q95. Finally, there are prospects of generalization to multi-mode EFC, by identifying other features in the normalized amplitude vs. normalized phase plane, and there are prospects of generalization from forcefully rotating LMs to spontaneously rotating NTMs and Quasi-Stationary Modes (QSMs). | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 288: Study NRMF driven torque in ECH-only heated plasma (ELMing H-modes) | ||
| Name: | Garofalo | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | 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 | 289: Test Transport Theory/Codes in Closed-Loop Profile Control XPs | ||
| Name: | Kolemen | Affiliation: | PPPL |
| Research Area: | Transport and Rotation | Presentation time: | Requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | There are many transport theory and codes. We have very limited tests of these theories with respect to experiments. One of the main reasons for this is that open loop testing these theories via perturbations is hard due to drift in many quantities and many effects that needs to be taken into account. Close loop testing gets rid of many of these issues because in this experiment we will be setting actuators and profiles we want to achieve.
I propose to test these against closed loop current profile control experiments. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | There needs to be high level of coordination with theorist on the optimal experiment design to test various aspects of theory. However, it is expected that profile controlled shots are setup and the request at various points of the profile are changed in time to identify the dynamics of transport. | ||
| Background: | Various transport codes are already working for the DIII-D tokamak. These codes generally have not been tested against each other independently. There is reluctance in experimentalist to use any of these tools since there is not any quantified results for the regime and task of their interest. | ||
| Resource Requirements: | |||
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| Other Requirements: | |||
| Title | 290: Integration of the 288 GHz R0 polarimeter measurements into EFITs | ||
| Name: | Zhang | Affiliation: | UC, Los Angeles |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Requested |
| Co-Author(s): | Tony Peebles, Troy Carter, Neal Crocker, Edward Doyle, Lang Lao, Terry Rhodes | ITPA Joint Experiment : | No |
| Description: | This experiment aims at using the 288 GHz polarimeter measurements to provide constraints on the EFITs, especially the current profile reconstruction. The polarimeter on DIII-D operates in a geometry of retroreflection along a horizontal radial direction, and outputs a phase determined by the density and magnetic field along the chord, so the final phase contains meaningful information, if interpreted correctly. A synthetic diagnostic code developed for the polarimeter can take the input of density and magnetic profiles, which are from Thomson scattering and EFITs, respectively, and calculate the expected polarimetry output. Preliminary comparison between the actual polarimeter measurements with its expectation shows reasonable for a few previous shots, where the polarimeter probing beam is not suffering too much refraction.
This proposal is not asking for dedicated experiment, but bring the awareness to the public that we prefer upper single null discharges to reduce the refraction of the polarimeter probing beam if this requirement does not compromise the purpose of the goal of experiments. We are particularly interested in piggybacking the ECCD or off-axis NBI experiments. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
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| Title | 291: ELM pacing with edge ECH | ||
| Name: | Jackson | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | J. Lohr, J. deGrassie | ITPA Joint Experiment : | No |
| Description: | Reduce ELM amplitude and W_ELM using modulated ECH deposited in, or near, the pedestal region. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Deposit ECH in the pedestal region. Vary the ECH modulation frequency and measure changes in ELM frequency and ELM energy. | ||
| Background: | ELM size in ITER is severely limited due to potential damage of PFCs. One scheme is to increase ELM frequency and decrease the energy per ELM. This has been demonstrated in DIII-D (J. Lohr, GAA20182, 1991) and in TCV (J. Rossel, APS11). DIII-D is the ideal machine to evaluate various ELM pacing techniques directly and determine which, if any, can meet ITER's needs. | ||
| Resource Requirements: | ECH, 6 gyrotrons | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | Need to ensure that ECH is not reflected into sensitive diagnostics or windows. This may require aiming ECH perpendicular to the LCFS. If so, it places a limit on plasma shape and Bt. | ||
| Title | 292: Test Drift Model for Heat Flux Width | ||
| Name: | Makowski | Affiliation: | AKIMA Infrastructure Services |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | A. Leonard, C. Lasnier, T. Eich | ITPA Joint Experiment : | No |
| Description: | In recent multi-machine studies it has been found that the heat flux width scales as a/Ip ~ 1/Bp and essentially independent of other plasma and engineering parameters. Two models have been proposed to explain this simple and strong dependence on Bp. The first posits that the grad-B drift carries particles into SOL with a characteristic scale length corresponding to the poloidal gyroradius. The second model posits that a critical pressure gradient extends up to separatrix and into the SOL and sets the SOL scale length.
If the drift model is correct, the heat flux width should be dependent on the direction of the grad-B drift as the SOL flows will be substantially different when the drift is away from the x-point rather than towards the x-point. If the critical gradient model is correct, there should be little dependence on the direction of the grad-B drift. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish LSN in reverse Bt plasma (grad-B drift away from x-point). Raise NB heating power in steps to determine H-mode threshold. Acquire heat flux profiles as a function of plasma current in ELMy H-mode phase of discharge. (Steady state heat flux is obtained by analysis of between ELM heat flux profiles). | ||
| Background: | Parallel heat flux to material surfaces is a critical design parameter for next step devices such as ITER, FNSF, and DEMO. The maximum steady-state heat load sustainable by material surfaces is in the range of ~10 MW/m2, giving rise to maximum parallel heat fluxes of ~1 GW/m2. In a multi-machine study, three devices, CMOD, DIII-D, and NSTX, undertook to measure heat flux profiles and other scrape off layer (SOL) parameters under similar plasma conditions. The divertor conditions were restricted to attached regimes in type I ELMy H-modes. The main result is that the parallel heat flux width is approximately proportional to a/Ip ~ 1/Bp.
Theoretical models have been developed to predict pedestal and SOL parameters including the heat flux width. Critical gradient models form one class descriptive of the pedestal. This type of model predicts that the pedestal pressure gradient rises until an MHD mode is destabilized, at which point, the turbulent transport arising from the MHD mode rapidly increases, limiting any further increase in the pressure gradient. The MHD modes are local so that the pressure limit is radially dependent. At a minimum, the saturated turbulent transport imposes a boundary condition on the fluxes into the SOL. However, it appears that the critical gradient model may be applicable all the way up to the separatrix and possibly into the near SOL. We examine whether gradients in electron temperature, Te, density, ne, and pressure in the immediate vicinity (1-2% in flux) of the separatrix can be related to the heat flux width in DIII-D. Detailed analysis is made possible by a new high spatial resolution Thomson scattering system. A class of SOL models proposes that guiding center drifts transport particles from the core into the SOL which then fuel parallel flows to the divertor targets and determine the heat flux width. The dependence of the heat flux scaling on a/Ip that is found experimentally suggests a scaling correlated with the inverse of the poloidal field consistent with this type of model. | ||
| Resource Requirements: | Reverse Bt operation | ||
| Diagnostic Requirements: | IRTV
Floor Langmuir probes Bolometers Midplane plunging probe Filterscopes Thompson core and divertor | ||
| Analysis Requirements: | Analysis of heat flux measurements
Profile analysis Pedestal analysis | ||
| Other Requirements: | |||
| Title | 293: Dynamic Pedestal Optimization | ||
| Name: | Snyder | Affiliation: | ORNL |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | R. Groebner, T. Osborne, K. Burrell | ITPA Joint Experiment : | No |
| Description: | Use predictions from the EPED model to optimize the pedestal height under various conditions in DIII-D. In particular take advantage of the predicted access to regions of very high pedestal pressure via dynamic modification of density during the discharge evolution (starting low, and then increasing). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Explore optimization of the pedestal height in a set of theoretically-motivated configurations with strong shaping and a range of q. Dynamically increase density via core fuelling to reach high pedestal pressures. | ||
| Background: | |||
| Resource Requirements: | Good wall conditions | ||
| Diagnostic Requirements: | high res Thomson, libeam desired | ||
| Analysis Requirements: | Modified version of EPED to allow detailed shape optimization | ||
| Other Requirements: | |||
| Title | 294: VH Mode Characterization and Control | ||
| Name: | Snyder | Affiliation: | ORNL |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Detailed characterization of VH mode plasmas with high resolution Thomson to determine if edge barrier is a wide single barrier or a pair of separate barriers. After establishment and diagnosis of high quality VH modes, attempt to control edge with RMP fields and tailored q95 values to prevent X-event and maintain steady state. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | High res Thomson, libeam desirable | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 295: Dependence of detachment of the outer divertor leg on R_tar in H-mode | ||
| Name: | Petrie | Affiliation: | GA |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | M.A. Mahdavi, D.N. Hill, P.C. Stangeby | ITPA Joint Experiment : | No |
| Description: | We assess how changes in the radial location of the outer divertor strike point (R_tar) affect the capability to reduce heat flux at the outer divertor target, as well as fueling and H-mode quality of the core plasma, under both attached and detached (radiating divertor) H-mode plasmas. Moreover, since there is variation in the lower DIII-D divertor structure, we will also be comparing the divertor and core plasma behavior under open and closed divertor conditions. Upon successful completion, we expect to have data relevant to making an informed appraisal of possible future divertor modifications to the DIII-D vessel. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Our base case H-mode plasma is a lower single-null divertor with the ion grad_B drift direction toward the X-point: Bt = -2.0 T, Ip = 1.0 MA, q95 = 3.2, and Pinj=6 MW. R_tar = 1.20 m, 1.30 m, 1.33 m, 1.50 m, 1.58 m, 1.695 m; add R_tar = 1.26 m and 1.64 m cases, if time allows. These are ??high X-point? plasmas and are approximately modeled by shot 146780; if necessary, X-point height can be lowered slightly in order to achieve the highest R_tar = 1.695 m case. The R_tar = 1.33 m and 1.695 m cases would be representative of the ??closed? divertor. One R_tar value per shot. Measurables include outer divertor target density n_tar, target temperature T_tar, target heat flux Q_perp, upstream density on the separatrix n_sep, and upstream temperature on the separatrix T_sep vs R_tar. To assess the quasi-baffling effect on the ease of detachment, R_tar scans at different core densities in H-mode for both floor and baffle top on shot-by-shot basis: R_tar = 1.20 m, 1.30 m, 1.33 m, 1.50 m, 1.58 m, 1.695 m. Fixed values of core density is obtained via feedback, with ne_V2 is =3.5-, 6.5 x 10^19 m-3. [Note that (4.0-4.5) x 10^19 m^-3 is already likely to have been obtained from the first part of this experiment.] After the 6.5 x 10^19 m^-3 density case has been completed, a steady gas puff will be used to achieve detachment. | ||
| Background: | Extremely high (and possibly destructive) values of heat flux can be anticipated for future highly powered, high performance tokamaks. Among the ways for dealing with this problem include increasing the ??toroidal flux expansion? (i.e., increasing R_tar in this experiment) and operating under detached or nearly-detached plasma conditions (e.g., exploiting the ??radiating divertor? concept). In this experiment we examine each of these approaches separately and in combination. What we will particularly watch for is not only how heat flux is reduced but also how these approaches feedback on plasma performance in the core.
Our previous studies have found for L-mode plasmas that the variations in the density n_tar and temperature T_tar at the outer strike point (OSP) were found in qualitative agreement with the two-point model when ??Floor? and ??Shelf Top? regions were considered separately, i.e., n_tar increased and T_tar decreased as R_tar increased. However, the data showed a clear discontinuity when the outer strike point was swept over the shelf top. We postulated that this discontinuity was due to less effective neutral trapping when the outer strike point was on the shelf top, leading to much reduced recycling. Our analysis further suggested that re-ionization of escaping neutrals upstream could drive a convected heat flux towards the target and that this heat flux tends to increase divertor temperature and reduced the divertor density. SOLPS-Eirene code modeling was consistent with this interpretation. So, in this experiment we want to closely examine these possibilities in the H-mode regime and under steady state conditions (and not the more transient conditions obtained by R_tar sweeps.) | ||
| Resource Requirements: | Machine time 1.0 day (in forward Bt), minimum 6 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 | 296: Detachment of the outer divertor separatrix with partial baffling | ||
| Name: | Petrie | Affiliation: | GA |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | D.N. Hill, P.C. Stangeby | ITPA Joint Experiment : | No |
| Description: | This experiment examines divertor detachment under open and closed divertor. Effort is made minimize the effects of radial location of the outer divertor separatrix strike point (R_tar), SOL parallel connection length (L_par), and poloidal flux expansion (f_exp). Upon successful completion, we expect to have data providing (1) an initial quantitative assessment of advantages, if any, that divertor closure in DIII-D provides for detachment, and (2) an evaluation of how open and closed divertor configurations affect core plasma behavior during detachment. Detachment in low and high power input cases is considered. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Our base case H-mode plasma is a lower single-null divertor with the ion grad_B drift direction toward the X-point. Bt = -1.95 T, Ip = 1.0 MA, q95 = 4.0, and Pinj=5-8 MW. R_tar = 1.32 m is the ??closed? case and R_tar = 1.41 m is the ??open? divertor case. The model plasma is shot 147739 @ t = 5.0 s. Flux expansion for both cases is 4.0. Although the angle the separatrix flux surface makes with the divertor surface can be somewhat greater in the former. This can be remedied by a small translation of the entire core and SOL outward by 8-9 cm. For each case, constant gas puffing pushes the divertor to detachment. ??Low? beam power (5 MW) and ??high? power (8 MW) cases are examined. Measurables include the time evolution of the target heat flux Q_perp, plasma energy confinement, and the upstream pedestal and separatrix density and temperature. Note that this experiment is complementary with an accompanying ROF proposal. | ||
| Background: | Extremely high (and possible destructive) values of heat flux are expected for future highly powered, high performance tokamaks. One of the possibilities of dealing with this problem includes operating in detached or nearly-detached plasmas. ??Baffling? the divertor is one way to facilitate detachment, although it is unclear how much (or if) this may affect pedestal and core plasma performance. In this experiment we will find this out. | ||
| Resource Requirements: | Machine time 0.2 day (4 good shots), ion grad_B drift direction is toward the X-point, minimum 6 co-beams. Note that this experiment can be done with fewer shots, if done in conjunction with experiment number 295. | ||
| Diagnostic Requirements: | Asdex gauges, core Thomson scattering, lower divertor fixed Langmuir probes, lower divertor IR camera, bolometer, and CER. | ||
| Analysis Requirements: | SOLPS/UEDGE and ONETWO | ||
| Other Requirements: | -- | ||
| Title | 297: Effect of the ion grad_B drift direction on divertor density, temperature, and heat flux | ||
| Name: | Petrie | Affiliation: | GA |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | D.N. Hill, P.C. Stangeby | ITPA Joint Experiment : | No |
| Description: | This experiment examines how reversing direction of the toroidal field affects divertor behavior in terms of density, temperature, and heat flux. The effort here is to reprised selected parts of a previously submitted experiment 295, which had focused on R_tar variation when the ion grad_B drift was directed into the divertor. Mini-scans in R_tar and plasma density (where the grad_B is directed away from the X-point) will be done in this experiment in order to compare with existing results (where the grad_B is directed toward the X-point). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Our base case in this experiment is a lower single-null divertor where the ion grad_B drift is directed away from the X-point: Bt = 2.0 T, Ip = 1.0 MA, q95 = 3.2, and Pinj=6 MW. R_tar = 1.20 m, 1.32 m, 1.50 m, and 1.695 m; add R_tar = 1.26 m and 1.58 m, if time allows. These are ??high X-point? plasmas and are approximately modeled by shot 146780; if necessary, X-point height can be lowered in order to achieve the highest R_tar = 1.695 m case. One R_tar value per shot. Measurables include outer divertor target density n_tar, target temperature T_tar, target heat flux Q_perp, upstream density on the separatrix n_sep, and upstream temperature on the separatrix T_sep vs R_tar. To assess the detachment, a steady gas puff is used to achieve detachment at individual R_tar values on shot-by-shot basis: R_tar = 1.20 m, 1.32 m, 1.50 m, 1.695 m; if time allows, add R_tar = 1.30 m and 1.58 m. | ||
| Background: | Previous studies of DIII-D divertor plasmas have shown that the distribution of the recycling at the inner and outer divertor targets depends strongly of the grad_B ion drift direction. Modeling of the SOL with UEDGE showed that these asymmetries in recycling were largely driven by the Er x B poloidal drift. The origin of the electric field (Er) arises mainly from the radial gradient in the electron temperature with respect to the flux surfaces in the private flux region and its direction is always into the PFR. For the ion grad_B directed toward the lower divertor, the Er x B drift transports ions through the private flux region from the outer plate to the inner plate in the lower divertor, thus increasing the recycling at the inboard plate. When the ion grad_B direction is reversed (i.e., when the toroidal field is reversed), the Er x B drift transport ions from the inner plate to the outer plate, thus increasing recycling at the outboard plate.
The bottom line: When the ion grad_B case toward the lower divertor, the stronger recycling at the inner divertor target for lowers temperature and raised density at that location, leading eventually (according to UEDGE) to more efficient fueling of the core plasma and detachment at lower core density, than if the ion grad_B drift direction was out of the divertor. How this translates into how easily detachment can be achieved for these large L_par and R_tar cases is the issue, when the ion grad_B drift direction is out of the divertor, even if the divertor is ??closed? (R_tar = 1.32 m and 1.695 m). | ||
| Resource Requirements: | Machine time 0.5 day (in reverse Bt), minimum 6 co-beams. | ||
| Diagnostic Requirements: | Asdex gauges, core Thomson scattering, lower divertor fixed Langmuir probes, bolometers, and CER. | ||
| Analysis Requirements: | SOLPS/UEDGE and ONETWO | ||
| Other Requirements: | -- | ||
| Title | 298: Tests of RMP Working Model at Higher Density | ||
| Name: | Snyder | Affiliation: | ORNL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Test the recently developed EPED-based model for RMP ELM suppression by attempting to raise the density of ELM suppressed discharges after suppression (in ECE optimized configuration). According to the model, this should allow ELM suppression at higher pedestal pressure, and should also make ECE more black near the edge to attempt direct study of field penetration effects on T_e (possibly including I-coil field oscillations) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Attain RMP ELM-suppression in strongly shaped discharges optimized for ECE. After ELM-suppression is attained, increase density via core fuelling while maintaining ELM suppression. Employ q ramps to optimize suppression and i-coil oscillations to allow detailed study of field penetration with ECE and ECE imaging. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | high res Thomson, ECE, ECE imaging, libeam if possible | ||
| Analysis Requirements: | EPED studies, M3D-C1 response calculations | ||
| Other Requirements: | |||
| Title | 299: Micro-Tearing Modes and Collisionality Scaling | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | W. Guttenfelder, C. Holland, T. Rhodes | ITPA Joint Experiment : | No |
| Description: | Look for direct evidence that micro-tearing modes are responsible for the strong collisionality scaling of confinement seen in H-mode plasmas. We will make turbulence measurements for the first time during a factor-of-4 collisionality scan in H-mode plasmas. Micro-tearing modes have small scale radial structure, so the UCLA high-k scattering diagnostic and the MIT PCI diagnostic are essential. The longer wavelength turbulence diagnostics can make contributions as a fiducial comparison. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will do a factor-of-4 collisionality scan in H-mode plasmas holding the other dimensionless parameters fixed. (1) Start with the high collisionality case, BT=1.4 T and Ip=1 MA. Establish a high beta (beta_N=2.4), high density (6e+19) ELMy H-mode plasma. Document fluctuations. (2) Measure the particle transport using deuterium gas puffs and Helium puffs. (3) Lower collisionality by increase field and current to 2.0 T and 1.4 MA. Keep density the same using cryopumping. Repeat fluctuation documentation. (4) Repeat particle transport measurement. | ||
| Background: | Numerous machines have shown that (normalized) H-mode confinement decreases with higher collisionality, the effect being strongest at highest collisionality. This strong scaling is difficult to reconcile with ITG/TEM transport, but it could be explained by micro-tearing modes. Micro-tearing modes should be important at high collisionality, high beta and flat density profiles. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 300: Check of gyro kinetic model | ||
| Name: | Prater | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | Chris Holland, Anne White | ITPA Joint Experiment : | No |
| Description: | A very large effort was made to test the ability of GYRO to model the turbulence and transport in an L-mode plasma. This work showed that the transport inside rho=0.7 could be understood, but not outside. The point of this experiment is to test whether the agreement in the core is fortuitous and whether the minor radius where the agreement begins to fail is universal. The way to make this test is to vary a plasma parameter that we know will affect the transport, and see whether the calculated turbulence and transport in the core match the measured turbulence and transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A good way to vary the core transport is to scale the plasma current (or the edge value of q at constant Bt). 128913 was done with Ip=1 MA, and by repeating the experiment at, say, 0.6 MA and 1.25 MA, the core transport could be varied by a factor 2. Some effort to keep some other dimensionless parameters constant, like the collisionality, could be made, but the experiment would be successful if the experimental core transport varied. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | All the turbulence diagnostics are critical, as well as MSE and the profile diagnostics, including reflectometers and Thomson scattering. | ||
| Analysis Requirements: | Running GYRO is essential. Using the knowledge gained from all the work with 128913, this might not be as difficult as before. Use of synthetic diagnostics is necessary. | ||
| Other Requirements: | |||
| Title | 301: Carefully limited Miultiharmonic Dynamic Error Field Reduction to Improve SS performance | ||
| Name: | Buttery | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | guess... | ITPA Joint Experiment : | No |
| Description: | Error fields and tearing modes play a key role in potentially limiting advanced scenarios. but there study has been rather disruptive in the pst - sometimes literally, sometimes using shots. However, by limiting excursions and rates of change in feedback system, it should be possible to routinely improve error correction by multiharmonic magnetic feedback (poloidal or toroidal), progressively improving a preprogrammed guess (STD correction *coef*betaN). the feedback itself shows through coil currents arrived at whether this is having significant effect. And it may improve performance and address stability physics questions without using any/much extra time. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Piggy back on SSI discharges with limited changes and very slow feedback (~100ms) DEFC, but enabling independent correction from different coil arrays or tor numbers. | ||
| Background: | Ultimately we must deal with improved EFC for low torque AT devices | ||
| Resource Requirements: | Audio amp control and spa's - Cs and Is. Improved control algorithms? | ||
| Diagnostic Requirements: | RWM sensors | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 302: Pedestal response induced by perturbations on n=3 static fields | ||
| Name: | Diallo | Affiliation: | PPPL |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | M. Podesta, T.Osborne, and others | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to determine the pedestal response induced by perturbations on n=3 static fields. To do so, we start with an ELMy regime, and record the corresponding coils currents and configuration as baseline. Then using carefully crafted waveforms (square pulses), we applied these waveforms on top of the now static n=3 fields. By varying the pulse height and width, and relative timing between pulses, systematic measurements of the electron density and temperature in synchronous with the pulse timings are performed. This approach will inform us on potential island formations time scales on the pedestal top and on the characteristic time scales of the edge profile response to 3D fields. With theses time characteristics, one can estimate/infer the diffusive and convective terms of transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We consider a DC n=3 fields as a reference waveform. We then use a scalpel approach whereby additive trains of square waveforms are generated. These waveforms are characterized by their height (coil current), width, and delay in time relative to the Thomson pulses. A judicious choice of this delay enables profile measurements that span the entire square pulse. Note that this approach will allow for profile reconstruction synchronous with the square pulse. Finally a scan of the pulse height and width is performed until the ELM suppression of achieved. | ||
| Background: | |||
| Resource Requirements: | Thomson timings, n=3 RMP waveforms | ||
| Diagnostic Requirements: | Thomson, CHERS, Edge turbulence diagnostics | ||
| Analysis Requirements: | Profiles, turbulence,ONETWO, ELITE | ||
| Other Requirements: | |||
| Title | 303: Snowflake divertor at DIII-D | ||
| Name: | Soukhanovskii | Affiliation: | LLNL |
| Research Area: | Plasma-material Interface | Presentation time: | Requested |
| Co-Author(s): | D. D. Ryutov, E. Kolemen | ITPA Joint Experiment : | No |
| Description: | This experiment will establish the baseline for understanding the snowflake divertor performance at DIII-D. Comparisons to an extensive physics database of snowflake divertor on NSTX will be made. We are particularly interested to look into details that were not characterized well in NSTX, namely, MHD pedestal stability with the snowflake (for possible ELM control), the snowflake divertor power balance during a Type I ELM cycle, and power balance and radiation distribution in the snowflake divertor between ELMs. Additionally, demonstration of heat flux mitigation with snowflake divertor and simultaneously particle control via cryopumping would be a giant contribution to NSTX Upgrade mission and research program. Experimental data from DIII-D would enable projections of snowflake to FDF and FNSF. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | After snowflake divertor configuration control is developed and demonstrated ( E. Kolemen et al., Proposal 248 ), perform pedestal and divertor measurements in two configurations: snowflake-plus and snowflake-minus.
