Title	1: Physics of Safety Factor Resonance for n=3 RMP ELM Suppression
Name:	Fenstermacher fenstermacher@fusion.gat.com	Affiliation:	LLNL
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Expand scope of physics studies of the safety factor resonance window for ELM suppression with the n=3 I-coils.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Perform a very systematic study of the physics determining the resonance window for n=3 ELM suppression with the I-coils. Scan plasma current up at fixed BT to produce a q95 down ramp from 4.2 to 3.2. Also scan Ip down to produce a q95 up ramp from 3.2 to 4.2 to look for differences due to build-up of edge current due to the Ip ramps. Change Ip ramp rate to look for differences in build-up of edge current. Do shot to shot fine q95 scan at edge of resonance window to eliminate the possibility of Ip ramps affecting the edge current. Carefully monitor edge plasma profiles (small wags and jogs for high resolution profiles) to determine if profiles remain the same during q ramps, ie. if window for suppression is due to something besides changes in pressure profile, vis. current transport (Snyder idea).
Background:	There is evidence from previous RMP ELM control experiments (see 128470, 472, 473, 474) that the physics that controls the density pumpout and therefore the significant changes to the edge pressure profile when the RMP is applied, may be less sensitive to a resonance window in edge safety factor than ELM suppression itself. Density pumpout is seen even for q95 far outside (q95=4.2) the resonance window for ELM suppression (q95=3.6 +- 0.1). Physics understanding here should go a long way toward ideas to expand the safety factor window for ELM suppression with new coil designs.
Resource Requirements:	Same resources as used for 2007 ISS ELM control experiments, see for example shot 128374 etc. I-coil maximum current with C_suplies, C-coil for optimum error field correction, 5 co-beams.
Diagnostic Requirements:	All pedestal and lower divertor diagnostics. Edge current measurements (especially simultaneously) with the Li-beam and co- plus counter-beam MSE would be highly desirable as would fast divertor IRTV.
Analysis Requirements:	All pedestal and lower divertor diagnostics. Edge current measurements (especially simultaneously) with the Li-beam and co- plus counter-beam MSE would be highly desirable as would fast divertor IRTV.
Other Requirements:	This is a 1 day experiment

Title	2: NTV versus Intrinsic Rotation
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	A. Garofalo, W. Solomon	ITPA Joint Experiment :	No
Description:	NTV predicts an offset rotation in the counter-Ip direction. H-mode Intrinsic Rotation is in the co-Ip direction. Which will win in ITER? ( This is a reprise of ROF 2007 #127.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Add the n=3 NTV I-coil recipe to ECH ELMing H-mode (truly intrinsic: no fast ions to be disturbed by the n=3 field). Measure velocity profile with a) standard first blip and b) the continuous, sparse balanced blips developed in 2008. Make ELMing ECH H-mode, ~ steady state. Add I-coil n=3. Vary I-coil current. Vary betaN with 3-6 gyrotrons. Test other perturbation mode effects upon intrinsic rotation, n=2 and resonant n=1.
Background:	The NTV counter-Ip offset velocity is reported to have been observed in DIII-D in 2008. Intrinsic Rotation is a known effect, especially in H-mode which seems to obey the Rice scaling, ~W/Ip, co-Ip. What will happen when both are in play? There is some indication that the offset velocity is not forthcoming at modest BetaN (very much below 2), perhaps due to a necessary plasma response to get the perturbed field big enough in the plasma. But ITER must be checked out at H-modes of such modest BetaN! And it must survive the transit from Ohmic BetaN to the scenario-dictated BetaN values!
Resource Requirements:	1 day experiment for n=3. 1 additional day to best both n=2 and n=1 effects. At least 4 gyrotrons available, 5 desirable. Co/Cn NBI. 30L. 33L. I-coils.
Diagnostic Requirements:	Standard for rotation experiments.
Analysis Requirements:
Other Requirements:

Title	3: RMP ELM suppression in ECH H-mode.
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	Test RMP ELM suppression (n=3) in q95 ~3.5 ECH ELMing H-mode.	ITPA Joint Experiment :	No
Description:	These are truly ITER relevant conditions; low rotation (intrinsic) and no fast NBI ions to add another species to respond to the n=3 field. BetaN will not be much above 1, perhaps 1.5 at most, given the foreseen ECH power available. If we canâ??t suppress ELMs below BetaN ~ 2, how will ITER manage to survive the needed extensive system check-out that will take place at reduced stored energy? If in this experiment we are successful in suppression of ELMs, this will be a big plus for ITER planning. We can eliminate any lurking fast ion effects as contributing to the phenomenon.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Make ELMing ECH H-mode, ~ steady state, at the established RMP q95 resonance, ~3.5. Add RMP I-coil n=3. Test RMP ELM suppression. If necessary, go to reduced BT to get as much deltaB/B as possible, at same q95.
Background:	RMP ELM suppression is established in DIII-D. The best low rotation target we can make is an ECH H-mode as used in intrinsic rotation experiments. Does the n=3 field penetrate much better at this lower rotation? But, additionally there is an indication that the high-BetaN plasma response is necessary to get RMP ELM suppression, so this could work against success here.
Resource Requirements:	1 day experiment to test RMP ELM suppression in ECH H-mode. If no suppression can be obtained there may admittedly be pressure to keep trying on another day. Gyrotrons, gyrotrons...
Diagnostic Requirements:	Standard for RMP and rotation experiments.
Analysis Requirements:
Other Requirements:

Title	4: The effect of collisionality on RMP ELM suppression
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Evans, M. Schaffer	ITPA Joint Experiment :	No
Description:	Vary the collisionality, nu*, by using multiple q95 RMP resonances for suppression. Access multiple resonances by using a single row I-coil.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Establish at least three q95 conditions for RMP ELM suppression with the single row I-coil. Top and Bottom can be tested for efficacy. The q95 values could be, for example, ~ 3.5, ~4.5, ~5.5. Adjust density a bit around the â??naturalâ?? pumped-out H-mode density to further adjust collisionality, if desired. Measure the threshold for ELM suppression. Make ELMing ECH H-mode, ~ steady state, at the established RMP q95 resonance, ~3.5. Add single row RMP I-coil n=3; obtain ELM suppression, as before. Use opposite row. Find the q95 resonance for ELM suppression at q95 ~4.5. Try to vary the density with strike point/pumping. Do one more q95 resonance, ~5.5. Try to vary the density with strike point/pumping.
Background:	A direct test of RMP ELM suppression in the low collisionality regime has been difficult to map out because of the link between RMP addition and density pump-out, and the resultant belief that low density is a prerequisite for ELM suppression in this even parity regime. We need to vary the collisionality by varying q95. With two row I-coil the outer major R field line pitch matching limits the resonance to q95 ~3.5. Using a single row means that the I-coil symmetry is simply n=3 and that there must be resonances at every m/3 value, until m gets so high so that the field line pitch is too shallow and a field line cuts adjacent I-coils. The perturbation upon this comes from the addition of the existing error fields. Presumably the q95 ~ 3.5 is the 10/3 I-coil resonance. We had one shot where we swept q95 with a single row and had indications of multiple resonances, but the ramp was too fast to clearly isolate the ELM-free windows. Of course nu*, which depends upon bounce frequency, may not be the operative collisionality, rather the collisionality as relates to an electron free streaming along the perturbed field to the wall.
Resource Requirements:	Standard for RMP experiments: 1 day experiment to test map out ELM suppression window vs q95. It may be necessary to operate at lower BT than standard RMP ELM suppression experiments in order to increase the available deltaB/B.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	5: Understand ECH density pumpout
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Requested
Co-Author(s):	T. Leonard, R. Groebner, T. Osborne	ITPA Joint Experiment :	No
Description:	Test the hypothesis that ECH density pumpout is related to the current density profile (and hence q profile) and is thus perhaps related to the Baker-Rosenbluth pinch. Or is it a purely pedestal-driven phenomenon? ITER will deploy ECCD for NTM suppression.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1. Compare density profiles in a steady state ECH H-mode, with deposition ~ 0.75, with a near-balanced beam H-mode at the same BetaN. Measure the current density profile in each. Vary q95. 2. Chop added ECH power into a beam-driven ELMing H-mode in order to modulate the density profile. Measure the electron density and current density profiles throughout the shot. Vary q95. The above are two separate ways to compare the ECH modification of the current density profile with the ECH modification of the electron density profile. The Baker-Rosenbluth pinch gives a direct relationship between the density profile and the q profile, inside the edge transport barrier in an H-mode discharge. In the first we want to see if the pure ECH H-mode has a here-to-for unobserved density pumpout, that is, a lower steady state density to Ip ratio, than in an NBI H-mode, at the same toroidal rotation.
Background:	This year several experiments saw, and utilized a large ECH density pumpout when ECH was added to a NBI-driven H-mode. This seems to be a greater pumpout than generally observed in the past, with the exception of QDB shots, perhaps because of the ECH being deposited further off-axis. C-Mod has observed LH-driven density pumpout with off-axis absorption. A common theme may be the modification of the density profile, simply due to electron heating.
Resource Requirements:	1 day experiment, with at least 4 gyrotrons. 30L, co/counter NBI.
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	6: Heat Pulse transport in the RMP perturbed boundary
Name:	Schmitz oschmitz@wisc.edu	Affiliation:	U of Wisconsin
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	T.E. Evans, M.E. Fenstermacher, M.W. Jakubowski (MPI Greifswald), C.J.Lasnier, W.P. West	ITPA Joint Experiment :	Yes
Description:	The search for experimental evidence for the edge stochastic layer used to capture the transport changes observed during ELM suppression is a high priority task for understanding of transport physics in edge control by RMP. In this experiment transport of small heat pulses deposited on the inner boundary of the vacuum predicted stochastic edge will be used to inspect if in comparison with an no-RMP discharge evidence for open field lines and therefore enhanced net-outward transport of the heat pulses is found.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Short ECH heat pulses in the order of 30-50 ms with a power of 1-4 MW will be injected radially as far outside as possible for the given density profile (we aim at r/a ~ 0.8). The radial propagation of these heat pulses will be monitored using the Thomson scattering systems and ECE as well as the BES arrays and reflectometers. The arrival of the heat pulse on the target will be inspected by carbon sputtering using CCD cameras and by application of fast IRTV (here shutter will be closed and only a small viewing line is available to secure hardware). This heat transport experiments will be performed in RMP ELM suppressed discharges, discharges outside the resonance window for ELM suppression and in no-RMP reference discharges. Comparison of those three H-mode regimes shall allow for resolution of the stochastic layer. In the tail of each discharge an L-mode phase is foreseen with I-coil application and ECH injection. This is intended giving additional information about screening/plasma feedback effects.
Background:	The baseline approach to describe the effect of non-axisymmetric, resonant magnetic perturbations (RMP) is so far the linear superposition of the external RMP field onto the stationary, 2D EFIT equilibrium. In this approach no plasma feedback is included and one crucial question also by the ITER IO is, in how far this paradigm is applicable at all or which boundaries it has. Evidence was found for the existence of the generic separatrix perturbation. However, so far no direct experimental proof of the existence of the stochastic region and the open field line, so called laminar region in the edge was found. Monitoring of the ECH heat pulses injected shall allow to see the heat transport along open field lines and comparison between ELM suppressed cases (inside of q_95 window) and no ELM suppressed cases (outside resonance window) and both compared to no-RMP references shall show enhanced heat losses in case of the stochastic layer is established. This experiment was successfully performed at TEXTOR-DED and contributed to the resolution of the perturbed edge layer and the laminar and stochastic domains inside of it. Comparison of the experimental results on both machines are suggested in the frame of ITPA task PEP-19.
Resource Requirements:	ECH with all 5 gyrotrons ready, standard ISS plasmas used for ELM suppression experiments (e.g. #132741), SPA supplies for I-coil, C-coil in standard n=1 EFC setting (same as #132741), 1.0ITER04 f-coil patch panel
Diagnostic Requirements:	ECE, Thomson Scattering (core, tangential and divertor systems would be desirable), BES, reflectometer, target Langmuir probes, DiMES_TV with C filters in filter wheel, tanTV system with C filters, IRTV (both cameras / LLNL+TEXTOR), MSE measurement combined with fast Li beam would be beneficial for edge current evolution during heat pulses, CER needed for rotation measurement
Analysis Requirements:	CER analysis and evaluation of other systems listed above
Other Requirements:	--

Title	7: RMP effects on boundary plasma potentials, transport and turbulence
Name:	Boedo boedo@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	Rudakov, Evans, Muller, Holland, Tynan, Moyer	ITPA Joint Experiment :	No
Description:	Obtain effect of RMP on electric fields and turbulence inside the LCFS.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use low power L-mode and H-mode discharges with varying collisionality. Insert fast probes past the LCFS and obtain plasma potentials, electric fields, density, temperature and turbulence and transport measurements. Vary collisionality and RMP parity
Background:	Existing publications and simulations.
Resource Requirements:	DIII-D, RMP coils, beams
Diagnostic Requirements:	edge diagnostics, turbulence diagnostics
Analysis Requirements:
Other Requirements:

Title	8: ELM velocity and size scaling
Name:	Boedo boedo@fusion.gat.com	Affiliation:	UCSD
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Rudakov, Leonard, Lasnier, Evans, Moyer, West, Muller, Tynan, Holland, Hollmann	ITPA Joint Experiment :	No
Description:	ITER needs an assessment of the power loading to the main walls mediated by ELMs and how that scales with plasma parameters.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce a parameter scan and measure velocity and size of the ELMs in the near, mid and far SOL. The parameter scan can be discharge performance as measure by the fusion product, input power and density.
Background:	Various pubs in the subject.
Resource Requirements:	DIII-D, NBI
Diagnostic Requirements:	Edge and fast diagnostics
Analysis Requirements:	data processing
Other Requirements:

Title	9: ELM characterization and dynamics under RMP
Name:	Boedo boedo@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	Rudakov, Lasnier, Evans, Moyer, West, Muller, Tynan, Holland, Hollmann, Leonard,	ITPA Joint Experiment :	No
Description:	Although ELMs can be suppressed during RMP, there are periods of small and infrequent ELMs under many circumstances and little is known about those at the microscopic level, velocity, Te and Ne, size, etc.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce ELM-ing discharges, supress the ELMs, study the remmants withe edge/pedestal fast diagnostics
Background:	Many pubs in the subject
Resource Requirements:	DIII- NBI, RMP coils
Diagnostic Requirements:	edge diagnostics, fast diagnostics
Analysis Requirements:
Other Requirements:

Title	10: Test q95 Resonance Window for ELM Suppression with Single I-coil Row
Name:	Fenstermacher fenstermacher@fusion.gat.com	Affiliation:	LLNL
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Scan q95 by Ip ramps during ELM suppression using a single I-coil row to test for q95 resonance window as in case with two I-coil rows	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Reproduce one of the good ELM suppression discharges using only the upper row of I-coils (131508 or 131484 candidate reference shots). Start with Ip up ramps to vary q95 from about 4.2 to 3.2 during application of n=3 RMP from single I-coil row. If q95 resonance window seen then vary I-coil current to look for effect on q95 resonance window. Also vary C-coil n=1 error field correction perturbation at fixed I-coil current. Also try Ip down ramps to vary q95 from 3.2 to 4.2 to test for effect of piling up edge current due to the Ip up ramp.
Background:	Experiments in January 2008 were successful in obtaining ELM suppression using n=3 RMPs from only a single row of the I-coils �?? vis with the upper row shots 131508, 131484 etc. and with the lower row of I-coils shot 131513. SURFMN analysis indicates that there is no localized maximum vs poloidal mode number, in the perturbation spectrum so that a simple picture would suggest that there would be no q95 resonance window with a single I-coil row. Testing this simple picture would help to understand the physics setting the q95 resonance window for ELM suppression with two rows of I-coils.
Resource Requirements:	I-coils in n=3 with capability for maximum current (6.4 kA). C-coils in optimal n=1 error field configuration Desireable for all C-power supplies operating simultaneously at full current capability (6.4 kA). All 5 co-beams. All cryopumps LHe cold.
Diagnostic Requirements:	All pedestal, SOL and divertor diagnostics especially pedestal Thomson, pedestal CER, midplane and divertor filterscopes, fast and slow divertor IRTV and visible cameras, divertor probes, midplane recip probe.
Analysis Requirements:	Control room SURFMN of applied mode spectrum on EFIT01. Post experiment profile analysis, kinetic efits, VARYPED �?? ELITE stability analysis, TRIP3D field line loss fraction analysis, divertor strikepoint pattern analysis from cameras and particle balance analysis.
Other Requirements:	This is a 1 day experiment

Title	11: Optimized Attempt at ELM Control with n=3 C-coil
Name:	Fenstermacher fenstermacher@fusion.gat.com	Affiliation:	LLNL
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Complete tests of potential for n=3 C-coil ELM control by applying C-coil RMP up to maximum C-coil current (6.4 kA, ie. 25.6 kAt).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Re-establish ELMing H-mode plasma shot 131526 with n=3 C-coil at 4.5 kA (18 kAt) and n=1 I-coil for error field correction. Increase C-coil current in steps to maximum available from C-power supplies (6.4 kA, for 18 kAt). Determine if any mitigation of ELMs can be obtained without locked modes.
Background:	A half day of experiments in January 2008 used the C-coil configured for n=3 with the I-coils configured for n=1 error field correction, to attempt ELM control from a large, on midplane, external RMP coil. A discharge at 4.5 kA in the C-coil ran to completion without any noticeable effect on the ELMs. No density pumpout was seen. Attempts to increase the C-coil current were made in subsequent discharges but each had early locked mode problems prior to application of the C-coil RMP. To complete the evaluation of the potential for ELM control by a large, on-midplane, external coil we should try to apply the maximum C-coil current available to plasmas with no locked mode activity prior to the application of the C-coil RMP.
Resource Requirements:	All C-power supplies operating simultaneously at full current capability (6.4 kA). All 5 co-beams. C-coils in n=3 configuration. I-coils in n=1 with optimal configuration for error field correction. All cryopumps LHe cold.
Diagnostic Requirements:	All pedestal, SOL and divertor diagnostics especially pedestal Thomson, pedestal CER, midplane and divertor filterscopes, fast and slow divertor IRTV and visible cameras, divertor probes, midplane recip probe.
Analysis Requirements:	Control room SURFMN of applied mode spectrum on EFIT01. Post experiment profile analysis, kinetic efits, VARYPED �?? ELITE stability analysis, TRIP3D field line loss fraction analysis, divertor strikepoint pattern analysis from cameras and particle balance analysis.
Other Requirements:	This is a 0.5 day experiment

Title	12: Intrinsic Rotation Scaling; Get the highest BetaN possible in an ECH H-mode
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	W. Solomon, K. Burrell, J. Rice (C-Mod)	ITPA Joint Experiment :	Yes
Description:	Use 5, or 6 gyrotrons to make an ECH H-mode and measure the intrinsic rotation profile.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Repeat past LSND discharges with greater ECH power to create an ELMing H-mode. This is purely intrinsic rotation, no NBI issues. The goal is at least BetaN = 1.5. We may need to try to lower BT to the minimum set by 3rd harmonic absorption in order to try for the highest BetaN.
Background:	John Riceâ??s international database really has no shots above BetaN ~ 1.25, with the exception of a couple of unique points from TCV, where it is not clear as to the conditions. It is important to get to ITER relevant BetaN, which is 1.8 for initial scenarios. DIII-D has used balanced NBI to get to BetaN ~ 2.1, but this condition showed a saturation in the intrinsic velocity scaling, not as large as the database regression would predict. This may be due to mhd activity that set in around BetaN ~ 2 in these DIII-D conditions. It is important to push toward BetaN of 1.8 with only rf, i.e. ECH in DIII-D, so that there is no question as to the effect of any remnant, albeit small, NBI torque.
Resource Requirements:	1 day experiment. minimum of 5 gyrotrons.
Diagnostic Requirements:	standard for intrinsic rotation experiments
Analysis Requirements:
Other Requirements:

Title	13: Measure Intrinsic Rotation Size scaling in DIII-D alone.
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	W. Solomon, K. Burrell, J. Rice (C-Mod)	ITPA Joint Experiment :	Yes
Description:	*Size Scaling NEEDED to confidently extrapolate to ITER. *Obtain steady state ECH H-modes in intrinsic conditions at the most extreme values of Rmagnetic with fixed Rmag/a, and vary aspect ratio, Rmag/a at fixed major radius. *For example, use Rm/a=3.2, with Rm = 1.45, 1.6, and 1.75m. Then pick an appropriate value to fix Rm, and scan Rm/a.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	see above
Background:	The regression analysis for John Riceâ??s intrinsic rotation database, in engineering parameters, predicts a favorable scaling with R, nearly R^2. However, if we simply compare the W/Ip Rice scaling slopes between C-Mod and DIII-D we get an unfavorable R scaling, more like 1/R^2. In DIII-D we can vary R of the magnetic axis at fixed aspect ratio enough to certainly detect R^2 versus 1/R^2 scaling. The above R ratio is 1.21, so R^2 = 1.46 and 1/R^2 = .68. The small R limit is set by obtaining outboard CER measurements far enough inward in minor radius.
Resource Requirements:	1 day experiment. minimum of 4 gyrotrons.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	14: Measure Intrinsic Rotation profiles for the bulk ion (He), especially poloidal intrinsic velocity.
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	W. Solomon, K. Burrell, J. Rice (C-Mod)	ITPA Joint Experiment :	Yes
Description:	Make steady state ELMing ECH H-mode discharges in helium and measure the bulk ion intrinsic velocity profiles, especially the poloidal velocity. Presumably the poloidal velocity will increase with the stored energy, so BetaN as high as possible is desired.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Repeat past LSND discharges with greater ECH power to create an ELMing H-mode in helium.. This is purely intrinsic rotation, no NBI issues. Deuterium NBI blips can be used for measurement, no need for helium NBI injection. We may need to try to lower BT to the minimum set by 3rd harmonic absorption in order to try for the highest BetaN.
Background:	Intrinsic rotation will never be understood, confidently, until the poloidal velocity of the bulk ion can be compared with theory. No one has such a set of data. We need to get the poloidal velocity well outside of the measurement error bars, if it exists at that large a value.
Resource Requirements:	1 day experiment. minimum of 5 gyrotrons.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	15: Sensitivity of One vs Two Icoils to Input Torque Variation
Name:	Fenstermacher fenstermacher@fusion.gat.com	Affiliation:	LLNL
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Compare response of ELM suppression with a single I-coil row to reduced input NBI torque vs response when both I-coil rows are used.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Reproduce one of the good ELM suppression discharges using only the upper row of I-coils (131508 or 131484 candidate reference shots) at about 4.5 �?? 5 kA. On shot by shot basis reduce the input NBI torque in steps starting at a time in the shot with good ELM suppression at full co-torque. At each condition repeat the shot with RMP from both I-coil rows using about 2.8-3.2 kA. If time remains repeat with RMP from lower I-coil row alone at 4.5 �?? 5 kA.
Background:	Experiments in January 2008 were successful in obtaining ELM suppression using n=3 RMPs from only a single row of the I-coils �?? vis with the upper row shots 131508, 131484 etc. and with the lower row of I-coils shot 131513. Initial indications were that the ELM suppression with a single I-coil row was less robust to reductions of input NBI torque than for suppression using both I-coil rows (compare 131517 vs 131518) in the sense that for the same torque reduction the rotation in the single coil case dropped rapidly and then we got a locked mode, while in the case with both I-coil rows the rotation only decreased slightly and the discharge ran ELM suppressed longer before generating a locked mode. Given the differences in the spectra between a single row RMP and an RMP from both coil rows, a comparison of the response to torque variation might improve our physics understanding of the effect of non-resonant components of RMP spectra on plasma performance.
Resource Requirements:	I-coils in n=3 with capability for maximum current (6.4 kA). C-coils in optimal n=1 error field configuration Desirable for all C-power supplies operating simultaneously at full current capability (6.4 kA). All 5 co-beams. All cryopumps LHe cold.
Diagnostic Requirements:	All pedestal, SOL and divertor diagnostics especially pedestal Thomson, pedestal CER, midplane and divertor filterscopes, fast and slow divertor IRTV and visible cameras, divertor probes, midplane recip probe.
Analysis Requirements:	Control room SURFMN of applied mode spectrum on EFIT01. Post experiment profile analysis, kinetic efits, VARYPED �?? ELITE stability analysis, TRIP3D field line loss fraction analysis, divertor strikepoint pattern analysis from cameras and particle balance analysis.
Other Requirements:	This is a 1 day experiment

Title	16: Te Gradient Modulation Experiments
Name:	DeBoo debooga@att.net	Affiliation:	Retired
Research Area:	Transport Model Validation	Presentation time:	Requested
Co-Author(s):	T. Rhodes, A. White, G. McKee, C. Holland	ITPA Joint Experiment :	No
Description:	Expand the successful experiment from last year. The goal of these experiments is to produce datasets where the local turbulence activity is varied by varying the local drive for the turbulence. Correlations between the two provide the detailed set of measurements required to perform turbulence stability code validation studies with GYRO and TGLF. Both a qualitative and more importantly a quantitative comparison of measurements and code predictions are the goals.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use ECH to systematically and repetitively modulate the local value of the electron temperature gradient and gradient scale length, a/LTe, during a discharge and look for correlations with turbulence activity measured with as many of our turbulence diagnostics as possible, looking for correlations in both temperature and density fluctuations across all wavenumbers we can measure. Based on a successful proof of principle experiment last year in an L-mode discharge with only ECH, several follow on experiments are: 1) Spatial scan: get datasets at r/a = 0.6,0.7,0.8 to include with last years experiment at r/a = 0.5. At each spatial location the CECE and DBS systems must perform a fine spatial scan to find the peak turbulence response. If 6 gyrotrons are available then a small scan of the strength of the drive term modulation can be performed and correlated with the turbulence level modulation. 2) Try a different turbulence regime: Ti~Te with low power NBI. Using the 150L source will allow using BES for additional local density turbulence measurements. The change produced in a/LTe is expected to be reduced since the ECH power is now working against a larger background heat flux associated with the additional NBI power. Spatial scan as above if can produce a turbulence response at r/a = 0.5. 3) Sensitivity to plasma shaping: try high triangularity and then high kappa individually to compare with last years experiment at low triangularity and low kappa. Favor a triangularity scan since it can be changed more than kappa. Favor spatial locations further out where triangularity and kappa can be changed the most.
Background:	Last years experiment showed that the local density turbulence level was well correlated with modulations in a/LTe at r/a=0.5 in an L-mode discharge. The turbulence level was varied by 15-20% by varying a/LTe by about 30%. A good dataset was obtained at this one spatial location and was compared to GYRO predictions. Some shots showed a very weak modulation of temperature fluctuations with the CECE system but not on the shots with detailed analysis performed. This may be due to being on the border of the CECE sensitivity limit. Measurements at larger radii would help in this regard as would having 6 gyrotrons rather than 4 to obtain a larger variation in a/LTe. GYRO predictions indicate that a larger modulation in temperature fluctuations as compared with density fluctuations should be observed.
Resource Requirements:	Experiment 1: about 2 days to do all 3 spatial locations and a small power scan Experiment 2: about ½ day to find out if we can produce a modulation in the turbulence even with the higher heat flux due to the added NBI, about 2 days if perform a full spatial scan as in 1 above. Experiment 3: one day At least 4 gyrotrons, 6 would be better (they must be used in pairs for these experiments).
Diagnostic Requirements:	CECE, DBS (both 2 and 4 frequency systems), FIR intermediate- and low-k systems, BES, perhaps the new PCI system. ECE and the usual profile diagnostic systems.
Analysis Requirements:	Although not essential it would improve the measurements/code comparisons if we have a synthetic diagnostic module in GYRO for the DBS system as we have for BES and CECE systems.
Other Requirements:	--

Title	17: High Collisionless NBI Torque Drive for GAMs, aka the VH-mode path?
Name:	deGrassie degrassie@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	G. Mckee, T. Rhodes	ITPA Joint Experiment :	No
Description:	Reprise of #78, ROF 2007. *Use high power NBI co-torque to transiently drive Geodesic Acoustic Modes and measure the plasma response and mode properties with BES. The model requires that â??enoughâ?? prompt NBI radial current be injected to raise the E field â??fast enoughâ?? so that the plasma â??ringsâ?? in this fashion (see Background below). Actually, the proposed target plasma and suddenly switched-on NBI level are reminiscent of the VH-mode recipe.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	*Select a target plasma with low collisionality, with q95 ~ 6. We probably want a DND biased up with normal BT to stave off the H-mode transition as long as possible. 3 NBI co-sources are turned on simultaneously and BES is deployed to look for a GAM response. Other turbulence diagnostics will be useful. If struck, perhaps the GAM response can be followed with only the one (150) beam for some time. Perhaps we will be able to do a number of measurements with various beam mixtures after the thump and ideally see if there is any correlation between the GAM response and any subsequent H-mode transition, or transport barrier formation.
Background:	*NBI torque injected by ions into promptly trapped orbits results in a radial fast ion current that delivers this torque via Jfast X B. The low collisionality plasma responds as a dielectric for times much shorter than the momentum transport timescale, that is, a return polarization current is generated in the bulk ions. This polarization is calculable for collisionless orbits, and depends upon the details of the orbit topology for an ion. For timescales much shorter than the thermal ion bounce time the gyro-orbits shift, giving the so-called classical polarizability. For timescales longer than a bounce time the banana orbits shift giving the neoclassical polarizability, about 100 times larger than the classical value. Passing-trapped ion collisions bring the plasma response to a common neoclassical value. *So, the plasma dielectric in this regime is a function of frequency (timescale). Striking the plasma â??fast enoughâ?? with a radial current source results in GAM generation as described in Hinton and Rosenbluth, PPCF vol 41, A653 (1999). These GAM oscillations are then collisionally damped. *We need to get the E-field to rise fast enough in a thermal ion bounce time in order to modify the orbit. An estimate shows that the prompt radial fast ion current scales with the local plasma beta, and Ip^2. So we want a low beta target (and low collisionality is important for longer GAM damping time), and low Ip, i.e. higher q95, say 5-6. The estimate indicates 3 co-sources would be enough. Hopefully, less will work to give a range to study.
Resource Requirements:	NBI. At least 3 gyrotrons also, to modify the Te profile.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	18: Impact of Rotation on Incremental Diffusivity in Hybrid Discharges
Name:	DeBoo debooga@att.net	Affiliation:	Retired
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Study the impact of rotation on the incremental thermal electron diffusivity in Hybrid discharges.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply electron heat pulses with modulated ECH to the outer region of the plasma and monitor the heat pulse propagation to the plasma core. Vary the plasma rotation with co/ctr NBI and characterize the impact on the heat pulse propagation.
Background:	Past experiments have shown that increased plasma rotation in hybrid plasmas increases the global confinement time and decreases the thermal diffusivity across the whole profile. Studying the impact of rotation on the incremental diffusivity, the change in diffusivity for a given change in temperature gradient, may help in the development of models and will certainly help further constrain any models developed to explain the key physics responsible for the improved transport.
Resource Requirements:	1 day experiment with at least 3 gyrotrons.
Diagnostic Requirements:	--
Analysis Requirements:	Will need to develop the capability to simulate the time dependent response of the plasma with say the TGLF transport code or other transport model in order to compare the simulations with measurements.
Other Requirements:	--

Title	19: Ion response to Te heat pulses
Name:	DeBoo debooga@att.net	Affiliation:	Retired
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Improve understanding of ion thermal transport by studying the parametric dependence of the Ti response to Te heat pulses from modulated ECH power in L-mode discharges.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply modulated ECH power to an MHD-quiet, sawtooth-free, L-mode discharge with the ECH resonant off axis, rho ~ 0.7. Monitor the perturbed ion response to the electron heat pulses with CER. Vary the source strength and coupling conditions to gain insight on the key physics responsible for the ion response. Is the ion response stiff and thus independent of the drive amplitude? Scans could include ECH power to see if the ion response is linear in that driving term, collisionality (density scan to where TEMs are stable) to test if response is tied to TEM activity, and L_Ti (P_nbi scan) and Te/Ti (vary ratio of P_ech to P_nbi)to vary ITG stability. Discussions should be held to determine the best set of parameters to vary in a systematic way to achieve the goal.
Background:	Many previous modulated ECH experiments have displayed an ion response to electron heat pulses that can not be explained by simple ion-electron collisional coupling. In the past the electron behavior was always the focus of the experiments. This experiment is meant to focus on the ion response and the key physics responsible for the ion response. Work was begun this summer on a simulation code for ion perturbations that includes ion coupling to electrons through an ion diffusivity that is dependent on Te and grad_Te as well as Ti and grad_Ti. This code can be used to help analyze the results of these experiments from the perspective of trying to identify which source terms are most important in order to best match the perturbed ion response.
Resource Requirements:	1 day experiment with at least 4 gyrotrons and at least 4 NB sources including those required for CER and BES
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	20: Momentum transfer across LCFS
Name:	Boedo boedo@fusion.gat.com	Affiliation:	UCSD
Research Area:	SOL Main Ion and Impurity Flows	Presentation time:	Not requested
Co-Author(s):	Rudakov, Lasnier, Evans, Moyer, West, Muller, Tynan, Holland, Hollmann, Leonard	ITPA Joint Experiment :	No
Description:	Study momentum transfer across the LCFS by changing the flows in the SOL and measuring the core (if any) response.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	establish L-mode and low power H-mode discharges. Pum strongly. Move strike point from pump to change SOL flows. See core response.
Background:	Various pubs on this
Resource Requirements:	DIII NBI ECH
Diagnostic Requirements:	boundary diagnostics, pedestal diagnostics, CER, fast edge diagnostics
Analysis Requirements:
Other Requirements:

Title	21: MOre SOL flows
Name:	Boedo boedo@fusion.gat.com	Affiliation:	UCSD
Research Area:	SOL Main Ion and Impurity Flows	Presentation time:	Not requested
Co-Author(s):	Rudakov, Brooks, Isler, Evans, Moyer, West, Muller, Tynan, Holland, Hollmann, Leonard, Lasnier,	ITPA Joint Experiment :	No
Description:	COntinue experiments and try to resolve discrepancies found	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	As before
Background:	Same
Resource Requirements:	As before
Diagnostic Requirements:	As before
Analysis Requirements:
Other Requirements:

Title	22: Heat transport in boundary
Name:	Boedo boedo@fusion.gat.com	Affiliation:	UCSD
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	Rudakov, Lasnier, Brooks, Isler, Evans, Moyer, West, Muller, Tynan, Holland, Hollmann, Leonard, Lasnier,	ITPA Joint Experiment :	No
Description:	Study the physics of parallel and perpendicular heat transport across LCFS and SOL. Compare IR camera profiles to Te and Ne probe profiles for parallel heat transport and measure perp heat transport both conductive and convective.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Continue existing work
Background:	There is no fundamental knowledge of all the processes
Resource Requirements:	DIII, NBI, ECH,
Diagnostic Requirements:	IR camera, probes, edge diagnostics
Analysis Requirements:	--
Other Requirements:	--

Title	23: Helium transport in advanced operating regimes
Name:	Weisen none	Affiliation:	CRPP - EPFL
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):	Punit Gohil?	ITPA Joint Experiment :	Yes
Description:	Measure He transport coefficients with focus on hybrid and steady state discharges. ELMy H-modes and L-modes still valuable for comparison, but historic data may be available already. This is part of an ITPA transport TG joint work proposal discussed at the Milano meeting, October 2008 (to my knowledge not yet endorsed). Knowledge of He transport coefficients will contribute to assessing fusion performance of these scenarios.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Select suitable discharges, preferably discharges considered as models for ITER. Use He puffs and measure evolution of He density profile using CXRS with sufficient time resolution. Obtain He transport coefficients from analyis of transient response. Alternatively or in combination with puffs, convert some of the NBI for He injection and obtain transport coefficients in steady state. This may be necessary for ITB's, since He penetration from outside may be too slow. May be combined with or extended to include measurements of He transport in SOL and divertor, He pumping, tau*_He/tauE.
Background:	DIII-D ,JET and JT60-U have published He transport investigations in standard H-modes, VH and plasmas with ITBâ??s (see latest ITER physics base for refs). Nonetheless He transport has received little attention in most devices, especially in advanced regimes such as envisioned for ITER. He ash transport is generally thought of as determining reactor performance via the ratio of helium residence time to the energy confinement time. Historical results indicate that He transport coefficients are similar to those of the background particles and also comparable to those of trace tritium ions investigated in JET. However particle convection may also be important. In many published cases, He profiles appear to be less peaked than the electron density. The possible existence of convection would impact reactor performance positively if V/D is larger for He than for D and T (i.e. weaker He inward convection or even outward convection) or negatively in the opposite case.
Resource Requirements:	NBI, some converted for He injection, gas valve with He, He pumping capability, such as Ar frosting on cryopump.
Diagnostic Requirements:	standard diagnostics, plus CXRS set for He, VUV spectroscopy to monitor He edge source
Analysis Requirements:	Standard plus Particle transport code for analysing transient response to puffs
Other Requirements:

Title	24: Powered VFI Operation
Name:	Hyatt hyatt@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Control/Operations	Presentation time:	Requested
Co-Author(s):	Mike Walker	ITPA Joint Experiment :	No
Description:	We propose to conduct experimental tests in which plasmas are generated and controlled with F-coils subject to the "VFI bus constraint" but with the voltage on the VFI bus regulated by a dedicated supply.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The ideal experimental setup would be to connect F-coils to the VFI bus as usual with the exception of the return coil(s). This coil would be connected to the VFI bus through a voltage regulated power supply that is set to maintain a specified VFI bus voltage. Preliminary tests will be done without plasma and a limited number of powered and return coils, and using a DC supply + choppers as the voltage regulated supply. After successful tests, a suitable SND plasma will be chosen to demonstrate the efficacy of a powered VFI approach. The results will inform a later decision to purchase and install a large (~10 kA/600V) voltage regulated supply that is engineered and dedicated to this purpose.
Background:	Recently completed analysis of MIMO experiment data demonstrates that it is essentially impossible to obtain routine robust model-based plasma shape control because of the combination of VFI bus constraint and chopper nonlinearities. In addition, operational experience with the present empirically tuned PID controllers has found the VFI connection to be problematic for shape control for a number of plasma equilibria. In large measure these difficulties arise because the VFI constraint as it is presently implemented â?? using simple coils to handle the net Fcoil current â?? forces some of the Fcoil supplies into saturation, and this in turn places some choppers deeply into nonlinear territory. If, however, the VFI bus voltage is regulated to a fixed small value, such as zero, each Fcoil is effectively decoupled from all others and all choppers can be maintained in relatively linear operation. This may well allow a robust MIMO implementation, and should address many of the shape control issues our present PID based control schemes encounter.
Resource Requirements:	½ day experiment. Need a few beams (1-3), 30L for MSE, standard error field correction.
Diagnostic Requirements:	Standard magnetics, coil currents, chopper and power supply voltages, CO2 interferometers for density control.
Analysis Requirements:	Auto EFITs
Other Requirements:	Initial tests (without plasma) must be OK'ed by Tokamak Operations

Title	25: Screening effect on recycling impurities of RMP-induced flows
Name:	Brooks brooks@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Tom Petrie	ITPA Joint Experiment :	No
Description:	Test whether the natural edge and SOL flows in ELM-suppressed, RMP plasmas are sufficient to screen recycling impurities from the core. Does deuterium pellet fueling during the RMP phase help maintain a reduced level of argon in the core? The result of this experiment will contribute toward our understanding of how RMP topology affects impurity screening.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Run an RMP plasma in the standard LSN configuration. Prior to turning on the RMP coils, introduce a trace amount of argon into the private flux region, through the PFX2 valve near the inside lower corner of the vessel. Monitoring core argon concentration with the SPRED spectrometer, determine whether the core argon concentration declines during the initial pump-out phase accompanying the RMP turn-on. If so, does it remain reduced during the steady state, ELM-suppressed phase? Use deuterium pellet injection during the steady state RMP phase to increase flow from core to divertor. Does the pellet-enhanced flow improve screening? Inject argon after ELM suppression is established with the RMP coils. Does core argon concentration evolve to the same level as with injection before RMP turn-on? In addition to monitoring core concentration of argon, use divertor diagnostics (spectrometer, visible imaging, IR camera, fixed Langmuir probes) in measure impact on plasma parameters in inner and outer divertor legs.
Background:	Empirically, it has been found that the RMP can suppress ELMs by modifying the edge pressure gradient through an enhanced flow of particles to the divertor targets. Perhaps, this continuous particle flow serves both to purge impurities from the edge, as ELMs transiently do, and to screen recycling impurities from re-entry. Whereas ELMs have a rotating 3-D structure, the RMP plasma has a stationary island structure. A priori, it is not apparent whether the flow in RMP will be sufficiently enveloping of the plasma surface to screen the core against penetration by recycling impurities. In its reliance on the intrinsic RMP-generated flows, this proposal differs from the puff-and-pump approach historically used on DIII-D to enhance compression of impurities in the divertor. This simple test of impurity screening with a trace level of argon has the advantage that it can be run in piggyback mode on a standard RMP plasma in LSN configuration. Research during the last year has concentrated on RMP plasma with low collisionality, where collisionality is defined in terms of plasma parameters in thetop of the pedestal after the density pump-out. Whereas deuterium-argon collisions should be sufficient to entrain argon during the initial density pump-out, it is uncertain whether collisionality remains high enough afterwards. Simple 1-D calculations of collisionality in the SOL/divertor plasma for typical, steady state RMP conditions suggest that the flow lies intermediate between the sheath- and conduction-dominated regimes (assuming ni,sep ~ 3x1019 m-3 and Te ~ Ti ~ 100 eV). Whereas, frictional entrainment in the downstream direction is reduced in this intermediate regime, the â??Ti force in the opposite direction is also reduced. The balance of forces might be sufficient; deuterium pellet fueling might tip the balance. Based on Petrieâ??s careful study of the influence of particle drifts on the ability of the puff-and-pump technique to concentrate impurities in the divertor relative to the core, one can say that the standard RMP configuration with Bxâ??B into the X-point is not the preferred arrangement. Nevertheless, at trace levels of argon, detachment caused by buildup of argon in the inner divertor leg, may not be an issue. If so, this experiment offers an opportunity, in piggyback mode, to gain useful insights.
Resource Requirements:	Machine Time: Piggyback on standard RMP discharge in LSN configuration. Number of Neutral beam sources: 3 HFS pellet injector Argon injection from PFX2
Diagnostic Requirements:	SPRED, MDS spectrometer, lower tangential TV, IR camera, fixed Langmuir probes
Analysis Requirements:
Other Requirements:

Title	26: Inboard Divertor Detachment Characterization
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Volume Recombination and Divertor Plasma Detachment	Presentation time:	Not requested
Co-Author(s):	J. Boedo, M.Groth	ITPA Joint Experiment :	Yes
Description:	Use all available divertor diagnostics to characterize the onset and evolution of detachment in the inboard divertor during H-mode. The goal is to assess the role of inboard divertor detachment for pedestal fueling and SOL flow. In addition collect enough data for testing code prediction of detachment onset.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Use magnetic configuration with both strike-points on the lower baffle. Use divertor sweep for characterization with divertor Thompson X-point probe. Carry out density scan in ELMing H-mode. Further characterization data should come from the IR camera, filterscopes, divertor spectometer, floor probes and tangential cameras. The Thomson, and X-point probes will measure the 2D profile of the divertor plasma, particularly the inboard divetor. The X-point probe can additionally provide cross-field drift particle fluxes. Floor probes will provide surface conditions, particularly particle flux. Spectroscopy with help with the 2D assessment of divertor conditions, and particularly the inboard SOL plasma above the X-point.
Background:	The state of the inboard divertor is a major uncertainty in assessing several aspects of pedestal, SOL and divertor operation in H-mode. The inboard divertor has been assessed as the major source of pedestal fueling in DIII-D, but the magnitude of that source is very uncertain. Assessing that source is dependent upon a comprehensive 2D measurement of the inboard divertor density and temperature. In addition, the flow of particles from the inboard divertor into the core plasma can be a significant driver of SOL flow through parallel pressure balance. An additional source of particle transport is due to cross-field drifts. The drift particle flux can be assessed with the X-point probe. These Finally, attempts with edge fluid codes to model detachment onset have not been successful. The additional data collected here will help with further study of the plasma conditions required for the onset of detachment.
Resource Requirements:	LSN running on lower baffle. Standard H-mode
Diagnostic Requirements:	Divertor Thomson, X-point probe, midplane probe
Analysis Requirements:	Total divertor diagnostic analysis, and Edge modeling
Other Requirements:	--

Title	27: Poloidal Asymmetry of Heat Transport in H-mode
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	J. Boedo, C. Lasnier, M. Groth	ITPA Joint Experiment :	No
Description:	Use a balanced double null configuration to assess the in/out asymmetry of energy and particle transport across the separatrix in H-mode compared to L-mode. Carry out a density scan for this experiment.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Setup a balanced double null configuration in ELMing H-mode. Use IR cameras, floor Langmuir probes and bolometry to carry out power balance measurements. The goal is to measure the in/out ratio of heat, and possible particles, across the separatrix. At a minimum this requires simultaneous measurement of both the inner and outer strike points at both the lower and upper divertors. If simultaneous up/down measurements are not possible then a fine scale DRSEP scan may be sufficient. The floor probe measurement of heat flux can be corroborated with IR camera measurements. Need both ohmic (or L-mode) and H-mode for comparison. In a balanced double null the heat flux to the inboard and outboard side should remain separated. Particle flux asymmetries are harder to assess due to complications such as recycling.
Background:	It has been well established that transport of heat and particles across the last closed flux surface is strongly localized to the outboard midplane in ohmic and L-mode plasmas. Theoretical considerations suggest the transport becomes more poloidally symmetric in H-mode. This has never been measured experimentally. By operating in a magnetically balanced double null configuration it becomes possible to separate the heat flux crossing the separatrix into the inboard and outboard contributions. The in/out heat flux can be measured in a relatively straightforward manner since there should be little in/out communication. Determining the in/out particle flux is more problematic due to recycling issues. This experiment will attempt to characterize the in/out asymmetry of the heat flux and how is changes from L-mode to H-mode.
Resource Requirements:	DN configuration in NBI H-mode
Diagnostic Requirements:	Upper and lower fixed Langmuir probes, IR camera
Analysis Requirements:
Other Requirements:

Title	28: In/Out ELM heat flux asymmetry with counter injection
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	C. Lasnier	ITPA Joint Experiment :	No
Description:	Measure the in/out asymmetry of ELM divertor heat flux deposition comparing co-NBI to counter-NBI injection.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Set up standard LSN ELMing H-mode discharges. Measure the ELM divertor heat flux with the fast IR camera. Additional useful diagnostics include the X-point probe for measuring Te on a fast timescale. Carry out an input torque scan to vary toroidal rotation from positive to negative with respect to the plasma current. Goal is to measure the ratio of inboard to outboard ELM heat flux as a function of toroidal rotation of the pedestal region. Fast Te measurements in the divertor will also help with understanding the ELM dynamics.
Background:	The in/out balance of ELM divertor heat flux is an important factor for determining a tolerable ELM size for ITER. Previous studies have typically found the ELM heat flux is greater to the inboard divertor by a factor 2-3 compared to the outboard divertor. This asymmetry changes direction, greater outboard ELM heat flux, with change in the toroidal field direction. A proposed explanation for these observations is that the ELM heat flux primarily flows with the ion convection. Co-NBI injection with the resulting pedestal rotation in the plasma current direction would result in preferential ion convection of ELM flux towards the inboard divertor for the case with the toroidal field direction with the GradB drift towards the X-point, and the opposite asymmetry for reversed toroidal field. This model also implies that the ELM asymmetry should also reverse if the pedestal toroidal rotation changes direction. This experiment is designed to test this conjecture.
Resource Requirements:	LSN H-mode with co and counter NBI injection
Diagnostic Requirements:	IR camera, X-point probe for fast Te, lower fixed probes
Analysis Requirements:	--
Other Requirements:	--

Title	29: Effect of elongation on disruption runaways
Name:	Granetz none	Affiliation:	MIT
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	Eric Hollmann, Dennis Whyte, Val Izzo, others?	ITPA Joint Experiment :	Yes
Description:	Avalanche growth of runaway electrons during the ITER disruption current quench (CQ) is predicted to result in up to 10 MA of current carried by relativistic electrons (RE). Impact of this population with the vessel wall and/or internal structures could result in severe damage. A survey of RE observations in present-day machines reveals that some tokamaks often have large RE current in the CQ, but others rarely do. One possible distinction noted between the groups is plasma elongation and/or vertical stability, i.e. machines with ITER-like elongations rarely observe significant runaways during the CQ. If we could verify this hypothesis and understand the underlying physics basis, then we might conceivable be able to convince ourselves that REs won't be a problem in the ITER CQ, thus greatly easing the mitigation requirements and the impact on the ITER cryopumps and tritium recycling plant.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	On machines with ITER-like elongation, such as C-Mod and DIII-D, first generate a runaway seed, and then trigger disruptions in both normal elongation (1.6-1.7) and low elongation (1.0-1.3) discharges. On C-Mod, the LHCD system is quite adept at producing a large runaway seed, and some experiments at normal elongation have already been done; low elongation experiments are planned. On DIII-D, it may be possible to use the ECCD system to generate REs, but this will probably require some development.
Background:	FTU, Tore-Supra, and TEXTOR often see large RE current during the CQ. All are circular, limited machines. JT-60U also observes REs in the quench, and although it's diverted, it has a relatively low elongation. In contrast, DIII-D, C-Mod, and ASDEX-U all have ITER-like elongations, and rarely see REs in the CQ (except in very unusual cases such as killer pellet injection, etcetera). This suggests that elongation and/or vertical stability is playing a role. Modeling of a low-elongation C-Mod gas jet disruption is currently being carried out to look at this, but we would also like to get experimental data from several tokamaks.
Resource Requirements:	Low elongation equilibria; reproducible disruption triggering system (gas jet or MGI); method of generating significant runaway seed (ECCD?)
Diagnostic Requirements:	HXR, synchrotron emission if possible, photo-neutrons; plasma current (plateau in CQ)
Analysis Requirements:	Calculation/estimation of driven runaway seed current. (On C-Mod, LH current drive is modeled with CQL3D and other codes.)
Other Requirements:	TBD

Title	30: Why don't wee see evidence for RMP impact on transport in L-mode?
Name:	Schmitz oschmitz@wisc.edu	Affiliation:	U of Wisconsin
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	T.E. Evans, M.E. Fenstermacher	ITPA Joint Experiment :	Yes
Description:	Common observations of transport changes (e.g. particle pump out and profile changes) and evidences for changes of the magnetic topology (e.g. striated heat and particle flux pattern) due to RMP application and possibly connected to the formation of a stochastic edge layer are not observed during application of comparable RMP spectra and amplitude to poloidally diverted L-mode plasmas. However, screening theory would predict that both, the lower toroidal rotation as well as the higher resistivity would hamper formation of screening currents and therefore shall facilitate RMP penetration. In last years experiments the standard I-coil n=3 and C-coil EFC spectrum was applied to L-mode plasmas in the tail of plasmas optimized for H-mode investigation. This proposal suggests to attribute a separate experiment in L-mode at DIII-D in order to reproduce and enhance stochatization indications in L-mode. The comprehensive data set to be obtained will allow to resolve (a) at wich level of current in each shape/confinement scenario stochastization occures (b) provide data for screening studies in very different shape/confinement regimes (c) allow for direct comparison to TEXTOR toroidal mode number n=4,2,1 results with the n=3 DIII-D RMP field, (d) will show the dependency of q_95 resonance effects on plasma shape (role of magnetic shear) and (e) will help to understand the beta_N dependence of ELM suppression and RMP transport effects. All of these topics are embedded into the ITER IO urgent requests on the physics background and therefore part of the ITPA task PEP-19 proposal.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Application of n=3 dominant spectra with C-coils in standard EFC setting is foreseen. We will use L-mode plasmas, both high field side (HFS) limited and poloidally diverted plasmas with different elongation/triangularity. The experimental sequence foresees to start with a plasma as circular as possible, limted at HFS and of as large extend as possible in order to get Thomson data in the plasma edge. Then we have two kinds of shots: (1) given RMP level, q_95 ramp for different power levels between low power L-mode to limiter H-mode (2) best q_95 with stepwise (~500 ms flat top) increase of the beam power. This shall allow to resolve the coupling of the RMP field at each power level and in the transition between low power L-mode and limiter H-mode. Consecutively we will increase for the same set of plasmas the elongation studying the effect of the plasma shape. The scan shall end for L-mode and "just" H-mode poloidally diverted plasmas at low and high triangularity.
Background:	Application of resonant magnetic perturbations (RMP) for ELM control is performed usually in the most ITER-like target scenarios, i.e. high performance, poloidally diverted H-mode plasmas, in particular ITER-similar shape (ISS) plasmas. Description of the observed transport effects applies magnetic field stochatization as one approach to capture the transport changes observed. Here we refer in particular to the particle pump out which appears to be characteristic in ISS plasmas at low pedestal electron collisionality. As evidence for the generic perturbation caused by the RMP field applied to poloidal divertor tokamaks, the perturbed separatrix structure was observed. Vacuum magnetic field modeling predicts that perturbation of the separatrix causes a generic structure, i.e. toroidally spiraling lobes which were experimentally identified in the heat and particle load pattern at DIII-D. However, application of the n=3 I-coil and n=1 EFC C-coil spectra to L-mode plasmas did not provide any evidence that the plasma experiences any changes at all by the RMP field. The only reaction observed was at a given RMP amplitude level the formation of locked modes and a disruption of the plasma. This finding is very counter-intuitive as screening theory would predict better field penetration in L-mode plasmas due to lower rotation and higher resistivity. These findings are also in contrast to circular, limiter L-mode plasmas like TEXTOR-DED where all signatures mentioned were observed and analyzed to be in fair agreement with the vacuum magnetic field topology. This experiments will focus on the resolution of tangle structures in L-mode plasmas at DIII-D. We will start with HFS limited L-modes in order to resolve the homoclinic tangle structures on the innerw all as predicted by TRIP-3D and the EMC3/EIRENE transport code. Then the manipulation of these structures and the connected transport effects with increasing elongation is studied leading to poloidally diverted plasmas as topological discrimination between limiter and divertor plasmas. These investigations are part of the ITPA task PEP-19 and essentially a comparisson to the results obtained at TEXTOR and MAST (here indication of pump out in L-mode was obtained).
Resource Requirements:	main goal is to refurbish patch panel for HFS limited plasmas and search for patch panel alternatives which will allow the shaping changes outlined at best with one patch panel only, all beam sources available, ECH for short heat pulses, I-coil driven by SPARs (current level would be enough but allows for toridal phases scan) and C-coil on C-supplies
Diagnostic Requirements:	Thomson scattering, core (would be edge for HFS limited plasmas) and divertor (would be another edge system), ECE system, reflectometer and BES, fast Li and MSE for edge current measurements, filtered camera observing center post (D_alpha + CII/CIII), for diverted case IRTV and lower divertor CCD cameras (Tan_TV and DiMES_TV), fast IR TVs would be desirable, fast UCSD CCD camera with D_alpha/CII/CIII filter
Analysis Requirements:	pre exp: detailed studies of tangles in different shapes (will be done from Juelich side), post exp: analyse plasmas with TRIP-3D and linearized 2 fluid ideal MHD model developed for TEXTOR analysis in Juelich, EMC3/EIRENE modeling of pattern and expected transport resposne (these data will generate a comprehensive data base for code validation also to understand the divertor ELM suppressed H-mode discharge modeling results obtained already)
Other Requirements:	--

Title	31: Real-time magnetic and kinetic profile control for advanced tokamak operation on DIII-D
Name:	Moreau didier.moreau@cea.fr	Affiliation:	CEA Cadarache
Research Area:	Integrated and Model-Based Control	Presentation time:	Requested
Co-Author(s):	D. Mazon (CEA), J.R. Ferron (GA), M.L.Walker (GA), E. Schuster (Lehigh University)	ITPA Joint Experiment :	Yes
Description:	The present proposal aims at developing real-time profile control methods to regulate the coupled evolution of several plasma parameter profiles (safety factor, plasma rotation velocity, ion and/or electron temperature) in advanced tokamak operation scenarios on DIII-D. Similar experiments using the same generic approach have started and will continue on JET (see background). The extension of the technique to DIII-D high-bootstrap-fraction discharges and the comparison between the experiments on the two devices will be fruitful for assessing the methodology (ITPA-IOS Joint Experiment). The description given here is to be taken as a mid/long-term collaboration program, and to which extent this program can be fully realized in 2009-2010 would depend on the experimental time allowed. However, partial results such as i) identifying a satisfactory model from existing data in a given scenario and, if necessary, from some new open-loop experiments in which various sets of actuators are modulated (1 session), and ii) controlling only a subset of the relevant plasma parameter profiles (e.g. the q-profile) in closed-loop (1 session), will constitute significant milestones. Tasks i) and ii) should then be repeated with increasing number of actuators (e.g. with/without all NB injectors, boundary flux control, ECRH, ICRH, etcâ?¦) and, concommitantly, with an increasing number of controlled, magnetic and kinetic parameter profiles, and a two-time-scale controller.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The initial task consists in selecting a particular plasma scenario which is not too close to power and performance limits (i.e. with some actuator headroom), and for which the performance is prone to be improved through real-time control. Discharges with a large fraction of non-inductive current drive and a high bootstrap-current component (ideally fully non-inductive plasmas) would be the prefered target for these experiments. Once a series of existing discharges corresponding to the chosen scenario have been selected, radial profiles of the inverse safety factor, toroidal rotation velocity, ion temperature and electron temperature will be processed and an approximate state-space model will be identified. The basic input data will consist of the co-current (on-axis/off-axis), counter-current (on-axis/off-axis) neutral beam power, and the line-averaged density. The latter is not a control actuator but it must be retained in the plasma model so that its variation can be taken into account as a disturbance during closed-loop real-time control experiments. Additional inputs such as the boundary loop voltage, ECRH or ICRH powers will be included if they are to be used in the selected scenario. Open-loop modulations experiments will then be performed if necessary to complete the database and improve the model identification. The details of these experiments will be defined once the preliminary analysis and processing of existing data has been done. In a second step, a series of closed-loop experiments will be performed. The first set of these experiments will address the control of the q-profile during the current ramp-up phase and/or during the high performance phase of the discharge in the aim of maintaining the current profile in steady state. In a further step, the simultaneous control of the q-profile and of a selected fluid/kinetic profile (e.g. toroidal rotation velocity or ion temperature) will be attempted. A comparison will then be made between the magnetic/kinetic profile control based on the simplest version of the controller and on the more sophisticated two-time-scale controller. Finally, depending on the available experimental time, the ultimate goal of these experiments would be to control simultaneously the safety factor, plasma rotation and ion/electron temperature profiles with the two-time scale controller.
Background:	Simultaneous real-time control of the magnetic and kinetic plasma profiles is an important goal to achieve for the stable operation of tokamaks in the so-called advanced steady state regime. A multi-variable approach based on a two-time-scale dynamical plasma model has been proposed in which the controller uses the combination of the available heating and current drive (H&CD) systems in an optimal way to regulate the evolution of the plasma [D. Moreau et al., Nucl. Fus. 48 (2008) 106001]. The controller design uses singular perturbation techniques and is quite generic (i.e. independent of the tokamak). It relies on a plasma state-space model which relates a set of (machine-dependent) input parameters or actuators (e.g. H&CD powers) to the measured output profiles to be controlled, namely, the current density (or safety factor) profile, which characterizes the magnetic state of the plasma, and one or several fluid/kinetic parameter profiles (plasma rotation velocity, ion and/or electron temperature, etc â?¦). It was suggested under the International Tokamak Physics Activity (ITPA-SSO/IOS) to test the relevance of the proposed model identification technique in a variety of situations, i.e. on tokamaks equipped with different sets of actuators and parameter profile measurements. A suite of codes has been developed to numerically identify the various elements of the model using experimental data from any device. It has been applied first to JET data where first closed-loop experiments to control the q-profile have then been successfully performed (see reference above). Recently, it has been also applied to JT-60U data and, despite the fact that, in the selected pulses, the input powers were not purposely modulated, a satisfactory model for the coupled evolution of the safety factor and plasma rotation profiles could be identified (D. Moreau et al., ITPA-IOS Topical Group Meeting, Lausanne 2008).
Resource Requirements:	Neutral beam heating at full power is needed including co-current (4 x 2.3 MW + 1 x 2.5 MW) and counter-current (1 x 2.3 MW + 1 x 2.5 MW) beams. Other additional heating and current drive systems will also be required at some stage (ECRH ~ 3MW) and ICRH (up to ~1.5 MW with ~4cm outer gap). Real-time control of the plasma boundary loop voltage would provide an additional actuator and would be desired. Analysis and model identification requires MATLAB computing environment and in particular the System Identification and Control tool boxes. Implementation of the controller algorithm on DIII-D would also benefit from the MATLAB/SIMULINK environment.
Diagnostic Requirements:	Real-time magnetic measurements and equilibrium reconstruction including the q-profile (EFIT and RTEFIT with MSE) are essential, and real-time measurements of the toroidal rotation profile, ion and electron temperature profiles, as well as line-averaged plasma density or density profile are required.
Analysis Requirements:	MATLAB software.
Other Requirements:

Title	32: Structure of the plasma response to kink resonant perturbations in LSN H-mode
Name:	Lanctot matthew.lanctot@science.doe.gov	Affiliation:	Department of Energy
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Garofalo, Reimerdes, Navratil, Bogatu, Schaffer, Soloman, Strait, In, LaHaye, Yu, Lasnier	ITPA Joint Experiment :	No
Description:	Plasma beta and rotation strongly affect the nature of the plasma response to non-axisymmetric fields. This experiment aims at documenting changes in the plasma response that result from varying the drive and damping sources for marginally stable n=1 resistive wall modes in LSN discharges. The main goal will be to measure the amplitude and structure of the plasma response under these conditions in order to validate RWM stability models in the MARS-F and MARS-K codes and to further characterize non-axisymmetric field effects.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Probe the plasma response using a strong slowly-rotating (~10 Hz) n=1 I-coil field at plasma beta values above and below the free boundary stability limit. Perform a beta scan at high rotation using only co-NBI. Measurements of the external and internal plasma response will be obtained using magnetic sensor arrays and soft x-ray cameras. Additional measurements will be attempted using the fast-framing camera, IR camera, ECE, CER, BES, and MSE. If possible, error field correction should be applied using only the I-coil. Vary q95 to change major radius of q=2 surface where the driven displacement is largest. Perform a second beta scan at a rotation value of less than 1% of the Alfven frequency at q=2 surface. Increase the rotation if the Icoil perturbation leads to immediate locked modes. Broadly deposited ECCD should be used for pre-emptive stabilization of rotating NTMs and locked modes at low-rotation.
Background:	Previous experiments using active MHD spectroscopy to drive a plasma response (RFA) in fast rotating plasmas above the no-wall limit have shown that the RFA amplitude at the wall depends on the field rotation frequency, the topology of the applied field and the normalized plasma beta. These measurements are qualitatively consistent with MARS-F modeling using fluid RWM stability models. MARS-F calculations of the stable driven plasma response are also consistent with internal measurements of the plasma displacement profile derived from soft x-ray measurements. Further measurements of the plasma response above and below the no-wall limit and at low rotation are needed to continue benchmarking the MARS-F code and to test recently implemented kinetic RWM stability models in the MARS-K code.
Resource Requirements:	Excellent vacuum conditions are required to minimize occurance of NTMs and locked modes at low rotation. I-coil should be configured to provide error field correction while applying a n=1 rotating field perturbation with 240 or 300 deg. phasing. 7 NB sources should be available to provide access to high-beta at low rotation and for diagnostic purposes. All available gyrotrons should be available to deposit ECCD in the vicinity of the q=2 surface as the q-profile evolves.
Diagnostic Requirements:	Magnetics, Thomson, MSE, SXR, ECE, CER, BES, fast-framing and IR cameras
Analysis Requirements:	CER
Other Requirements:	Testing of I-coil configuration and calibration time during startup for SXR diagnostic

Title	33: Scaling of the q95 ELM suppression window with beta_N
Name:	Evans evans@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	A key requirement for RMP ELM suppression in ITER is to understand and expand the q95 window for ELM suppression while minimizing the amplitude of the resonant and non-resonant magnetic field perturbations. The goal of this experiment is to determine if the q95 resonant window can be increased as betaN is increased between 1.4 and 2.2.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Using an n=3, even parity 60 degree, I-coil configuration with 4 kA, establish a reproducible resonant q95 ELM suppression window with betaN=1.4 by ramping q95 from 4.2 to 3.2 with either a slow +dIp/dt or -dBT/dt (the later is preferred if the -dBT/dt rate is fast enough to cover the desired q95 range in 3.5 s). This will result in a relatively small delta_q95 resonant window. Obtain a full set of diagnostic data at this condition along with at least one I-coil off reference discharge and one case with a 0 degree toroidal phasing. Increase the NBI power and establish reproducible discharges at betaN=1.6 then repeat the discharge sequence used for the betaN=1.4 case. Repeat the sequence again with betaN = 1.8, 2.0 and 2.2. Identify the betaN value with the largest delta_q95 resonant window and do an I-coil current scan with both 0 and 60 degree toroidal phasing to see if the window can be expanded beyond 0.5 (the largest window achieved previously in ISS plasmas).
Background:	Previous experiments have shown that the q95 resonant window for ELM suppression can be increased by increasing the current in the n=3 I-coil field in combination with the n=1 C-coil field. A betaN threshold was also found in low collisionality RMP H-modes with low triangularity. This data suggests that as betaN increases less n=3 I-coil current may be needed to obtain the same delta_q95 ELM suppression resonant window. It is assumed that if this hypothesis can be verified then the it may be possible to obtain a relatively large ELM suppression resonant window at high betaN with a smaller I-coil current than at low betaN. This data will help us understand the relationship between the the I-coil mode spectrum and effects such as resonant amplification of the RMP field as a betaN is varied.
Resource Requirements:	ISS plasmas with I-coils and C-coils. 5 co-NBI sources and 2 counter-NBI sources (possibly 3-4 ECH gyrotorns if available).
Diagnostic Requirements:	A full set of RMP H-mode diagnostics including a fast IR camera viewing the lower divertor.
Analysis Requirements:	Control room TRIP3D/SURFMN. Kinetic EFITs, CER (Ti, rotation and fZ) and python profiles. SOLPS5-EMC3-EIRENE modeling.
Other Requirements:

Title	34: RMP coil pattern variation to investigate ELM and pump-out physics
Name:	Mordijck smordijck@wm.edu	Affiliation:	The college of William and Mary
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	--	ITPA Joint Experiment :	Yes
Description:	Based on predications calculated with the vacuum field line tracing code TRIP3D, the width of the field line loss fraction (an indication for stochasticity) is varied for different experiments. Previous experiments show that even a small region of island overlap can lead to ELM suppression. On the other hand, the perturbation strenght on a given run-day has been correlated with the amount of density pump-out. By controlling the stochastic layer width we try to identify 3D magnetic physics that are important towards ELM suppression with at the same time limiting density pump-out. This experiment will help validate physics concepts for the new coil design.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	We change the I-coil pattern from simple even/odd parity with different phases to up-down assymetric configurations, to control the width of the stochastic layer. The experiments will be based upon theoretical predictions from the vacuum field line tracing code TRIP3D. By allowing up-down asymmetry in the coil currents strengths, we can change the stochastic layer width without having to change the plasma shape, q95 or Beta_n (each of which leads to more complicated changes than just the changes in 3D field physics).
Background:	--
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	35: Test and verify safety factor dependence of turb./transp. simulations via multi-scale turb. meas.
Name:	Rhodes rhodes@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport Model Validation	Presentation time:	Requested
Co-Author(s):	TMV Task Force	ITPA Joint Experiment :	No
Description:	Full title: Test and verify safety factor dependence of turbulence/transport simulations via multi-scale and multi-field turbulence measurements Description: The detailed predictions of turbulence and transport simulations will be tested and verified by utilizing a scan in safety factor keeping all other relevant dimensionless quantities fixed.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Utilizing a scan in plasma current the safety factor q will be varied over a range of ~2 while keeping all other relevant dimensionless quantities fixed. This approach is an excellent match to the restrictions placed upon the magnetic field by the ECH and the various millimeter wave diagnostics (ECE, CECE, DBS, reflectometry). The discharge will be a sawtooth free L-mode. As a second, but important aspect, after we have matching conditions for the different q values and obtained all relevant data (in particular radial fluctuation profiles), we will vary one or more other parameters (e.g. collisionality, kappa, etc.) to determine how the observed agree-ment/disagreement between experiment and simulation is affected. This experiment will make full use of the multi-scale (ITG to ETG scales) and multi-field (ñ, Ttilde, flow) turbulence measurements that are now available on DIII-D (i.e. BES, CECE, DBS, FIR, high-k MBS, PCI, re-flectometry, correlation reflectometry/DBS).
Background:	Both GYRO and TGLF have been shown to have a strong dependence of chi_e,i on the safety factor (see Kinsey, Staebler, and Waltz, Physics Of Plasmas 15, 055908 (2008)). Thus there is a firm expectation that a safety factor scan (all other parameters constant) will result in testable predictions. On the experimental front, in 2001 we performed a q scan over the range q_95=3.5-6.3 keeping the other dimensionless quantities (scale lengths, collisionality, mach number, etc.) reasonably constant (Petty, et al, Phys. Plasmas, 1011, 11, 2004). In this experiment, the plasma current was varied and the resulting Te and Ti variation was compensated by adjusting the NBI appropriately. Sawtooth free operation was obtained during the period of interest by utilizing early beam injection. The q scaling of the thermal diffusivities (assuming a form chi~q^alpha) was found to be fairly strong, with an average alpha of 0.7 and 0.9 for the electrons and ions respectively. Note that at the time of this experiment only a very few of the turbulence measurements were available as compared to today.
Resource Requirements:	Machine Time: TBD Number of Neutral Beam Sources: required ECH: required
Diagnostic Requirements:	All fluctuation and profile diagnostics.
Analysis Requirements:	GYRO, TGLF, XPTOR, plus others. Extensive use of TGLF/XPTOR/similar is expected in order to facilitate the design of the experiment/s, e.g. required measurement error levels, sensitivity to parameters, �?�
Other Requirements:	--

Title	36: Data collection for validation of current and temperature profile evolution models
Name:	Walker walker@fusion.gat.com	Affiliation:	GA
Research Area:	Integrated and Model-Based Control	Presentation time:	Requested
Co-Author(s):	J.R. Ferron	ITPA Joint Experiment :	No
Description:	Obtain experimental data of plasma current and temperature profile responses to DIII-D heating and current-drive actuators, by modulating ECH/ECCD, neutral beam, and perhaps ICH actuators.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The ideal experimental approach would be to combine acquisition of data for profile response validation with data acquired during the course of other (e.g. q-profile control) experiments during which profiles are being modified by heating and current drive actuators. However, since many of these experiments emphasize use of neutral beam power, it may be necessary to execute a number of shots dedicated to obtaining data for validation using other heating and current drive sources. Since the methodology for this validation is not yet mature, it is expected that some portions of initial experiments would need to be repeated and therefore it is desireable to minimize initially the number of shots specifically designed for validation, perhaps by performing them in piggybacks. Most of this data collection would take place during the plasma current flattop, with a smaller amount during the current rampup or rampdown. The amount of data taken under each condition would depend on the ability to see a detectable response to actuator variations. Some types of experiments anticipated include response to steps or oscillations in ECH power, which could be performed under a number of conditions: current drive or heating, heating at different radial locations, and under different (co/counter) beam conditions so as to validate both the current profile response and the Te evolution. Some preliminary tests using ICH may be attempted to determine if sufficient plasma response can be detected for useful model validation. Beam power variation will also be used to fill in beam response data that is not available during plasma current flattop.
Background:	Development of current profile targets and current profile control methods to achieve those targets has been going on for the past few years. Much of the early development was empirical. More recent work has focused on use of current profile evolution models to predict requirements for open-loop evolution of heating and current-drive actuators to achieve the desired current profiles and for development of closed-loop profile controllers. Use of some simple electron thermal transport model appears to be necessary to incorporate into the flux diffusion model that is currently being used for developing current profile controllers, to achieve reliable predictions. To support this effort, comparisons of model-predicted profile evolution with experimental data have been evaluated to determine how well existing models are able to predict actual profile evolution. Up to this point, most data used in these comparisons has been produced from experiments designed for other objectives and comprises primarily variations in beam power. The model versus data comparison work will continue, but there is a need for data that allows evaluation of portions of the model that have not been evaluated previously because of data limitations. These comparisons may require specially designed experiments to generate the needed data, especially in plasma current flattop where there is less of this type of validation data available.
Resource Requirements:	½ day experiment + piggybacks. Small number (2-4) beams, 30L for MSE, all available gyrotrons, all available ICH power, standard error field correction.
Diagnostic Requirements:	Standard magnetics and coil currents, MSE, interferometry, Thomson, CER, ECE
Analysis Requirements:
Other Requirements:

Title	37: VFI-less operation
Name:	Walker walker@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Control/Operations	Presentation time:	Requested
Co-Author(s):	Al Hyatt	ITPA Joint Experiment :	No
Description:	We propose to conduct experimental tests in which plasmas are generated and controlled with F-coils unconstrained by the "VFI bus constraint".	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The ideal experimental setup would be to connect the VFI bus to the E- bus, with all F-coils except 6 and 7 connected between these two buses. However, this type of connection could only be done if modifications were made to DIII-D operational systems to ensure the safety of the F-coils. (A process for determining the necessary modifications has already been initiated.) If that setup is not possible, we would connect all F-coils between the E+ bus and E- bus, so that there is no VFI constraint on the coil currents. A patch panel configuration and plasma similar to 97581 would likely be used. A modified shape control algorithm would be used in the PCS to control the plasma with this unusual-for-D3D patch panel. The objective of this work is to demonstrate that DIII-D can be operated routinely without the VFI constraining F-coil currents.
Background:	Previous experiments (shots 94303-94324, 97563-97589) have shown plasma operation is possible on DIII-D without connecting the F-coils to the VFI bus, although routine operation using such configurations has not been attempted. Previous work has demonstrated an unwanted drift of the common reference flux. One purpose of this experiment is to remedy that control flaw. Recently completed analysis of MIMO experiment data demonstrates that it is essentially impossible to obtain routine robust model-based plasma shape control because of the combination of VFI bus constraint and chopper nonlinearities. In addition, operational experience with the present empirically tuned PID controllers has found the VFI connection to be problematic for shape control for a number of plasma equilibria.
Resource Requirements:	½ day experiment. Need a few beams (1-3), 30L for MSE, standard error field correction.
Diagnostic Requirements:	Standard magnetics, coil currents, chopper and power supply voltages, CO2 interferometers for density control.
Analysis Requirements:
Other Requirements:

Title	38: Operations development time.
Name:	Walker walker@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Control/Operations	Presentation time:	Not requested
Co-Author(s):	Al Hyatt, Dave Humphreys, Jim Leuer	ITPA Joint Experiment :	No
Description:	Establish regular 2 hour experimental slots on Thursday 5-7 for initial testing of experimental use of new operational capabilities. Possible uses include shape development, diagnostic calibrations, control development, tests of ICRF or ECH systems, etc.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Define a method of proposing experiments for each Thursday after OPS sessions a week or more in advance. Define a process for prioritizing and scheduling competing tests. We recommend that priorities be based on demonstration of readiness to proceed, need for capability in upcoming physics experiments, relevance to goals of DIII-D program, and available manpower and budget constraints (an option should be available to not conduct a particular 2 hour experimental session if constraints are too severe).
Background:	Presently, any experiment that uses a new or modified capability must take on the substantial risk associated with the lack of experimental testing of that capability. In the past, this has sometimes resulted in significant portions of an experimental day or even the whole experimental day being consumed by difficulties associated with a lack of readiness of the new capability. In addition, the time it takes for new systems to come on-line can be increased significantly when scheduled experiments are not willing to take on that risk. In recent years, the use of regularly scheduled 2 hour experimental sessions for operational development has significantly accelerated the development of new capability.
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	41: Poloidal asymmetry of SOL flows in DN plasmas
Name:	Brooks brooks@fusion.gat.com	Affiliation:	GA
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	A. Leonard	ITPA Joint Experiment :	No
Description:	Measure impurity flow velocities in lower divertor of balanced DN discharges, in L-mode and H-mode, with tangential viewchords of multi-chordal divertor spectrometer (MDS), using magnetic splitting of spectral lines to distinguish contributions from inner and outer legs. In H-mode, measure times of arrival of heat and particle pulses at inner and outer strike points, top and bottom, using 100 kHz filterscope system. C III channels are sensitive to local changes in Te associated with thermally conducted heat pulse, Dα to changes in particle flux. Look at C III for differences in time of arrival of heat pulses on inner and outer legs; look at Dα intensities for differences in size of particle pulses at inner and outer strike points. This proposal complements ROF proposal 27 submitted by A. Leonard.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Set up a balanced double null configuration in ELMiing H-mode, per ROF 27. Use high temporal resolution of real-time filterscopes to distinguish time of arrival of heat pulses on four legs of DN configuration. Unfortunately, some light from the near SOL is always superposed on the divertor signal of the inner leg filterscope views, limiting their capability to detect a complete absence of signal from the inner leg. Differencing of adjacent filterscope chords can be used to increase discrimination of inner leg signal. Use of the midplane tangential filterscope array in combination with the vertical systems provides a sequence of temporal markers for following the heat flux pulse associated with each ELM. The heat transport between ELMs, one focus of ROF 27, can not be addressed with filterscopes; only IR cameras and Langmuir probes can access this information. The tangential views of the MDS provide a measure of impurity flows on the inner and outer legs in the lower divertor, but only in a time-averaged manner over many ELMs.
Background:	As stated in ROF 27, no careful measurements exist of the decoupling between inner and outer SOL in DN configurations.
Resource Requirements:
Diagnostic Requirements:	MDS, filterscopes
Analysis Requirements:
Other Requirements:

Title	42: Exposure of BW sample
Name:	Wong wongc@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Dmitry Rudakov, Phil West, Todd Evans and Bill Wampler	ITPA Joint Experiment :	No
Description:	We propose to expose a boron (B) loaded W-disc under different plasma discharges including transient events, and different methods of boronization by using the DiMES system.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We would like to expose a BW-disc to begin with piggyback discharges and then dedicated exposures to ELMs and disruption. Similar disc will be exposed during normal boronization and real time boronization to study the re-coating of B onto the disc.
Background:	W has been proposed as the plasma facing material for the ITER divertor and for DEMO. Unfortunately, due to high power density, W surface would melt under Type-I ELMs and disruption, and the surfaces of the chamber wall and the divertor will also be damaged when interact with in coming alpha particles. The BW-disc approach is proposed to mitigate these concerns. A thin (e.g. 2 mm) thick W-sheet is to have ~50% of its surface covered with 1 mm diameter holes. These holes are to filled with B and the plasma facing surface is to be coated with B. During normal discharges, the plasma would only see B. During Type-1 ELMs and disruption, due to much lower melting point of B than W, only the implanted B will be vaporized without damaging the W disc. At the same time the coating of B will protect the W from alpha particles penetration. In order to maintain a uniform B surface for steady state operation of DEMO, the technique of real time boronization, i.e. boronization during plasma operation, will also need to be demonstrated.
Resource Requirements:	DiMES and MiMES system
Diagnostic Requirements:	All lower divertor diagnostics and also core diagnostics to monitor the migration of B and W, and the change of the level of core impurities.
Analysis Requirements:	Before and after surface diagnostics of the BW-disc. Modeling of the experiment
Other Requirements:

Title	43: Real time boronization
Name:	Wong wongc@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Dmitry Rudakov, Phil West, Todd Evans and Bill Wampler	ITPA Joint Experiment :	No
Description:	We propose to perform real time boronization of different types, by B-gas injection, B-pellets injection and B-dust injection.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We propose to perform real time boronization of different types, by gas injection, B pellet injection and B-dust injection. Coating on the chamber wall and at the divertor will be monitored by optical diagnostics and by the MiMES and DiMES systems. The B loaded W-disc can also be used for the monitoring of the coating before and after the boronization.
Background:	W has been proposed as the plasma facing material for the ITER divertor and for DEMO. Unfortunately, due to high power density, W surface would melt under Type-I ELMs and disruption, and the surfaces of the chamber wall and the divertor will also be damaged when interact with in coming alpha particles. The BW-disc approach is proposed to mitigate these concerns. A thin (e.g. 2 mm) thick W-sheet is to have ~50% of its surface covered with 1 mm diameter holes. These holes are to filled with B and the plasma facing surface is to be coated with B. During normal discharges, the plasma would only see B. During Type-1 ELMs and disruption, due to much lower melting point of B than W, only the implanted Be will be vaporized without damaging the W disc. At the same time the coating of B will protect the W from alpha particles penetration. In order to maintain a uniform B surface during steady state operation of DEMO, the technique of real time boronization, i.e. boronization during plasma operation, will also need to be demonstrated. This proposal is to focus on the development of real time boronization.
Resource Requirements:	Depending on the sequence of the experiments, gas injectors, pellet injector and B-dust injectors, also MiMES and DiMES systems will be needed.
Diagnostic Requirements:	All edge diagnostics and core diagnostics to measure the ingress of B to the plasma core and the change in the level of core impurities.
Analysis Requirements:	Surface diagnostics before and after exposure of DiMES amd MiMES samples.
Other Requirements:	When diborane gas is used for the experiment, all established safety precautions and procedures from DIII-D will have to be followed and observed.