Perform basic parameter scans, e.g. input power (i.e. P_SOL) , SOL/divertor collisionality (via D2 puffing). Further ideas - add argon puffing, use LD cryopumping. | ||
| Background: | Snowflake divertor was proposed theoretically (D. D. Ryutov, Phys. Plasmas 14 (2007) 064502; D. D. Ryutov et al., Phys. Plasmas 15 (2008) 092501) and studied experimentally on NSTX (V. A. Soukhanovskii et al., Nucl. Fusion 51 (2011) 012001; EPS 2011; Phys. Plasmas, to be submitted (2012)) and on TCV (Piras, F. et al., PPCF 51 (2009) 055009; PPCF 52 (2010) 124010; PRL 105 (2010) 155003). | ||
| Resource Requirements: | Demonstrated snowflake control by PCS is a prerequisite.
High-to-medium triangularity LSN plasmas, NBI heating with 1-2 MW increments, divertor gas. | ||
| 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 | ||
| Other Requirements: | |||
| Title | 304: Effect of the Ion Drift Orbits on the Main Ion Temperature Profile in the Pedestal | ||
| Name: | Kagan | Affiliation: | Los Alamos National Lab |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Perform direct measurements of the background ion temperature profiles for a wide range of the pedestal width to the poloidal ion gyroradius ratios. Determine if the discrepancy between the ion temperature and plasma density scales grows as this ratio goes from ~5 to ~0.5. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | |||
| Background: | A first-principle based analysis finds that in a banana regime pedestal the main ion temperature profile must be much wider than the drift ion orbit; i.e. its characteristic scale must be noticeably greater than the poloidal ion gyroradius. Plasma density does not have such a limitation and, in fact, is found to have a scale comparable to the poloidal ion gyroradius in many experiments. Hence, when the pedestal width to rho_pol ratio is small the ion temperature profile must be much wider than that of the plasma density, whereas once this ratio becomes larger the two profiles are allowed to have similar scales. Direct measurements of the main ion temperature by deGrassie supports this point in the pedestal as wide as (1/2)rho_pol, but comparing the two profiles in the series of shots with pedestal width to rho_pol ratio ranging from ~5 to 0.5 would provide a more solid evidence for the mechanism underlying temperature equilibration in a banana regime pedestal. Clarification of this mechanism is necessary for adequate theoretical description of pedestals. Currently the ion temperature profile is often taken to be as narrow as the pedestal itself when modeling H-Mode, which is justified by impurity ion temperature measurements. However, impurity ion species is more collisional than the main one, making physics behind establishing its temperature profile quite different. It is therefore crucial to measure the temperature of main ions directly rather than to deduce it from that of impurities to elucidate the issue. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 305: Effect of the Background Electric Field on the Poloidal Flow in the Pedestal | ||
| Name: | Kagan | Affiliation: | Los Alamos National Lab |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the main ion species poloidal flow near the electric field maximum in a low collisionality pedestal, as well as the electric field itself. By looking at different spatial locations/shots find the net poloidal velocity dependence on the radial electric field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | It is highly desirable to measure the net poloidal velocity of background ions directly. However, even if it is only the toroidal component of the main ion flow that can be measured directly, but at the same time both toroidal and poloidal flow components can be measured for impurities, the main ion poloidal velocity can be recovered quite robustly through the pressure balance equation. | ||
| Background: | Poloidal flow of background ions is neoclassical in nature. In other words, it is due to the drift motion that ion gyrocenters undergo in the tokamak magnetic field line geometry. In a subsonic pedestal of a width comparable to the poloidal ion gyroradius a strong radial electric field arises to maintain pressure balance, making the corresponding gyrocenter orbits substantially different from their core counterparts. As a result, in banana and plateau regime pedestals, neoclassical phenomena become dependent upon the electric field. Most interestingly, a recent first principle study predicts that in the banana regime pedestal the poloidal flow of background ions is reduced in magnitude, or even reversed, compared to what is seen in the core. This prediction was indirectly confirmed by comparing the net poloidal velocity of boron impurities observed in the C-Mod pedestal with the first-principle based expression accounting for the electric field. Either measuring the main ion poloidal flow directly or deducing it through the pressure balance as described in the previous section should provide a more accurate knowledge of this flow. Hence, the proposed experiment would allow further verifying of the electric field effect on neoclassical flows in the pedestal. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 306: NRMF driven torque in ECH-only heated plasma | ||
| Name: | Garofalo | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | 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: | Repeat the May 13, 2009 experiments, this time using the C-coil for n=3 application instead of the I-coil. The C-coil does not produce the strong density pump-out that the I-coil produces. The C-coil, therefore, should keep betan unaffected across the n=3 switch-on time. | ||
| Background: | Effect of n=3 fields on ECH H-modes was observed for the first time in experiments on May 13, 2009 [Miniproposal # 2009-02-03]. Only odd
parity n=3 I-coil fields were used. Strong density pump-out was observed, accompanied by strong reduction of beta. Since the intrinsic torque is expected to depend strongly on beta, to disentangle the intrinsic torque from the neoclassical NRMF torque, constant beta is desired across the application of the NRMF. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 307: Control of Major Disruptions in DIII-D | ||
| Name: | Sen | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | Robert Granitz, MIT | 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 â??pressure,â?? 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 availible in:
"Feedback control of major disruptions in ITER" PHYSICS OF PLASMAS 18, 082502 (2011) 10.1063/1.3598449 | ||
| Resource Requirements: | Gating of ECH beams. | ||
| Diagnostic Requirements: | ECE imaging system based on shotkey barrier diodes. | ||
| Analysis Requirements: | Image reconstruction and localization of the rough centroid of the relative MHD modes. | ||
| Other Requirements: | None | ||
| Title | 308: Deuterium toroidal rotation measurement | ||
| Name: | Stacey | Affiliation: | Georgia Tech |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure to toroidal rotation in matched pairs of shots (e.g. L and H mode phase, before and after an ELM, Elming H-mode and RMP) using the methodology descripted by Grierson et al. in Rev. Sci. Instr. 81, 10D735 (2010). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Select match pairs of shots/timeslices in which the edge differences are associated with some particular phenomenon (e.g. an ELM, a L-H transition, RMP turned on). Measure densities, temperatures, rotation velocities, radial electric fields. | ||
| Background: | Backgroundâ??Developing a Predictive Edge Pedestal Profiles Model
W. M. Stacey Georgia Tech, Atlanta, GA 30332; The physics of the edge pedestal may be categorized as: A) the physics of the transport processes which determine the density and temperature profiles in the absence of or between ELMs; and B) the physics of the MHD instabilities which set a limiting pressure gradient for the onset of ELMs. The transport processes in category-A arise from three different types of phenomena: 1) collisional or fluctuation phenomena which drive particle, momentum and energy fluxes: 2) force (momentum) and energy flux balance requirements that determine pressure and temperature gradients; and 3) the loss of charged particles from the plasma by other mechanisms. There have been ongoing investigations1-3 of the type-2 and type-3 transport processes in DIII-D over the past few years and of their implications for priority in identification of the mechanisms causing type-1 transport processes. (1 PoP 17, 112512 (2010), 2NF 51, 013007 (2011), 3 NF 51, 063024 (2011)) Radial and toroidal momentum balance require that the pressure gradient be balanced by the and smaller radial forces, which provides a requirement that the ion radial pressure gradient scale length be equal to the difference in the radial ion velocity and the â??pinch velocityâ?? divided by a â??diffusion coefficientâ??. The pinch velocity involves the radial electric field, the poloidal and toroidal rotation velocities, and other terms. The toroidal angular momentum transport frequency (due to viscoscous, charge-exchange, inertial and anomalous torques) is contained in both the â??pinch velocityâ?? and the â??diffusion coefficientâ??. The heat conduction relation requires that the ion temperature gradient scale length equal the conductive heat flux divided by the product of the ion density and temperature and thermal diffusivity. The ion density gradient scale length then follows subtracting the temperature and pressure gradient scale length. The ion radial velocity can be determined determined by solving the continuity equation. This formalism provides a first-principles predictive model for the edge pedestal pressure, density and temperature profiles, once the type-1 thermal diffusivities and angular momentum transport frequencies are known. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Rev. Sci. Instrum. 81, 10D735 (2010) | ||
| Analysis Requirements: | In order to extract the toroidal momentum transport frequencies from the measure deuterium and carbon toroidal rotation velocities, it will be necessary to solve the associated toroidal momentum balance equations backwards at each radial location, using the measured toroidal rotation velocities as input and solving for the momentum transport frequencies. | ||
| Other Requirements: | |||
| Title | 309: Enhancement of the Bootstrap Current in the Pedestal | ||
| Name: | Kagan | Affiliation: | Los Alamos National Lab |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the bootstrap current near the electric field maximum in a low collisionality pedestal, as well as the electric field itself. By looking at different spatial locations/shots find this current dependence on the radial electric field. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Lithium beam diagnostics similar to the one described in D. Thomas et al Phys. Rev. Lett. 93, 065003 (2004). | ||
| Background: | The strong radial electric field, inherent to a subsonic tokamak pedestal, cannot modify drift orbits of electrons as it does for ions, because the poloidal gyroradius of the former is much less than that of the latter. Indirectly, however, electrons do feel the electric field through their friction with ions, whose net flow is substantially modified by this field. A revised expression for the bootstrap current including the effect of the electric field predicts that in a banana regime pedestal this current is larger than it is given by conventional neoclassical formulae. Due to indubitable practical importance of the bootstrap current direct observation of the described effect could strongly impact major tokamak experiments. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 310: Test stability model of divertor heat flux width | ||
| Name: | Leonard | Affiliation: | GA |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | During a power scan measure divertor heat flux profile and upstream SOL plasma profiles. Compare separatrix pressure gradient scaling to that expected from ideal MHD stability. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Carry out a power scan, roughly 2 MW to 12 MW of input power in a configuration optimized for divertor and SOL diagnostics. The divertor strikepoint should be on either the shelf, or floor, in a location where there is plenty of space to measure at least 2 e-folding lengths of the divertor heat flux from the IR camera. The Thomson location should be adjusted to make sure the high spatial resolution measurement chords extend at least two cm into the SOL. The power scan should be on a shot to shot basis in order to collect enough Thomson data for the highest quality profile data possible. The Thomson separatrix location should be scanned across at least a couple of channels. The lower divertor should have a diagnostic sweep to determine the target plate electron temperature from the target plate Langmuir probes. The scaling of the upstream separatrix pressure gradient will be compared with that expected from ideal MHD stability calculations. | ||
| Background: | Measurement of the divertor heat flux width across a number of tokamaks has revealed a consistent scaling. However, the physical processes leading to that scaling has not been established, lending uncertainty in a applying the empirical scaling to ITER, FNSF and other future devices. A critical pressure gradient at the separatrix that scales like the ideal ballooning mode has been put forward as a possible mechanism controlling the width of the heat flux profile. The separatrix pressure is expected to vary with a number of parameters, but the strongest dependence should be on the parallel heat flux density. The observation that the heat flux width does not change with power implies that the separatrix pressure should increase linearly with power and weakly with target plasma temperature. And thus the separatrix pressure gradient should increase with power. The MHD stability limit would not be expected to increase similarly. On DIII-D we have now measured the upstream plasma profile scaling with plasma current and density. High quality measurements of the scaling of the SOL profiles with power is now needed. | ||
| Resource Requirements: | 6 NBI sources, LSN, standard BT | ||
| Diagnostic Requirements: | Core Thomson, Divertor Thomson | ||
| Analysis Requirements: | Thomson profile analysis, Ideal Ballooning stability analysis with Baloo | ||
| Other Requirements: | |||
| Title | 311: Power balance at divertor detachment | ||
| Name: | Leonard | Affiliation: | GA |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Document both inboard and outboard divertor as a function of density up to detachment. Use power balance measurements to document where power is flowing and compare with fluid modeling of detachment. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Carry out a density scan while fully measuring the divertor plasma. The divertor plasma measurements will include IR camera heat flux, ion flux from the target probes, divertor Thomson for the outboard divertor and the new center post probe for the inboard divertor. The X-point probe should be used to measure the cross field drifts. Some divertor sweeping will be required for full documentation of the divertor. Carry out density scan for a second higher power if time is available. | ||
| Background: | Fluid codes have not adequately modeled divertor detachment in that they require a higher upstream density than is observed at detachment onset. Detachment should be set essentially by power balance. When density rises to the point where power can no longer support a target temperature of about 5 eV, detachment onsets. This is one way to track down and isolate where the discrepancy between model and experiment. Is power flowing to a different region than expected? Or are radiation rates different than expected for the measured conditions? | ||
| Resource Requirements: | LSN and forward Bt. 6 NBI sources should be available. | ||
| Diagnostic Requirements: | Divertor Thomson should be available and working well. The new center-post insertable probe should be available. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 312: SOL profiles and main wall ion fluxes during RMP ELM suppression | ||
| Name: | Leonard | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Document the far SOL profiles and ion flux to the main chamber during RMP ELM suppression. Toroidal rotation of the applied fields would allow measurement of toroidal symmetry in the far SOL | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Setup a standard case to document the divertor and the far SOL with and without applied RMP fields for ELM suppression. The configuration should be set up optimally for diagnostic measurements. The SOL should be measured with high resolution edge Thomson extending a cm or two into the SOL. The edge reflectometer should be collecting data, and the mid-plane probe should be inserted as far as possible for far SOL profiles. The divertor strike-point should be located so the full profile of the heat flux can be measured by the IR camera and the target plate conditions measured by the target plate Langmuir probes. Even if full ELM suppression cannot be obtained in this configuration the resulting profiles should still be measured and compared to a non-RMP case. Rotating the applied RMP fields should allow determination of the toroidal asymmetry in the SOL and divertor. | ||
| Background: | ITERâ??s specification for main chamber wall fluxes is based measurements without applied 3D fields. There has been some indication that SOL fluxes in between ELMs can increase when RMP fields for ELM suppression are applied. The effect of RMP fields on the far SOL has not yet been documented. ITER critically needs this information if any first wall design changes are needed to handle any additional wall fluxes from RMP ELM suppression. This is particularly true if wall fluxes become highly localized in the poloidal and toroidal directions. This experiment also allows for measuring the toroidal symmetry of fluxes to the divertor. | ||
| Resource Requirements: | I-coils in a configuration that can rotate toroidally the applied ELM suppression magnetic field. | ||
| Diagnostic Requirements: | All of the boundary and SOL diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 313: VH-mode core impurity control with small 'pacing' pellets | ||
| Name: | Unterberg | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, N. Commaux | ITPA Joint Experiment : | No |
| Description: | Recent results using small pellet (~ 1.3mm) at high frequency showed a large reduction in the core nickel content. This new tool can be used to stabilize (prevent x-events) and extend the duration of the VH-mode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Characterization of the VH-mode pedestal would be desired to document the potential changes with pellets. | ||
| Background: | Can/should be combined with ROF# 294 (VH-mode characterization and control by P. Synder). | ||
| Resource Requirements: | 60Hz, 1.3 mm, pellets | ||
| Diagnostic Requirements: | "high-quality" Thomson | ||
| Analysis Requirements: | Osborne profile analysis. PB stability analysis. | ||
| Other Requirements: | -- | ||
| Title | 314: VH-mode core impurity control with small 'pacing' pellets | ||
| Name: | Unterberg | Affiliation: | ORNL |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, N. Commaux | ITPA Joint Experiment : | No |
| Description: | Same as ROF#313.
Recent results using small pellets (1.3mm) at high frequency (~ < 60 Hz) showed a large reduction in the core nickel content. This potential new tool can be used to stabilize (prevent x-events) and extend the duration of the VH-mode. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Characterization of the VH-mode pedestal would be desired to document the effects of the pellets. | ||
| Background: | Can/should be combined with ROF# 294 (VH-mode characterization and control by P. Snyder). | ||
| Resource Requirements: | 60 Hz, 1.3 mm pellets | ||
| Diagnostic Requirements: | hi-res Thomson | ||
| Analysis Requirements: | Osborne profile analysis. P-B stability analysis. | ||
| Other Requirements: | |||
| Title | 315: Degradation of Off-axis Neutral Beam Current Drive due to Microturbulence | ||
| Name: | Pace | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Off-axis NBI Physics Group, and Energetic Particles Group | ITPA Joint Experiment : | Yes |
| Description: | Off-axis current drive with neutral beams is susceptible to degradation due to microturbulence because the off-axis ion deposition places them in a region of significant turbulent fluctuation amplitude. The best off-axis neutral beam current drive is achieved when the magnetic field is directed opposite the typical orientation, i.e., counter-clockwise from above. Previous work (MP 2011-51-03) focused on measuring turbulent fluctuations in standard magnetic field direction, at the sacrifice of off-axis current. It remains to perform an experiment with optimal off-axis current drive parameters. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A reversed-Bt L-mode will be developed based on shot 145183. This shot is dominated by ion temperature gradient-type turbulence. Companion discharges that add ECRH power at rho = 0.2 will produce a larger Te/Ti ratio that leads to a trapped electron mode-type turbulence. This provides a fundamental difference in the beam ion/microturbulence interaction.
The reversed field and the innjection of ECRH at rho = 0.2 are major changes compared to previous discharges. | ||
| Background: | A relevant previous experiment was conducted under MP 2011-51-03. That experiment focused on characterization of the turbulent fluctuations, which required operating with standard toroidal magnetic field direction. While excellent turbulence measurements were obtained, the effect on off-axis current drive might not be determined. While we tried to reproduce the operating space developed successfully by the former TMV Group, our diagnostic needs required an additional neutral beam compared to their past experience. This meant that we went into H-mode more readily, especially with the addition of ECRH. Furthermore, the ECRH deposition at rho = 0.4 did not produce enough of a Te increase to shift the plasma into a trapped electron mode-type turbulence regime. | ||
| Resource Requirements: | - off-axis NBI at maximum injection angle
- ECRH power > 3 MW | ||
| Diagnostic Requirements: | Turbulence set: DBS, CECE, BES, PCI
Current Drive set: MSE | ||
| Analysis Requirements: | Current Drive: NUBEAM and other modeling as developed by the Off-axis NB Physics Group in 2011 | ||
| Other Requirements: | Simulation efforts to extend with GYRO, GTC, and offered to other groups. | ||
| Title | 316: Reduction of Fast Wave Antenna Coupling due to Parametric Decay Instabilities | ||
| Name: | Pace | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Energetic Particles Group | ITPA Joint Experiment : | No |
| Description: | High-harmonic fast wave heating can suffer parasitic loss due to the excitation of parametric decay instabilities near the antenna strap. This is most commonly an issue when a resonance is located near the strap (e.g., an 8th ion cyclotron harmonic). Fast wave heating and current drive is efficient because the waves, once past the edge, easily travel to the plasma core and are absorbed. Avoiding power loss in the edge is the key to further improving high-harmonic fast wave performance. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | ++ This experiment utilizes the same plasmas as necessary for ROF submission 1: Investigate Disagreements Between Thomson Scattering and ECE Measurements in High Te Discharges ++
L-mode discharges such as 140544 serve as the initial operating space. This proposed experiment can largely be performed in piggyback with ROF #1 mentioned above. The unique discharges required include a magnetic field scan that will move the 8th ion cyclotron harmonic across the FW strap. This provides simple and direct confirmation that a parametric decay process is responsible. | ||
| Background: | Experiments in 2010 produced the surprising result that scrape-off layer (SOL) ions accelerated at the 180-degree fast antenna were measured by the fast ion loss detector at 225R-1. Analysis showed that these ions remained in the SOL at all times, and they reached a maximum energy (22 keV) equivalent to the maximum total voltage across the straps (22 kV). | ||
| Resource Requirements: | Fast wave system with P_rf ~> 3 MW | ||
| Diagnostic Requirements: | - FW antenna Langmuir probes (spectral analyzer to identify parametric decay spectra)
- SOL probes to obtain plasma density and temperature - fast ion loss detectors | ||
| Analysis Requirements: | - constructions of plasma profiles across SOL | ||
| Other Requirements: | -- | ||
| Title | 317: Validation of RF Wave Codes through Fast Wave Damping on Beam Ions | ||
| Name: | Pace | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Energetic Particles Group, Fast Wave Team | ITPA Joint Experiment : | No |
| Description: | Codes that solve for the propagation of injected ICRF waves through tokamak plasmas are capable of producing advanced, non-axisymmetric profiles of the wave's electromagnetic fields, but it is very difficult (or even impossible) to validate these results directly. Most diagnostics measure the effectiveness of the injected waves by observing the response of the tail ion distribution. A synthetic diagnostic used for validation purposes therefore couples both the wave solving code (e.g., AORSA) and a particle response code (e.g., CQL3D). This complicates the identification of the source of any inconsistencies between measured and synthetic results. <br><br>The idea proposed here is to use the damping of fast waves on neutral beam ions to remove some of the sources of error in this type of work. The initial distribution due to the NBI is well known through modeling by NUBEAM. This means that the fast wave damping is studied over a narrower parameter space as it increases the energy of the highest energy phase space of the beam distribution (i.e., the delta-change in the ion distribution is much smaller and less error prone compared to cases in which the ions are heated out of the bulk plasma). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | A validation effort requires coverage across a wide parameter space. This is achieved by iterating through a the set of simultaneous beams and fast wave antenna that inject (including changing the timing between initial NBI in the presence of ongoing FW injection). | ||
| Background: | -- | ||
| Resource Requirements: | Simultaneous NBI and FW, with the ability to alternate sources on each shot. | ||
| Diagnostic Requirements: | Fast ion diagnostic set (FIDA, FILD, etc.). | ||
| Analysis Requirements: | Significant time devoted to running RF and particle codes such as AORSA, ORBIT-RF, and CQL3D. | ||
| Other Requirements: | -- | ||
| Title | 318: L-H transition dependence on asymmetry in divertor detachment | ||
| Name: | Leonard | Affiliation: | GA |
| Research Area: | L-H Transition | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure the H-mode power threshold as a function of density and the resulting inboard divertor detachment and SOL flow. Test the conjecture that the increase in power threshold at lower density is due to inboard divertor reattachment and a resulting drop in SOL flow. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Measure the H-mode power threshold as a function of density in an USN plasma with the Grad B drift towards the upper divertor. Monitor the SOL flow with the X-point Mach probe and upper inboard divertor detachment with the upper fixed languir probes, filterscopes, divertor spectrometer and divertor tangential cameras. Correlate the increase in threshold at low density with changes to the upper inboard divertor plasma and resulting SOL flow. | ||
| Background: | The H-mode power threshold is observed to increase at lower density. This trend is important for designing ITER operational scenarios and for determining the mix of power required for ITER to access H-mode. This experiment tests the conjecture that SOL flow from the outboard divertor towards the inboard side aids the H-mode transition. Also conjectured is this SOL flow is enabled by a detached inboard divertor plasma that allows easy access to the core plasma for neutrals born in the inboard divertor. The final part of this conjecture is that at low density the inboard divertor reattaches shutting off the SOL flow and thus inhibiting the H-mode transition. If this conjecture is born out, then the optimal density for ITER to achieve H-mode can be determined by accurate modeling of the ITERâ??s inboard divertor detachment. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Upper divertor langmuir probes, upper tangential camera, midplane and X-point insertable Mach probes | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 319: physics of nonlocality | ||
| Name: | SUN | Affiliation: | National Fusion Research Institute |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | P.H.Diamond,G.McKee, B.Tobias | ITPA Joint Experiment : | Yes |
| Description: | The purpose of this experiment is to determine what physics underlies the so-called â??non-localityâ?? phenomonen. By â??non-localityâ??, we mean that a rapid response in the core is observed to follow from an edge perturbation on a time scale far shorter than any standard approximation to a global or diffusion model confinement time.
On HL-2A, some preliminary results of density fluctuation show that the suppression of certain frequency fluctuation plays a role in the nonlocal phenomenon. There is only core flucation measurement on HL-2A. With better performance and advanced fluctuation diagnostics, more detailed information can be easily obtained on DIII-D for the study of this topic.And also preliminary simulation results by CRONOS show that EXB shear may attribute to the formation of this "transient ITB"(nonlocality), detailed information of fluctuation is urgent to do furthur study. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We would like to perform all these experiments in Ohmic heating and low density plasma. It could be the start shots in the morning or the ending shots in the afternoon. In our first approach, pellet injection will be used to trigger nonlocal transport response in low density plasma (ï½?0.1-0.2Ã?1020m-3). The key diagnostic system for the experiment will be ECE Te measurement to follow the electron temperature evolution. full profile of density to follow the density change during nonlocalty. All available fluctuation diagnositcs will be main tools to study to physics behind the effect, like the turbulence spreading. | ||
| Background: | The understanding of electron heat transport in magnetically confined plasmas is an important issue for achieving a good predictive capability for ITER, where the electron heating is dominant as a result of the interaction between electrons and alpha particles as a fusion reaction product. However, a number of dramatic observations involving transient behaviour pose serious challenges to the understanding of heat transport in tokamaks. Among them, a widely observed phenomena â??nonlocalityâ?? has puzzled scientists for more than 15 years. Phenomena suggesting fast, apparently â??non-localâ?? responses of core plasma parameters to edge perturbations have been observed in many tokamak experiments. The observation of an unusual fast heat pulse generated by the Lâ??H transition in JET was the first clear non-local transport response. There are two unusual features of this heat pulse: a very fast propagation (within 1 msec) of the temperature rise and the observation that heat pulse amplitude doesnâ??t decay in the core. Then later, the â??promptâ?? responses to ECRH have been observed in W7-AS. In these cases, the core electron temperature responses were of the same polarity as the changes in edge electron temperature. The most striking evidence for non-locality is from cold pulses experiments, in which a transient core Te rise is observed in response to peripheral cooling. This phenomenon is charactered by a simple but challenging picture: a strong cooling in the edge plasma induces significant heating in the central plasma on a timescale much shorter than a diffusive propagation time scale. This opposite response of the plasma core occurs before the edge cooling pulse reaches the central region of the plasma. Since this extremely fast response is a formidable challenge to standard transport models, the development of an understanding of the non-local effect could lead to significant new directions in research on anomalous transport. | ||
| Resource Requirements: | Pellet injection | ||
| Diagnostic Requirements: | ECE, reflectometer, CECE, mm backscattering, alll fluctuation diagnostics. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 320: Main Chamber and Divertor ELM Characterization for BOUT++ Validation | ||
| Name: | Fenstermacher | Affiliation: | LLNL |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | C.J. Lasnier, S.L. Allen, X. Xu, I. Joseph | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment would be to fully characterize the 3D structure of ELM filaments with all available edge diagnostics, including in particular the simultaneous main chamber and divertor IR and visible emission imaging from the new LLNL periscope, and use that comprehensive dataset to validate new non-linear ELM simulations in realistic experimental collisionality now available with the BOUT++ code. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish LSN H-mode plasma with all pedestal and SOL aspects optimized for ELM characterization (vis. optimize outer gap, ZUPERTS, divertor strike-points etc. etc for diagnostics), all fast pedestal/SOL/divertor optimized for ELM dynamics and new LLNL periscope set for main chamber ELM filament measurements. After fully characterizing base case the most important scans to test BOUT++ simulations are density (collisionality), Te at constant collisionality (to test FLR and ion diamagnetic drift stabilization effects) and q95 (to vary resonant magnetic surface spacing). | ||
| Background: | Recent improvements to the BOUT++ code now allow fully non-linear simulations of ELM growth from the linear instability phase through the full non-linear crash to be done for realistic, experimentally achievable pedestal collisionality. Previously the non-linear simulations at experimental collisionality could not be completed due to the development of thin sheet currents (electron Larmor radius scale) that would develop in the simulations. Simulations at artificially elevated collisionality predicted larger ELM energy loss than seen in experiments. A hyper-resistivity model added to BOUT++ produces a physically reasonable mechanism to prevent the thin sheet currents and initial simulations suggest ELM energy loss much closer to experimental ranges. Validation of the model now requires a comprehensive dataset including the new capability to measure the ELM structure throughout the main chamber with the LLNL periscope. | ||
| Resource Requirements: | 1.0 day, LSN H-mode plasma with co-Ip NBI (probably moderate power to keep ELM freq low and size large). | ||
| Diagnostic Requirements: | LLNL IR + Visible periscope for main chamber ELM filament measurements and all fast pedestal/SOL/divertor diagnostics for measuring ELM dynamics.
Upgraded edge MSE (with co+counter beams for Er correction) and Li-beam (if available) both for edge bootstrap current measurement. | ||
| Analysis Requirements: | Possibly BOUT++ ELM simulations in advance to predict the spatial structure of ELM filaments for the experimental conditions proposed. | ||
| Other Requirements: | -- | ||
| Title | 321: Joint CMOD / DIII-D QH-mode for 2013 JRT | ||
| Name: | Fenstermacher | Affiliation: | LLNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, A. Garofalo, D. Whyte | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment would be the DIII-D contribution to a joint QH-mode experiment with CMOD as part of the 2013 Joint Research Target on stationary enhanced confinement regimes without large Edge Localized Modes (ELMs). Recent results from DIII-D suggest that QH-mode using wave heating alone (no input torque) might be possible with sufficient counter Ip torque from NRMF. The CMOD A-coils (midplane, external to the vacuum vessel) would be used to try to generate the needed NTV torque to achieve and study QH-mode in CMOD, and compare with similar wave heated QH-mode plasmas using the C-coils for NTV torque obtained in new DIII-D experiments. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish previous QH-mode plasmas with wave heating alone (developed in a separate set of DIII-D experiments for 2012) using NTV counter Ip torque from C-Coil NRMFs. Then modify the shape, q95 and collisionality to match QH-mode plasmas obtained in CMOD using wave heating and their A-coils. Fully diagnose the pedestal characteristics especially those of the density and thermal transport barriers, turbulence and particle transport, EHO or other edge modes and compare with CMOD results. | ||
| Background: | The Joint Research Target for 2013 will focus on understanding the physics mechanisms that can allow stationary enhanced confinement regimes without large Edge Localized Modes (ELMs). In particular the JRT13 seeks 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; in the case of this proposal the EHO that is important to QH-mode. The JRT proposes that 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 also be investigated. This experimental proposal addresses input to the JRT that could be obtained by joint DIII-D / CMOD QH-mode experiments. | ||
| Resource Requirements: | 0.5 - 1.0 day, Very clean wall conditions (possible recent boronization), counter IP, all NBI sources for torque scans, all available ECH and FW power for QH-mode with wave heating alone, maximum C-coil (n=3) and I-coil n=1 + n=3) currents for maximum NTV torque from NRMF components, DN patch panel biased down for high shaping capability. | ||
| Diagnostic Requirements: | All available pedestal/SOL/divertor diagnostics optimized for EHO and profile measurements, pedestal fluctuations and particle transport. | ||
| Analysis Requirements: | IPEC predictions of NTV torque possible with A-Coils in CMOD vs C-Coils in DIII-D. | ||
| Other Requirements: | -- | ||
| Title | 322: QH-mode Performance with NTV from C-coil NRMFs | ||
| Name: | Fenstermacher | Affiliation: | LLNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, A. Garofalo | ITPA Joint Experiment : | No |
| Description: | The goals of this set of experiments are to 1) determine the input co-torque limit for QH-mode operation with NTV torque provided by the C-coil alone and compare with theory (IPEC and other calculations), and 2) extend the G=betaN H98y2/(q95)2 =0.4 performance of QH mode with NTV from the C-coil to stationary pulses of many energy confinement times. These are both extensions of results already obtained in the 2011 campaign to performance that would be more convincing as a proposed alternate ELM control scenario for ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | For experiment 1, re-establish QH-mode shot 145066, including the torque ramp, with NTV from the C-coil alone. Maximize the current possible in th en=3 C-coil. Gradually increase the level of co-Ip torque at the end of the torque ramp and look for a jump in the toroidal rotation from negative NTV offset rotation to positive rotation indicating that the maximum co-input torque has been exceeded.
For experiment 2, re-establish QH-mode high performance shot 147351, including BT ramp and input torque ramp to achieve G=0.4 conditions. Apply various locked mode prevention techniques (eg. ECCD at the 2/1 surface, vary the BT ramp rate etc.) to try to prevent the core rotation from continuing to decrease and thereby avoid the locked mode and extend the stationary G-0.4 performance of this shot to long duration. | ||
| Background: | High normalized performance (G=betaN H98y2/(q95)2 = 0.4 )QH-mode was established in the 2011 campaign (shot 147351). The high performance phase was stationary (after BT and input torque ramps) for on the order of 100 ms before the shot terminated due to a locked mode. At the time of the locked mode it appeared that the core rotation was continuing to decrease. To strengthen the argument that QH-mode is capable of the normalized performance needed in ITER, we need to show a stationary G=0.4 condition for several energy confinement times. | ||
| Resource Requirements: | 1.0-1.5 days, Reversed Ip, all co and counter NBI, maximum current possible for C-coil (n=3) and I-coil (n=1 EFC and n=3 combined), extremely good wall conditions for QH-mode, ECCD available for core locked mode avoidance | ||
| Diagnostic Requirements: | All pedestal, SOL and divertor diagnostics, core rotation, ECEI in the edge | ||
| Analysis Requirements: | IPEC and other NTV torque analysis and predictions of the toroidal rotation as a function of the input torque to compare with measurements. | ||
| Other Requirements: | -- | ||
| Title | 323: QH-mode in Wave Heated Plasmas | ||
| Name: | Fenstermacher | Affiliation: | LLNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, A. Garofalo | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to establish QH-mode operation in plasmas heated by wave alone, without any fast ion injection. The edge rotation shear needed for QH-mode will be provided by NTV torque from NRMFs from n=3 C-coil and I-coil fields. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Re-establish the QH-mode conditions from the shot series 145094, 098, 109, 117 using a torque ramp with co plus counter NBI injection. At the end of the torque ramp at near zero input torque, gradually replace NBI power with ECH power to maximum available. Then reduce remaining NBI power to zero while maintaining zero input torque. | ||
| Background: | Experiments in the 2011 campaign firmly established QH-mode operation at near zero input torque using torque ramps to balanced NBI injection and counter NTV torque from I-coil and C-coil NRMFs. In principle then QH-mode should be possible in wave heated plasmas using the same NTV torque techniques. This experiment would test this and determine if there is anything special about balanced NBI plasmas with substantial fast ion component that is needed for QH-mode operation. If successful it would also provide the basis for an attempt at QH-mode in CMOD wave heated plasmas with NTV torque from the CMOD midplane A-coils. | ||
| Resource Requirements: | 1.0 day, Reversed Ip, all co and counter NBI, maximum current possible for C-coil (n=3) and I-coil (n=1 EFC and n=3 combined), extremely good wall conditions for QH-mode, maximum ECH and FW powers to replace balanced NBI for wave heated QH-mode. | ||
| Diagnostic Requirements: | All pedestal, SOL and divertor diagnostics, core rotation, ECEI in the edge | ||
| Analysis Requirements: | IPEC and other NTV torque analysis and predictions of the toroidal rotation as a function of the input torque to compare with measurements. | ||
| Other Requirements: | -- | ||
| Title | 324: Joint CMOD / DIII-D I-mode for 2013 JRT | ||
| Name: | Fenstermacher | Affiliation: | LLNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | D.G. Whyte, A. Hubbard, R. Groebner, J. W. Hughes, E. Marmar, Tony Leonard, A. White, D. Pace | ITPA Joint Experiment : | Yes |
| Description: | The goal of this experiment would be the DIII-D contribution to a joint I-mode experiment with CMOD as part of the 2013 Joint Research Target on stationary enhanced confinement regimes without large Edge Localized Modes (ELMs). Recent results from DIII-D (145556) suggest that plasma operation with many of the signatures of the CMOD I-mode have been made in DIII-D. A MP for a half day exploration of I-mode in DIII-D was approved in 2011 but the experiment was not executed to lack of time. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Re-establish conditions of shot 145556 and document. Then scans of the important operational parameters that control access to I-mode namely heating power (eight 0.5 MW steps of 500ms each in each shot), target plasma density (shot-to-shot in increments of 1.e19 starting at 5.e19 both higher and lower) and q95/current (1.0, 1.25 and 1.5 MA) | ||
| Background: | The Joint Research Target for 2013 will focus on understanding the physics mechanisms that can allow stationary enhanced confinement regimes without large Edge Localized Modes (ELMs). In particular the JRT13 seeks 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; in the case of this proposal the Weakly Coherent Mode that is important to I-mode. The JRT proposes that 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 also be investigated. This experimental proposal addresses input to the JRT that could be obtained by joint DIII-D / CMOD I-mode experiments.
On C-Mod, (and consistent with the ASDEX-Upgrade experiments, the I-mode regime is defined by three observations [Whyte Nucl Fusion 2010], which are (in order of importance) 1. The formation of a temperature pedestal in the absence of any strong densitypedestal or particle transport barrier. 2. A change in the fluctuation characteristics of the pedestal region including a reduction in broadband fluctuations and/or the appearance of a weakly-coherent highfrequency edge fluctuation. 3. The absence of standard L-H transitions such as a sudden drop in D-alpha or uncontrolled increases in density and/or core radiation. A candidate I-mode has been identified on DIII-D based on these criteria in a recent shot #145556, which occurred during an experiment to produced ??Opaque SOL? (Figs. 1- 3). This shot followed the standard ??recipe? to date for achieving I-mode: single-null shape with ion grad-B drift pointed away from the primary X-point (Fig. 1) and strong central heating (~7 MW, Fig. 2) at levels many times the standard power to achieve Hmode in with ion grad-B towards the X-point. C-Mod experiments have also indicated that shaping, particularly triangularity, is important to I-mode access and performance; divertor topology (closed better than open) may also be important. | ||
| Resource Requirements: | 0.5-1.0 day, Single-nulls, LSN Plan I (#145556) USN Plan II (#145174), Beams At least 4, including 30 left. Cryopumps Plan I (lower pump used) Plan II (Both Upper pump) | ||
| Diagnostic Requirements: | All available pedestal/SOL/divertor diagnostics optimized for edge MHD modes, fluctuations, profile measurements, and particle transport. | ||
| Analysis Requirements: | With successful documentation of I-mode in DIII-D, CMOD would then attempt to match the DIII-D conditions as much as possible in follow-on experiments. We would also propose follow-on experiments in DIII-D from the pool of Director??s Reserve time. | ||
| Other Requirements: | -- | ||
| Title | 325: Investigate Disagreements Between Thomson Scattering and ECE Measurements in High Te Discharges | ||
| Name: | White | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | M. E. Austin, R. Prater, P. Bonoli, R. Harvey, S. Scott, B. Bray, C.C. Petty, R. Pinsker, D. C. Pace | ITPA Joint Experiment : | Yes |
| Description: | These very high temperature plasmas (Te > 10 keV), when heated with NBI and FW, provide the opportunity to study the Parametric Decay Instability (PDI) with the FILD diagnostics. See D. C. Pace (NF, submitted -- see also science meeting talk: Meeting, August 19, 2011).
This experiment is a great combined experiment of H+CD group and Energetic Particles group. The goal of this experiment is to search for a discrepancy between Thomson scattering (TS) and ECE measurements of Te on DIII-D in discharges with high electron temperature. To carry out the experiment, L-mode discharges are used to attain central electron temperatures of Te(0) = 9 keV with NBI + FW and Te (0) = 15 keV with NBI + ECH (achieved in 2009 in MP 2010-55-01). This year, we will be targeting high power, off-axis ECH discharges to reach Te(0) > 12 keV, while maintaining a Maxwellian f(v) in the core plasma. Electron temperature measurements from TS and the absolutely calibrated Michelson interferometer would be compared to look for any discrepancy between the two diagnostics. With the 2009 data, a gate timing issue with TS was discovered several weeks after the run. This limited the time resolution/available statistics for the experiment (Bray, Friday Science Meeting, May 7th , 2010) and is a main reason the experiment must be repeated. A second motivation for this new experiment this year is that based on using NBI on DIII-D to investigate a specific hypothesis for the cause of the discrepancy (R. Harvey): cold electrons deposited from the NBI may flatten the electron f(v) in the correct vicinity to produce the discrepancy (per modeling by de la Luna and Krivenski) |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The hypothesis is that cold electrons deposited from the NBI may flatten the electron f(v) in the correct vicinity to produce the discrepancy (per modeling by de la Luna and Krivenski). Indeed, it is well known that after the NBI is shut off and the influx of cold electrons (cold is born at the ionization energy, 10s of eV) ends, Te increases over several confinement times. This is seen in DIII-D (moderate rises of 100s of eV) and was quite large in TFTR supershots (1-2 keV). At DIII-D, with combination of NBI and ECH, it will be possible to make high Te plasmas with phases where the beams are on and where beams are off. TS and ECE will be compared in both cases. If the cold electrons are responsible for the discrepancy, the discrepancy should only be seen when beams are on.
Reference shot 140715 (beams and FW to high Te(0) > 9 keV). Obtain high quality TS and ECE data for detailed comparisons and modeling. Apply high power ECH near r=0.25 or 0.3 starting when the current ramp is nearly complete, generate plasmas with weak negative shear for good confinement but avoid the very strong eITB that comes with strong negative central shear. Prater??s recipe for using ECH in these experiments: Leave the plasma center free of ECH to maintain a Maxwellian distribution there and use large enough k_parallel that little EC power is deposited via relativistic downshift; split the power equally into positive k_parallel and negative k_parallel to avoid driving excessive local current . In this experiment, we can look with the beams on and then turn the beams off during the highest Te phase of discharge to get the beam off data. | ||
| Background: | One hypothesis from JET was that ICRH-generated fast ions may be related to the cause of the discrepancy, because the discrepancy was seen above 5 keV in NBI+ICRH plasmas but not NBI only plasmas [de la Luna 2008]. However, recent experiments at C-Mod produced high Te plasmas with ICRH (fast ions present) and Mode Conversion heating (no fast ions) and there was no evidence of the discrepancy for Te(0) < 8 keV in either case [White, et al. to be submitted Nuclear Fusion]. This would indicate that fast ions may not responsible for the discrepancy.
In the new experiment at DIII-D we will explore a a different hypothesis related to NBI cold electron influx that may help explain the TFTR discrepancy in particular, where the discrepancy was seen in NBI plasmas [Taylor, 2009] Note that in the core of typical tokamak plasmas with Te(0) < 7 keV the TS and ECE measurements of electron temperature are in very good agreement. Also note that TS and ECE measurements of electron temperature often disagree in high Te(0) discharges that are strongly heated via ECH, but in these cases the disagreement can be explained by a well understood perturbation of the electron energy distribution function caused by the ECH [3]. In contrast to these cases, the cause of the TS/ECE discrepancy in discharges heated with only NBI or NBI + ICRH on JET and TFTR where Te(0) > 5 keV is not known. Such a discrepancy has been observed in NBI discharges and discharges heated with both NBI and Ion Cyclotron Resonance Heating (ICRH) discharges in TFTR and JET [1,2]. In these cases, the central electron temperature Te(0) measured with ECE diagnostics is 10-20% higher than the TS measurement of Te(0) The discrepancy starts at Te(0) ~ 7 keV and increases approximately linearly with electron temperature. Theoretically, a non-Maxwellian electron distribution f(v) with distortion near the thermal velocity may create such a measurement discrepancy between TS and ECE measurements [4], however, no known mechanism can sustain that type of distribution with finite heating power As an ITPA joint experiment, either a positive or a negative result on this topic from DIII-D can significantly impact international efforts to understand the past discrepancies that have been reported on TFTR and JET [1,2]. For example, a positive result, the observation of the discrepancy between TS and ECE on DIII-D, would verify the discrepancy on an additional machine and would therefore motivate a new and detailed investigation of the phenomenon. However, a negative result, the observation of no discrepancy under a variety of conditions with high electron temperature produced with NBI and FWH, would be equally beneficial as it would show that agreement between TS and ECE can be obtained in high temperature tokamak discharges. In both cases, the experimental results and associated modeling will improve the understanding of heating and diagnostic techniques in plasmas relevant for ITER and reactors. | ||
| Resource Requirements: | All gyrotrons
FW heating systems All available NB sources | ||
| Diagnostic Requirements: | Thomson scattering, Michelson interferometer, 40-channel ECE radiometer, ECEI, CER and MSE, fast magnetics, all fast ion diagnostics, all available fluctuation diagnostics. If available, oblique ECE. | ||
| Analysis Requirements: | EFIT, gaprofiles, ECESIM, ONETWO/autoonetwo, GENRAY, TORAY,and CQL3D | ||
| Other Requirements: | (references are here)
[1] E. de la Luna, et al., Rev. Sci. Instrum. 74, 1414 (2003) [2] G. Taylor, PPPL report 4202 (2006) [3] C. C. Petty et al. GA Report A25804 (2007) [4] V. Krivenski et al. 29th EPS Conference on Plasma Phys. and Contr. Fusion Montreux, 17-21 June 2002 ECA Vol. 26B, O-1.03 (2002) | ||
| Title | 326: Study of H-mode pedestal penetration and transport | ||
| Name: | Osborne | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | R.J. Groebner, J. Canik, P.B. Snyder, A. Leonard | ITPA Joint Experiment : | No |
| Description: | Study the physics of pedestal penetration, and transport in the pedestal steep gradient region. Experiment will try to identify the instability at the H-mode barrier propagating front and the mechanism for the suppression of this instability. Experiment will study the development of turbulence in the steep gradient region of the pedestal as the pressure gradients rise. In this regard attempts will be made to produce discharges which are not KBM limited, and also look for KBM turn on in the iter-ELM period. Understanding the physics of H-mode barrier propagation might allow control of the penetration for ELM avoidance through other means than the RMP. Experiment will also give another test of PB mode ELM trigger physics; if some other ingredient was needed to actually trigger the ELM at the PB limit this might offer another pathway to ELM control. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Produce very low power NBI and/or ECH heated H-mode discharges with very slow evolution of profiles between ELMs. Probably select for good ballooning stability (high shaping) for KBM control. All fluctuation diagnostics to look for turbulence turn on and try to follow the turbulence suppression front propagation. Low power operation/ECH may allow plunging probe across the separatrix for local turbulence transport measurements. The experiment will provide turbulence measurements, detailed profile evolution data, and detailed divertor data for comparison with turbulence simulations and transport modelling. | ||
| Background: | Previous analysis of low heating power H-mode dicharges (T.Osborne HMWS 2011) has demonstrated pedestal gradients more consistent with paleoclassical than KBM. Clearly demonstrating that this is the case with data from fluctuation diagnostics and larger differences in expected gradients could support both theories. In particular it might be possible to more clearly demonstrate the KBM turn on as the critical gradient was approached in the inter-ELM period. Many years ago it was shown from reflectometry measurements the the fluctuation suppression zone near the H-mode transition corresponded to the region of high ExB velocity shear. Although ExB suppression of ion scale turbulence is also suggested by the barrier propagation in VH-mode, a number of other possibilities exist: drift reversal suppression of TEMs, or density gradient (etae) penetration suppression of TEM, ETG, micro-tearing, might also be important. Advances in fluctuation diagnostics could provide more insight the instability involved in barrier propagation as well as to what is controlling the gradient in the transport barrier region. Results on ASDEX-U and previous current density profile evolution simulation results on DIII-D have indicated pedestal current density profile saturates before the ELM onset in contrast to what might be expected for PB ELM trigger suggesting another ingredient might be required. | ||
| Resource Requirements: | Low recycling conditions. At least two days separated by an analysis period. | ||
| Diagnostic Requirements: | High accuracy TS calibration. All fluctuation diagnostics, plunging probes. Lithium beam for current density profile and/or density profile measurements. | ||
| Analysis Requirements: | Standard profile, kinetic EFIT, ELITE, KBM. Transport simuliation of curent density profile evolution. SOLPS modeling for particle source and transport. Comparison of fluctuations and profile evolution with turbulence codes. | ||
| Other Requirements: | |||
| Title | 327: Energy Transport in the DIII-D L-mode Near Edge | ||
| Name: | Kinsey | Affiliation: | CompX |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | Petty, Luce, Burrell, McKee, Waltz, Staebler, Holland, Doyle, Rhodes | ITPA Joint Experiment : | No |
| Description: | Dedicated experiments are needed to specifically study the energy transport in the DIII-D L-mode near edge region. TGLF and GYRO both show clear evidence of the energy transport being undepredicted beyond rho=0.7 in DIII-D L-modes. The origin of the underprediction remains unknown. Without a clear understanding of the transport in the L-mode edge, we are significantly in our ability to predict H-mode pedestal formation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Perform safety factor and nustar scans holding all other dimensionless parameters fixed while obtaining fluctuation data. Any deviation from the expected ITG/TEM scaling with safety factor and nuei may suggest a different transport mechanism is present. | ||
| Background: | Previous L-mode dimensionless similarity focused on the core region and did not have fluctuation measurements. Onetwo analyses suggest the ion and electron energy transport may scale differently in the near edge region. | ||
| Resource Requirements: | DIII-D L-modes with NBI and possibly ECH. | ||
| Diagnostic Requirements: | edge profile measurements, core profiles, CER, MSE, BES | ||
| Analysis Requirements: | ONETWO, TRANSP, XPTOR, GYRO | ||
| Other Requirements: | |||
| Title | 328: Exploration and Characterization of I-mode | ||
| Name: | Whyte | Affiliation: | Massachusetts Institute of Technology |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | A. Hubbard, M. Fenstermacher, A. Leonard, A. White | ITPA Joint Experiment : | Yes |
| Description: | Search for the stationary "I-mode" transport regime on DIII-D to determine parameters necessary access to I-mode, avoid H-mode and characterize the pedestal conditions. These results will be compared with the C-Mod I-mode results [Whyte, et al. Nucl Fusion (2010)]. I-mode is possibly a highly favorable regime: it features a stationary edge temperature barrier without an edge particle barrier. This leads to stationary discharges with high T pedestal, H-mode energy confinement, and no core impurity accumulation (L-mode particle confinement) without the requirement for ELMs. Almost all I-mode discharge on C-Mod are stationary and ELM-free.
In addition to its attractiveness as an ELM-free regime, I-mode has important scientific contributions to make in transport studies since the regime clearly separates the energy and particle/impurity channels from one another. C-Mod reports correlations between I-mode and changes in edge fluctuation changes including a weakly coherent EM mode existing in the pedestal region. It is speculated that this mode is responsible for enhanced particle transport in the edge. Detection of this mode and comparison to C-Mod results, as well as the EHO, is desirable. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The experimental approach is to use the operational and physics insights gained from C-Mod. In addition the 2011 campaign may have accidentally triggered an I-mode on DIII-D so this may be a good starting point. The I-mode regime is most readily observed with unfavorable grad-B drift topology and medium to high triangularity. As q95 is decreased towards ~3 sudden transitions are found from L to I-mode, with a rapid increase of edge T (within < sawtooth pulse). In addition the required power to access I-mode in this condition increases with Ip (contrary to the LH threshold in favorable grad-B). This tends to lead to high absolute performance at ITER's q95~3 but without ELMs: a very favorable result for ITER.
We will use the identification parameters as defined by C-Mod 1) Formation of a T pedestal without a n pedestal 2) Observation of reduced broadband edge fluctuations and/or presence of a high frequency (>100 kHz) weakly coherent EM mode 3) No standard H-mode transition signatures of D-alpha drop and/or increasing density Based on the C-Mod experience we will explore for I-mode by scanning within a shot the heating power in small steps, and then change other global parameters of interest on a shot to shot basis namely: Ip, Bt (or q95) and density. A clear result from C-Mod was that the I and H power threshold has different dependences in Ip, Bt (or q95) than favorable direction. Namely the power threshold clearly increases with Ip. Also the threshold increased with lower B, the opposite trend of standard LH scaling. Shaping also played a role in avoiding H-mode access, e.g. x-point spacing, triangularity. In a way this is not news for D3D which used a high delta, upper null shape to avoid H-mode transitions in high fusion gain experiments of 90's (Lazarus et al) Another key feature in accessing I-mode is density. C-Mod found optimal density windows using cryopumping. There is an intriguing possibility that the heating gap between I and H could be made much larger by going below the "low-density" LH threshold. This has not been possible yet on C-Mod due to L-mode impurity limitations with ICRF. However it may be possible to explore this limit with NBI heating. | ||
| Background: | Excerpts from abstract of C-Mod paper on I-mode (Whyte, et al. Nucl. Fusion 2010)
"An improved energy confinement regime, I-mode is studied in Alcator C-Mod, a compact high-field divertor tokamak using Ion Cyclotron Range of Frequencies (ICRF) auxiliary heating. I-mode features an edge energy transport barrier without an accompanying particle barrier, leading to several performance benefits. H-mode energy confinement is obtained without core impurity accumulation, resulting in reduced impurity radiation with a high-Z metal wall and ICRF heating. I-mode has a stationary temperature pedestal with Edge Localized Modes (ELMs) typically absent, while plasma density is controlled using divertor cryopumping. I-mode is a confinement regime that appears distinct from both L-mode and H-mode, combining the most favorable elements of both. The I-mode regime is investigated predominately with ion grad-B drift away from the active X-point. The transition from L-mode to I-mode is primarily identified by the formation of a high temperature edge pedestal, while the edge density profile remains nearly identical to L-mode. Laser blowoff injection shows that I-mode core impurity confinement times are nearly identical with those in L-mode, despite the enhanced energy confinement. In addition a weakly coherent edge MHD mode is apparent at high frequency ~ 100-300 kHz which appears to increase particle transport in the edge. The I-mode regime has been obtained over a wide parameter space (BT=3-6 T, Ip=0.7-1.3 MA, q95=2.5-5). In general the I-mode exhibits the strongest edge T pedestal and normalized energy confinement (H98>1) at low q95 (<3.5) and high heating power (Pheat > 4 MW). I-mode significantly expands the operational space of ELM-free, stationary pedestals in C-Mod to Tped~1 keV and low collisionality ν*ped~0.1, as compared to EDA H-mode with Tped< 0.6 keV, ν*ped>1. There is a good chance that plasmas that are "I-mode-like" have been previously obtained on DIII-D. It seems reasonable to explore this further. Based on the Cmod results, the "new" parts of these studies for DIII-D is that a) The T pedestal can have a sudden, clear bifurcation from L-mode rather than just being a continuous slow modification from L-mode profiles b) The confinement mode can be made stationary, rather than intermediate of L to H. c) The sudden changes in edge fluctuations associated with the T pedestal without a density pedestal have been identified. | ||
| Resource Requirements: | NBI heating
Cryopumping Unfavorable grad-B drift topology | ||
| Diagnostic Requirements: | Full suite of pedestal diagnostics
Full suite of edge fluctuation diagnostics - BES - reflectometry - magnetics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 329: High Performance I-mode Dicharges | ||
| Name: | Osborne | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | J. Boedo, L. Schmitz | ITPA Joint Experiment : | No |
| Description: | Experiment will explore the possibility of I-mode discharges at high beta and H factor. Experiment could also provide more understanding of the physics of the I-mode and KBM. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Search for an I-mode regime at higher heating power using guidance from low density Type III ELM results. Obtain fluctuation measurements. | ||
| Background: | Recent results on DIII-D (Boedo,Schmitz) have demonstrated an ELM free regime at low heating power and low density similar to CMOD I-mode. The regime also has similar characteristics to the low density Type III ELM regime studied in the mid-1990s (Osborne). 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. The pedestal pressure gradient is generally low in both the Type III and recent I-mode regime compared to the Type I regime but the pedestal width is greater. During the type III studies however 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). 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. If the recent I-mode discharges are indeed ELM free examples of the old Type III results these scalings could act as a guide to obtaining high performance ELM free I-mode discharges. The fact that the pedestal widths are greater in the Type III/I-mode regimes suggests the gradients are not limited by KBM. | ||
| Resource Requirements: | 1 day experiment to test the possibility for good performance I-mode discharges | ||
| Diagnostic Requirements: | Profile and fluctuation diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 330: Low Torque Advanced Inductive Plasma with QH-mode Edge | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | ITER 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. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | There are at least 3 approaches that should be considered for this experiment. The first involves starting 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 exasperate this, so EC will likely be needed. This, however tends to be unfavorable for QH-mode operation. Still it may be possible to remove the EC once the NRMF is turned on. A second option is to start with a standard co-Ip hybrid and again work on the early density (early NTMs should be avoidable with this approach). In fact, #117738 may be a suitable starting point, which is a co-NBI hybrid with a brief QH-phase. Then ramp down the torque (QH-mode will certainly be lost), and when near balanced, apply NRMF and EC if necessary. Finally, a third option is to start with an existing low counter-NBI torque QH-mode and modify the early heating etc to be more consistent with AI operation. | ||
| 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 | 331: Test of paleoclassical transport in ELM-free H-mode | ||
| Name: | Groebner | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | S. Smith, T. Osborne | ITPA Joint Experiment : | No |
| Description: | Attempt to produce pedestal conditions in which experimental pedestal electron thermal diffusivity is well below the prediction of paleoclassical theory. Want a condition where fluctuation-driven electron thermal transport is most likely to be absent and the early ELM-free phase of a slowly evolving H-mode might be the best chance to do this. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use shape with good stability against ELMs. Apply heating power only slightly above H-mode threshold. These conditions should give long ELM-free period and slow evolution of pedestal. Measure Te and ne profiles in pedestal with good resolution. Monitor low and medium-k fluctuations and ensure that there is a ~100-300 ms period early in H-mode during which fluctuations are very low or absent. | ||
| Background: | Analysis by S. Smith shows that the paleoclassical transport model predicts gradients of the pedestal electron temperature that are generally greater than or equal to the measured values. This result is consistent with paleoclassical transport providing the minimum level of observed electron thermal transport in the pedestal. Most of the analyzed data were obtained in the last 20% of the ELM cycle for the discharges in the study. During this part of the ELM cycle, density fluctuations are normally observed in the pedestal, and it is plausible that they contribute to some transport. Early in the ELM-free H-mode (period after the L-H transition), it is often observed that low-k fluctuations are very low or absent in the pedestal. Thus, this time period is expected to have the lowest level of fluctuation-driven transport in the pedestal of any period in the H-mode and should be an ideal time to test the paleoclassical model. If paleoclassical physics is not the correct physics, this time period might provide the best opportunity to observe chi_e values well below paleoclassical predictions. Due to the fact that the pedestal usually builds up at a significant rate during this time, it may be necessary to perform time-dependent transport analysis on the data to obtain the underlying chi_e values for comparison with the paleoclassical predictions. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | High resolution edge TS and CER. Fluctuation measurements at low and medium k. | ||
| Analysis Requirements: | Profile and transport analysis to extract time-dependent pedestal chi_e values. Predictions of the chi_e values from paleoclassical model. | ||
| Other Requirements: | |||
| Title | 332: ELM Suppression at Lower Collisionality with Non-Resonant n=3 I-coil | ||
| Name: | Osborne | 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. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | 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. | ||
| Background: | 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. | ||
| Resource Requirements: | 1 Day experiment. I-coil in odd parity. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 333: Study importance of "plume" effect in carbon density measurements | ||
| Name: | Groebner | Affiliation: | GA |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, J. Munoz, B. Grierson | ITPA Joint Experiment : | No |
| Description: | Correlate CER signals with different beam sources. Determine if we have evidence of a significant "plume" effect (defined in background) producing a larger inferred carbon density from tangential views than from vertical views. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Choose a discharge which has shown a significant difference (50% or more) in carbon densities as computed from the vertical vs the tangential CER chords. Individually pulse on each neutral beam source (or beamline) and monitor the response of different CER chords. Need to look for evidence that the tangential chords are seeing signal that is correlated with beams that are not directly observed. Such a signal would be a signature of the plume effect. Compute carbon densities for all chords and determine if the tang vs vert densities can be reconciled with the results from the beam pulsing. | ||
| Background: | Analysis to generate carbon impurities from CER data obtained for the C VI 5290 line generally show higher densities computed from tangential views than from vertical views. These differences are often as large as 50% and have been as high as a factor of two. Our usual interpretation has been that the tangential chords are seeing a significant signal due to the "plume" effect and that the densities from the vertical chords are the most reliable. The "plume" effect arises from the fact that ions, which undergo charge exchange recombination with a neutral beam, will drift some distance along field lines before being re-ionized to the next charge state. During this time, there is a finite probability that the ions can be excited by electron impact excitation to produce the line used for CER measurements (5290 for C VI). However, if these ions have a sufficiently long life-time, they could have drifted into the sight of chords which do not directly observe the beam that caused the CER reaction. Nevertheless, these chords could see signal from these ions that was not produced by the CER process, as assumed in impurity density computations. This process is a problem for the measurement because the ions can be distributed over a range of radii and because they inflate the impurity density computed along the observing chord. Due to the mostly toroidal drift of these ions, the plume effect is a much bigger problem for tangentially-viewing chords than for vertically viewing chords. However, recent work casts some doubt on this explanation and raises other explanations for the differences in densities from the two sets of chords. Due to the fact that we do not have a separate measurement of carbon density, it is difficult to empirically determine if one set of views is more reliable than the other. The experiment discussed here should be able to determine if the plume effect is an important systematic issue, at least for the discharges studied. | ||
| Resource Requirements: | At least one neutral beam from each beamline. | ||
| Diagnostic Requirements: | CER, Thomson, CO2 | ||
| Analysis Requirements: | GAprofiles, IMPcon to compute impurity densities from CER chords | ||
| Other Requirements: | |||
| Title | 334: Driving Alfvenic Activity with Scanned Frequency Fast Wave Injection | ||
| Name: | Pace | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Energetic Particles Group | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to develop an operational scheme by which the fast wave system excites Alfven eigenmodes through a beat wave process. The successful realization of this goal will enable the study of Alfven eigenmodes, and their impact on energetic ion confinement, in DIII-D plasmas that are otherwise incapable of producing these modes (e.g., current flattop at normal magnetic field amplitude). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Two different operating modes are anticipated for the Fast Wave system.
(1) The fast wave 285-antenna will operate with an input waveform composed of two frequencies: A) 60.0 MHz B) 60.0 - f_IF [MHz] where f_IF is the intermediate frequency desired (i.e., an Alfven eigenmode frequency in the lab frame). The value of f_IF will be repeatedly scanned between 20 and 200 kHz during the entire discharge in order to identify any Alfven eigenmode drive as the resonance condition is reached. (2) The 180-antenna will attempt a similar operation centered on 90 MHz. This system is being upgraded in 2011 with an expected power increase of 30% to approach 1.3 MW injected. The initial attempts to drive Alfven eigenmodes (AEs) using beat wave excitation from fast wave injection will occur in plasmas of general interest to the Energetic Particles Working Group. Neutral beam injection during the earliest stages of the current ramp produce significant Alfvenic activity in typical DIII-D shapes (122117) as well as in oval plasmas (142111). Fast wave injection of beat waves in these plasmas will be used to increase the amplitudes of the already existing AEs. A successful result in this situation serves as a proof-of-principle for this process. Other discharges will focus on exciting AEs in plasmas for which they would not otherwise be observed, but in which they are thought to be only marginally stable or weakly damped. A series of discharges that feature properties conducive to Alfven eigenmode production will be identified, including reduced magnetic fields (B_T ~ 1.0 T) in which the neutral beam ions are super-Alfvenic. Should observations indicate that the fast wave system is driving AEs in a variety of situations described above, then the experiment will transition to discharges for which the modes would never be observed (standard magnetic field, delayed neutral beam injection, etc.). | ||
| Background: | Energetic ion transport due to Alfven eigenmodes is suspected as a mechanism that may reduce the fusion power of ITER [1]. Recent work at DIII-D has quantified the transport of beam ions [2,3] in experiments that utilize early neutral beam injection during the current ramp to excite AEs. A natural progression of this work requires the ability to study wave-particle interactions in steady state plasmas that better approximate reactor conditions in which super-Alfvenic fusion alphas drive the modes. This also enables DIII-D to study fundamental features of AEs, including damping rates that are investigated elsewhere through the use of active MHD antenna systems [4]. An advantage of the fast wave excitation method compared to active MHD antennas is that it is theoretically capable of driving large amplitude and core localized modes.
Excitation of toroidicity induced AEs through beat wave interactions from ICRH systems has previously been achieved on JET [5] and more recently in AUG [6]. In the AUG work it was possible to measure the radial displacement eigenfunction of the resulting mode, which was then compared to theoretical predictions, enabling the identification of the mode and confirmation of the ICRH drive. At DIII-D, the excitation of these modes in discharges for which they were previously unavailable will enable the study of many new physics phenomena of relevance to ITER, including AE induced transport and losses of off-axis injected neutral beam ions and the resulting degradation of off-axis neutral beam current drive. [1] A. Fasoli, et al., Nucl. Fusion 47, S264 (2007) [2] M.A. Van Zeeland, et al., Phys. Plasmas 18, 056114 (2011) [3] W.W. Heidbrink, et al., Phys. Rev. Lett. 99, 245002 (2007) [4] A. Fasoli, et al., Plasma Phys. Control. Fusion 52, 075015 (2010) [5] A. Fasoli, et al., Nucl. Fusion 36, 258 (1996) [6] K. Sassenberg, et al., Nucl. Fusion 50, 052003 (2010) | ||
| Resource Requirements: | - Fast Wave system at > 1 MW injected power per antenna
- NBI system | ||
| Diagnostic Requirements: | - synchronous detection electronics on Fast Wave systems
- all FILD units (midplane and R-1): measurement of any fast ion losses at frequencies coherent with the driven modes - ECE and ECEI: detection of AEs - all FIDA systems: measurements of fast ion distribution to identify any adjustments due to the presence of AE activity | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 335: Generation of unstable RWM to verify kinetic RWM stabilization physics | ||
| Name: | Sabbagh | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | J.W. Berkery, J.M. Hanson, et al. | ITPA Joint Experiment : | Yes |
| Description: | Specifically target the generation of an unstable resistive wall mode by altering the kinetic RWM stabilization properties of a high betaN target plasma. This specific goal and would significantly add to the verification of the kinetic RWM stabilization physics model initially demonstrated in NSTX and presented in 2008, and since supported by follow-up experiments in NSTX (that have generated unstable modes) and DIII-D (that have shown increased resonant field amplification (RFA)). | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use results from past DIII-D experiments to determine the least stable target plasma for RWM stability. The instability drive can be a combination of both pressure gradient and current â?? the goal being to minimize the ideal stability of the plasma. Once generated, the energetic particle population (EP) and plasma rotation profile would be altered to decrease the stabilizing deltaW_kinetic term to drive the RWM unstable. Off-axis NBI could be used as one tool to change the EP pressure profile. Non-resonant magnetic braking could be optionally used to further alter the rotation if near maximum NBI power would be needed to provide the mode instability drive for ideal deltaW. | ||
| Background: | Experiments on NSTX, with publication at the IAEA FEC 2008, showed a correlation of RWM destabilization at relative high plasma rotation that correlated with weakened RWM stability for the experimental rotation profile at the marginal RWM stability point. Further experiments have supported this theory and publications have been made for NSTX (e.g. Sabbagh, et al., IAEA FEC 2008 paper EX/5-1, Sabbagh, et al., Nucl. Fusion 50 (2010) 025020, Berkery, et al., Phys. Rev. Lett. 104 (2010) 035003, Berkery et al., Phys. Plasmas 17 (2010) 082504, etc. ), and also for more recent experiments on DIII-D (Reimerdes, et al., Phys. Rev. Lett. 106 (2011) 215002). The DIII-D experiment D3DMP No. 2011-14-01 (run in 2011) showed clear evidence of a substantial move toward reduced RWM stability (maximum RFA was generated), but the given run time was insufficient (time lost due to device technical issues) to bring the plasma past the RWM marginal stability point as planned. | ||
| Resource Requirements: | 8 NB Sources are required. Pellet fueling. I-coil and C-coil fields for n = 1 correction and potential n = 2 or 3 fields for non-resonant magnetic braking. Experiment should follow boronization. | ||
| Diagnostic Requirements: | Equilibrium diagnostics including kinetic profiles (Thomson, CER, MSE) and fast ion diagnostics, RWM sensors. | ||
| Analysis Requirements: | TRANSP/ONETWO calculations of fast ion pressure, Kinetic EFIT, MISK (MARS-K, and HAGIS as available). Analysis will contribute to the ITPA MHD Stability group MDC-2 Joint Experiment and Analysis on RWM Stability Physics. | ||
| Other Requirements: | |||
| Title | 336: Alteration of RWM stability to verify kinetic RWM physics by varied NBI mix + off-axis plasma target | ||
| Name: | Sabbagh | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | J.W. Berkery, J.M. Hanson, et al. | ITPA Joint Experiment : | Yes |
| Description: | Specifically alter the kinetic RWM stabilization properties of a high betaN target plasma by changing the energetic particle (EP) distribution, measuring the change in the resonant field amplification (RFA), and comparing to quantitative calculations of kinetic RWM stability theory. The specific alteration would come from generating a higher poloidal beta plasma (lower Ip), and by shifting the plasma vertically and changing the mix of midplane and off-axis NBI. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Use results from past DIII-D experiments to determine the least stable target plasma for RWM stability. The instability drive can be a combination of both pressure gradient and current â?? the goal being to minimize the ideal stability of the plasma. Once generated, the energetic particle population would be altered by changing the NBI mix and the plasma target vertical position. Plasma rotation profile would also be altered to minimize the stabilizing deltaW_kinetic term to drive the RWM unstable. Non-resonant magnetic braking could be optionally used to further alter the rotation (to reduce RWM stability) if near maximum NBI power would be needed to provide the mode instability drive for ideal deltaW. | ||
| Background: | Experiments on NSTX, with publication at the IAEA FEC 2008, showed a correlation of RWM destabilization at relatively high plasma rotation that correlated with weakened RWM stability for the experimental rotation profile at the marginal RWM stability point. Further experiments have supported this theory and publications have been made for NSTX (e.g. Sabbagh, et al., IAEA FEC 2008 paper EX/5-1, Sabbagh, et al., Nucl. Fusion 50 (2010) 025020, Berkery, et al., Phys. Rev. Lett. 104 (2010) 035003, Berkery et al., Phys. Plasmas 17 (2010) 082504, etc.), and also for experiments on DIII-D (Reimerdes, et al., Phys. Rev. Lett. 106 (2011) 215002). Experiment and theory show that NSTX and DIII-D results can be unified under the present kinetic RWM stability theory (Sabbagh, et al., IAEA FEC 2010 paper EX/5-5). The DIII-D experiment D3DMP No. 2011-14-01 (run in 2011) showed clear evidence that off-axis NBI altered the measured RFA. The present experiment aims to focus on the dependence of RFA vs. EP distribution. RFA will be the primary measurement tool. Generation of an unstable RWM is desired, but not required. | ||
| Resource Requirements: | 8 NB Sources are required. Pellet fueling. I-coil and C-coil fields for n = 1 correction and potential n = 2 or 3 fields for non-resonant magnetic braking. Experiment should follow boronization. | ||
| Diagnostic Requirements: | Equilibrium diagnostics including kinetic profiles (Thomson, CER, MSE) and fast ion diagnostics, RWM sensors. | ||
| Analysis Requirements: | TRANSP/ONETWO calculations of fast ion pressure, Kinetic EFIT, MISK (MARS-K, and HAGIS as available). Analysis will contribute to the ITPA MHD Stability group MDC-2 Joint Experiment and Analysis on RWM Stability Physics. | ||
| Other Requirements: | |||
| Title | 337: Attempt to influence runaway electrons with an n = 1 non-axisymmetric control field | ||
| Name: | Sabbagh | Affiliation: | Columbia U |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | N. Eidietis, J.W. Berkery, J.M. Hanson, et al. | ITPA Joint Experiment : | Yes |
| Description: | Attempt to influence runaway electrons (RE) with an n = 1 non-axisymmetric control field. The overall goal would be a shutdown of the RE beam in a controlled manner. A partial goal would be the controlled reduction of the current in the beam. | ITER IO Urgent Research Task : | Yes |
| Experimental Approach/Plan: | Follow on the success of N. Eidietis, et al. (APS DPP 2011 invited talk) by applying a n = 1 field by the I coils to generate a kinked perturbation of the beam that could be moved to a limiter surface, or otherwise controlled to reduce the RE current. Influence on the RE beam would first be afforded by a pre-programmed n = 1 field, which would be static, then toroidally propagating. If n = 1 global modes are detected and become unstable during this process, n = 1 active feedback could be enabled to attempt to control the perturbation. | ||
| Background: | N. Eidietis, et al. showed success in controlling a RE beam in DIII-D with the plasma control system by axisymmetric means (APS DPP 2011 invited talk). The present experiment aims to build on those results using n = 1 control fields, normally used for RWM active control. | ||
| Resource Requirements: | Setup similar to past RE beam experiments. I-coil and C-coil fields for n = 1 correction and control. | ||
| Diagnostic Requirements: | Equilibrium diagnostics including kinetic profiles (Thomson, CER, MSE) and fast ion diagnostics, RWM sensors. | ||
| Analysis Requirements: | Mode stability analysis if modes are generated. | ||
| Other Requirements: | |||
| Title | 338: Can we obtain ELM suppression without substantial loss of density? | ||
| Name: | Rhodes | Affiliation: | UC, Los Angeles |
| Research Area: | ELM Control: 3-D Field Induced Transport | 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 | 339: Is deficit in gyro-kinetic predicted transport towards edge of Lmode resolved by higher Te, Ti? | ||
| Name: | Rhodes | Affiliation: | UC, Los Angeles |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Test hypothesis that deficit in gyro-kinetic predicted transport/turbulence towards edge of Lmode is ameliorated at higher Te, Ti. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Start with known Lmode condition, verify previously observed transport predictions. Increase Te, Ti independently to determine if predicted transport/turbulence deficit remains in same ratio, increases, or decreases. A positive result is a significant change in the discrepancy indicating that Te, or Ti, holds a key to the puzzle. | ||
| Background: | -- | ||
| Resource Requirements: | ECH, fast wave, turbulence and transport diagnostics. One expt day. | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 340: Do differences in co/cnter GAM/ZF behavior affect deficit in gyrokinetic predicted Lmode transport? | ||
| Name: | Rhodes | Affiliation: | UC, Los Angeles |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | J Hillesheim | 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 | 341: "Born-locked" mode control by ECCD and magnetic perturbations | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye, M. Lanctot, R. Prater, E.J. Strait, A. Welander | ITPA Joint Experiment : | No |
| Description: | Use Magnetic Perturbations (MPs) in real-time to either cancel the Error Field (EF) that penetrated and caused a Locked-mode (LM), or to bring it in view of the ECCD, which will stabilize it. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat the January-February 2010 LM-control experiment at even lower rotation. This can be obtained with reduced, zero or slightly negative NBI torque. The lower rotation will reduce the rotational shield and promote the EF penetration. If not sufficient, apply a deliberately imperfect EF-correction to operate in presence of a finite EF, or reduce q95 to move the q=2 surface closer to the edge, where it will experience reduced shielding and be more prone to EF penetration. The reduced q95 will also make the experiment more relevant to ITER and disruptions.
Once the born-locked forms, a threshold in the DUSBRADIAL radial field amplitude will be exceeded and cause the dud detector to trip and trigger changes in the ECCD and MPs similar to the past experiments. The main difference will be that the MPs will obviously be applied after locking. In the rotating precursor case, the dud detects the rotating precursor approximately 30ms before the actual locking, so that the mode is made lock directly in the desired position. Here, first the mode forms (locked, already, in the "wrong" position), then the MPs are applied. This opens up the choice of whether to apply a static MP or a rotating one, that drags the mode from one position to the other. The two will be compared experimentally, and various rotation velocities will be attempted in the dynamic approach: the rotation needs to be rapid, in order to rapidly start the ECCD stabilization, but not too rapid, otherwise the mode will "slip". Finally, among possible MPs to be applied at the dud trip, we suggest applying a MP equal and opposite to the n=1 EF detected in real-time. This might result in unlocking or stabilization. Most signals (e.g., the LM amplitude) are already available in real-time. Some work might be required to analyze in real time the LM phase from the internal saddle loops, and some logic needs to be implemented in the PCS to translate the desired MP amplitude and phase into the appropriate I-coil currents. | ||
| Background: | In January-February 2010, magnetic + ECCD experiments resulted in the complete stabilization of locked modes (LMs) for the first time. Those experiments concentrated on LMs with rotating precursors. Earlier experiments (2006-08) showed that a slow (0.66Hz) rotating n=1 field can successfully control the LM position, both in the case with and without precursor. At that time, however, the ECCD power was not sufficient for complete stabilization.
In brief, the 2006-08 results suggest good position control and the 2010 results suggest that we have sufficient ECCD power for the success of the proposed experiment. It is important to learn how to stabilize ??born-locked? modes at DIII-D, where they represent approximately half of the LM population, and it will be even more important in ITER where, due to slow rotation, the fraction is expected to be even higher. | ||
| Resource Requirements: | -- | ||
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| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 342: Unlocking by NBI and Study of Locking-Unlocking Hysteresis | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye | ITPA Joint Experiment : | No |
| Description: | At locking, increase the NBI torque. Repeat for fixed NBI torque but higher NBI power, to assess the highest beta at which the mode can be unlocked. Generalize to a 2D scan of NBI torque and power. An economical way to perform such scan would consist in prescribing to the NBI feedback a fixed torque input and decreasing beta in every discharge, then scan the torque from shot-to-shot. Similarly, one can prescribe a fixed beta and increasing torque in every discharge, then scan beta from shot-to-shot. Either way, the mode will unlock for a certain combination of beta and torque. Later in the same discharges, perform the reverse ramps of beta and torque and yet again the original ramps (basically, pre-program triangular waves of beta and torque). This will allow to identify the thresholds for unlocking and relocking. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce 141501-502. Repeat with more or less NBI torque. The mode is expected to unlock sooner (and possibly not relock) or later (or not unlock at all).
Perform a 2D scan of torque and beta as indicated above. Compare the NBI torque at which the mode locks with the minimum torque to apply in order to unlock it. These should be asymmetric (in particular, in absolute terms, more torque should be necessary for unlocking than for locking), due a difference between dynamic and static friction. | ||
| Background: | The â??dud detectorâ?? detects locked modes and their rotating precursors in real time. The PCS is usually instructed to respond by dropping the NBI power and thus beta, to make the locked mode less disruptive. Different real-time changes are proposed here.
This work will take advantage of J. Ferronâ??s simultaneous and independent control of beta and torque by acting on the total NBI power and co/ctr mix. This work will also provide useful input to â??calibrateâ?? the torque balance equation solver TORBA under development at UW-Madison. The latter will help to translate mode unlocking requirements (the torque to apply to the mode) into an NBI torque request (the torque to be imparted to the plasma). Mode unlocking by a change of co/ctr mix has already been obtained in shots 128903, 141501 and 141502 during other locked mode control experiments. The scope of this proposal is to assess the pros and the limits of this technique and to â??automateâ?? it, for future dial-in. | ||
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| Title | 343: q95 Scaling of Turbulence and Transport | ||
| Name: | McKee | Affiliation: | U of Wisconsin |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | K. Burrell, C. Holland, T. Rhodes, L. Schmitz, G. Wang, Z. Yan, L. Zeng | ITPA Joint Experiment : | No |
| Description: | Confinement is known to vary strongly with plasma current and/or q95 as the relevant dimensionless parameter. Connecting this dependence to the scaling of turbulence will be important for understanding the Ip scaling of transport as well as to validating transport simulations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We will perform this experiment in steady L-mode and H-mode conditions and vary q95 systematically. Most likely, we would keep toroidal field constant (2.0 T) and vary current with q95 varying from ~3.0-8.0, and obtain comprehensive fluctuation data with multiple repeat discharges at each condition. We will aim for a moderately co-current rotation plasma and adjust co/ctr beam power to maintain pressure, density and rotation nearly constant. We may consider hybrid discharges for the H-mode phase to avoid sawteeth and obtain long, steady conditions. | ||
| 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 [McKee-PPCF-2006]. This experiment would seek to obtain comprehensive turbulence measurements with multiple fields, a wide wavenumber range, and over a wide radial extent in well-characterized, well-behaved discharges. | ||
| Resource Requirements: | Co and counter NBI | ||
| Diagnostic Requirements: | BES, DBS, CECE, UF-CHERS, PCI, profile | ||
| Analysis Requirements: | TGLF, GYRO,... | ||
| Other Requirements: | |||
| Title | 344: Integrated Disruption Control | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Develop and test an integrated system for disruption avoidance/mitigation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Technical tests and â??calibrationsâ?? (e.g. of the dud detectors) in 2-hour slots on Thursdays. Physical tests initially as piggybacks on shots at risk of disruptions, then in dedicated experiments where disruptions are generated by diverse methods such as n_e ramp-down, EF ramp-up, beta ramp-up for mode onset and locking, laser blow-off or other impurity seeding. | ||
| Background: | See block diagram at slide 21 of the presentation given by F. Volpe at the Friday Science Meeting on Feb.1, 2008.
A new PCS disruption-specific piece of software will consist of 4 parts: 1) STABILITY BOUNDARIES: monitor (a) q95 from EFIT, (b) n_e from the interferometer, (c) betaN from EFIT. At the same time, compute their stability limits with DCON in real time. If needed and if requested, act on Ip, density control or beams. 2) PRECURSORS: analyze in real-time magnetics, optical and edge diagnostics in order to promptly identify (a) locked mode or its rotating precursor, (b) MARFE, (c) detachment. 3) AVOIDANCE: (a) solve torque balance equation. Calculate whatâ??s most efficient among the following: (i) drop NBI to make mode less disruptive, (ii) maximize NBI torque to unlock the mode, (iii) apply static n=1 Magnetic Perturbations (MPs) and continuous ECCD or (iv) rotating MPs and modulated ECCD, for mode suppression. (b) increase NBI ( c ) ? not clear how to respond to detachment 4) MITIGATION: keep monitoring signals of 1) and 2). If 3) has failed, drop NBI and deploy one of the following: MGI, killer pellet, ECH (â??a la FTU and AUG). | ||
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| Other Requirements: | Real-time DCON.
Changes to the PCS. | ||
| Title | 345: Compare ECCD and ECH control of disruptions | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Test/confirm in other types of disruptions (density-limit, high-beta, low q95, laser blow-off) what has been found in locked-mode disruptions: that ECCD is more effective than ECH at avoiding disruptions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Cause disruptions of 2-4 types: 1) by ramping ne and hitting the Greenwald density limit, for comparison with AUG and FTU, 2) by ramping beta up, 3) by ramping q95 down, 4) by injecting impurities (for example by laser blow-off, if available, for comparison with FTU), which increases the resistivity and cools the plasma, resulting in current and thermal quench. Approaches 2) and 3) and, by radiative destabilization, approach 4), are likely to lead to the formation of an island. If so, operate at high rotation to prevent locking, in order to differentiate from the high-beta locked-mode disruptions studied so far.
Dud on Vloop (disruption detector, as in FTU and earlier AUG experiments) or on DUSBRADIAL etc. (locked mode detector, as at DIII-D and, more recently, AUG), whichever trips first. For comparison, also dud on Vloop only. On average, this will react ~400ms later but, unlike the locked mode detector, will also pick disruptions neither caused nor aggravated by LMs. Apply ECH. Disruption should be delayed or avoided compared to no-ECH reference. Similar to AUG, scan radial location of ECH from shot to shot. Maximum effect is expected for deposition at q=2. A smaller peak is expected at q=3. Repeat with ECCD. | ||
| Background: | EFFECT of ECCD
It has been already demonstrated experimentally at DIII-D [Volpe 2010] that ECCD is more effective than ECH at stabilizing locked NTMs and thus at avoiding locked-mode disruptions. This is not surprising, because it is the same physics of rotating NTM stabilization, where the superior stabilization efficiency of ECCD over ECH is well-known. In fact, stabilization in ITER is expected to be entirely dominated by ECCD. EFFECT of ECH For efficient stabilization it is required that ECCD is deposited in the island O-point. In the case of locked islands, this is easily and reproducibly achieved at DIII-D by means static n=1 magnetic perturbations. Thus, the simplicity of ECH, of not requiring magnetic control, is only a minor asset. Besides, although not as much as ECCD, ECH would also benefit from the magnetic control of the toroidal phase of locking. The real potential asset of ECH is that by heating the plasma it can prevent, counteract or reduce the thermal quench. Heating, however, is always present, even during ECCD. In fact, ECCD is a consequence of heating and its asymmetry in the velocity space. Besides, FTU and AUG scans show the maximum effect for ECH at q=2, not in the core. This suggests that ECH stabilizes an MHD instability or other q-resonant phenomenon, and that the effect of compensating for the thermal quench is secondary. In summary, ECH is suspected to have minor effects. All these ECH effects are also present during ECCD, which offers additional advantages. On the other hand, FTU avoided more complicated disruptions than at DIII-D, with a locked 3/1 mode followed by a rotating 2/1 which eventually locked too. MARFEs were also observed in those disruptions. LM and DISRUPTION STATISTICS According to JET statistics, nearly 100% of disruptions originate from a LM, or develop a LM at some point in their evolution. Preliminary results [Mao, Volpe] seem to confirm similar statistics also for DIII-D. Hence, controlling the LM is equivalent to avoiding LM disruptions and to mitigating most other disruptions. If this is correct, ECCD is expected to be more efficient than ECH at avoiding, delaying or mitigating disruptions. | ||
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| Title | 346: Locked mode control at low (ITER-like) q95 and with external C-coils | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate magnetically-assisted ECCD stabilization of locked-modes under two ITER conditions, namely: 1) low q95 and 2) using external coils only. The latter would free up the internal coils for other tasks needing proximity to the plasma and/or fast response. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat 141492 but with the C-coils, same toroidal phase. Might require adjusting the C-coil currents. Repeat lowering q95 from shot to shot in steps of 0.3 and changing BT consistently to keep the q=2 location approximately fixed. Repeat the q95 scan again, but without consistently varying BT. As a result, the q=2 location will move towards colder regions, and ECCD be less efficient. Hence, we will increase the current drive density by switching to narrow ECCD (in 141492 it was deliberately broad, for ease of alignment). This will make the alignment more challenging, but we will use the Search and Suppress or Active Tracking to adjust the plasma position, BT or steer the mirrors in real time, if ready. | ||
| Background: | ECCD stabilization of disruptive locked modes assisted by static n=1 magnetic perturbations succeeded at q95=4.5 but should be tested at low, ITER-like values of q95=3-3.5. As a result the mode will lie at outer radii, closer to the wall and error field, and thus interact more intensively with them, i.e. will lock more rapidly. It will also be less shielded by rotation. After locking, the main differences will be proximity to the edge (thus, stronger perturbation from the EF, and bigger island) and the fact that locally the plasma will be colder, and ECCD less efficient.
The other rationale for this experiment is that so far the technique used the internal I-coils, but to date the installation of internal coils in ITER is still uncertain. Because the technique only needs static n=1 perturbations comparable in strength with the machine EF, it is likely to work with external coils as well, but it is important to confirm it experimentally. As a by-product, this experiment will also test the efficacy of ECCD control of locked modes in preventing or avoiding low q95 disruptions. | ||
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| Title | 347: Make locked mode and disruption control more routinely available by real-time steering of ECH/ECCD | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Disruption Characterization and Mitigation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Make locked mode and disruption control more routinely available by programming the dud and responding with magnetic perturbations and with ECH/ECCD steered in real-time. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In every discharge featuring ECH/ECCD, set the dud detector to detect a locked mode and/or disruption (from Vloop), whichever comes first and, instead of the usual drop in NBI power, respond to the dud by steering the ECH/ECCD towards the q=2 location, as tracked by the â??Search and suppressâ?? or â??Active trackingâ?? algorithms. Also, at the dud, apply an n=1 static perturbation (with the I-coils or the C-coils, whichever was used; in the I-coil case, apply an n=1 regardless of the relative phase between the upper and lower row).
In shots where ECH/ECCD is not required, we can turn it on later in the discharge at a â??stand-by (10% of maximum) power level. ECH/ECCD can only be modulated (for example from 10 to 100%) but not turned on (from 0 to 100%) by the PCS. NOTE: permission will be asked to individual session leaders every day, but some form of co-ordination and authorization at an upper level would be appreciated. | ||
| Background: | Locked mode and disruption control was demonstrated at DIII-D in discharges where the NBI was ramped to generate an NTM and rotation was kept low, to let it lock. These high-beta, highly disruptive discharges are sometimes tagged as â??idealizedâ??, but should rather be referred to as â??worst case scenariosâ??. Regardless, it is agreed that the automatic extension of control to a wider class of locked modes and disruptions is highly desirable to test its applicability and versatility in view of ITER. | ||
| Resource Requirements: | We are not asking for dedicated time, but for permission to program the dud detector and, if tripped, use magnetic perturbations and steer the ECH/ECCD in as many experiments as possible, especially if they foresee a risk of locking or disruptions and if they plan to use ECH/ECCD anyway. | ||
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| Title | 348: Excitation and characterization of Quasi-stationary Modes (QSMs) | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye | ITPA Joint Experiment : | No |
| Description: | Understand the physics of QSMs, their excitation and their rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Restore both shots with â??spontaneousâ?? QSMs (see Mao-Volpeâ??s database) and shots which initially featured a LM that later evolved into a QSM (F. Volpeâ??s experiments, e.g. 141058-60, 64 etc.). In both types, scan the NBI torque from slightly co- to slightly counter, the n=1 EF (thus, the associated torque) and possibly, super-imposed, the n=3 magnetic braking.
Repeat with ECCD, that was observed to reduce the QSM frequency, probably as an indirect result of affecting its amplitude (141058-59). For comparison repeat with ECH only, which has a weaker dependence on toroidal phase. If time, try to trigger a QSM by pellet injection, similar to J. Snipes at JET [NF 28, 1085 (1988)]. Snipes also observed QSMs being destabilized by large sawteeth. This however, is more easily reproducible as a piggyback on a sawtooth or giant sawtooth experiment than as part of the dedicated experiment. Finally, apply a magnetic perturbation (MP) rotating opposite to the QSM. Apply different rotation frequencies, to confirm that the rapid oscillations and slow growth of the QSM amplitude observed in 141064 and other discharges are beating phenomena occurring at the small difference and large sum of the QSM and MP frequencies. Note that the slow growth was periodically reset by rapid sawtooth-like amplitude crashes, flashes of light and bursts of particle and heat losses. | ||
| Background: | Quasi-stationary modes (QSMs) are naturally slowly rotating modes (10-30 Hz), intermediate in nature between locked modes (LMs, 0 Hz) and rapidly rotating NTMs (5-30 kHz). Their slow rotation is probably a stable solution of the torque balance equation, just like LMs and NTMsâ?? are. However, QSMs are observed much less frequently (and reproducibly?) than LMs and NTMs. This proposal seeks to understand why.
Apart from their fundamental interest, QSMs are appealing for two main reasons: they are â??imperfectâ?? locked modes, not really locked but slowly rotating, thus easier to stabilize (e.g. by ECCD only, with no need for magnetic control of their phase); they are â??imperfectâ?? NTMs, rotating at unusually slow frequencies, which makes them easy to diagnose in detail even with diagnostics, such as CER and MSE, that normally lack the necessary time resolution. For this reasons, it is important to understand how to convert LMs and NTMs in QSMs. Specific questions to address are: is the QSM an energetically less-favorable solution of the torque balance equation? Or it only appears under certain conditions, for example depending on the competition of distinct torques (from the wall, from the error field (EF), viscous etc.) having different dependences upon rotation frequency? It is well known that the magnetic friction from the wall and the torque from the EF counteract and partly cancel each other at frequencies of the order of the inverse resistive wall time, consistent with QSMs. It is also theoretically predicted that stronger/weaker EFs can move this energetically favorable torque minimum to lower/higher frequencies. In fact, very strong/weak EF might prevent the QSM from being observed by letting it â??degenerateâ?? in a LM or rapid NTM. This is confirmed by various LM control experiments where QSMs were prevented by brute force (large MPs). By contrast, small (but not too small) MP led to QSM, for certain phases (shot 141052) but not for others, for which probably it was partly canceling the EF (141051). It is speculated that typical EFs are high enough to observe LMs but too small for the QSM solution to exist or be observable. An alternative interpretation is that the PLASMA rotation â??transitsâ?? too rapidly through the range of frequencies at which the MODE likes to rotate. This is analogous to a mechanical system having several modes of vibration. Different eigenmodes will have different eigenfrequencies. Hence, if the frequency of a perturbation (for example a motor, or a piston) is ramped up (e.g. when a car is accelerating), the system will pass through several resonances. However, if the acceleration is rapid enough (=if the system spends at a certain frequency less than a period) the system will rapidly go out of resonance, and a certain mode will not be observed, or be barely observed. Talking of timescales, QSMs often strike 0.3-1s after locking, suggesting that some slow, global resistive effect might be involved in their onset. | ||
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| Title | 349: Mutual alignment of gyrotrons: a new technique based on steering and modulation | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Exploit the fact that two gyrotrons aiming at the exact same target and modulated out-of-phase are equivalent to a single gyrotron operated in cw, to solve the problem of how to align 24 beams in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate 2 gyrotrons at 10-100 Hz and out-of-phase. Keep one beam fixed and steer the other. When both deposit at the same rho, the situation is indistinguishable from continuous ECH from a single gyrotrons. Hence, no heat-waves will propagate in the plasma. The corresponding signature will be a zero or minimum of heat-pulse height in the ECE.
Repeat for the other two pairs of launchers. Finally, align each pair to the other. To assess the benefits of the improved alignment, aim all gyrotrons at the same target and modulate them in phase with each other. A broader or narrower deposition width is expected in the height pulse analysis of the ECE, depending whether the old or the new angular calibration is used. | ||
| Background: | ITER will be equipped with at least 24 gyrotrons delivering at least 20MW to the plasma. An earlier remote-steering design raised some concern on the ECH/ECCD deposition being too broad and the driven current density too low. This concern propelled a large, international modelling effort, that culminated in a new front-steering design. The new design was more than satisfactory as far as the deposition width of an INDIVIDUAL beam was concerned. ITER, however, will feature 24 gyrotrons. Failing to correctly align them to each other will result in a broadening of the TOTAL driven current that might make the improved design vain, unless an extremely high angular precision in the gyrotrons aiming (0.2deg) is achieved [F. Volpe, J. Phys.: Conf. Series 74, 1409 (2003)].
DIII-D has the highest number of gyrotrons in operation (6), and has been recently equipped with remote steering launchers, hence it is the best environment where to test the mutual alignment of gyrotrons under conditions close to ITER. | ||
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| Analysis Requirements: | toray, heat-pulse analysis | ||
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| Title | 350: Effect of impurities and wall conditioning on NTMs | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Measure beta threshold for NTM onset for different wall conditions and under controlled core/edge cleanliness or impurity seeding. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Develop a small database of 10-20 discharges with NTM onset during NBI (hence, beta) ramp. A 2/1 mode is preferred, as it is closer to the wall and possibly more dependent on impurities and wall conditioning. The otherwise identical shots should differ only by wall conditions (e.g. be taken after a boronization, disruption, or glow discharge) and/or deliberate impurity contamination (impurity pellet, puffing, laser blow-off). There are various ways of building the database.
Approach 1: modify the plasma test shot run every morning at DIII-D (the same used for long term monitoring of wall conditioning and impurities) by adding an NBI ramp at the end. Make additional, small changes, if needed for NTMs but then leave them fixed for the duration of the campaign. Approach 2: try building the database in a single day or half-day of experiment. However, this would only tackle the dependence on impurities and the short-term dependence on wall conditions, but fail to do any monitoring in the long term. | ||
| Background: | Dedicated experiments at NSTX (F. Volpe, L. Delgado-Aparicio, S. Gerhardt, S. Sabbagh et al.) showed that the onset of NTMs is delayed by Lithiumization and anticipated by Neon puffing. Post-disruption data at DIII-D also suggest an effect of wall conditioning and impurities on NTMs.
Here we propose to systematically characterize these effects at DIII-D, by operating otherwise NTM-unstable discharges under controlled conditions of core/edge cleanliness or impurity seeding. We also propose to compare with the NSTX results. This is important in view of ITER, where it might be preferable to wait for good wall conditioning before trying high-beta operation, if this poses the risk of a major NTM, possibly resulting in locking and disruption. | ||
| Resource Requirements: | Pellet, impurity pellet, laser blow-off? | ||
| Diagnostic Requirements: | All edge diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 351: o First experiments with the centerpost Swing-probes | ||
| Name: | Tsui | Affiliation: | UCSD |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | o CK Tsui, AW Leonard, R Boivin, PC Stangeby, JG Watkins, JA Boedo | ITPA Joint Experiment : | No |
| Description: | o Two centerpost swing-probes are being installed for the 2012 campaign. Each probe swings out a mach-probe tip on a 20cm arc, which is long enough to reach the LCFS in most plasma shapes. The design is new and untested, so a number of proving experiments will be necessary. Subsequently, the swing-probes will be employed together with the rest of DIII-Dâ??s edge diagnostics to characterize a simple-as-possible-plasma as well as possible. A well characterized plasma with the additional measurements on the under-diagnosed inboard side of the SOL will be interpreted using the OEDGE code. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | o Proof of concept expts:
â?¢ B-field only, no plasma experiments. â?¢ To confirm that systems and data acquisition are operating correctly. Itâ??s possible this test could be complete in air before the end of the vent. â?¢ Sheath limited L-mode, to confirm probe measurements are correct, by comparing with measurements made by the reciprocating probe in the outer SOL and target probes â?¢ Early in the campaign, so that if any unforeseen problems with the assemblies arise, there is a chance to repair them if there is a mid-campaign vent. â?¢ Can be run in piggyback so long as the plasma shape and conditions are appropriate. o Particle Transport SAPP expt â?¢ Can be combined with an impurity flow speed experiment, or a PMI experiment. â?¢ A multi-shot experiment 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) â?¢ Ideally, the plasma should be single null L-mode with no risk of disruptions. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | o 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 experiments. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 352: Electron Critical Gradient in the Edge "Shortfall" Region | ||
| Name: | Petty | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | J. E. Kinsey | 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 over prediction of the critical gradient value and/or transport stiffness. The critical gradient in the electron temperature and stiffness 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 will be determined from (1) by measuring the power balance and heat pulse transport, and (2) from the turbulence behavior. | 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=0.85. with one gyrotron modulated at 25 Hz. (3) Shot to shot, move one gyrotron to rho=0.65 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 over predicting the electron temperature critical gradient; the experiment will also measure the stiffness in the conductive electron heat flux, which may provide additional clues as to the origin of the disagreement. | ||
| Resource Requirements: | ECH: 7 gyrotrons desired, 6 gyrotrons essential.
NBI: 30LT and 330LT beams essential, 150LT and 210RT 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 | 353: Pedestal-centric Tokamak Optimization | ||
| Name: | Snyder | Affiliation: | ORNL |
| Research Area: | Torkil Jensen Award | Presentation time: | Requested |
| Co-Author(s): | K.H. Burrell, R.J. Groebner, T.H. Osborne | ITPA Joint Experiment : | No |
| Description: | 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.
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. 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. 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: | In a clean machine, ideally shortly after boronization, attain 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. 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 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: | (above) | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | high res Thomson, CER, BES, libeam if possible | ||
| Analysis Requirements: | EPED studies before and after expt | ||
| Other Requirements: | boronization | ||
| Title | 354: Search for magnetic fluctuation-induced transport during RMP operations | ||
| Name: | Zeng | Affiliation: | UC, Los Angeles |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | J. Zhang, E. Doyle, T. Rhodes | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to search for the evidence of magnetic fluctuation-induced transport during RMP operations by microwave polarimeter and density fluctuation measurements. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | By using microwave polarimeter and density fluctuation meansuremnets (DBS and BES), we can investigated the coupling of magnetic and density fluctuations , in order to search for magnetic fluctuation-induced particle transport during RMPs. | ||
| Background: | In MST, it has been observed that magnetic fluctuation-induced transport in magnetic stochastic layers. Possibly we can also use single channel microwave polarimeter to search for this kind of transport. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | microwave polarimeter, DBS, BES and profile reflectometer | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 355: Investigation of particle transport and turbulence variations when RMP phase flipping | ||
| Name: | Zeng | Affiliation: | UC, Los Angeles |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | E. Doyle, T. Rhodes | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate particle transport and turbulence during RMP phase flipping (0 and 60 deg), in order to further understand RMP-induced transport and turbulence. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | By using small modulations (0.5 or 1 kA) in I-coil current in different RMP phases (0 and 60 deg) , measure density fluctuation behavior , particle diffusivity and inward pinch velocity via profile reflectometer. We can make these measurements with q95 scan. | ||
| Background: | In last year RMP experiments, it was observed that density fluctuations, profiles, beta_n and rotations significantly change when n=3 RMP phase flipping. It is interesting to further study. Specially by using small I-coil current modulation, profile reflectometer will obtain D and V variations with different RMP phases. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | DBS, BES and profile reflectometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 356: Simultaneous ne and Te measurement by reflectometer via relativistic effects | ||
| Name: | Zeng | Affiliation: | UC, Los Angeles |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | E. Doyle, A. Peebles, T. Rhodes, M.E. Austin | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to test a novel simultaneous measurement of core ne and high Te (> 5 kev) by using refletometer via relativistic effects. This also will provide a demo for future profile reflectometer measurement in ITER. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using high power ECH and FW heating, obtain high Te (> 5 kev) in a wide radial range (up to 0.5 kev ?), with low BT (1.6 - 1.65 T), and ne(0) < 6 e13 cm^-3. Via analysis of O- and X- mode relativistic effects of profile reflectometer with dual mode operation, core ne and Te will be simultaneously measured. The results will compare to TS and ECE measurements. Also make density and Te fluctuation measurements (DBS, BES, cece and ecei), for turbulence study in high Te. | ||
| Background: | In high Te plasma, profile reflectometer measurement should consider the relativistic effects. By using the difference of O- and X- mode relativistic effects, the dual mode operative reflectometer can simultaneously obtain ne and Te profiles. The design and simulations for this kind measurement for ITER have been published. I hope this novel measurement can be demonstrated in DIII-D. | ||
| Resource Requirements: | ECH, FW | ||
| Diagnostic Requirements: | profile reflectometer, Thomson scattering, ECE, DBS, BES, CECE and ECEI | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 357: Particle transport measurement in QH mode via gas puff modulation | ||
| Name: | Zeng | Affiliation: | UC, Los Angeles |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | E. Doyle, T. Rhodes, L. Schmitz, G. Mckee, K. Burrell | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to measure particle transport in QH mode regime by using main gas puff modulation technique. Compare particle diffusivity and inward pinch in QH mode with ELMy H-mode results. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In QH mode discharges, with or without nonresonant magnetic fields, apply D2 gas puff modulation. Use profile reflectometer measurement, to obtain D and V. Also proceed fluctuation measurements (DBS, BES, and ECEI) in order to investigate turbulence behavior. | ||
| Background: | |||
| Resource Requirements: | D2 gas puff modulation | ||
| Diagnostic Requirements: | profile reflectometer | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 358: Particle transport and turbulence investigations in core transport barrier plasmas | ||
| Name: | Zeng | Affiliation: | UC, Los Angeles |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | E. Doyle, M. Austin, T. Rhodes | ITPA Joint Experiment : | No |
| Description: | The goal of this work is to measure particle transport and turbulence in core transport barrier plasmas, e.g., "bat-eared" Te profile plasma. Comparing to theoretical models, e.g. TGLF , GYRO, further study turbulent transport. This work also can help to understand the physics of internal barrier. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In core barrier discharges, e.g. with "bat-eared" Te profile plasma, by using main gas puff modulation technique, profile reflectometer will measure particle diffusion coefficient D and pinch velocity V. Also make fluctuation measurements (DBS, BES, and ECEI) in order to investigate turbulence behavior. Also make GYRO and TGLF simulation. | ||
| Background: | -- | ||
| Resource Requirements: | ECH, D2 gas puff modulation | ||
| Diagnostic Requirements: | profile reflectometer , DBS, BES, cece and ecei | ||
| Analysis Requirements: | TGLF and GYRO | ||
| Other Requirements: | -- | ||
| Title | 359: EFC with multiple toroidal harmonics | ||
| Name: | Strait | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | First test in DIII-D of combined n=1 and n=2 error field correction.
This goal is to test the hypothesis that multiple toroidal harmonics of the error field are important, and possibly to improve plasma performance through improved error correction. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The basic idea is simply to superimpose an n=2 current distribution on the normal n=1 I-coil correction current, with the same pitch as the n=1 correction field. The n=2 field then has two degrees of freedom (amplitude and toroidal phase).
Use Ohmic plasnas for the experiment, with the figure of merit being the low-density locked mode threshold as usual. (1) Evaluate the low-density locked mode threshold with standard n=1 EFC. (2) Keeping the standard n=1 EFC fixed, optimize the n=2 field with the usual 4-quadrant current ramp method. (3) Apply the optimum n=2 contribution, and again evaluate the low-density locked mode threshold. The toroidal mode number of the locked modes induced with large n=2 fields will also provide information on whether the influence of the n=2 field is through braking or direct drive of an n=2 instability. | ||
| Background: | In-vessel measurements in 2001 showed that there is an n=2 contribution to the DIII-D error field from elliptical distortions of the coils. We have never attempted to correct this error field
Recent experiments have emphasized that single-mode n=1 error field correction provides only partial correction. It is not known what the important uncorrected fields are, but the n=2 contribution could be important This experiment is the first test of this hypothesis. Although not an ITPA joint experiment, this experiment is highly relevant to the ITPA working group on error field correction criteria. | ||
| Resource Requirements: | The I-coil must be operated as 12 independent circuits, using sub-SPAs. Due to limited I-coil current capability in this configuration, the experiment will probably need to be done at reduced Ip and Bt.
Does operation with 12 independent circuits require any capabilities to be added to the I-coil patch panel or to the PCS? | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Control room analysis tools for locked mode detection and EFC calculation will need to be updated for n=2 modes. | ||
| Other Requirements: | |||
| Title | 360: EFC with multiple poloidal harmonics | ||
| Name: | Strait | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | First test in DIII-D of n=1 error field optimization with different upper and lower I-coil current distributions.
This goal is to test the hypothesis that multiple poloidal harmonics of n=1 are important, and possibly to improve plasma performance through improved error correction. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The basic idea is add a second I-coil current distribution that has a 60-degree upper-lower toroidal phase difference.
-- This distribution is orthogonal to the standard â??240 degreeâ?? n=1 configuration, and thus should not couple to the mode that the 240 degree configuration is optimized for. It may, however, couple to other poloidal modes. -- From another point of view, this distribution is up-down anti-symmetric where the original distribution is up-down symmetric. If properly aligned with the standard EFC current, it simply varies the relative amplitudes of the two rows of I-coils while keeping a constant sum of the two amplitudes. If the error field source is not up-down symmetric, this should improve the error field correction. Use Ohmic plasnas for the experiment, with the figure of merit being the low-density locked mode threshold as usual. (1) Evaluate the low-density locked mode threshold with standard EFC. (2) Keeping the standard 240 degree EFC fixed, optimize the 60-degree field with the usual 4-quadrant current ramp method. (3) Apply the optimum 60-degree contribution, and again evaluate the low-density locked mode threshold. Any changes in the poloidal structure of the locked mode that is induced with a large antisymmetric n=1 field will also provide information on whether the influence of additional poloidal harmonics is through non-resonant braking or resonant destabilization of a second n=1 mode. | ||
| Background: | Recent experiments have emphasized that single-mode n=1 error field correction provides only partial correction. It is not known what the important uncorrected fields are. This experiment tests the hypothesis that multiple poloidal harmonics of n=1 are important.
Although not an ITPA joint experiment, this experiment is highly relevant to the ITPA working group on error field correction criteria. | ||
| Resource Requirements: | The I-coil will be connected as n=odd pairs, with 6 independent circuits. These can be powered with some combination of SPAs and C-supplies in order to have a 5 kA capability for all circuits. | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 361: Characterize pedestal turbulent modes in early ELM-free H-mode phase | ||
| Name: | Yan | Affiliation: | U of Wisconsin |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | G. McKee, R. Groebner, P. Snyder, T. Osborne | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to look for the onset of turbulent modes (low to intermediate k) and characterize such modes observed in the pedestal region in early ELM-free H-mode phase | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The main idea is focusing on the early ELM-free H-mode phase right after L-H transition to watch the evolution of the pedestal profiles and turbulent modes onset in a set of current (collisionality, if possible) scans in low and high triangularity shape. Beam power just above the L-H transition power threshold will be used to achieve ELM-free H-mode phase as long as possible. May have several transitions in one shots to improve the statistics. Diagnostics like BES, UF-CHERS, DBS, CECE will be used to measure the turbulence spatial and temporal structure across the pedestal region. | ||
| Background: | Detailed characterizing turbulent modes evolution during pedestal building up is of great interest. Previous measurements from BES have shown KBM like mode [1,2], which has been predicted to limit pedestal pressure. However, definitive experimental evidence of KBMs is still lacking. It has been observed the pedestal profiles evolve very fast after ELM crashes [3], within the first several ms the pedestal has built up to ~80% of saturated level. Therefore it is difficult to see the onset of the modes when pressure gradient exceeds some critical value. The early ELM free H-mode phase, however, is much longer, and the pedestal builds up much slower. In the 2011 campaign, an interesting process of the modes evolving from onset to coherent, then to broadband turbulence has been observed during this ELM free H-mode phase. So this period of time will help to definitively characterize the turbulence mode during the pedestal building up.
[1] Z. Yan, et al., PRL, 107, 55004,2011 [2] Z. Yan, et al., PoP, 18, 056117, 2011 [3] R. Groebner, et al., NF, 50, 064002, 2010 | ||
| Resource Requirements: | 7 neutral beams | ||
| Diagnostic Requirements: | High resolution TS, CER, BES, UF-CHERS, DBS, CECE, PCI | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 362: Correction of a spatially localized error field | ||
| Name: | Strait | Affiliation: | GA |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Test the use of localized coil currents to correct a localized error field source (i.e. the TBM mockup). This proposal is closely related to #194 by Rob La Haye.
The goal is to test the hypothesis that multiple poloidal and toroidal harmonics are important for plasma performance. This is a general test of error field physics, and is NOT just specific to ITERâ??s test blanket modules. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The TBM is located at 270 degrees,
(1) Use C259 and C319 with opposite polarities to generate a local toroidal field perturbation, counter to that of the main TBM coil. (same as in Robâ??s proposal) (2) Use IU270 and IL270 with opposite polarities to generate a local vertical field perturbation, counter to that of the TBM solenoid coil. (new feature) In H-mode plasma, optimize these two correction fields (independently and together) with respect to plasma rotation -- as was done for n=1 error field correction in 2011. The two correction coil sets have only one degree of freedom each, so optimization should be straightforward. | ||
| Background: | TBM mockup experiments in 2011 showed that n=1 error field correction did not fully recover the decrease in rotation caused by the TBM mockup field. This strongly suggests that other harmonics of the TBM field are important and must be corrected.
This experiment tests local correction of a local error field, where both the error field and the correction field contain a broad spectrum of spatial harmonics. Although not an ITPA joint experiment, this experiment makes a further contribution to the ITER issue of the effects of the TBM error field, and is also highly relevant to the ITPA working group on error field correction criteria. | ||
| Resource Requirements: | Requires re-installation of the TBM mockup module, for at least a few days of operation. This module is a unique tool for error field physics studies.
In order to add the local field for TBM correction along with the standard n=1 error field correction, at least 5 independent I-coil circuits are needed. This can be done at full I-coil current with a combination of SPAs and C-supplies. A non-standard I-coil patch panel may be required. Similar remarks apply to the C-coil. If used only for TBM correction, one non-standard circuit is needed. If also used for n=1 error field correction, 5 independent circuits are needed. | ||
| Diagnostic Requirements: | Preparation for this experiment should include modeling with SURFMN to determine the range of currents that will reduce the resonant and non-resonant fields from the TBM, for n in the range 1-3. IPEC or MARS-F modeling may also be useful. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 363: Measure q95 window and I-coil threshold at as a function of rotation (including counter rotation) | ||
| Name: | Solomon | Affiliation: | GA |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | The goal of this experiment is to measure and document the q95 window and I-coil threshold current as a function of core/pedestal rotation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Using ISS, vary initial torque *PRIOR* to applying the RMP (this was planned as an option last year, but we seemed to need the RMP early to prevent the density from getting too high early on, preventing successful ELM suppression - we must avoid that temptation!) Do q95 scans at fixed I-coil current of ~4 kA to establish the q95 window for each rotation level. Then, repeat discharge with fixed q95 near the center of the window, and ramp up the coil current to establish the threshold. Rotation should be pushed as far in the counter-Ip direction as is practical. Consider using C-coil for NRMF torque to provide additional counter torque, and potentially decouple the core and pedestal rotation. | ||
| Background: | One candidate picture for explaining RMP ELM suppression is that an island-like structure is formed near the top of the pedestal, corresponding to regions where vperp_e are "small". A consequence of this hypothesis is that reducing the rotation from rapid co toward balance should shift the vperp_e crossing point further inward, and increase the magnitude of vperp_e near the top of the pedestal. This should increase the screening, and result in more difficult ELM suppression, which in turn should be manifested either as a requirement for larger I-coil current, or at fixed I-coil current, a reduction of the q95 window for ELM suppression. Both aspects of this should be tested. A first attempt was made at this in FY11, but locked modes at reduced rotation proved problematic. Starting at reduced rotation before applying the RMP may alleviate this. | ||
| Resource Requirements: | 1 day expt, 210 beams, 6 gyros for NTM control, RMP I-coil | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 364: Radial structure and propagation of GAM oscillations in electron temperature, density, and ExB flow | ||
| Name: | Wang | Affiliation: | UC, Los Angeles |
| Research Area: | General BPP | Presentation time: | Not requested |
| Co-Author(s): | Peebles, Rhodes, etc | ITPA Joint Experiment : | No |
| Description: | The purpose of the experiment is to obtain a simultaneous measurement of Geodesic acoustic mode oscillations in electron temperature, density, and ExB flow via ECE, ECEI, BES and DBS in L-mode plasma with various heating power levels, and seek to compare with Gyrokinetic simulations to test and validate code predictions. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Generate (known) L-mode target plasma with added steps of different NBI and/or ECH power levels. Document GAM features with ECE, ECEI, BES and DBS. | ||
| Background: | GAMs are believed to play an important role in plasma turbulence and transport. Radial structure and other features (e.g. interplay with turbulence) of GAMs in multiple fields are yet to be understood. Recently GAM oscillations in electron temperature and eigenmode character have been observed on DIII-D for high power L-mode discharges. A dedicated experiment is desired to obtain a simultaneous measurement of GAMs in electron temperature, density, and ExB flow in L-mode plasma with various heating power levels, and compare with Gyrokinetic simulations to test and validate code predictions. | ||
| Resource Requirements: | most NB sources and gyrotrons | ||
| Diagnostic Requirements: | ECE, ECEI, BES, DBS, CECE, and other routine diagnostics | ||
| Analysis Requirements: | GYRO | ||
| Other Requirements: | |||
| Title | 365: Correlation length of electron temperature turbulence and comparison with gyrokinetic simulations | ||
| Name: | Wang | Affiliation: | UC, Los Angeles |
| Research Area: | General BPP | Presentation time: | Not requested |
| Co-Author(s): | Rhodes, Peebles | ITPA Joint Experiment : | No |
| Description: | This experiment seeks to investigate radial correlation length of electron temperature turbulence and compare with gyrokinetic simulations. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Generate 2 steady-state plasma phases with ECH and ECH+NBI L-mode in one target shot. At 3 mid-radius locations take CECE measurements from shot to shot: fix one channel location and program to scan the location of another channel around it to collect correlation data sets. To ensure signal-to-noise ratio in correlation analysis, slow enough scan should be applied. | ||
| Background: | The turbulence correlation length provides a robust alternative to fluctuation levels to compare with turbulence modeling predictions. Due to the much larger thermal noise superimposed in the electron temperature fluctuations, it has been challenging to obtain a radial correlation length of electron temperature turbulence using the CECE measurement. A dedicated experiment will allow for a detailed investigation. | ||
| Resource Requirements: | most gyrotrons and NB sources | ||
| Diagnostic Requirements: | CECE, DBS, BES, PCI, profile reflectometer and routine profile and other diagnostics | ||
| Analysis Requirements: | ONETWO, TGLF, GYRO | ||
| Other Requirements: | |||
| Title | 366: counter-OANBI destabilized AEs | ||
| Name: | Tobias | Affiliation: | Los Alamos National Laboratory |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Attempt to demonstrate unstable RSAEs propagating counter to plasma current with off-axis NBI injected in the counter-current direction | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Repeat off-axis NBI discharges of the previous experimental campaign, but in reversed Ip, such that the 150 sources are now counter-injection. It should be possible in this scenario to retain a reversed shear q-profile, and this may be somewhat confirmed by MHD spectroscopy though MSE will not be available with only the 150 sources on. The counter-injection angle should produce a large population of counter-going passing and trapped ions and test our hypothesis as to what role the magnetic moment distribution of these particles plays in eigenmode stability. | ||
| Background: | During the previous campaign, ECEI and BES have been used to look in detail at the rotation and 2D phase of unstable Alfvén eigenmodes on DIII-D. By reversing Bt, we showed that both had a correlated relationship to the direction of diamagnetic drift. Using the off-axis beams, we hoped to show that the modes could also be "flipped" by using a local inversion of the pressure gradient near q_min to reverse the diamagnetic flow direction. Codes that were successful benchmarked for the normal, on-axis NBI cases (TAEFL, Gyro, GTC) predicted that plasma profiles within the error bars of those measured should have had unstable RSAE-like modes propagating counter to the plasma current. However, we observed that the normal RSAEs were stabilized and no new ones showed up. We hypothesize that this is because the fast ion population tends toward co-propagating in these cases, and now hope to see if injecting a large counter-going fast ion population will be sufficient to destabilize the modes.
The data will be invaluable as we continue to develop predictive capability which includes a more realistic, 5D phase-space picture of the fast ion population. Analysis of relevant scenarios with M3D-k, for example, is already underway and will need experimental data for validation. In addition, this configuration will provide the opportunity to better understand why the MHD frequency of BAAEs was so dramatically changed by OANBI, though the phase structure of the modes stayed the same, without "flipping" as we expected to see had they been true counter-propagating modes. BAAEs will be discussed more in my other proposal. | ||
| Resource Requirements: | both off-axis NBI sources | ||
| Diagnostic Requirements: | ECEI, BES, reflectometers, interferometry, core Thomson | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 367: Acoustic-Alfvén eigenmodes with counter OANBI | ||
| Name: | Tobias | Affiliation: | Los Alamos National Laboratory |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | This requires essentially the same stuff as proposal #366 (reversed Ip and off-axis beams), plus ECH (~3 sources) | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | In reversed Ip and with all NBI injected off axis, attempt to destabilize BAAE-like modes. Done for shape and current profile similar to 142111 (up-down symmetric, L-mode), this should provide a telling comparison with BAAEs in normal Ip that were strongly modified by on vs. off-axis NBI. | ||
| Background: | In the normal Ip discharges mentioned above, BAAEs are destabilized in the late current ramp for both on and off-axis NBI cases. This is different than was observed for RSAEs, which were stabilized by OANBI. Furthermore, the BAAEs during OANBI appear to have had negative frequency in the plasma frame, i.e. they propagate counter to the plasma current. This is peculiar in light of the fact that the phase structure of the mode does not change when the direction of propagation appears to do so. By this we mean that ECEI reveals the same 2D twisting, in the same direction, for both types of modes.
Because these modes are acoustically coupled to the thermal plasma, they are of great interest and very challenging from a physics perspective. We have several hypotheses that may explain the mode behavior we've observed, and there has been great progress in the last year towards developing a code that can produce unstable modes of this kind. Obtaining data for the very different fast ion distribution obtained with counter-injection away from the axis will be invaluable. | ||
| Resource Requirements: | 2 beams off axis, ECH (~3 sources). ECH is key for diagnostic purposes in that BAAEs are often of small amplitude and ECE diagnostics have an easier time observing them when the Te gradient is steepened in the core. | ||
| Diagnostic Requirements: | ECEI, BES, reflectometry | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 368: A lithium capability for DIII-D | ||
| Name: | Jackson | Affiliation: | GA |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | R. Maingi, D. Mansfield, A. Garofalo | ITPA Joint Experiment : | No |
| Description: | The capability to inject lithium into DIII-D would greatly increase our ability to control edge density and recycling and might have many benefits:
1. Reduced recycling to improve VH and QH-mode, Maingi, ROF 41 2. Low edge collisionality in a true ITER shape (lower triangularity than with the DIII-D cryopump) - ELM suppression experiments, (Loarte ROF193) - ITER baseline scenario and steady state work 3. Particle control in ELM-free I-mode 4. ELM pacing 5. Low collisionality Super H-mode |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Install lithium delivery capability on DIII-D | ||
| Background: | Early experiments with the Lithium pellet injector, LPI< (e.g. Jackson, JNM 1996) lacked the capability of depositing sufficient lithium to substantially affect recycling. Deposition rates were usually .le. 10 mg/shot. Subsequent work at PPPL has developed devices to increase lithium 10 x or more above this work (120mg/s with the Li shaker) | ||
| Resource Requirements: | Need the PPPL lithium delivery devices (the "flapper" and the "shaker") | ||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 369: High beta, 2 < q < 3 everywhere | ||
| Name: | Wade | Affiliation: | ORNL |
| Research Area: | Torkil Jensen Award | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Weak magnetic shear is predicted to be favorable for reducing transport. The ultimate incarnation of weak negative shear is a case in which the safety factor is maintained between two low order rational surfaces. With the advent of off-axis NBI and direct feedback of n=1 RWMS, DIII-D is in a position to attempt to see if such a configuration (even tranisently) can be produced and maintained. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Utilize target plasmas from 2012 with off-axis NBI that have qmin > 2 and ramp down the toroidal field until edge q < 3. Will likely need significant time to tune RWM feedback and probably will have to be more aggressive in front-end to maintain qmin > 2 due to the lower toroidal field. | ||
| Background: | |||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 370: Investigate the connection between fast-ion losses and (mitigated) ELM during RMP phase (Dup 371) | ||
| Name: | Chen | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | Energetic Particle Group | ITPA Joint Experiment : | No |
| Description: | ---In AUG, during RMP phase, the level of fluctuations in fast-ion losses is observed to be connected to ELM dimensions (size & frequency).<br> * High level broad-band turbulence of fast-ion losses (only) during mitigated phase<br> * In ELM-mitigated phase, fluctuations in fast-ion losses are comparable to ELM induced fast-ion losses during unmitigated phase.<br> * ELMs seem to be killing turbulent fast-ion losses rather than ejecting ions.<br>[M. Garcia-Munoz, IAEA TM Austin, TX 2011]<br>---No similar results are obtained on DIII-D yet. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This proposal is not asking for dedicated experiment, but interested in piggybacking the experiments with plasma conditions similar to AUG runs:
ELMy H-Mode with 5MW NBI and 1MW ECRH, Bt= -1.5T, Ip=0.8MA, q95=6; And BES, DBS and/or Li-beam are running. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 371: Investigate the connection between fast-ion losses and (mitigated) ELM during RMP phase on DIII-D | ||
| Name: | Chen | Affiliation: | GA |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | Energetic Particle Group | ITPA Joint Experiment : | No |
| Description: | ---In AUG, during RMP phase, the level of fluctuations in fast-ion losses is observed to be connected to ELM dimensions (size & frequency).<br> * High level broad-band turbulence of fast-ion losses (only) during mitigated phase<br> * In ELM-mitigated phase, fluctuations in fast-ion losses are comparable to ELM induced fast-ion losses during unmitigated phase.<br> * ELMs seem to be killing turbulent fast-ion losses rather than ejecting ions.<br>[M. Garcia-Munoz, IAEA TM Austin, TX 2011]<br>---No similar results are obtained on DIII-D yet. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This proposal is not asking for dedicated experiment, but interested in piggybacking the experiments with plasma conditions similar to AUG runs:
ELMy H-Mode with 5MW NBI and 1MW ECRH, Bt= -1.5T, Ip=0.8MA, q95=6; And BES, DBS and/or Li-beam are running. | ||
| Background: | -- | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 372: AE Induced Electron Transport | ||
| Name: | Van Zeeland | Affiliation: | GA |
| Research Area: | Energetic Particles | Presentation time: | Not requested |
| Co-Author(s): | EP Group | ITPA Joint Experiment : | No |
| Description: | The goal of this experiment is to investigate Alfven eigenmode induced electron particle and heat transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This experiment will investigate particle and heat transport through gas puffing and ECH pulses at the plasma edge (in separate discharges) in a series of similar discharges with and without Alfven eigenmodes. The target discharge will begin with the typical DIII-D Alfven eigenmode current ramp discharge in which many RSAEs and TAEs are observed. In a series of discharges, the level of Alfvenic activity will be reduced by several techniques. First, the beam injection timing will be delayed by ~150-200 ms which has been observed to reduce the level of modes. Second, on-axis heating will be replaced with off-axis beam heating (150s tilted to 16.4 Deg.) which was also observed to significantly reduce the drive for modes near mid-radius. Finally, beam voltage will be reduced while maintaining constant power. Particle and heat pulse transport will be monitored via reflectometry and ECE respectively. In all discharges, the AEs and broadband fluctuations will be monitored via BES, DBS, and reflectometry as well as ECE and ECEI. Also, beam modulation will be kept to a minimum to avoid additional heat and particle source oscillations. | ||
| Background: | Alfven eigenmodes are common in DIII-D plasmas and their ability to cause fast ion transport is well documented. The impact of these modes on the electron particle and heat transport has not been investigated on DIII-D and only recently been addressed on other devices, most notably NSTX, where experiments (Gorelenkov, Tritz) have shown the ability of GAEs to cause electron transport. One technique to investigate electron transport is through modulation of particle and/or heat sources. In DIII-D, particle transport has been investigated by repetitively puffing gas at the plasma edge and diagnosing the inward propagating density pulse with reflectometry. Heat transport has been studied by pulsing localized ECH and observing the heat pulse propagation with ECE. | ||
| Resource Requirements: | All beams except 210L, ECH | ||
| Diagnostic Requirements: | Reflectometry, FIDA, FILD, BES, ECE,ECEI | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 373: gyro-kinetic L-mode | ||
| Name: | Staebler | Affiliation: | GA |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | Yes | |
| Description: | in order to create an L-mode plasma that oly has gyro-kinetic turbulence we propose to use high ECH heating in a very low density (1.5x10^13/cm^3) L-mode. This has been done on ASDEX-U. The electron temperature profile looks like an H-mode and in fact does not change at the L/H transition. Using the complete diagnostic and fluctuation measurements on DIII-D to study this regime could shed light on the nature of the missing L-mode transport. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Reproduce the ASDEX-U discharge and measure it to death. | ||
| Background: | All L-mode that have been simulated with GYRO or TGLF to date have underpredicted transport in the core/edge transition region (r/a>0.7). If we can make a discharge where gyro-kinetic transport is sufficeint it would help to establish the reason why it is not under other conditions. The ASDEX-U case looks like a good candidate for an gyro-kinetic L-mode. | ||
| Resource Requirements: | 3MW ECH +some NBI | ||
| Diagnostic Requirements: | all possible fluctuations. CER, Thompson, ECE, MSE | ||
| Analysis Requirements: | Transport analysis with TGLF and GYRO | ||
| Other Requirements: | |||
| Title | 374: Examine the Dependence of broadband ion temperature and density fluctuations on eta_i | ||
| Name: | Uzun-Kaymak | Affiliation: | METU - Middle East Technical U |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | R. Fonck, G. McKee | ITPA Joint Experiment : | No |
| Description: | Investigate the eta_i dependence of localized density and ion thermal fluctuations by varying the ion temperature scale length at a nearly constant density profile. We will develop experimental conditions that maximize the fluctuation signal levels observed with UF-CHERS. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish long-pulse, low current, low density, USN or inner wall limited L-mode discharges to make careful measurements of fluctuation multifield characteristics. Vary the ion temperature profile, and thus gradient using NBI power scans while maintaining the toroidal rotation velocity nearly constant by utilizing a mixture of co and counter current neutral beams. These discharges will be long pulsed to improve the signal-to-noise for photon noise limited measurements ion temperature fluctuations. To get a good background emission comparison, these shots will be repeated with the 150L/R beam off but at the same total beam power. In addition to a power scan, repeat these shots at two different density regimes to provide a range of eta_i to study turbulence driven transport models. Repeat discharges will be used for a spatial scan. Deuterated methane injection will be considered to increase the carbon content for higher SNR. ECH will also be employed to vary Te/Ti, control rotation and density as well as to increase carbon content. | ||
| Background: | It is expected that as the normalized ion temperature gradient scale length (R/L_T,i) is increased, heat flux will increase rapidly as the critical gradient is exceeded. Fluctuations are likely to increase as well, though temperature and density fluctuations should respond differently. Furthermore, linear eigenmode calculations predict that as the gradient scale length ratio is reduced, the nature of the modes may change from ITG dominated to TEM.
In this experiment, the primary goal is to obtain experimental conditions to measure ion thermal fluctuations with UF-CHERS. The UF-CHERS is designed to measure broadband low-k ion temperature fluctuation as well as the toroidal rotation velocity fluctuations at 1 us time resolution. These shots are particularly tuned for low rotation velocity, stationary discharges to obtain adequate signal to noise for the fluctuation data. Scans of collisionality and temperature will be employed to investigate the turbulence characteristics. These measurements will allow us to quantify the low-k turbulence and n-Ti phase correlations across the plasma mid-radii. They will be also used for verification and validation purposes. The UF-CHERS diagnostic has obtained prototype measurements of broadband ion temperature and toroidal velocity fluctuations with a two-channel correlated measurement; cross-correlation between UF-CHERS and BES (n-Ti) yielded improved signal. An upgraded dual-channel UF-CHERS detector array will be deployed for 2012 DIII-D operations which will provide significantly improved SNR for small-amplitude fluctuation measurements. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Fluctuation diagnostics | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 375: What causes density buildup at the pedestal top? | ||
| Name: | Callen | Affiliation: | U of Wisconsin |
| Research Area: | Pedestal | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | The basic proposal is to reduce the core density fueling from neutral beams to see if this can reduce (further) the slow rate of rise of the density at the pedestal top long after (> 10 ms) an L-H transition, or perhaps between long period ELMs. It is proposed to do this by exchanging NBI heating with ECH while keeping the overall core electron heating in the plasma approximately constant. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This type of experiment needs to achieve an L-H transition with relatively low NBI (or ECH?) power. Then the heating power should be reduced to the smallest possible level that just barely keeps the plasma in an H-mode, which apparently allows a long slow buildup of the pedestal top after an L-H transition and low frequency ELMs. This is the scenario used for the Groebner at al, NF 49, 045013 (2009) shots and for obtaining long period ELMs this past year (2011 experimental campaign).
The new element would be to vary the ratio of the NBI to ECH power and thus the amount of core fueling during the long slow buildup of the top of the pedestal. It would be critical to deposit the ECH near the plasma center to avoid or minimize ECH-induced density pump-out. To facilitate comparisons with paleoclassical pinch-type processes it would be best to operate at a plasma current and toroidal field where the paleoclassical processes are most likely dominant -- >~ 1 MA and ~ 2T? Also, it would probably be important to keep the electron power flow from the core into the pedestal region relatively constant -- so the paleoclassical transport in at least the steep gradient region of the pedestal is nearly constant. The following ROFs propose related or similar types of experiments: 045 (Lore), 189 (Canik), 326 (Osborne), 361 (Yan). | ||
| Background: | H-mode pedestals always seem to grow in height and width until they reach the peeling-ballooning stability boundary and precipitate an ELM. The cause of this growth to the P-B boundary is not understood. If it was understood, we might be able to develop measures to impede it and thereby prevent ELMs.
Pedestals seem to reach a transport quasi-equilibrium about 10 ms after an L-H transition or an ELM [Groebner at al, NF 49, 045013 (2009)]. By this time the steep gradient region of the pedestal (rho > 0.96) has saturated. However, the top of the pedestal keeps growing slowly in magnitude and inward. The DIII-D data shows that it is mainly the electron density at the top of the pedestal that increases. The T_e at the pedestal top saturates fairly quickly while the density at the pedestal top continues to grow for times up to 100s of ms until an ELM is precipitated. Recent results with the upgraded Thomson system (Osborne et al, 2011 H-mode workshop) indicate the nearly flat pedestal top density profile continues to slowly grow about linearly in magnitude with the knee of the density profile monotonically moving slowly inward Thus, the key physics question is apparently: what causes the density at the pedestal top (rho <~0.96) to increase slowly. A "back of the envelope" estimate indicates the rate of growth of the density at the pedestal top is of the order of the fueling rate of neutral beams in the plasma core. Could it be that the pedestal region is providing a "perfect" transport barrier and the plasma retains all the core NBI fueling? Or is there a weak density pinch at the top of the pedestal that causes the continuing pedestal density growth by allowing the pedestal/plasma to "suck up" all the available plasma fueling from outside the separatrix? Or perhaps there are other possibilities -- very slowly evolving changes in the turbulence-induced density transport at the pedestal top? Resolving which it is could make an important contribution to understanding pedestals. It might also indicate how ELMs could be prevented or slowed down in their evolution to the P-B stability boundary and thereby suppress or delay ELMs. | ||
| Resource Requirements: | Mainly a very slowly evolving H-mode pedestal with sufficient, but not too much ECH and NBI power -- just a few MW. Then vary the mix of ECH versus NBI heating to see if the NBI core fueling is responsible for the rate of rise of the electron density at the top of the pedestal. | ||
| Diagnostic Requirements: | Good Thomson measurements of the long slow evolution of the top of the pedestal. Also, it would be good to get data from diverter region diagnostics to facilitate quantitative modeling of the neutral density using SOLPS or other codes. Fluctuation diagnostics that could measure net density transport fluxes would also be useful. | ||
| Analysis Requirements: | ONETWO interpretive analysis and probably SOLPS for estimating the neutral fueling at the top of the pedestal. | ||
| Other Requirements: | -- | ||
| Title | 376: Scaling Of Pedestal Plasma Transport With RMP I-coil Current | ||
| Name: | Callen | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | -- | 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 an 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-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 ONETWO, TRANSP, SOLPS etc.) 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 slowly, or perhaps in steps, throughout a given discharge. In addition to fiducial discharges with no RMPs, the I-coil current could also be initiated at a low level (2 kA?) and slowly increased into an RMP-suppressed regime. It would be important to look for any abrupt changes in the plasma transport characteristics in the pedestal region, particularly the carbon toroidal rotation there, as the ELM suppression I-coil current threshold is exceeded. Also, diagnostic measurements (e.g., by ECEI, DBS and BES) that could be used to look for island-like signatures or increased kink-like distortions near Psi_N ~ 0.95 at these I-coil thresholds would be helpful.
The following ROFs propose related or similar types of experiments: 14 (Maingi), 61 (Boedo), 363 (Solomon). | ||
| Background: | The best and most 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 the many papers on pedestal density transport that have resulted from Saskia Mordijck's Ph.D. thesis (at UCSD, January 2011). It has also has been very useful in the development of the magnetic flutter model of RMP-induced pedestal plasma transport (Callen et al., "RMP effects on pedestal structure and ELMs," UW-CPTC 11-13, 19 December 2011, available via http://www.cptc.wisc.edu). 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 ONETWO, SOLPS, TRANSP etc. | ||
| Resource Requirements: | Mainly a set of similar parameter RMP-suppressed discharges are needed with increasing I-coil currents from 0 (for fiducial discharges) to over 6 kA in small steps -- say in units of 1 kA. 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, fast T_e response from ECE and CER carbon rotation measurements of the plasma parameters at the pedestal top (0.85 < 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. Fluctuation diagnostics that could measure net density and electron heat transport fluxes would also be useful. | ||
| Analysis Requirements: | ONETWO, TRANSP interpretive analysis plus perhaps iterative SOLPS modeling will be needed for estimating the changes in the electron density and thermal transport plus plasma toroidal rotation as a function of the I-coil current in RMP-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 | 377: Steroscopic imaging of pellet ablation during pellet ELM pacing | ||
| Name: | Unterberg | Affiliation: | ORNL |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | L. Baylor, N. Commaux, S. Allen, M. VanZeeland, R. Moyer | ITPA Joint Experiment : | No |
| Description: | 3-D visualization of pellets entering the plasma have been used on other devices [1,2] to help localize the pellet ablation process. This information can be used to determine/confirm the pellet trajectory in 3D as well as the shape of the ablation cloud in 2D. <br> <br>[1] R. Sakamoto and H. Yamada, Rev. Sci. Instrum. 76 (2005) 103502. <br>[2] B. Bose et al., 51th APS-DPP (2009), http://meetings.aps.org/link/BAPS.2009.DPP.GO4.4. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This effort will be mostly piggyback measurements on planned pellet experiments this year. | ||
| Background: | Pellet ELM pacing was shown successful in the last run campaign to increase the ELM frequency by ~ x10 and decrease the ELM magnitude in impulse particle and heat flux significantly. This is a very promising technique to control ELMs because it is potentially independent of the discharge scenario (unlike, to-date, QH-mode or RMP ELM suppression). Understanding of the pellet ablation process will help project this technique to other machines. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | Fast cameras (e.g. the VI Phantoms) at two locations -- ideally the new LLNL periscope location and the 225R-1 location. | ||
| Analysis Requirements: | Transform techniques to triangulate the images of two cameras. Potentially pellet ablation code analysis from JOREK or similar. | ||
| Other Requirements: | -- | ||
| Title | 378: Boronization during plasma discharges | ||
| Name: | Rudakov | Affiliation: | UCSD |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | O. Buzhinskij, R. Doerner, C.P.C. Wong | ITPA Joint Experiment : | No |
| Description: | This experiment will attempt boronization during plasma discharges in DIII-D using non-toxic, non-explosive metacarborane C2H12B10. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Inject metactrborane near attached OSP in LSN L-mode discharges. Monitor effect on the discharge parameters. Use DiMES to study B deposition. | ||
| Background: | Non-toxic, non-explosive metacarborane C2H12B10 was successfully used in T-11 for boronization during a plasma discharge. Tests have also been conducted in PISCES | ||
| Resource Requirements: | DiMES, LSN L-mode with OSP near DiMES, 1/2 day experiment | ||
| Diagnostic Requirements: | All edge diagnostics, SPRED, CER, core and divertor Thomson | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 379: Parametric Dependence of Turbulence and Transport on rho-star in low-rotation H-mode plasmas | ||
| Name: | McKee | Affiliation: | U of Wisconsin |
| Research Area: | Transport and Rotation | Presentation time: | Not requested |
| Co-Author(s): | C. Holland, L. Schmitz, T. Rhodes, G. Wang, A. White, Z. Yan | ITPA Joint Experiment : | No |
| Description: | Perform a rho* scan (rho*=rho_I/a, normalized ion gyroradius) in moderately low-rotation hybrid H-mode plasmas for validation studies, as well as to obtain data set for future similarity aspect ratio scan with NSTX-U (where similar rho* scan to be proposed and performed). | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Establish a low-current (~1 MA/2 T) hybrid H-mode discharge as low-rho* condition; high rho* would be 0.5 MA/1T). Hybrids are desirable for their long duration and lack of sawteeth (142019 could be a reference).
These experiments will also support validation efforts by comparing measured turbulence/transport response with predictions from TGLF, GYRO and other codes. (more details to be provided, this is more or less of a placeholder...) | ||
| Background: | Basic characteristics of turbulence scale with rho* in a predictable way according to gyrokinetic equations (e.g., correlation lengths scale with rho_I, normalized turbulence intensity, ñ/n, scales with rho*, and decorrelation rates scale as a/c_s). Previous L-mode scaling experiments (ca, 2000) showed this dependence to hold (McKee et al., Nuclear Fusion 41, 1235 (2001)). This experiment would seek a similar comparison in H-mode discharges, where transport scaling is not the same as in L-mode (e.g., ion thermal transport appears gyro-Bohm, while in L-mode looks Bohm like based on previous work by Petty & Luce). Compare with GYRO/TGLF will be essential. This will feed into the 2012 JRT, if performed. | ||
| Resource Requirements: | All NBI, ECH | ||
| Diagnostic Requirements: | BES (8x8), UF-CHERS, DBS, CECE, FIR, PCI, B~ interferometer (J. Zhang), etc. | ||
| Analysis Requirements: | TGLF, GYRO, ... | ||
| Other Requirements: | |||
| Title | 380: Dependence of ICRF antenna loading on ELM frequency, type and size | ||
| Name: | Diem | Affiliation: | U of Wisconsin |
| Research Area: | Steady State, Heating and Current Drive | Presentation time: | Not requested |
| Co-Author(s): | E. Unterberg, P. Ryan, M. Murakami, E.F. Jaeger, D. Green, D. Rasmussen, M. Kaufman, R. Pinsker, A. Nagy | ITPA Joint Experiment : | No |
| Description: | An ITER requirement for the ICRF antenna is to couple 20 MW while dealing with edge localized modes (ELMs), which can result in very fast changes in the ICRF antenna loading [1]. The goal of this proposal is to investigate the impact of edge profile modifications due to ELMs on fast wave antenna loading. How the antenna loading changes with ELM frequency, size and type will be investigated.
[1] D.W. Swain, R. Goulding, â??ITER Ion Cyclotron System: Overview and plansâ??, Fusion Engineering and Design Volume 82, Issue 5-14, pages 603-609, October 2007. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | This will be a piggyback experiment. It will require low power injection (~Watts) from the fast wave antennas at the end of ELMy discharges. Specifically interested in piggybacking at the end of ELM-pacing experiments with pellet injection and 3D field ELM modification experiments. | ||
| Background: | See description. | ||
| Resource Requirements: | Three FW systems, one at 60 MHz and two at 90 MHz. This experiment will require low power injection (~Watts) at the end of a discharge. | ||
| Diagnostic Requirements: | The edge reflectometer adjacent to the 285/300 FW antenna in addition to the UCLA profile reflectometers would aid in this experiment. | ||
| Analysis Requirements: | Edge profile analysis will be performed as well as a comparison of the loading measurements to AORSA calculations. | ||
| Other Requirements: | |||
| Title | 381: Dependence of upstream Te on target plate potential | ||
| Name: | Watkins | Affiliation: | Sandia National Lab |
| Research Area: | SOL Physics | Presentation time: | Not requested |
| Co-Author(s): | Stangeby
Rudakov |
ITPA Joint Experiment : | No |
| Description: | In a carefully constructed magnetic field geometry, use the divertor Thomson to measure upstream Te as the target plate potential is changed at the Langmuir probe. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The divertor Thomson laser pulse would have to be phased with the Langmuir probe bias programming such that the upstream Te could be measured on the same field line with and without bias. To have a better chance of alignment, the current or toroidal field could be ramped slowly near the desired value with the strike point near the divertor Thomson radius of 1.49 m. | ||
| Background: | Target plate heat flux calculated from Langmuir probe particle flux and electron temperature agrees with the IR camera heat flux when using a theoretical value of 7 for the sheath power transmission factor in the far SOL but drops to ~1 near the strike point. One theory is that the electrons in the Langmuir probe flux tube are reflected back upstream by the negative bias resulting in a higher upstream temperature. | ||
| Resource Requirements: | very stable plasma shape | ||
| Diagnostic Requirements: | divertor Thomson
divertor Langmuir probes DIMES special probe | ||
| Analysis Requirements: | |||
| Other Requirements: | need to make special DIMES collector | ||
| Title | 382: xpt gas puff for ELM control | ||
| Name: | Watkins | Affiliation: | Sandia National Lab |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | By locally fueling the core plasma in a flux expanded region near the primary or secondary x-point, it may be possible to form a localized pressure gradient that will trigger rapid ELMs in the flux compressed region at the outer midplane. The key idea is that in the flux expanded region, the fueling is much more localized in flux space and would allow very narrow local pressure gradients at the outer midplane high enough to trigger small rapid ELMs and avoid the larger and more damaging type 1 ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | make a shape conforming to the upper baffle and puff small amounts of gas directly into the x-point. Reverse Bt may be necessary. | ||
| Background: | Rapid ELMs have been observed when placing the outer strike point on the upper dome which moved the neutral recycling source very near the xpt. The effect appeared to be an xpt fueling effect but no dedicated investigation has been performed. | ||
| Resource Requirements: | dedicated and calibrated gas controller for the dome xpt gas line | ||
| Diagnostic Requirements: | upper xpt camera | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 383: "Catching" rotating modes with rotating fields before locking | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Apply rotating magnetic perturbations (MPs) to the rotating precursor of a locked mode to ??catch it? and entrain it while it slows down. After the mode locks to the rotating MP, the rotation can either be kept constant, at a safely high level to avoid locking to the static error field or to the walls, which is one of the main causes of disruptions, or accelerated. As a result the mode would accelerate too, and be stabilized by rotational shielding and rotation shear. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | As soon as mode rotation frequency f<1kHz (dud), apply intense (I-coil current 1.5kA) n=1 traveling wave at frequency <=f, to account for dwell time, then slow down at the same rate as the mode. Can be pre-programmed or in f/back (rtnewspec). Alternatively, use magnetic feedback, for the I-coils to feed back on Mirnov. While the mode slows down, reduce its amplitude accordingly, to account for reduced rotational shielding.
If rotating mode locks to rotating field at time t, repeat with pre-programmed changes in MP rotation after t, for example keep the rotation steady, or accelerate it again. | ||
| Background: | So far, MPs successfully controlled initially locked modes at DIII-D. Recently there has also been some success in "catching" with a 20Hz rotating field a rotating precursor which was slowing down from few kHz to 0Hz. This generated promising results when combined with ECCD modulated in phase and in synch with the driven rotation. However, per se (without ECCD), a mode locked to a MP rotating at 20Hz is not different from a Quasi-Stationary Mode (QSM) and not very different from a mode locked to the static Error Field: the only important advantage is the toroidal spread of localized effects such as heat loads. Here we propose to pre-emptively apply sufficiently intense MPs rotating at higher frequencies (>100Hz), which would avoid locking altogether and rotationally stabilize the mode. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | PCS changes involving real-time newspec and/or adaptation of magnetic feedback to NTMs | ||
| Title | 384: Operate at optimal q95 and simply turn n=3 RMPs on/off to pace ELMs | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | -- | ITPA Joint Experiment : | No |
| Description: | Operate at optimal q95 and simply turn n=3 RMPs on/off to pace ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Recreate standard ELM control discharge. Find broadest q95 window and repeat shot with q95 fixed in the middle of that window. Simply modulate RMPs on/off. Repeat for faster and faster modulation, to cause shorter and shorter ELM-ing and ELM-free periods. If/when arrived to a single ELM or only a few ELMs per period, change duty cycle to change the inter-ELM period.
Apart from representing a potential new ELM-pacing technique, this experiment would improve our understanding of how rapidly the plasma edge gradients, currents and transport respond to the RMPs. Modulation will allow high-quality measurements by lock-in and coherent averaging analysis techniques. Finally, note modulating the edge gradients is interesting for transport studies in general, regardless of ELM control, in that it would test transport models, TGLF23 simulations and ETG and ITG theory. For this reason, in addition to the discharges described above, that would alternate ELM-ing and ELM-free periods, it is proposed to take some full-modulation shots at non-optimal q95 (always ELM-ing) and some discharges at optimal q95 in which the RMPs are not turned off completely, but only modulated in amplitude (ELM-free). Differentiating between two ELM-ing periods or two ELM-free periods will give the transport coefficients in the ELM-ing and the ELM-free plasma, respectively. | ||
| Background: | ELMs are not always undesirable: small, frequent "grassy" ELMs might help control the plasma density and Helium ashes without affecting too severely the tokamak walls. Learning how to "pace" ELMs would be sufficient in this respect, given the inverse proportionality observed at AUG and other machines between the ELM frequency and the energy that they carry.
Modulation of RMP control is not new, but the proposal here is to force it to the ??natural? inter-ELM period and below, for faster, smaller ELMs. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | BES for n~, CECE for T~ and fast magnetics for B~. They all showed correlation or anti-correlation with the RMPs. | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 385: Modulated central ECH for heat transport studies during ELMs and ELM-control | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Modulate central ECH to generate heat waves. Use ECE to study their propagation across the plasma, in particular across the edge. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Modulate the ECH at 8-80Hz in: 1) an ELM-suppressed discharge at optimal q95, 2) an ELMy one at non-optimal q95, 3) an ELMy discharge with the RMPs turned off, 4) a q95 ramp and, if time, 5) an ELM-free H-mode without RMPs. In the latter, use 5-6 gyrotrons to trigger the ELM-free H-mode and 1-2 to generate the heat waves. | ||
| Background: | The Pulse-height analysis of "heat waves" generated by time-modulated ECH is a well-known transport analysis technique providing the electron heat diffusivity profile and other transport coefficients. It is proposed to use it to document the heat transport in the stochastic edge of DIII-D during in ELM-suppressed discharges s well as, for comparison, in ELM-ing ones. It will be interesting to diagnose how q95 affects the heat transport, and compare with particle transport (density pump-out). | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | Height Pulse Analysis (HPA) | ||
| Other Requirements: | |||
| Title | 386: Oblique-ECE-assisted MECCD suppression of 2/1 NTM | ||
| Name: | Truong | Affiliation: | Sandia National Lab |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | F. Volpe, M. Austin, R. La Haye, R. Prater, A. Welander | ITPA Joint Experiment : | No |
| Description: | Use oblique ECE to radially align and correctly modulate ECCD in 2/1 island. Stagger poloidally the launching directions of the various gyrotrons to reproduce ITER-like broad deposition and further enhance the relative benefits of modulation. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | -Dial up a discharge with a 2/1 NTM rotating at f<5kHz, for example #135861.
-Interface oblique ECE and ECCD as in #132113 -Add new oblique ECE capabilities developed at UW-Madison: *new ultra-low-noise video-amplifiers *new broader-band phase-shifter, to compensate for the oblique ECE being collected and ECCD being injected at different locations *(if ready) new radiometer with lower noise and more channels (16 instead of 2) *(if ready) new hardware ECE-Mirnov correlator. Mirnov will provide the correct frequency, oblique ECE the correct phase. -Compare *narrow vs. broad ECCD *modulated vs. continuous *deposition in the O-point, X-point and in between. | ||
| Background: | The system has been already applied with success to the alignment and modulation of narrow ECCD to a rotating 3/2 island, resulting in its complete stabilization and in saving 30% of average power compared to continuous ECCD. Further improvements, recognizable also in a reduced demand of peak power, are expected for broad ECCD.
So far, the more malicious 2/1 mode has been successfully â??trackedâ?? at TEXTOR but has never been stabilized by modulated ECCD, neither using Mirnov drive, nor oblique ECE. It will be important to do this for the first time, on the way to ITER. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 387: Fine scale stepwise q-scan to determine role of low-order rationales in ELM suppression | ||
| Name: | McKee | Affiliation: | U of Wisconsin |
| Research Area: | ELM Control: 3-D Field Induced Transport | Presentation time: | Not requested |
| Co-Author(s): | Evans, Nazikian, Rhodes, Schmitz, Wade, Yan | ITPA Joint Experiment : | No |
| Description: | Determine how transport and turbulence in the near pedestal region vary as q95 is varied in very small increments inside and outside of the ELM suppression windows, 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 characteristics. BES and DBS would be used to measure turbulence characteristics (150L flat is desired beam configuration, 150R not on) | ||
| Background: | During q95/Ip ramp discharges in RMP ELM-control experiments (e.g., 145019), 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-coil in n=3 configuration, 150L beam@0º | ||
| Diagnostic Requirements: | BES, DBS, PCI, CECE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 388: ELM-pacing by pre-programmed modulated ECH/ECCD | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Demonstrate that modulated ECH/ECCD can periodically destabilize the edge pressure and gradient, and so trigger ELMs. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Begin by setting a value of rho which is known or expected to maximize ECH/ECCD effect on ELMs. Modulate ECH/ECCD 10% slower/faster and a factor of 2 slower/faster than the natural ELM frequency. Compare perpendicular, co- and ctr- launch.
Initial tests can piggyback on RMP control of ELMs. ECH pulses of 200ms towards the end of the discharges are brief enough not to damage diagnostics and other equipment in case absorption is only partial and stray is high. At the same time, they are long enough to see effects on ELMs, if any, including slow changes following transport and relaxation phenomena. Dedicated run-time could be shared with ELM control experiments deploying continuous ECH/ECCD, by adding modulation at the end of the continuous pulses. The advantage of exclusively dedicated time, not shared with others, is that different ECH frequencies can be tested in the same discharge, at different times. Finally, if ELM-frequency control succeeds, the other important parameter to control is the energy loss per ELM. The knobs in this case are the radius of ECH deposition and the amount of driven current. This is why, as a final experiment we propose to fix the ECH frequency and scan, on a shot-to-shot basis, the radius of deposition and the toroidal direction of launch. | ||
| Background: | In 2004, AUG modulated ECH at 100Hz at the edge of an ELM-ing plasma. The ELM frequency, initially of 150Hz, changed accordingly. Also, as expected, the ELMs slightly intensified. The opposite change, i.e. making the ELMs smaller and more frequent, has not been demonstrated yet, and would be highly desirable for ITER. Moreover, in AUG the ECH effects were suspected to dominate over ECCD, but the two were never really disentangled. Although seminal, AUG results leave much room for improvement and for the first demonstration of ELM-pacing by modulated ECH/ECCD at HIGHER frequencies. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 389: ELM-suppression by ECH/ECCD modulated in the rotating ELM filament, in f/back with D_alpha | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | ELM Control | Presentation time: | Not requested |
| Co-Author(s): | ITPA Joint Experiment : | No | |
| Description: | Modulate ECH/ECCD in phase and in synch with the rotating ELM filament, in search for enhanced stabilization. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Consider a D_alpha or other diagnostic of ELMs. If necessary, change optics to narrow its view and resolve one ELM filament at the time. As the ELM filament rotates, it modulates this D_alpha signal. The latter can be used as a driver for ECH/ECCD modulation in synch and in phase with the ELM, similar to oblique ECE for NTMs. The same electronics which interfaced the oblique ECE to the gyrotrons can be adapted to this purpose. | ||
| Background: | ECCD modulated by Mirnov probes at AUG and by oblique ECE at DIII-D in phase and in synch with a rotating islands has been effective in stabilizing 3/2 NTMs.
On the other hand, continuous ECH has been shown to affect, and in some cases completely stabilize ELMs at DIII-D, AUG and JT-60. Fusing these results, the present proposal intends to investigate the possible benefits of modulating the ECH/ECCD in synch with the rotating ELM filament. The scope is to selectively pump-down the ELM filament, or drive a current in it, or heat the space in between two filaments. The idea is that, by doing so, one might apply a perturbation equal and opposite to the ELM, and directly suppress it, similar to the ECCD compensating for the missing bootstrap current in a neoclassical island. The cw ECH/ECCD approach, instead, aims at making the plasma less unstable. In other words, it moves j_par and/or grad P away from the peeling-ballooning stability boundary. It removes the unstable condition, it doesnâ??t suppress the instability. The downside is a cost in plasma performances. Conversely, active control of the instability enables operation in a nominally unstable, possibly higher performance region. Modulated ECH is more likely to have an effect, but modulated co- and ctr-ECCD should be tried too, especially on considering that ELM filaments have been demonstrated to carry current. ECCD might enhance or reduce these currents, much like it compensates for the bootstrap current deficit in neoclassical islands. Finally, ECCD at extreme radii or even outside the separatrix, might affect, possibly cancel the SOL currents. As a bonus, the method also has a potential as an indirect, comparative diagnostic of SOL currents, provided ne and Te in the SOL are known and ECCD can be calculated. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | Change optics in front of a D_alpha filterscope to narrow its view and resolve single ELM filaments. Filaments are clearly visible in UCSD Phantom camera, which, however, buffers data and cannot transfer them in real-time. Acquire D_alpha in real-time and pass signals to oblique-ECE â??boxâ?? in the annex, with modified voltage thresholds and pass-bands. | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 390: Add modulated ECCD to magnetic feedback control of locked modes | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | M. Okabayashi | ITPA Joint Experiment : | No |
| Description: | Completely suppress by means of modulated ECCD a locked mode that is only unlocked or prevented to lock, and forced to rotate, by magnetic feedback. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Add modulated ECCD to M. Okabayashi's discharges for #237. Modulation can be pre-programmed, if mode rotation is uniform and reproducible. Simple repetitions of the discharge and small shot-to-shot differences would give an automatic scan of the relative phase between the modulated ECCD and the rotating island. Alternatively, modulation can feed back on magnetics. In fact, it can adopt the I-coil real-time waveforms. For this purpose, connect magnetic-feedback real-time computer to ECH opto-isolators in the annex: apart from imposing 0-10V, the waveforms are the same! Choosing the real-time waveform for one set of I-coils or the other to feed the gyrotrons will also scan the relative phase. | ||
| Background: | Previous results by M. Okabayashi showed that magnetic f/back can either unlock a 2/1 mode initially locked to the EF, or prevent locking altogether. In both cases the mode rotated at 15-40 Hz and had a finite amplitude, comparable with a perfectly locked (0 Hz) mode. Hence in many respects, such as confinement and beta, this slowly rotating mode is as bad as a locked one. ECH/ECCD has the potential to fully suppress the mode. For best results, ECCD should be modulated in phase with the transit of the island O-point in the ECCD deposition region. Otherwise, for continuous ECCD and such a slow mode rotation, the X-point might be exposed to ECCD for too long, resulting in destabilization. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
| Title | 391: ECCD modulated by horizontal ECE | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | M. Austin, R. La Haye, D. Truong, A. Welander | ITPA Joint Experiment : | No |
| Description: | Show that horizontal ECE can replace oblique ECE as a driver for ECCD modulation in phase with NTM O-point. Coincidentally, no phase correction is required for the 2/1 mode, because the ECE and the ECCD happen to be about 180deg out-of-phase. Some correction â??to optimize experimentally- is required for the 3/2 mode. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | The approach will be similar to shots where ECCD was modulated by oblique ECE, e.g. #132113, except that the analogue interface in the annex (â??the boxâ??) will read the horizontal ECE signals.
The horizontal ECE is located at phi=60deg, i.e. approximately 180deg apart from the gyrotrons (phi=240-270deg) and thus the ECCD deposition regions (which differ from the launching positions by <10deg). Hence, the horizontal ECE is measured in a ideal position from which it can directly modulate the ECCD in phase with a rotating 2/1 mode, with no need for phase correction. To stabilize a 3/2 mode, instead, we will need the phase-shifter already used with success in the past, or the new shifter with flatter response which we recently developed. | ||
| Background: | Horizontal and oblique ECE have the advantage, over Mirnov probes, of being local, internal diagnostics of NTMs. This simplifies the phase correction when they are used to modulate the ECCD in phase with a rotating island. Oblique ECE simplifies this phase correction even more than horizontal ECE, if collected along the ECCD launch direction, or an equivalent one.
In the last campaign, the new oblique ECE radiometer was successfully interfaced, by an analogue circuit in the annex, to the gyrotron power supplies, and ECCD was modulated in synch and in phase with the O-point of a rotating 3/2 NTM. Complete stabilization was obtained. The analogue interface worked very reliably and correctly manipulated the signals that it was receiving from the radiometer. Those signals, however, were not perfect NTM measurements, partly because of intrinsic reasons (ECE is sensitive to all Te fluctuations, not just to NTMs) and partly because of the signal-to-noise ratio of the radiometer, which needs to be improved. While these improvements are under way (a new oblique ECE radiometer is under construction at UW-Madison), we suggest to connect the analogue interface to the horizontal ECE. Its superior signal-to-noise ratio will make up for the slightly more difficult phase correction. Should it work, it would be a more robust, reliable and easy-to-use driver for ECCD modulation: more robust and reliable because horizontal ECE has been tested for years, is one of the main DIII-D diagnostics, regularly maintained, easier to use because it is always available and does not require various shutters and an optical switch to be opened or closed, as in the case of oblique ECE, which shares a transmission line and a launcher (receiver) with one of the gyrotrons. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | |||
| Analysis Requirements: | |||
| Other Requirements: | Cables or optical links between horizontal ECE radiometer and annex, where same electronics formerly used for oblique ECE will interface it to gyrotrons. | ||
| Title | 392: Slow driven rotation of islands, with diagnostic and modulated ECCD applications | ||
| Name: | Volpe | Affiliation: | Columbia U |
| Research Area: | Stability and Disruption Avoidance | Presentation time: | Not requested |
| Co-Author(s): | R. La Haye, J. Hanson | ITPA Joint Experiment : | No |
| Description: | Apply rotating n=1 error field to unlock and spin up to 200-1000Hz an initially locked mode. Resolve forcefully rotated islands with diagnostics such as CER and MSE which normally have too little temporal resolution. Later in the same discharges, add ECCD modulated at the frequency of driven rotation. Scan relative phase (between ECCD and I-coils) from shot to shot. For O-point phasing, it should be stabilizing. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Use dud detectors to trigger an I-coil travelling wave right after mode locking. Pre-program ramps of frequency and amplitude of I-coil current in such a way that the strength of the magnetic perturbation at the island location remains approximately the same, even at high frequency (i.e., for high shielding from the wall). Optimize ramps and clamp them at highest frequency at which the rotating perturbation is still coupled with the mode, without phase-slipping. Collect CER and MSE data. Later in the same discharges, add ECCD modulated at the frequency of driven rotation. Scan relative phase (between ECCD and I-coils) from shot to shot. For O-point phasing, it should be stabilizing. | ||
| Background: | Sustained rotation (a.k.a. entrainment) has been already demonstrated at DIII-D and elsewhere, at frequencies of the order of ~100Hz. These frequencies are attractive for diagnostic purposes, to coherently average island measurements with various diagnostics, resulting in high-precision measurements. Furthermore, this would enable measurements with relatively slow diagnostics such as CER and MSE. Finally, preliminary stabilization results have been obtained with ECCD modulated at the same frequency as the rotating perturbation, but more work is needed to investigate the potential and ITER attractiveness of this stabilization technique. The technique combines the efficiency of modulated ECCD with the ease of modulation at a pre-determined frequency. In other words, the island rotation is adapted to the ECCD modulation, and not vice versa. | ||
| Resource Requirements: | |||
| Diagnostic Requirements: | CER, MSE | ||
| Analysis Requirements: | |||
| Other Requirements: | |||
| Title | 393: ITER Langmuir probe tests | ||
| Name: | Watkins | Affiliation: | Sandia National Lab |
| Research Area: | Plasma-material Interface | Presentation time: | Not requested |
| Co-Author(s): | Guangwu Zhong, Southwestern Institute of Physics, Chengdu, China, Dean Buchenauer, SNL, David Donovan, SNL, Clement Wong, GA, Peter Stangeby, Univ. Toronto | ITPA Joint Experiment : | No |
| Description: | There are a couple of issues with the ITER Langmuir probe design that can be tested on DIII-D using the DIMES platform. The first issue involves the performance of annealed pyrolytic graphite (APG) for Langmuir probe tips. Also, the orientation of the highly conducting planes may , for design reasons, need to be different that the orientation used on the DIII-D pyrolytic graphite probe tips. The mechanical strength of the APG is another part of this issue. The second issue involves the design of the probe tip geometry. The interpretation of data from flat probe tips has been an uncertainty for some time and needs to be tested against proud probe tips and the divertor Thomson system. We have a unique capability on DIII-D to resolve these issues. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Build a DIMES probe test module for testing probe geometry and materials. Take plasma data with this test module by running with the strike point over DIMES and the divertor Thomson port (R=1.49m) and compare different material, PG plane orientation, and probe tip geometry. Compare measurements with divertor Thomson. | ||
| Background: | The ITER Langmuir probes are listed as a necessary diagnostic for determining plasma conditions in the divertor and degree of detachment for heat flux control. They are the responsibility of G.Zhong of the Southwestern Institute of Physics in China. Several Issues have been identified that need to be tested on a tokamak before building and installing the probes on ITER. We think DIII-D is the perfect platform to conduct these tests and we have several people here that are experts in probe design as well. | ||
| Resource Requirements: | DIMES, plasma shots with outer strike point on DIMES. | ||
| Diagnostic Requirements: | Divertor Langmuir probes, Divertor Thomson, DIMES ITER probe test module | ||
| Analysis Requirements: | Some divertor modelling may be needed. | ||
| Other Requirements: | need to design and build an ITER Langmuir probe test module for DIMES. | ||
| Title | 394: q95 < 2 tokamak operation via active control of MHD stability | ||
| Name: | Piovesan | Affiliation: | Consorzio RFX |
| Research Area: | Torkil Jensen Award | Presentation time: | Not requested |
| Co-Author(s): | RFX-Âteam: M. Baruzzo, T. Bolzonella, L. Marrelli, P. Martin, P.Piovesan, L. Piron, P. Zanca. DIII-ÂD team: J. Hanson, R. La Haye, Y. In, M. Okabayashi, T. Strait | ITPA Joint Experiment : | No |
| Description: | Based on recent successful RFX-mod results [M. Baruzzo et al. EPS 2011, paper P2.091, G. Marchiori et al. EPS 2011, paper P2.110], we propose to run tokamak discharges with q95 around or below 2 in DIII-D by exploiting active control of MHD stability via I- and C-coils and in particular of the 2/1 mode.
We propose to start with Ohmic or low-beta plasmas and then to move to advanced scenarios at higher beta. If successful, this type of operation may open access to unexplored tokamak regimes with potentially higher normalized fusion performance and provide therefore a transformational result. Expanding the operational space towards very low edge safety factor values would in fact open a path toward the increase of normalized fusion gain in future fusion reactors, given the (1/q95)^2 dependence. In addition this experiment will add significant knowledge to the field of optimized feedback control via active coils. |
ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | We plan to demonstrate low-q operation with MHD control in medium current (IP~1MA) Ohmic or low-beta plasmas first. This may be done by ramping the plasma current during the flattop at fixed toroidal magnetic field. The mode growth near q95â??3 will be minimized with mild current ramp, so that we will explore the q95â??2 operation in a reproducible manner. Some co-injected neutral beam may be used to drive plasma rotation and for diagnostic purposes.
DEFC should be optimized for these type of plasma, to avoid n=1 tearing mode locking during the first part of the discharge. The effectiveness of DEFC near q95â??3 was demonstrated by Y. In et al., in the study of current-driven kink experiment in 2008 [Y. In et al. 2010 Nucl. Fusion 50 042001]. The possibility to use ECCD for pre-emptive mode suppression should be investigated, but it could be prevented by the value of BT. Other options to avoid mode locking when going from q95=3 to 2 could be the application of complex gains or of pre-programmed rotating magnetic perturbations to rotate the mode. Fast feedback with I-coils should be then used to control the 2/1 external kink mode, which is expected to grow unstable when q is lowered around 2. We propose to start with the feedback scheme normally used in DIII-D, which computes the plasma response, possibly using the AC compensation algorithm recently implemented by the RFX-mod team [L. Piron et al. 2011 PPCF 53 084004]. As a second step, we plan to investigate control schemes more similar to the ones developed in RFX-mod, e.g. sideband correction. If this operation will be successful, we propose to extend these experiments to higher plasma beta. Recent work showed that the presence of a partially suppressed 2/1 NTM is compatible with hybrid mode operation [J.D. King et al. 2012 Phys. Plasmas 19 022503]. We propose to extend this to even lower q, taking advantage of the results obtained in Ohmic plasmas. Developing a low-q hybrid scenario with MHD control done with active coils and ECCD is the final goal of the proposal and would be a great achievement for the implications this would have on tokamak performance. Other open points: - can we control to some extent q on axis? is it better to have it above or below 1? - in the high beta scenario, could we expect kinetic stabilization effects of the 2/1 mode? | ||
| Background: | RFX-mod has recently been run also as a low-current Ohmic tokamak, with the purpose of providing an experimental test bed for advanced studies on feedback control of MHD stability. Recent RFX-mod experiments done in collaboration with DIII-D (M. Okabayashi, Y. In) and JAEA (M. Takechi) showed that it is possible to access low-q regimes with q95 below 2 by feedback control of the 2/1 external kink, which otherwise grows unstable and disrupt the discharge [M. Baruzzo et al. EPS 2011, paper P2.091, G. Marchiori et al. EPS 2011, paper P2.110].
Key to the success of these experiments was proper cleaning of the n=1 sensor signals [P. Zanca, MHD workshop 2011]. In particular, removal of sidebands from active coils is necessary for successful feedback action in the RFX-mod case. This procedure is to some extent analogous to use the plasma response as feedback variable, as normally done in DIII-D. The results mentioned here are very robust and have been reproduced also with partial coil coverage, down to 3.1% of the plasma surface (using 6 out of 192 coils) [T. Bolzonella, ITPA/MHD 2011]. An exploratory proposal similar to the present one was proposed by Y. In et al for the TJA in 2008, but did not get experimental time. Based on the recent results of RFX-mod, on the work done in these years in DIII-D, and on some experience on DIII-D stability control experiments gained over the years by the RFX team, we think that now the chances for success of this experiment and therefore for the access to an unexplored scenario like that with q95â??2 are much higher. | ||
| Resource Requirements: | DEFC, fast MHD control with I coils, ECCD pre-emptive suppression of 2/1 mode.
Stability modelling before the experiment may be desirable, in order to assess plasma shapes that have wall stabilization of the 2/1 external kink mode. Feedback modelling may also be desirable, to assess the merits of using audio amplifiers (high bandwidth) or SPAs (high current). We plan one day to realize the experiment. | ||
| Diagnostic Requirements: | The best possible diagnostic coverage. In particular, magnetic, ECE, and soft x-ray diagnostics to detect the 2/1 mode. MSE to properly measure the q profile. | ||
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| Title | 395: Creation of thick carbon co-deposit layers on tiles for ex situ tests of D (tritium) release | ||
| Name: | Stangby | Affiliation: | U of Toronto |
| Research Area: | Boundary and Pedestal Physics. Plasma-Material Interface. | Presentation time: | Not requested |
| Co-Author(s): | Peter Stangeby, Jim Davis, Chris Chrobak, Tony Leonard | ITPA Joint Experiment : | No |
| Description: | The ITER Organization proposes to drop the C+W divertor and to have an all-W divertor from day-one. This has been questioned by the ITER STAC. The final decision will not be made until 2014. The main argument against use of graphite at the targets is the tritium retention due to co-deposition with carbon. The estimates for the rate of D(T) retention by this process may be too high, however, in light of recent findings, see Background. In order to generate the database required to adequately inform the 2014 decision about the ITER divertor, further lab tests are required as soon as possible using thick carbon co-deposits created in a tokamak. | ITER IO Urgent Research Task : | No |
| Experimental Approach/Plan: | Near the end of the 2012 campaign, a ½ day of high power shots will be used with the outer strike point placed near the entrance to the lower pumping duct, to create thick co-deposits on the tiles in the pump duct entrance. When the machine is opened, 10 tiles with thick co-deposits will be removed for tests in Toronto, and perhaps also in Sandia and MIT. These lab tests will be used to confirm and further quantify the results recently reported by Tore Supra (Tsitrone, 2012) and Toronto (Davis, 2012), see Background. | ||
| Background: | 1. 1. Tore Supra [E Tsitrone, ?DITS project: Closing the particle balance in Tore Supra?, 2012 Divsol ITPA meeting, Juelich]: 70-90% of D retained in C co-deposits outgases over 1-2 years when tiles simply remain under vacuum at 120 C. 2. Toronto [TJ Finlay, JW Davis, AA Haasz, abstract submitted to the 2012 Hydrogen Workshop ?Hydrogen Isotope Exchange in DIII-D Co-deposits?]: ~ 50% of D in thin co-deposits on DIII-D tiles is released by H2-baking (10 hrs, 150 torr, 350 C). We will create thick co-deposits in DIII-D as (1. then H2-bake samples and tiles in Toronto. | ||
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