Title	44: Nonlinearity of Thermal Conductivity
Name:	Gentle k.gentle@utexas.edu	Affiliation:	U of Texas, Austin
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	Max Austin	ITPA Joint Experiment :	No
Description:	An adequate model must not only predict the state of the plasma but also how that state will change in response changing conditions. This is difficult to extract from models that are computationally intensive and somewhat fluctuating for a single state. Experimental guidance would be helpful. For example, if we could establish linearity, a few calculations would allow interpolation and even limited extrapolation. There are tests of system linearity that are much more sensitive than an experimental or computational effort to measure a transport coefficient as a function of some parameter with sufficient precision and range to determine the details of the function. One of the best is modulated ECH. Specifically, if a symmetric triangular waveform is used, symmetry implies that only even harmonics exist in the drive. Odd harmonics in the response imply nonlinearity in a strict sense. (For example, the familiar offset-linear relation -- above the break point where DeBoo has shown one always operates -- is linear in this sense. Odd harmonics imply a flux proportional to the temperature gradient to an exponent different from unity, which could be extracted from the details of the response. The nonlinearity in the electron thermal conductivity will be determined from modulated ECH experiments.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The first test is proposed for the typical sawtooth-free L mode discharges often used for modulated ECH, cf. 117942. On successive shots, ECH powers of 1, 2, and 4 MW would be applied for approximately 1 s at rho of 0, 0.2, 0.4, and 0.6. Later, a similar approach could be applied to other states of interest.
Background:	On shot 117942, the triangle waveform often used for gyrotron conditioning was applied as ECH at rho ~0.2. Analysis of the Te(t) response showed a clear asymmetry about the maxima and minima, indicating odd harmonics. There is clearly a nonlinearity. Further experiments are needed to determine how the nonlinearity is distributed between the absorption process and thermal conductivity, although either nonlinearity would have important consequences for our understanding.
Resource Requirements:	4 MW of ECH
Diagnostic Requirements:	ECE
Analysis Requirements:
Other Requirements:

Title	45: High-resolution heat flux measurement at the strike point
Name:	Lasnier Lasnier@fusion.gat.com	Affiliation:	LLNL
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	Lasnier, Boedo, Watkins	ITPA Joint Experiment :	No
Description:	Obtain higher-resolution profiles of the outer strike point heat flux using IRTV, with optics modified by removing existing lenses and/or inserting a telescope.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Modify and test optics of IRTV system before the experiment. Run ELMing H-mode with the OSP on top of the shelf. Slowly sweep the strike point across a floor Langmuir probe in the field of view to get high-resolution measurements of the sheath factor.
Background:	There are predictions of a very-high heat flux peak near the strike point due to either electron or fast ion parallel transport. Some data shows sharply peaked profiles that do not appear to be resolved by the 2-mm resolution of the existing IRTV configuration.
Resource Requirements:	Calibration bake of DIII-D
Diagnostic Requirements:	Modified IRTV optics, Floor Langmuir probes, divertor diagnostics, divertor Thomson
Analysis Requirements:	IRTV calibration, floor probe analysis
Other Requirements:

Title	46: Plasma exposure of tungsten nano-scopic 'fuzz' layers on DIMES
Name:	Baldwin none	Affiliation:	UCSD
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	D Rudokov, R. P. Doerner, G. Tynan, C Wong	ITPA Joint Experiment :	No
Description:	In the current ITER design, tungsten divertor plasma facing surfaces are to operate in the temperature range below 700 K [1] with the majority of the divertor heat and particle load being incident on carbon fiber composite (CFC) strike-plates. In this scenario, several studies have identified only minor deleterious effects to the tungsten, such as blistering, which is caused by H isotope and He ion bombardment. For the most part, tungsten surfaces perform well and erosion is minimal in a detached plasma regime because incident ions do not have sufficient energy to contribute towards physical sputtering. However, there is now considerable momentum in the fusion community towards replacing the CFC strike-plates with castellated tungsten plates, thus producing an all W divertor to be implemented from the outset, or by the beginning of the DT phase, of ITER operation. In this regime the tungsten strike-plates will be exposed to considerable He ion loads at an operational temperature closer to 1100 K. Unfortunately, at this temperature a new type of He ion induced damage has been observed. He ion bombardment leads to the formation of a surface layer of tungsten nano-rod like structures that contain nanometer scale bubbles or voids [4]. Further, this layer is not limited to the near surface, but grows in thickness with the kinetics of a diffusion-limited process in both pure He and mixed D2-He plasmas [5]. A question of vital importance is the fate of such nano-structured layers. Due to reduced thermal conductivity with the bulk tungsten, and the scale of individual nano-structures, it is very plausible that the action of a divertor plasma heat load and/or momentum flux, could easily contribute to the removal of such layers. Such a material loss process contributes to the erosion of the tungsten component, high Z dust generation, and possibly ablated W that could potentially reach the core plasma. Presently there are no tokamak experiments that have explored this effect in spite of the importance of the problem in an ITER scenario. [1] G. Federici, R. A. Anderl, R. Andrew, et al. J. Nucl. Mater. 266â??269 (1999) 14 [2] W.M. Shu, G.-N. Luo and T. Yamanishi, J. Nucl. Mater. 367â??370 (2007) 1463 [3] S. Nagata, B. Tsuchiya, T. Sugawara, N. Ohtsu and T. Shikama, J. Nucl. Mater. 307â??311 (2002) 1513 [4] S. Takamura, N. Ohno, D. Nishijima, S. Kajita, Plasma and Fusion Research 51 (2006) 1, N. Yoshida, H. Iwakiri, Proc. 1st Int. Symposium and 1st Korea-Japan Workshop on Edge Plasma and Surface Component Interactions in Steady State Magnetic Fusion Devices, May 20â??22 (2007) Toki, Japan [5] M.J. Baldwin et al 2008 Nucl. Fusion 48 035001	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	An experiment is proposed using the DIII-D DiMES probe to explore this effect. A DiMES button sample of W would be exposed to He plasma in the PISCES-A device at UCSD. We have considerable experience in producing such layers to a specified thickness and propose to produce a DiMES target with a nano-structured layer of 1 $$$$$#956;m thick (~60 $$$$$#956;g.cmâ??2 of nano-scopic W is measured on a 1 $$$$$#956;m thick layer). Using DiMES, the button with nano-structured layer will be exposed to the strike point of the plasma. The removal of the layer by control of the DIIID divertor plasma will be followed by optical spectroscopy by observation of the line radiation from W and to learn about its transport. Subsequent weight loss can be measured by scanning electron microscopy at UCSD. Other diagnostics such as RBS could be used to assess the amount of any deposited material on the W surface of DiMES and the corresponding impact of removal of the W, which could be transported to the plasma core in DIII-D.
Background:
Resource Requirements:	DIMES
Diagnostic Requirements:	OES
Analysis Requirements:	Ex-situ SEM, RBS, TDS at UCSD.
Other Requirements:

Title	47: ECCD triggered 4/3 NTM for predictable high performance Hybrid access
Name:	Hyatt hyatt@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Use targeted ECCD to destabilize a 4/3 NTM to preferentially access a high performance hybrid. 4/3 Hybrids generally have higher performance than 3/2 hybrids due to decreased transport, but are less often seen. Creating them on demand could benefit hybrid physics studies in general and ITER scenario performance in particular.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Start with a standard Hybrid trajectory. At some time after the intial betan rampup that results in a hybrid, but before the NTM appears, start counter ECCD aimed at destabilizing a 4/3. It may be enough to just start one off, or it may require constant counter ECCD for some time. Investigate the initial ECCD start timing, power profile and application period to achieve a robust 4/3 hybrid operation scenario.
Background:	Hybrid operation results when a suitable NTM, either a 3/2 (most likely) or a 4/3, grow and then saturate at a tolerable level at reasonably high values of betan. The NTM increases transport a bit, but also keeps q(0) a bit above 1.0 so sawteeth don't happen. The present hybrid trajectories leave the choice of NTM mode up to the plasma. A 4/3 is preferred because these islands tend to be smaller at saturation and thus cause less transport. Hence 4/3 hybrids tend to have higher performance than 3/2 hybrids. Often the existence of hybrid operation is dependent on wall conditions; sometimes cleaning shots are necessary after disruptions or after several failed hybrid attempts have loaded the walls. Forcing a 4/3 NTM to start up could widen the accessible hybrid operations window and increase the number of physics grade hybrids during an experiment day.
Resource Requirements:	ECCD - at least 3 gyros, suitable hybrid target plasmas, NBI.
Diagnostic Requirements:	MSE, Thomson, CO2 interferometers, fast magnetics
Analysis Requirements:	EFIT, MSE, MHD analysis
Other Requirements:

Title	48: Pellet pacing of ELMs in low toroidally rotating plasmas
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	L. Baylor, T. Jernigan	ITPA Joint Experiment :	Yes
Description:	Determine the effectiveness of pellet pacing of ELMs as a function of the toroidal rotation by injecting pellets at varying frequencies into plasmas with different plasma rotation.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce a set of standard ELMing H-mode plasmas with different toroidal rotations and inject pellets at relevant frequencies for ELM pacing and document the effectiveness at different rotation values.
Background:	The sensitivity of the ELM properties to toroidal rotation may be significantly affected by pellet injection since the the pellets will change the toroidal rotation through changes in the edge density. This experiment aims to resolve how the interplay between the pellets and the edge rotation affect the ELMs, especially at low toroidal rotation which is very ITER relevant.
Resource Requirements:	ELM pacing Pellet dropper, co and counter beams
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	49: Dependence of the H-L transition on toroidal rotation
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	G. McKee	ITPA Joint Experiment :	No
Description:	The LH transition has been determined to be sensitive to the edge toroidal rotation.The objective of this experiment is to determine the dependence of the back transition (H-L) to the edge toroidal rotation. This can be resolved by determining the power threshold for the back transition at various edge toroidal rotation values and plasma parameters, whilst examining the turbulence dynamics.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce and maintain H-mode plasmas at different edge toroidal rotation values. Then decrease the input power at constant torque and observe the conditions at which the back transition to L-mode occurs. Focus all availabale turbulence dagnostics at the edge during these measurements. The key is to examine the turbulence behaviour prior to teh H-L transition and compare it to the forward L-H transition to get a better understanding of the key factors affecting the turbulence and the forward and back transitions. Repeat these scans for different plasma parameters and configurations to map out the operatinal space for the back transition.
Background:	Given the sensitivity of the H-mode transition to the edge toroidal rotation, an important issue for H-mode physics then becomes the dependence of the back transition (H-L) to the edge rotation and the determination of new key physics by observing the turbulnce dynamics across the back transition as a function of rotation. This is an especially important issue for ITER for which the conditions in which H-mode plasmas can be maintained at low toroidal rotation is now an open question.
Resource Requirements:	Co and counter beams
Diagnostic Requirements:	All turbulence and profile diagnostics
Analysis Requirements:
Other Requirements:

Title	50: The scaling of the low density limit for threshold power
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	Y. Andrew (JET), D. McDonald (JET)	ITPA Joint Experiment :	Yes
Description:	Determine the scaling of the H-mode low-density limit, i.e. the minimum density below which the H-mode power threshold increases rapidly, as a function of applied torque and plasma rotation and of machine parameters such as toroidal field and plasma current.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Determine the H-mode threshold power as function of applied torque at various target densities and for different plasma parameters such as toroidal field and plasma current.Examine the corresponding variation in the edge turbulence for these parameters.
Background:	Previous results from DIII-D indicate a strong dependence of the H-mode threshold power with applied torque.C-Mod results in D plasmas suggest that the low-density limit increases with increasing toroidal field and has little or no dependence on plasma current. These experiments will determine these effects in DIII-D with a further dependence on applied torque and emphasis on turbulence measurements.
Resource Requirements:	co and counter beams
Diagnostic Requirements:	ALL turbulence and profile diagnostics
Analysis Requirements:	--
Other Requirements:	--

Title	51: Confinement and H factors at input powers close to Threshold power
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	G. Wang, Y. Andrew (JET), D. McDonald (JET)	ITPA Joint Experiment :	Yes
Description:	Determine the behaviour of the energy confinement and H-mode H factors at input powers close to the H-mode threshold power (i.e. Pin ~ 1.1-1.3 Pth).The influence of applied torque on confinement at these low powers will also be determined.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Perform power ramps close to the threshold power and determine affect on confinement. Peeform scans with different values of applied torque and at different plasma parameters (e.g. Bt, Ip, ne) and compare with energy confinement scaling relationships.
Background:	In most tokamaks, operation in H-mode is usually performed at additional power level much higher than the L-H transition threshold power. In ITER, the available power will force operation close to the L-H threshold power, especially at high density (~1020m-3). It is important, therefore, to determine the level of confinement that would be expected at powers close to threshold power and then to compare this with confinement levels predicted from confinement scaling relationships which have previously been determined at power levels much greater than the threshold power.
Resource Requirements:	co and counter beams
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	52: FW-only L/H Transition Power Study
Name:	Pinsker pinsker@fusion.gat.com	Affiliation:	GA
Research Area:	General ITER Physics	Presentation time:	Requested
Co-Author(s):	F.W. Baity, J.C. Hosea, A. Nagy	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 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.
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 would be a very helpful addition to the usual diagnostic set for this experiment.
Analysis Requirements:	--
Other Requirements:	--

Title	53: FW coupling and electron heating in ELM-stabilized H-modes with RMPs - continued
Name:	Pinsker pinsker@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	C.C. Petty, T.E. Evans, M. Porkolab, F.W. Baity, A. Nagy, J.C. Hosea	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. Poor B-supply regulation 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 would be a significant plus to these studies.
Analysis Requirements:
Other Requirements:

Title	54: 4th and 6th harmonic FW synergy studies
Name:	Pinsker pinsker@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	W.W. Heidbrink, M. Porkolab, J.C. Hosea, A. Nagy, F.W. Baity	ITPA Joint Experiment :	No
Description:	In 2008 piggyback experiments, a distinct synergy between 4th harmonic absorption of Fast Waves (FWs) on an injected beam and 6th harmonic absorption was discovered: while absorption of 90 MHz power at 2 T on a beam was weak (6th harmonic absorption) the addition of 60 MHz power (4th harmonic absorption) as a preheater greatly enhanced the effect of the 6th harmonic power, particularly in the neutron rate. In a 2-hour period of dedicated time late in the run, the effect was reproduced, but the result was somewhat clouded by an unwanted L/H transition in the combined beam/4th harmonic/6th harmonic phase. Furthermore, the FIDA measurements in the piggyback experiment that first detected the effect were of low quality, for reasons that are still unclear. We need to improve the dataset on the synergistic effect both by increasing the parameter range in which the effect is observed and, even more importantly, fully documenting the effect with FIDA measurements.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In the previous work, we compared the effects of 60 MHz and 90 MHz separately with the effect of 60 MHz and 90 MHz combined, with the 60 MHz coming on first, followed a few hundred msec later by the addition of the 90 MHz power. We need to scan the 60 MHz preheating power upwards with a fixed timing, and to vary the length of time that the 60 MHz is on alone before the addition of the 90 MHz power to probe the time development of the fast ion distribution function. Fine-scale scan of BT to maximize effect, density scan as time permits at optimized 60/90 MHz relative timing and toroidal field.
Background:	See description.
Resource Requirements:	One day experiment. NB sources: all sources necessary for full diagnostic information, particularly those needed for the most complete FIDA coverage. 3 FW systems (60 MHz plus 2 at 90 MHz).
Diagnostic Requirements:	Best FIDE coverage possible is crucial. Neutron measurements PLASTIC, ZNS also critical. Usual profile diagnostics in order to be able to complete transport runs to calculate predicted neutron rates.
Analysis Requirements:	FIDA data analysis; TRANSP runs to compute predicted neutron rates without FW acceleration of fast ions. Modeling of results with time-dependent CQL3D runs incorporating two FW frequencies simultaneously.
Other Requirements:

Title	55: Heat transport in the tokamak SOL - FY10 Joule MIlestone
Name:	maingi none	Affiliation:	ORNL
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	J.A. Boedo(UCSD), C.J. Lasnier(LLNL), V.A. Soukhanovskii(LLNL), J. Canik(ORNL), R. LaBombard (MIT), J. Terry(MIT)	ITPA Joint Experiment :	No
Description:	The main idea is to conduct the set of coordinated experiments mentioned in the FY 2010 Joule milestone. The scope and details of those experiments are being worked out. In DIII-D, there is a wide body of existing divertor heat flux and midplane SOL width data. The main part suggested here is anticipated generation of data not already in hand, or that needed to consitute overlap with NSTX and C-Mod. Examples are NBI/RF comparisons, detailed comparisons with fast probe, ohmic discharges, etc.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Specifically there is a desire to obtain data with the same heating method in all 3 machines: 1. dedicated RF H-modes 2. dedicated ohmic discharges 3. dedicated RF and NBI comparisons
Background:	All three fusion facilities have agreed to the FY2010 Joule milestone: Conduct experiments on major fusion facilities to improve understanding of the heat transport in the tokamak scrape-off layer (SOL) plasma, strengthening the basis for projecting divertor conditions in ITER. In FY2010, FES will measure the divertor heat flux profiles and plasma characteristics in the tokamak scrape-off layer in multiple devices to investigate the underlying thermal transport processes. The unique characteristics of C-Mod, DIII-D, and NSTX will enable collection of data over a broad range of SOL and divertor parameters (e.g., collisionality, beta, parallel heat flux, and divertor geometry). Coordinated experiments using common analysis methods will generate a data set that will be compared with theory and simulation. This proposal is intended to be one of the main experimental executions toward that milestone. Discussions have begun between between the C-Mod, DIII-D, and NSTX teams. One guiding physics question to be answered is the mechanism for heat transport from the the midplane to the X-point and then the divertor. Whereas in NSTX and DIII-D, it appears that parallel transport dominates all the way, there may be additional radial dispersal of power at the X-point in C-Mod, motivating analysis with the same 2-D fluid and also turbulence codes. Practically C-Mod will get most of their data early in the FY2009 run for a few technical reasons. NSTX will also dedicate several days in FY2009 toward this milestone, with more time in FY2010. Both C-Mod and NSTX have installed new diagnostics to fulfill this milestone.
Resource Requirements:	Difficult to assess, but envision that up to 3 days may be needed, probably aimed more toward FY10. More details will be available in the coming months.
Diagnostic Requirements:	All boundary diagnostics.
Analysis Requirements:	Analyze data with 2-d codes, such as UEDGE and SOLPS, as well as turbulence codes, such as SOLT and BOUT.
Other Requirements:

Title	56: Exposure of tungsten nano-scopic 'fuzz' on DIMES
Name:	Baldwin none	Affiliation:	UCSD
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	M. J. Baldwin (UCSD), D Rudokov (UCSD), R. P. Doerner (UCSD) and G. Tynan (UCSD) â?? C Wong (GA)	ITPA Joint Experiment :	No
Description:	In the current ITER design, tungsten divertor plasma facing surfaces are to operate in the temperature range below 700 K [1] with the majority of the divertor heat and particle load being incident on carbon fiber composite (CFC) strike-plates. In this scenario, several studies have identified only minor deleterious effects to the tungsten, such as blistering, which is caused by H isotope and He ion bombardment. For the most part, tungsten surfaces perform well and erosion is minimal in a detached plasma regime because incident ions do not have sufficient energy to contribute towards physical sputtering. However, there is now considerable momentum in the fusion community towards replacing the CFC strike-plates with castellated tungsten plates, thus producing an all W divertor to be implemented from the outset, or by the beginning of the DT phase, of ITER operation. In this regime the tungsten strike-plates will be exposed to considerable He ion loads at an operational temperature closer to 1100 K. Unfortunately, at this temperature a new type of He ion induced damage has been observed. He ion bombardment leads to the formation of a surface layer of tungsten nano-rod like structures that contain nanometer scale bubbles or voids [4]. Further, this layer is not limited to the near surface, but grows in thickness with the kinetics of a diffusion-limited process in both pure He and mixed D2-He plasmas [5]. A question of vital importance is the fate of such nano-structured layers. Due to reduced thermal conductivity with the bulk tungsten, and the scale of individual nano-structures, it is very plausible that the action of a divertor plasma heat load and/or momentum flux, could easily contribute to the removal of such layers. Such a material loss process contributes to the erosion of the tungsten component, high Z dust generation, and possibly ablated W that could potentially reach the core plasma. Presently there are no tokamak experiments that have explored this effect in spite of the importance of the problem in an ITER scenario. [1] G. Federici, R. A. Anderl, R. Andrew, et al. J. Nucl. Mater. 266â??269 (1999) 14 [2] W.M. Shu, G.-N. Luo and T. Yamanishi, J. Nucl. Mater. 367â??370 (2007) 1463 [3] S. Nagata, B. Tsuchiya, T. Sugawara, N. Ohtsu and T. Shikama, J. Nucl. Mater. 307â??311 (2002) 1513 [4] S. Takamura, N. Ohno, D. Nishijima, S. Kajita, Plasma and Fusion Research 51 (2006) 1, N. Yoshida, H. Iwakiri, Proc. 1st Int. Symposium and 1st Korea-Japan Workshop on Edge Plasma and Surface Component Interactions in Steady State Magnetic Fusion Devices, May 20â??22 (2007) Toki, Japan [5] M.J. Baldwin et al 2008 Nucl. Fusion 48 035001	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	An experiment is proposed using the DIII-D DiMES probe to explore this effect. A DiMES button sample of W would be exposed to He plasma in the PISCES-A device at UCSD. We have considerable experience in producing such layers to a specified thickness and propose to produce a DiMES target with a nano-structured layer of 1 ï?­m thick (~60 ï?­g.cmâ??2 of nano-scopic W is measured on a 1 ï?­m thick layer). Using DiMES, the button with nano-structured layer will be exposed to the strike point of the plasma. The stability and/or removal of the layer by control of the DIIID divertor plasma will be followed by optical spectroscopy by observation of the line radiation from W in order to learn about its transport. Subsequent target weight loss can be measured to study erosion and scanning electron microscopy at UCSD can be used to perform post mortem analysis. Other diagnostics such as RBS could be used to assess the amount of deposited materials, if any, on the W surface of the DiMES button.
Background:
Resource Requirements:	DiMES
Diagnostic Requirements:	OES
Analysis Requirements:	Ex-situ SEM, RBD, EDX at UCSD
Other Requirements:

Title	57: Access to H-mode during ramp up/down phases
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine the power required for H-mode access during plasma current ramp-up/down phases for conditions and parameters applicable to ITER (normalised dIp/dt, plasma shape evolution, etc.). Control of various plasma profiles (e.g. Te, Te, current profile, etc.) are dependent on the type of confinement regime during the discharge evolution.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Determine the H-mode power threshold and energy confinement during plasma current ramp-up/down phases (for ITER relevant conditions) using power ramps.
Background:	Control of the current profile shape in ITER during ramp up/down phases may require operation in H-mode in conditions with varying plasma current, shape, etc. Therefore, the ability to access H-mode during these phases will be critical for controlling the evolution of these profiles.
Resource Requirements:	co and counter beams, ECH
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	58: Direct Measurement of E_rad Corrugation at Rational Surfaces
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	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.	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.
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.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:	Need GYRO simulations.
Other Requirements:

Title	59: Core transport barriers and "bat-eared" Te profiles in EC-heated discharges
Name:	Austin austin@fusion.gat.com	Affiliation:	U of Texas, Austin
Research Area:	Transport	Presentation time:	Requested
Co-Author(s):	K. Gentle	ITPA Joint Experiment :	No
Description:	We want to investigate unusual transport properties observed in recent DIII-D discharges with off-axis electron cyclotron heating. In experiments in 2007 employing ECH, steady-state hollow Te profiles with sharp changes in gradients were seen in low density, low current discharges. These profiles are reminiscent of those on the RTP tokamak which were indicative of core electron transport barriers. We propose to investigate the thermal electron transport in these shots with modulated ECH, looking at heat pulse propagation across the implied barrier region. Preliminary analysis indicates the discharges are developing negative central shear and the radii with the sharp gradient changes are possibly located near low order rational q_min surfaces. This is a unique opportunity to study an ECH discharge without an apparent heat pinch.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish a low density, high q discharge like 129534 and apply heating with four gyrotrons to produce hollow Te profiles. Modulate one of the four gyrotrons, or add modulated power from a 5th gyrotron at a different radius to create heat pulses for transport analysis. Add neutral beam blips to some discharges to measure q profile with MSE data. Increase the plasma current to change the target q profile. Vary ECH deposition radius either during a shot with Bt ramp or shot-to-shot with launchers and look for step-wise changes in transport. Collect ne and Te fluctuation data for radii near the apparent transport barriers.
Background:	Discharges with strong ECH exhibiting steady-state hollow Te profiles and step-wise changes in transport were first seen it the RTP tokamak. The observed electron transport behaviour was attributed to good surfaces near low order rational q values in a reverse shear profile. This ties in with recent DIII-D experiments on changes in transport seen near integer q_min traversals. With the excellent diagnostic set on DIII-D including ECE, MSE, and fluctuation diagnostics, it should be possible to attain improved understanding of the role of special values of the safety factor in electron transport. Observation of zonal flow structures and changes in turbulent fluctuations would lead to validation of the rational-q transport model.
Resource Requirements:	Three neutral beam sources, including 30L. At least four gyrotrons, nom. 2.4 MW total power and 1 sec pulse length.
Diagnostic Requirements:	Essential: Thomson scattering, ECE radiometer, MSE Desirable: FIR scattering, Correlation ECE, BES
Analysis Requirements:	ONETWO, Toray, possibly GYRO
Other Requirements:

Title	60: High Beta, Steady State Hybrids
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Assess Steady-State Current Profiles for Optimum Performance	Presentation time:	Requested
Co-Author(s):	S. Ide	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 2009, the addition of a sixth gyrotron and optimization of the five co-beams will allow us to eliminate the residual 10 mV loop voltage of our best previous case. We will also examine the effect of lower rotation on the confinement and stability properties of these plasmas (while sacrificing state-state operation). A sixth co-beam source will allow steady-state conditions to be achieved at higher beta and lower q_95 in 2010. This experiment will demonstrate that H-mode (hybrid) discharges with q_min~1 are capable of high beta (beta_N~4) operation with >50% bootstrap current fraction. The remaining noninductive current will will be supplied by on-axis sources at high efficiency. The poloidal magnetic flux pumping that is self-generated in hybrid will suppress the sawteeth despite the strong on-axis current drive, which is important for avoiding the 2/1 mode. The higher efficiency for on-axis current drive will offset the modest bootstrap current fraction such that this scenario will satisfy the requirements for FDF as well as (or better than) the high q_min scenario with strong off-axis current drive.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Start by repeating shot 133881. (2) Inject all six gyrotrons with central current drive. For the five co-NBI sources, increase the injection voltages as much as possible while maintaining a plasma pulse length of at least 5 seconds. (3) Optimize the dynamic error correction, adjust the plasma shape for optimal pumping. (4) Reduce plasma rotation by injecting counter-beams. Evaluate the effect on the transport and stability properties. (5) When sixth co-beam becomes available, attempt to increase beta_N using this additional heating power. If plasma current is overdriven (i.e. negative loop voltage), then increase plasma current to compensate.
Background:	The current proposal for FDF envisions a high q_min advanced tokamak scenario with 70% bootstrap current fraction. While this is compatible with the US view of DEMO, the physics of the high q_min AT scenario is still being developed. There is also an issue regarding the high off-axis current drive efficiency needed for FDF in this proposal. Here I propose that the low q_min hybrid scenario is compatible with the requirements of FDF, and it has several advantages. First, the physics basis is well advanced. Long duration hybrid discharge with high beta and high confinement are routinely achieved. Second, because q_min=1 in the hybrid scenario, all of the external current drive can be deposited near the plasma center where the current drive efficiency is the highest (because of the lack of trapped particles and the high electron temperature). While the bootstrap current fraction will be lower in this low q_min hybrid scenario (50% rather than 70%), the increase in the current drive efficiency for central deposition more than makes up for this. Experiments on DIII-D have come very close to demonstrating this scenario using five co-beams and five gyrotrons. Hybrid plasmas with beta_N=3.4 were stably produced with a loop voltage of 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. The remaining question regarding reactor relevance will the be role of high toroidal rotation. While we cannot separate the roles of torque injection and current drive in the NBI system, we can evaluate the role of rotation in the plasmas by using the counter beams; it is doubtful the the resulting increase in the loop voltage will have a major impact on the transport or stability properties of these plasmas.
Resource Requirements:	NBI: 5 co sources in 2009, 6 in 2010. 2 ctr sources will also be used. ECH: 6 gyrotrons with 3.7 MW of injected power. FW: It is desirable to couple 1 MW or more, but core absorption needs to be demonstrated. I-coils: Dynamic error field correction will be used.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:	--
Other Requirements:	--

Title	61: Measurement of Particle Transport Coefficients in L-mode and RMP plasmas
Name:	Gentle k.gentle@utexas.edu	Affiliation:	U of Texas, Austin
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	T. Rhodes, Lei Zeng	ITPA Joint Experiment :	No
Description:	The development of fast, broadband reflectometry opens a new source of accurate n(r,t) data that is necessary for particle transport analysis. This is an opportune time to revisit particle transport because of its importance for advanced tokamak scenarios and bootstrap current, maximizing core fusion output within the Greenwald constraint, fueling and ash removal, and validation of codes that are now beginning to include the particle channel. With a suitable edge density perturbation, n(r,t) data can be analyzed to extract a D(r) and V(r) with an accuracy and spatial resolution that depends on the quality of the n(r,t) data. The proposal is to apply this approach using a modulated gas feed to two cases: (1) A common sawtooth-free L-mode as often used for modulated ECH experiments, and (2) a plasma with a RMP, which is likewise suitably quiescent and is known to have changes in particle transport that will be important for the development of this important technique. This proposal specifies the use of modulated gas feed because that technique has been used successfully in L-mode plasmas. However, any edge perturbation can provide the essential time dependence for the analysis. Gas modulation should be compared with shallow pellet injection as proposed by Baylor. Pellets may give larger, more accurate signals and be easier to control.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The experiments will proceed in three phases. In the first, gas modulation will be applied to one sawtooth-free L-mode for the development of the gas modulation technique. The amplitude must be adjusted for a suitable signal, and the frequency matched to the target plasma. (If the modulation frequency is too high, there will be only a skin effect determined by the edge particle D; if the frequency is too low, the profile will remain static and the solution separates as n(t)n(r). The frequency must be chosen between these limits to provide data suitable for analysis.) In the second phase, the optimized gas modulation will be applied to a sequence of L-modes from ohmic to strongly NBI-heated to determine the scaling of the transport coefficients. Finally, after development of an effective tool, the experiment will be performed on an RMP state. The RMP has been observed to make global changes in density profiles. Knowing the changes in particle transport will be important for the development and extrapolation of the technique. Once the technique is established, other applications may certainly follow.
Background:	Particle transport is one of the primary plasma transport channel, but it has received far less attention than the energy transport channels. This is not for lack of importance or interest, but because it is harder to study experimentally. Lacking a strong central particle source, the net flux is nearly zero; techniques are required to separate the different contributions to the flux that sum to zero. The principal approach is time-dependent analysis, which requires a drive for time dependence and accurate data for the full n(r,t). The latter has not been generally available since the employment of modern divertors made full radial coverage by interferometers rather difficult.
Resource Requirements:	6 MW NBI
Diagnostic Requirements:	UCLA Reflectometer
Analysis Requirements:	--
Other Requirements:	--

Title	62: Runaway electron strike point determination
Name:	James jamesan@fusion.gat.com	Affiliation:	Facebook
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):	E. M. Hollmann	ITPA Joint Experiment :	No
Description:	An experiment to determine the strike point of runaway electrons in both the early 'prompt loss' phase and the late VDE phase of fast shutdowns and disruptions, via observation of SXR and gamma radiation in SXR arrays and scintillator array.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Initiate a fast shutdown with Argon or Neon Killer Pellets from a flat-top plasma to generate runaway electrons, then observe and try to locate their strike points. Switch between USN/LSN configuration to verify hypothesis that RE always escape via x-point in prompt phase. Once a strike point is identified, position small array of variably shielded colocated scintillators for an energy measurement.
Background:	Runaway generation and transport have been shown to damage the first wall and vacuum vessel of large machines (eg JET), and will pose an operational hazard to future machines (eg ITER) unless they are prevented or mitigated. This experiment helps us understand where and how runaways are deconfined, possibly enabling future prevention or mitigation.
Resource Requirements:	.5-1 day of runtime
Diagnostic Requirements:	Scintillator array upgrade required for amplitude based measurement of runaway strike point, and subsequent energy measurement.
Analysis Requirements:	Believable q contours from jfit during disruption would make energy measurement easier, but not necessary.
Other Requirements:

Title	63: H-mode power threshold in helium plasmas
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	Determine the power required to produce H-mode plasmas in helium discharges for a range of plasma conditions and configurations. Determine the behaviour with changing input torque for all conditions.Compare this with deuterium and hydrogen discharges at same plasma conditions and configuration	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Determine the threshold power for H-mode plasmas by performing power scans for different plasma conditions and configurations. Change the input torque for different conditions and evaluate how the threshold power varies.
Background:	One of the operating gases suggested for the first phase of ITER is helium because of its expected lower H-mode power threshold than for hydrogen. However,detailed knowledge of the differences between helium, hydrogen and deuterium and the behaviour as a function of plasma conditions , configurations and input torque is severely limited, especially for extrapolations to ITER conditions. This experiment will aim to resolve these issues.
Resource Requirements:	co and counter beams, ECH, NB system and plasma operations in helium
Diagnostic Requirements:	All fluctuation diagnostics
Analysis Requirements:
Other Requirements:

Title	64: Fueling in helium plasmas
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):	L. Baylor	ITPA Joint Experiment :	No
Description:	Determine the ability to fuel helium plasmas with hydrogen pellets from the inboard launch location. Perform experiment with different plasma conditions and pellet rates and determine effectiveness of fueling for ITER purposes. Also, the effect on the H-mode power threshold as the H/He ratio changes will be determined.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Inject hydrogen pellets into helium H-mode plasmas and determine effectiveness of pellets for fueling the plasma. Examine the effects on dilution and confinement for different plasma conditions.Perform power scans to determine the H-mode power threshold for different levels of fueling and H/He ratios.
Background:	Fueling of helium plasmas in ITER with pellet injection will have to be performed with hydrogen pellets. The expected effectiveness and behaviour for ITER relevant conditions, as well as how the H-mode power threshold changes with the H/He ratio, are presently unknown and this experiment will aim to answer these questions.
Resource Requirements:	pellet injector, helium plasmas, helium and hydrogen beams
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	65: Use of ECCD to create non-thermal seed for runaway generation studies
Name:	James jamesan@fusion.gat.com	Affiliation:	Facebook
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):	Rick Moyer	ITPA Joint Experiment :	No
Description:	Observe susceptibility of plasma to ECCD seed creation of runaway electrons (RE) at various radii throughout shutdowns, by adjusting radial position of ECCD resonance.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use ECCD to create a non-thermal electron population, then inject Neon or Argon Killer Pellet to induce a fast-shutdown, creating runaways from the non-thermal population. Amplitude of radiation emitted by RE impact with the first wall will be used as a diagnostic for susceptibility to RE generation at the selected location.
Background:	The precise location of RE generation during a fast shut down/disruption may yield clues to the generation mechanisms. Knowledge of said mechanisms may enable development of schemes to prevent or mitigate RE formation more effectively.
Resource Requirements:	At least one gyrotron necessary for this experiment. KP shutdown also required.
Diagnostic Requirements:	Upgraded scintillator array for improved diagnosis of runaway properties.
Analysis Requirements:	--
Other Requirements:	--

Title	66: Pellet induced H-mode in hydrogen or helium plasmas
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):	L. Baylor, N. Commaux, T. Jernigan	ITPA Joint Experiment :	No
Description:	Produce pellet induced H-mode plasmas below the nominal threshold power for different sets of plasma conditions in hydrogen or helium plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	nject inside and outside launched pellets in to L-mode plasmas at input powers well below the H-mode power threshold for a given set of conditions. Determine the power threshold with power scans using the pellets and document the plasma edge with edge turbulence diagnostics.
Background:	The ability to lower the H-mode power threshold using pellets has been proven from previous studies in deuterium plasmas. However, the behaviour and the ability to reduce the threshold power in hydrogen or helium plasmas has now to be determined for these plasmas and this experiment aims to resolve this issue.
Resource Requirements:	co and counter NBI, ECH, pellet injector
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	67: RMP ELM-control in hydrogen or helium plasmas
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):	T. Evans, M. Fenstermacher	ITPA Joint Experiment :	No
Description:	Determine the effectiveness of Resonant magnetic perturbations for ELM control in hydrogen or helium plasmas. Evaluate the effectiveness of ELM control when compared with deuterium plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use the Icoils to control the ELMs in hydrogen or helium plasmas for a range of different plasma conditions and configurations. Compare the effectiveness of ELM control with that for deuterium plasmas.
Background:	RMP ELM-control in deuterium plasmas has been well documented. However, the effectiveness of this technique in hydrogen or helium plasmas has not been determined and requires detailed studies.
Resource Requirements:	co and counter NBI, Icoils
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	68: QH-mode in hydrogen or helium plasmas
Name:	Gohil gohil@fusion.gat.com	Affiliation:	GA
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Produce QH-mode plasmas in hydrogen or helium plasma discharges. Determine how the QH-mode and EHO properties compare with deuterium plasmas and also with standard ELMing H-mode plasmas in hydrogen or helium plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use counter-injected or all co-injected beams to produce QH-mode plasmas in hydrogen or helium plasmas. Document all possible edge quantities.
Background:	The QH-mode plasmas in deuterium discharges have certain distinctive properties. How do these properties vary in hydrogen or helium plasmas and can they provide an insight in to the physics of QH-mode plasmas? Resolving these issues is an important concern, especially for ITER in which the early operational phase will be with hydrogen or helium plasmas. Resource Requirements: --
Resource Requirements:	co and counter NBI, ECH
Diagnostic Requirements:	All edge diagnostics
Analysis Requirements:
Other Requirements:

Title	69: Higher Beta ELM-Suppressed Hybrids
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	B. Hudson	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.
Resource Requirements:	NBI: All 7 sources are required. EC: Minimum of 3 gyrotrons, prefer to have 6 gyrotrons. I-coil: Required with n=3 setup.
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	70: Hybrid Beta Limit at Low Rotation
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine the beta limit (most likely due to the onset of a 2/1 mode) for low rotation hybrid plasmas. Compare the beta limit to the ideal no-wall limit, and to the limit in rapidly rotating plasmas. First study q95=4.2, and if time permits study q95=3 and q95=5 as well.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Establish hybrid plasma with q95=4.2 using PCS feedback control of beta_N and rotation. (2) Use counter NBI to reduce rotation rate to minimal value. (3) Program a slow ramp up in beta_N to determine the stability limit (likely limit is onset of 2/1 mode). (4) Repeat beta_N ramp using only co-NBI. (5) If time permits, repeat at q95=3 and q95=5.
Background:	The bulk of hybrid study on DIII-D has been with co-NBI, producing rapid plasma rotation. For q95~4, the beta limit appears to coincide with the ideal no-wall stability limit. At q95~3, the beta limit is about 80% of the ideal no-wall limit. At q95~5, some hybrid plasmas have exceeded the ideal no-wall limit (probably owing to rotational stabilization). However, it is expect that in future larger devices like ITER the normalized rotation rates will be much smaller. Experiments in standard H-mode plasmas indicate that the beta limit decreases at lower rotation. Therefore, we need to measure the beta limits for low rotation hybrid plasmas at different q95 as part of our validation of this scenario for ITER.
Resource Requirements:	NBI: All 7 sources are required.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	71: ECCD in High Beta Poloidal Plasmas
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	P. A. Politzer, R. Prater, M. Choi	ITPA Joint Experiment :	No
Description:	Use off-axis ECCD for current profile control in high beta poloidal plasmas with large bootstrap current fraction. Also test the effect of beta poloidal on the ECCD efficiency.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Establish high beta poloidal plasmas with high bootstrap current using two gyrotrons for central heating. (2) Inject ECCD off-axis (4 gyrotrons) and measure the effect on the current profile evolution. Scan the resonance location radially, and scan the toroidal injection angle to compare the effects of co and counter ECCD. (3) Use an open loop method of adjusting the ECCD power and location to give the optimal current profile for the high bootstrap current experiments. (4) Keeping ne and Te fixed, raise the plasma current to lower beta poloidal. Measure the effect on the ECCD efficiency at various radial locations.
Background:	Demonstrating current profile control is an important next step for showing the utility of the ECCD system on DIII-D. High beta poloidal plasmas are a good candidate for this since they have little ohmic current (thus the back EMF effects are small). Current profile control may be a benefit to optimizing the high bootstrap current experiments. Furthermore, we can use these plasmas to extend our ECCD database by comparing the ECCD efficiency in plasmas with the same ne*Te but different Ip. This will be an important test of the effect of beta poloidal on the ECCD efficiency (high beta poloidal is expected to reduce trapping effects and increase the current drive efficiency).
Resource Requirements:	NBI: At least 4 sources. EC: Minimum 5 gyrotrons.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:

Title	72: Dependence of Stiffness on Elongation
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Scan the temperature gradient at fixed density using a power scan to determine the stiffness of transport. Repeat this at various values of plasma elongation. Compare results with GYRO, TGLF, GLF23, and MM theory-based models.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Establish H-mode plasmas with density control. (2) Scan NBI power to vary the ion temperature gradient. (3) Try to maintain constant ExB shear and Ti/Te ratio. (4) Repeat NBI power scan at elongations between 1.5 and 2.0.
Background:	Confinement databases have not been able to clearly distingush between the GLF23 and Multimode theory-based models, probably because the most sensitive parameters are not clearly varied. Perhaps the largest difference between GLF23 and MM is the level of transport stiffness, especially in the outer regions of the plasma. In addition, these models have a very different elongation scaling of the stiffness value. Studying this experimentally should help us to validate (or disprove) these two theory-based models.
Resource Requirements:	NBI: At least 5 sources. EC: Minimum 4 gyrotrons.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	73: Measurement of Inductive Poloidal Current
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	P. A. Politzer, T. Suzuki	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	74: Extreme Off-Axis ECCD
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	R. Prater, J. Lohr, M. Choi	ITPA Joint Experiment :	No
Description:	Use the higher ECH power available this year and MSE system to test the physics of ECCD in the outer regions of the plasma, 0.5< rho_ec < 0.9, especially for cases where large trapping effects are expected. The current drive determined from MSE signals will be compared to theoretical models.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	These experiments can be done in either L-mode or H-mode plasmas, but H-mode is preferred since it is the standard operating mode for DIII-D. Co/counter current drive comparisons should be done. (1) Scan the ECCD location across the midplane radius on the high magnetic field side of the plasma from 0.5 < rho_ec < 0.9 by varying the toroidal magnetic field. (2) Scan the poloidal angle of the ECCD location at fixed rho by varying BT and the antenna steering for rho=0.5 and rho=0.8. (3) Scan the ECCD location vertically at a poloidal angle of 90 deg from 0.5 < rho_ec < 0.9 by varying the antenna steering.
Background:	Previous experiments on DIII-D have studied the effects of electron trapping and beta for ECH locations between 0.0 < rho_ec < 0.4 owing to limited power. However, many important applications of ECCD (such as NTM stabilization and AT sustainment) require current drive locations that are further off-axis. It is important to test the physics of ECCD in this situation so that predictive theoretical models can be used to guide the application of ECCD in future experiments. Furthermore, it is expected that the Ohkawa current drive mechanism (reverse ECCD from electron trapping) may dominate the Fisch-Boozer current drive mechanism (forward ECCD from reduced collisionality) far off axis, especially when the power deposition is on the low field side. It is important to test our theoretical models in this limit.
Resource Requirements:	NBI: Co and counter MSE beams, and at least 2 other sources. EC: Minimum 5 gyrotrons.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:

Title	75: Sustained Monster Sawteeth
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	R. I. Pinsker, W. W. Heidbrink	ITPA Joint Experiment :	No
Description:	Sustain monster sawteeth for > 2 s by using a combination of ICRF for sawtooth stabilization and off-axis ECCD for q(0) control.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Create monster sawteeth using fast wave absorption on beam ions. This could be done at 60 MHz in low density plasmas. (2) Apply off-axis co ECCD outside the q=1 surface (around r/a=0.3-0.4) to prevent q(0) from dropping to 0.8. Try a few different locations and power scans. (3) If any fast wave power is not being used for beam ion heating, try counter FWCD to observe effect on q(0) (not expected to work as well as off-axis ECCD).
Background:	Monster sawteeth crash when q(0) drops to a critical value. Using off-axis ECCD to halt the decreasing evolution of q(0) should allow monster sawteeth to be sustained indefinitely.
Resource Requirements:	NBI: At least two sources. EC: Mimimum 6 gyrotrons. FW: Need >1 MW at 60 MHz.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:

Title	76: Simulation of Alpha Channeling Current Drive
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	K. L. Wong, W. W. Heidbrink	ITPA Joint Experiment :	No
Description:	Demonstrate the practicality of using TAE's to selectively sweep co-moving energetic particles to larger radius, resulting in localized current drive to reduce or reverse the central magnetic shear even if the initial fast particle population is isotropic. Use the fast D-alpha diagnostic to verify the redistribution of energetic ions.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Inject balanced beams into low current L-mode plasma. Optimize the beam sources for the collection of MSE data. (2) Decrease the plasma density until the TAE's become unstable. Document the change in the current profile. (3) Compare with different NBI power levels and different mix of left and right sources. (4) Compare with co-only NBI and counter-only NBI.
Background:	Previous experiments on DIII-D with co-NBI showed that TAE modes can displace the energetic particles to larger radius, broadening the noninductuve current drive profile. This can be thought of as a simulation of the alpha channeling effect in a burning plasma. Since only co-moving fast ions can resonate with the TAE's and get ejected from the core towards the edge, this process allows local noninductive current drive to be obtained from an initially isotropic alpha particle distribution. To better simulate this alpha channeling effect on DIII-D, we will inject balanced beams and then use the TAE's to spatially separate the co-moving and counter-moving fast ions. While this may not result in much net current drive, it should dramatically change the magnetic shear profile. The spatial redistribution of the energetic ions can also be verified by the total pressure profile determined by the MSE diagnostic, and the fast D-alpha diagnostic.
Resource Requirements:	NBI: Co and counter MSE beamlines (4 sources total).
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:

Title	77: Reducing the power loads to the divertor in ELM-supressed phase
Name:	Jakubowski marcin.jakubowski@ipp.mpg.de	Affiliation:	Max-Planck Institute for Plasma Physics
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Todd Evans, Max Fenstermacher, Mathias Groth, Charles Lasnier, Oliver Schmitz, Phil West; Jonathan Watkins	ITPA Joint Experiment :	Yes
Description:	Application of the RMP to the plasma boundary at DIII-D leads to complete ELM elimination, which reduces the heat peak loads significantly. At low collisionalities it is unfortunately accompanied by an increase of total heat flux to the divertor and decrease of radiated power. The hypothesis is that it is caused by strong decrease of the sheath potential due to hot electrons from the pedestal area reaching the divertor surface. Our expectation is that seeding of small amount of radiating gas into the scrape-off layer would enhance Prad and reduce the energy deposited to the target plates.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Application of n=3 dominant perturbation with C-coils to the ITER Similar Shape and ITER-like collisionality plasma is envisaged with q95 in the range compatible with complete ELM suppression. Using the gas inlet located at the plasma midplane a small amount of Helium would be puffed into the scrape-off layer during the RMP phase. One needs to find a proper amount of injected gas, which would not cause MARFEs, but still keep the target power loads on the low level. Analysis of the probe data together with spectroscopic diagnostic can help to confirm the hypothesis of changes in the divertor power loads due to decreasing sheath potential
Background:	Heat loads to the plasma facing components are one of the key questions for the future safety of the fusion devices like ITER or DEMO. As one of the most promising methods to control the power exhaust in poloidally diverted tokamak is application of RMP there were numerous experiments performed at DIII-D for different plasma configurations including low and high triangularites and for different electron pedestal collisionalities. Independent on the plasma triangularity all discharges at electron pedestal collisionality below 0.5 show increase of the heat flux to the target plates (about 15% of total heating power). It seems to be caused by the strong decrease of the floating potential due to very hot electrons from the pedestal area hitting the divertor surface. At very low nu^star they are able to follow the stochastic field lines up to the target plates. Lower floating potential leads to the higher rate of energy transfered to the ions in the sheath and thus increase of the power loads is observed. At the same time we observe decrease of the radiated power in the scrape-off layer. It is expected that enhancing the radiation in the scrape-off layer would allow to keep the averaged power loads on the pre-RMP level and still benefit from significantly reduced peak heat loads. Proper balance needs to be found between puffing and pumping rate such that plasma would not colapse due to MARFEs, but still maintain Prad on the proper level. At higher collisionalities high energy electrons dissipate energy by collisions and do not affect the sheet potential so effectively.
Resource Requirements:
Diagnostic Requirements:	Filtered camera observing center post (D_alpha, D_gamma + CII/CIII), for diverted case IRTV and lower divertor CCD cameras (Tan_TV and DiMES_TV), fast IR TVs would be desirable, fast UCSD CCD camera with D_alpha/CII/CIII filter; Langmuir probes and fast thermocouples
Analysis Requirements:	Post exp: Reconstruction of bolometric and spectroscopic data; Use of TRIP3D to analyze connection of the highest heat flux/lowest floating potential area to the stochastic boundary near the pedestal.
Other Requirements:

Title	78: Separating Rotational Shear and rho* Scaling Effects on Transport
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	T. C. Luce, R. E. Waltz	ITPA Joint Experiment :	No
Description:	Measure the rho* scaling of transport and fluctuations for both L-mode and H-mode plasmas as a function of the toroidal Mach number to ascertain the importance of the rotational shear on question of Bohm vs. gyro-Bohm scaling.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Study both limiter L-mode and divertor H-mode plasmas on separate days. (1) Start at BT=2 T and q95=4. Scan the beams from all co to all counter in 5 steps (including a balanced case). (2) Go to BT=1 T at the same q95 and same plasma shape. Reduce density by factor of 2.5. NBI power should be adjusted to match beta values of first step. (3) Repeat scan of beams from all co to all counter in 5 steps. (4) Fine tune the mix of co and counter NBI to precisely match the Mach numbers between the low and high rho* cases.
Background:	Transport modeling of rho* scaling experiments, using either theory based modeling or turbulence simulation codes, have predicted that changing rotational shear during a rho* scan can make intrinsic gyro-Bohm transport appear to be Bohm-like. Experiments were begun (but not finished) on TFTR to separate the two effects by measuring the rho* scaling at different toroidal Mach numbers. It is proposed to use the balanced NBI capabilities on DIII-D to measure the rho* scaling of transport and turbulent fluctuations for both L-mode and H-mode plasmas as the Mach number is scanned for co-rotating to counter-rotating. This should allow multiple pairs of rho* scans at fixed rotational shear to be obtained.
Resource Requirements:	NBI: All seven sources required (not simultaneously).
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	79: Modulation of Bootstrap Current
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	B. Hudson	ITPA Joint Experiment :	No
Description:	Directly measure the bootstrap current profile near the H-mode pedestal by modulating either (1) the edge ECH heating power, or (2) the I-coil current, and observing the oscillating MSE/LIB response.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Establish QBD discharge using 210 beamline with good MSE and Lithium Beam diagnostics at relatively high field. (2) Aim ECH for power deposition near the top of the H-mode pedestal. Modulate all gyrotrons using several different frequencies (2-10 Hz). (3) Vary some parameter that should effect the magnitude if the bootstrap current (such as q95) and repeat. (4) Try modulating the H-mode pedestal pressure by modulating the I-coil current.
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. The best method of modulating the bootstrap current is to apply 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]. QBD plasmas make good target discharges since the modulation effects of the ELMs are not present and the discharges last a long time.
Resource Requirements:	NBI: Co and counter MSE beams, and at least 2 more sources. EC: Mimimum 5 gyrotrons.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:	Modulated I-coil current.

Title	80: Electron Heat Pinch
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	T. C. Luce, M. E. Austin	ITPA Joint Experiment :	No
Description:	(1) Demonstrate the existence of an electron heat pinch with convincingly small error bars. (2) Determine if the electron heat pinch is dependent upon the sign of the magnetic shear as predicted by some theories.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Reproduce the previous best cases of the heat pinch using the maximum ECH power available this year (up to 6 gyrotrons). The poloidal steering of the ECH antenna will allow variation of the heating location at constant target parameters. (2) Modulate the ECH power to obtain direct evidence on inward heat transport. (3) Use early off-axis ECH in low density plasmas to form negative shear discharges with rho_qmin > rho_ech. Measure the behavior of the Te profile as the q profile relaxes to a low shear state. (4) Repeat at several different densities (different Te) to vary the amount of shear reversal.
Background:	Understanding electron transport requires that the effects of diffusion and convection be distinguished, and the heat pinch issue falls squarely into this area. The inward transport effect seen with off-axis ECH remains a severe challenge to the theoretical community. Previous experiments using low field side, second harmonic launch found a negative electron heat flux near the plasma center, in support of the old 60 GHz ECH data from 1989-1990. However, due to higher ion temperatures, even in the ohmic phase, than 1989-1990, the error bars include positive (although small) values for the electron thermal diffusivity. This experiment will use the higher ECH power available this year to increase the separation betweem Te and Ti and thus reduce the error bars. Furthermore, the theoretical heat pinch model of coupled transport between Grad-J and Grad-T can be tested by comparing the non-diffusive electron transport for positive and negative shear plasmas.
Resource Requirements:	NBI: At least 2 sources. EC: Mimimum of 6 gyrotrons.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	81: Pedestal width scaling with toroidal field direction
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	R. Groebner, T. Osborne, P. Snyder	ITPA Joint Experiment :	No
Description:	Compare the H-mode pedestal width and general structure with toroidal field direction. Particularly compare include the electron density height and width and ELM characteristics. Also carry out a global beta scan.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Setup standard ELMing H-mode in a configuration shape that can be repeated for both upper and LSN discharges. Previous experiments have shown that changes to the pedestal density due to toroidal field changes were not dependent on the geometry of the divertor surface and baffles. Carry out a global beta scan in LSN with GradB towards the divertor to make sure the pedestal pressure increases with global beta as in previous experiments. Carry out beta scan in USN with exact same geometry and all other input parameters fixed.
Background:	This experiment is designed to investigate two previous observations that do not fit the EPED1 pedestal height scaling. The first of these is the height of the pedestal dependent on the toroidal field direction. Previous controlled experiments have found that while the pedestal temperature may remain fixed the pedestal density and the pedestal pressure can be less in a configuration with the GradB drift direction away from the primary X-point. This result was found not to be dependent on the divertor surface and baffle geometry. The second observation was for a scan of global beta in a LSN hybrid scenario the pedestal height did not increase with global beta. This experiment would examine these to observations to isolate the controlling aspects that led to the observed scaling.
Resource Requirements:	DN configuration with upward and downward biased single null configurations. Beam heating required
Diagnostic Requirements:	Full pedestal diagnostic complement
Analysis Requirements:	Profile and stability analysis
Other Requirements:

Title	82: ITER pedestal dependence on separatrix shape
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	T. Osborne, R. Groebner, P. Snyder, T. Luce	ITPA Joint Experiment :	No
Description:	This experiment will examine in detail pedestal dependence on shaping parameters in the ITER configuration. Upper shape parameters, including triangularity, squareness, DRSEP and location of the upper secondary X-point will be varied over ranges that are obtainable in ITER. Lower triangularity will also be varied to also test that dependence. Finally RMP ELM control effectiveness can also be examined as a function of these shape parameters.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	First a series of target shapes must be created. These will be shapes of varying upper triangularity that would be compatible with the ITER plasma facing components and poloidal field control coils. Triangularity is one shape parameter to be varied, but also squareness and DRSEP should also be explored. Preliminary analysis with ELITE will determine which shapes should produce a significant variation in edge stability. The lower triangularity dependence should also be examined. About 4-6 shapes should be chosen as targets for the scan. The experiment would then reproduce these shapes at the ITER proposed q and the pedestal and ELM characteristics measured. RMP coils should also be utilized to examine the tradeoffs of ELM mitigation with shape variations. The resulting pedestal profiles for all cases should then be examined with ELITE to determine consistency of experiment with the models.
Background:	ITERâ??s current baseline operational scenario utilizes a single LSN shape. This shape was optimized for the highest upper and lower triangularity that could be reasonable obtained. The proposed upper triangularity in ITER will produce a DRSEP of approximately 4cm. This may require significant shielding for the upper plasma facing components near the upper secondary X-point. The shape should be optimized in a more quantitative fashion. This experiment will attempt to determine the trade off between pedestal height and the requirement for shielding due to the location of the secondary upper X-point. The lower triangularity will also be examined to determine if the advantages of reducing the lower triangularity may outweigh any reductions in pedestal height. RMP ELM control may also need quantitative optimization with regard to shape and this experiment could also shed some light on that issue.
Resource Requirements:	ITER demonstration configuration with beam heating
Diagnostic Requirements:	Full pedestal diagnostic complement
Analysis Requirements:	Profile and stability analysis.
Other Requirements:

Title	83: RMP enhanced particle transport threshold dependence on collisionality
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Osborne	ITPA Joint Experiment :	No
Description:	Characterize enhanced pedestal transport and ELM suppression as a function of Ip at constant q. By varying plasma current and toroidal field a factor of 3 variation in pedestal collisionality at constant Greenwald density fraction may be obtained. The RMP current would be varied at each field strength to determine the collisinality threshold for onset of pedestal transport enhancement and ELM suppression.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Setup ELMing H-mode in an ITER similar shape at full toroidal field, 2.1 T, and optimal q for RMP enhanced pedestal transport and ELM suppression. Vary the toroidal field to 1.4 T and 0.7 T with plasma current for constant q. At each field strength adjust the density to maintain constant Greenwald density fraction. This should lead to a factor of 3 variation in pedestal collisionality with the usual pedestal scalings. For each field strength carry out an RMP current scan to determine the relative field strength perturbation required for enhanced pedestal transport and ELM suppression. All the available diagnostics should be used to characterize the pedestal over the scans.
Background:	In scaling present RMP experiments towards ITER the collisionality dependence remains an open question. In current devices RMP ELM suppression requires low density. If collisionality is the most important factor determining the onset of enhanced transport then ITER may be able to operate with suppressed ELMs at higher normalized density than current devices. This scaling will be a critical factor in designing an operational scenario for ELM control in ITER.
Resource Requirements:	RMP coil operation in ITER similar shape
Diagnostic Requirements:	Full pedestal diagnostic complement
Analysis Requirements:	profile and stability analysis. Edge fluid and neutral Monte Carlo analysis.
Other Requirements:

Title	84: Density dependence of pedestal width scaling
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	R. Groebner, T. Osborne, P. Snyder	ITPA Joint Experiment :	No
Description:	Measure pedestal characteristics at high density over a wide range of conditions by varying plasma current and shape. This experiment will be the high density complement to last yearâ??s pedestal scaling experiment.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Target discharges include high and low plasma current at constant toroidal field and constant q. High and Low triangularity discharges should also be targeted. In each configuration a moderate density and high density should be obtained to complement the low density points obtained last year. Slow separatrix sweeps should be included to obtain the best pedestal data. Care should be taken to get the best data at conditions where the narrowest pedestal and widest pedestal should be expected. The narrower pedestals should occur at low shaping and low q and low global beta. The wider pedestals are expected at high shaping, high q and higher global beta.
Background:	Good agreement was obtained in a data set last year that examined the scaling of the pedestal width with the square root of pedestal poloidal beta. However over this data set taken at the low end of the natural H-mode density, the pedestal poloidal beta was strongly correlated with the ion poloidal gyro-radius. For this data a poloidal gyro-radius scaling fit the width data as well as the pedestal poloidal beta scaling. This experiment is designed to break that correlation so that a pedestal beta scaling can be distinguished from a gyro-radius scaling.
Resource Requirements:	Patch panel change required at midday between DN and LSN configuration
Diagnostic Requirements:	Pedestal diagnostics
Analysis Requirements:	Profile and stability analysis
Other Requirements:

Title	85: Wide density pedestal beyond the neutral penetration depth
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	R. Groebner, T. Osborne, P. Snyder	ITPA Joint Experiment :	No
Description:	Obtain a density pedestal much wider than the neutral penetration depth, thus demonstrating the important of a density pinch in determining the pedestal density structure.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Setup conditions that maximize pedestal width and minimize neutral penetration length. Wide pedestals are obtained with strong shaping, high q and high plasma beta. Minimum neutral penetration occurs for high density that can be obtained at high current. Optimizing these conditions leads to a high triangularity LSN discharge at high global beta at maximum toroidal field. Plasma current should probably be less than 1.2 MA but will be scanned to maximize pedestal density width. Density would then be scanned, through gas puffing, to obtain the widest pedestal density. Other pedestal characteristics would also be obtained through slow separatrix sweeps. ECH power may also be applied to lower the neutral beam power required to achieve the target beta and thus reduce the central fueling. The data would eventually have to be modeled with fluid codes and neutral Monte Carlo codes to interpret the neutral and density behavior.
Background:	At this time is unknown what role a particle pinch may play in determining the pedestal density structure. A number of observations find the pedestal density width expanding as the neutral penetration width is decreasing. A density pinch could explain this observation. The presence of a particle pinch in the pedestal will have significant implications for pellet fueling requirements and divertor heat flux control at the core target density.
Resource Requirements:	Maximum ECH power available
Diagnostic Requirements:	Pedestal characterization diagnostics
Analysis Requirements:	Profile and stability analysis, Edge fluid and neutral Monte Carlo analysis
Other Requirements:

Title	86: L-H transition dependence on divertor detachment
Name:	Leonard leonard@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Measure the 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 :	Yes
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:	SN configuration
Diagnostic Requirements:	upper divertor characterization and X-point Mach probe
Analysis Requirements:
Other Requirements:

Title	87: Test of Neoclassical Toroidal Viscosity theory using modulated I-coil currents
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	A. Garofalo, W.M. Solomon	ITPA Joint Experiment :	No
Description:	Use modulated I-coil currents to investigate the theory of braking of plasma toroidal rotation by non-resonant error fields	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Modulate the I-coil currents to modulate the non-resonant drag on the plasma. Investigate the effects as a function of modulation frequency, background plasma rotation, collisionality and I-coil parity.
Background:	This experiment was given 1/2 day in 2008. Unfortunately, there were issues of machine cleanliness since it was run after an experiment with significant gas puffing. Accordingly, the QH-modes were poor. Attempts to perform this experiment in ELMing H-mode lead to locking of the ELM frequency to the I-coil modulation. Although this locking was a significant discovery, the ELM effects on the rotation masked the direct I-coil effects. We need to perform this experiment in high quality QH-mode plasmas, since this avoids the ELM problem while still allowing us to probe H-mode plasmas.
Resource Requirements:	I-coil system connected to do both error field correction and n=3 braking. QH-mode will probably require reversed plasma current unless the development of co-NBI QH-mode provides long, steady QH-mode with co-rotation.
Diagnostic Requirements:	All profile diagnostics. CER at high enough speed to have 10 samples per I-coil modulation period.
Analysis Requirements:
Other Requirements:

Title	88: Feedback Control of q_min Using BT Ramping
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Sustain the high qmin target profile by (1) combined Ip and BT ramps to control the inductive current profile, and (2) better aligning the sources of noninductive current drive and reducing the plasma density. The real time MSE EFITs in the PCS will be used to track the evolution of q_min in these AT discharges. Downward ramps of BT will be used by the PCS to apply an off-axis parallel electric field to drive an off-axis inductive current to maintain q_min at the target value. An Ip ramp can be added to keep q_95 fixed, if desired. The central NBCD over-drive will be eliminated by using a more balanced beam injection. To increase the effectiveness of off-axis ECCD, the lower divertor cryopump will be used for density control in this high triangularity plasma shape. The off-axis location of the ECCD will be determined by code analysis to give a flat loop voltage profile with the desired high qmin current profile.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Two days are required for this experiment, with high power ECCD only being required on the second day. The first day will concentrate on developing feedback control of the "flat shear" scenario with combined Ip and BT ramps, as well as improved density control. The mix of co and counter NBI will also be adjusted to eliminate the central current over-drive. Between shot analysis of the loop voltage profile will guide the experiment. After the first experimental day, off-line analysis of the current profile evolution will determine the optimal location for the ECCD to achieve a flap loop voltage profile which will be close to zero for 100% noninductive current drive. The choice of which BT value to end the toroidal field ramping is part of this consideration. On the second experimental day, the off-axis ECCD will be added to sustain the desired current profile with high q_min. Again between shot loop voltage analysis will help guide the experiment. High beta will help increase the bootstrap current for more off-axis current drive. Since the tayloring of the noninductive current profile will better decouple the heating profile from the q profile, a push to high beta values in the second day of this experiment may be more successful than previous attempts.
Background:	Analysis of the "flat shear" experiments in the AT group, which include the effect of ramping the toroidal field on Ohm's law, showed that the sources of noninductive current drive were not well aligned with the target current profile. The loop voltage profile was negative near the axis and positive near the edge, which indicates that there was too much central current drive and not enough off-axis current drive. As a result, even though the integrated noninductive current fraction was close to 100%, the current profile continued to evolve until a q=2 surface appeared in the plasma which lowered the beta limit. If a better aligned noninductive current profile is utilized, then the targer high qmin profile can be sustained. The central NBCD overdrive can be easily corrected in 2006 by substituting some counter NBI. To increase the off-axis current drive, it will first be necessary to lower the plasma density. The previous experiments were essentially unpumped because of the high triangularity plasma shape; however, the modification to the lower divertor will now make it possible pump on these plasmas. (The optimal direction of BT for optimal pumping will need to be determined in separate experiments.) Using six long-pulse gyrotrons will then allow increased off-axis current drive to be obtained.
Resource Requirements:	NBI: At least 5 co-sources. EC: For 2nd day of experiment, need 6 gyrotrons.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:

Title	89: Main ion poloidal rotation measurements in helium plasmas
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	J.S. DeGrassie, W.M. Solomon	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to test the neoclassical theory of poloidal rotation by measuring the rotation of the main (helium) and impurity (carbon) ions under a variety of conditions. These will be in Ohmic, ECH and NBI H-modes.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	These experiments will build on the ECH H-mode developed in helium plasmas during the 2003 and 2004 campaigns. In addition, they will utilize the ability of the CER system to make high time resolution measurements of the plasma rotation just after beam turn on. This allows us to use the beams to make the measurement before they have had time to alter the rotation. Using combinations of ECH and beams, we will make L-mode and H-mode plasmas, some totally RF dominated and others beam dominated. The beam dominated shots will cover cases with significant co-injection, balanced injection and counter injection.
Background:	Because of its role in affecting E x B shear stabilization of turbulence, developing a predictive understanding of plasma rotation is an essential part of transport studies. In addition, the theory of the neoclassical toroidal viscosity contains an offset velocity which is directly related to the poloidal rotation. Having a predictive understanding of this offset is essential if we are to predict the plasma rotation in shot with significant nonresonant error fields. Measurements in 2004-2005 presented by W. Solomon at the 2005 APS DPP meeting showed that the impurity poloidal rotation did not agree with neoclassical predictions in QH-mode and RMP ELM-suppressed H-mode plasmas. However, to date, there has not been a good test of the theoretical predictions for the main ion rotation. One possible explanation for the discrepancy seen by Solomon et al is the effect of friction with the neutral beam ions. Performing experiments both with and without NBI will allow us to investigate this possibility. In addition, having measurements of the main and impurity ions allows provides significantly more information, since the coupling to the fast ions is different. During the LTOA, we installed a improved detectors for the poloidal views for the CER system; we are now in a position to make much better measurements of poloidal rotation.
Resource Requirements:	4 gyrotrons. CER beams plus 2 more. Helium beams required
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	90: Simultaneous Suppression of 3/2 and 4/3 NTM with ECCD
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Use co-ECCD at both the q=1.33 and q=1.5 surfaces to simultaneous suppress both the 4/3 and 3/2 tearing modes. Determine if an improved confinement state can be obtained, relative to plasmas with EC heating but no current drive (or NTM suppression). Determine if the beta limit is significantly affected. If sufficient gyrotron power exists, this idea can be extended to the simultaneous suppression of the 3/2 and 2/1 modes.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Divide the gyrotrons into two groups. One group is aimed at the q=1.33 surface near the inner midplane. Its location is controlled by varying either R or BT. The second group is aimed at the q=1.5 surface by poloidally steering the antennae upwards. Its location is controlled by either real-time steering or by shifting the plasma vertically. (2) Prepare target hybrid plasmas with beta_N=3 at either q95=4.4 or q95=3.1. (3) First suppress only the 3/2 NTM to determine the minimum amount of ECCD power required. Compare these cases with fiducial discharges with EC heating only. (4) With the 3/2 NTM suppressed, use remaining gyrotrons (but no more than necessary) to also suppress the 4/3 NTM that becomes dominant after the 3/2 mode is gone. Compare this with a fiducial case where the gyrotrons aimed at the q=1.33 surface are changed to radial injection. (5) Compare the plasmas with simultaneous 3/2 and 4/3 suppression to fiducial plasmas with no NTM suppression but same amount of EC heating. (6) For plasmas with both 3/2 and 4/3 suppression, determine beta limit for 2/1 NTM.
Background:	When the 3/2 NTM is suppressed by ECCD, the 4/3 NTM becomes unstable. While the 4/3 mode is likely more benevolent than the 3/2 mode (not proven for the case of ECCD suppression because of the deleterious effects of electron heating on confinement), the 4/3 mode is still expected to have some negative consequences. Therefore, it would be interesting to suppress both the 3/2 and 4/3 modes using ECCD. Since it only requires 3-4 gyrotrons to stabilize the 3/2 NTM, it is expected that the remaining gyrotrons can stabilize the 4/3 NTM which is at smaller radius. This may result in higher confinement, but to prove this fiducial discharges with the same electron heating but no current drive need to be studied.
Resource Requirements:	NBI: All 5 co sources are required. EC: Minimum of 6 gyrotrons.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:

Title	91: Prompt torque and zonal flow damping
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	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 for QH-mode. 7 NBI sources. 3-4 ECH gyrotrons
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	92: Test neoclassical poloidal rotation prediction as a function of toroidal rotation speed
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	W.M. Solomon, S.K. Wong	ITPA Joint Experiment :	No
Description:	Test predictions of an extended version of neoclassical theory which says that the poloidal rotation of ions in the plasma should vary as the the toroidal rotation changes	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Utilize QH-mode plasmas with modulated neutral beams for best rotation measurements in a high temperature, H-mode plasma. Scan toroidal rotation by varying the co-counter beam mix. Investigate poloidal and toroidal rotation speed of various impurities: helium, carbon, neon and argon.
Background:	Measurements of carbon and neon poloidal rotation by Solomon et al. [Phys. Plasmas 13, 056116 (2006)] showed a significant discrepancy between experimental measurements and the predictions of neoclassical theory embodied in the NCLASS code. Recently, S.K. Wong has pointed out that the standard neoclassical prediction for poloidal rotation needs to be modified when the main ions or the impurities are rotating at speeds comparable to their individual thermal speeds. In addition to Wong's analytic derivation, this physics is now embodied in the NEO code written by E. Belli and J. Candy. In the experiment by Solomon et al, the carbon and neon impurities were rotating at near their thermal speeds. The goal of this experiment is to make a detailed test of this extension of neoclassical theory.
Resource Requirements:	Reverse Ip. 7 NBI sources.
Diagnostic Requirements:
Analysis Requirements:	Comparison of experimental results with NEO code
Other Requirements:

Title	93: Diagnostic spatial cross calibration using edge sweeps in QH-mode
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	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 so that we can finally include the MSE views of the 210 beam. This MSE portion of the calibration is particularly important, since the relative location of the 210 system relative to the other MSE systems has never been established. Establishing this location is an essential first step in using the co plus counter MSE views to determine the edge current density.
Resource Requirements:	Reverse Ip. 7 NBI sources
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	94: Dependence of Intrinsic Rotation on Density/Safety Factor
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Using balanced NBI, determine whether the "onset" of intrinsic rotation is related to the evolution of the density profile or the safety factor profile. The effect of EP modes should be avoided. Hopefully with will yield some insight as to the physics behind the intrinsic rotation. When injecting balanced beams early in the current ramp phase of discharges on DIII-D, the central rotation will initially be close to zero, or perhaps slightly negative, Around 0.6 s, the central rotation will transition to a strongly positive value, i.e. intrinsic rotation appear. Since the plasma is still strongly evolving during this period, it is not clear what the important physics is behind the development of co rotation. In general, the density is increasing, the minimum in safety factor is decreasing (passing through 3 at around this time), the central magnetic shear is evolving from negative to positive, and EP modes may be present. A systematic experiment needs to be done to sort out the possibilities.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Inject balanced NBI early during the current ramp up (typically around 0.3 s) of L-mode plasmas. Vary the current evolution using ECH; the goal is to vary the time that q_min passes through rational values such as 3,2, etc. Vary the density evolution with gas puffing. Ideally the ECH and gas puffing can be traded off such that the safety factor evolution is similar to different densities.
Background:	TCV has reported that in ECH plasmas, spontaneous co rotation (aka intrinsic rotation) develops after the density is ramped up beyond a threshold value. Experiments on C-Mod did not confirm this, and perhaps even contradicted the TCV results. In fact, the dependence of the co rotation on LHCD suggests to the author that the current profile may be playing the dominant role rather than the density profile.
Resource Requirements:	NBI: 30LT and 210RT required. ECH: Need 6 gyrotrons.
Diagnostic Requirements:	MSE is critical.
Analysis Requirements:
Other Requirements:

Title	95: Te/Ti scaling of electron heat transport in ETG-dominated H-mode plasmas
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport Model Validation	Presentation time:	Requested
Co-Author(s):	T.L. Rhodes, A.E. White, W.W. Peebles, J.C. Hillesheim, C. Holland	ITPA Joint Experiment :	No
Description:	Determine Te/Ti scaling of ETG turbulence level and associated electron heat transport in ELM-free H-mode/QH-mode plasmas where ITG/intermediate scale turbulence is suppressed by ExB shear. Test predictive capability of TGLF/XPTOR in an ETG-dominated H-mode regime. Determine ETG critical gradient experimentally and compare with model predictions.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Measure the local intermediate/small-scale turbulence level in counter-injected, Elm-free H-mode/QH_mode plasmas (r/a ~0.5) where electron heat transport is predicted to be ETG-dominated (2 < k rho_s < 5 is accessible by DBS in the core, k rho_s ~15 by the high-k backscattering system). 2) Vary Te and Te/Tias much as possible by adding core ECH and reducing P_NB. Obtain turbulence level and compare predicted electron heat flux with transport analysis results. 3) Flatten the local electron temperature gradient at r/a~0.5 by using off-axis ECH (r/a>0.6). Measure turbulence level and attempt to determine the ETG critical gradient. 4) Starting in counter-injected QH-mode, decrease torque and observe re-appearance of intermediate scale and ITG-scale turbulence as the ExB shearing rate in the core is reduced.
Background:	Predicting H-mode electron heat transport in burning plasma is very important since a-particles mainly heat the electrons. Gyrokinetic simulations indicate that electron heat transport is dominated by high-k turbulence (ETG/TEM and ETG, k rho_s once ITG-scale turbulence is suppressed by ExB shear (Goerler and Jenko, PRL2008, Kinsey, PoP 2008). Recent measurements in DIII-D using Doppler Backscattering (DBS) have demonstrated significantly reduced large scale and intermediate-scale turbulence (reduced by a factor >10) in the core and edge of low density, counter-injected H-mode and QH-mode plasmas (Te/Ti ~2, P_NB ~ 7 MW). Core turbulence suppression occurs at the L-H transition as the local core ExB shearing rate exceeds the linear ITG/TEM growth rate (calculated by TGLF). A simultaneous, prompt reduction in core electron heat diffusivity is found from TRANSP analysis. The residual electron heat transport in these plasmas is likely due to ETG modes; the ion transport is found to be neoclassical.
Resource Requirements:	5 Gyrotrons, Beams
Diagnostic Requirements:	DBS, high-k Backscattering, BES (low-k turbulence in outer plasma), BES, PCI, MSE, ECE Optionally CECE to acquire QH-mode data in outer plasma (long time windows)
Analysis Requirements:	TGLF/ XPTOR will be used to determine optimum experimental conditions. Local (flux-tube, r/a~0.5) GYRO runs will be made. Resolution of high-k ETG modes is essential.
Other Requirements:	--

Title	96: Divertor ELM energy deficit and first wall ELM energy fluxes
Name:	Pitts richard.pitts@iter.org	Affiliation:	ITER Organization
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Requested
Co-Author(s):	C. Lasnier, P. C. Stangeby, J. A. Boedo, D. Rudakov	ITPA Joint Experiment :	No
Description:	Attempt to indirectly estimate the ELM-wall interaction using measurements of the divertor ELM energy deficit obtained with IRTV. Measure ELM energy and/or particle load to the upper baffle or divertor regions in single null lower discharges or the lower baffle in single null upper pulses in support of ITER estimate for thermal load specification on first wall surfaces in the second X-point vicinity	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	This experiment requires regular Type I ELMing H-mode with ELMs of sufficient size to allow reasonably accurate estimates of the plasma energy loss per ELM. The idea is to estimate the wall energy deposition during the ELMs by measuring the energy deposited on the divertor targets with high time and spatial resolution IRTV and comparing with the stored energy loss per ELM. At the same time, estimates will be required, if possible, of the radiated energy loss per ELM. Discharges should be run with the outer-wall gap varied so as to observe the change in the relative energy deficit at the divertor as the wall is approached. For these experiments, interaction with the top of the machine should be minimised (choice of triangularity) such that ELM energy is primarily deposited on the outboard main walls. A further important issue is the ELM energy deposited on surfaces intersecting the first limiting flux surface near the second X-point. Discharges to study this should be of the correct triangularity to ensure that this occurs in the upper divertor region. Alternatively, if diagnostics do not allow it, then equivalent (or as similar as possible) ELMs could be created in single null upper plasmas (not clear how possible this is with the grad B drift changing direction) and the ELM energy on the first limiting flux surface at the bottom of the machine could be observed with the existing vessel floor viewing IRTV (though complications may arise due to SOL drifts or rotation affecting the in-out ELM energy balance).
Background:	ITER has just completed an important exercise in thermal load specifications for the divertor and first wall surfaces, both steady state and during transients (notably ELMs and disruptions). These specifications are based on extrapolations from existing tokamak data but many are in need of confirmation/refinement by further an improved experiments. One such issue is that of energy deposition on the first wall during Type I ELMs. By first wall is meant primarily that part of the main chamber surface which intersects with the first open flux surfaces beyond the separatrix (which can be an outboard limiter or the second divertor region in unbalanced double null plasmas). In the ITER standard H-mode scenario the greatest threat comes from ELM energy conducted along field lines striking the upper dump plate area which is the region intercepting the strong fluxes outside the lowser divertor. The ITER first wall shaping design is in progress now (procurement in 2011) and the upper dump plate region, near the second X-point is of particular concern. There are very few experiments which have looked directly main chamber ELM-wall loads, mainly due to the diagnostic issues involved. In the absence of adequate (or any!) main chamber IR views, one approach, which has been attempted at JET, is to look at the deficit in energy between that lost from the plasma and that observed in the divertor using high quality IR measurements of the target loads. Some idea of radiation losses on the ELM timescale is also useful, but may not be a necessity if the absolute ELM energy is not so high. Hardly any experiment has addressed the issue of ELM energy loads deposited in the second X-point vicinity and ITER extrapolations have been based largely on worst case scenarios extrapolated from a small handful of indirect measurements. It is urgent from the ITER first wall design point of view that tokamaks attempt to proivide direct measurements of these ELM fluxes.
Resource Requirements:	Single null lower NBI heat H-mode with regular Type I ELMs and limited upper divertor interactions. Similar ELMs but in a discharge tailored to arrange for the first limiting flux surface to intersect the upper baffle. Or upper single null in H-mode with interaction at the lower baffle
Diagnostic Requirements:	Assuming that main chamber IR viewing is not possible, then the second most ideal situation would be IRTV viewing both upper and lower divertor regions simultaneously. If only lower IRTV is possible then experiments in SNL and SNU can be attempted. All fixed divertor Langmuir probes and any available midplane SOL diagnostics (for far SOL ELM energy fluxes). Best possible pedestal measurements of density and temperature for input to any subsequent modelling. Radiation losses, if possible on ELM timescale, ELM plasma stored energy loss. The new visible camera diagnostics viewing the outside wall regions would be extremely useful.
Analysis Requirements:	High quality IRTV analysis will be required to make this experiment work. Idem for divertor langmuir probes. Experiments with outer gap variations have potential for the provision of excellent data on ELM radial transport using the reciprocating turbulence probes and/or MiMES.
Other Requirements:	--

Title	97: Initial development of an integrated approach for rapid shutdown
Name:	Walker walker@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	Ted Strait, Dave Humphreys	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to use existing capabilities for a first, simple demonstration of an integrated, multi-layer approach to stability control and disruption avoidance. - NTM suppression with ECCD - remedial action if ECCD suppression fails - mitigation of an impending disruption The focus will be on developing the scenario and algorithmic methods for responding to a 2/1 NTM. The response scenario will necessarily contain many decision branch points, for example (parts of this scenario are not yet proven): - NTM suppression with ECCD (either always-on or detect-then-suppress) - If ECCD suppression fails and can't recover, then ... - move plasma to neutral point - if island is small and saturated, then ... - continue to apply ECCD - transition to soft shutdown - execute soft shutdown: drop kappa, beta, Ip - if other faults detected during soft shutdown, then ... - ... follow another set of fault response logic ... - if island is too large (danger of disruption), then ... - apply RMP - apply counter ECCD to further destabilize mode - apply massive gas injection - if large runaway current arises, then ... - hold plasma at neutral point - them move plasma toward sacrificial surface - otherwise ... - apply heating and/or current drive (proposed) to reduce: - drive voltage for runaways - current quench rate The eventual target is to demonstrate routine detection and mitigation of all disruptions, but the plan is to build this capability in a stepwise manner over several operational campaigns.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	- Implement logic in the PCS to handle this type of response scenario with many decision points - Many portions of this algorithmic approach would be implemented and tested in a non-interfering manner. E.g. detection of impending disruption from NTM near end of discharge in piggyback experiments. - For a dedicated experiment: - Establish a plasma in which the NTM is avoided by pre-emptive ECCD at the q=2 surface. - Use the dud detector to reduce the beam power if an n=1 rotating mode appears. Demonstrate this during programmed or unprogrammed reductions of ECCD power. - Configure the dud detector to trigger massive gas injection if the rotating n=1 mode locks. If such events do not occur naturally, induce them by adjusting the NB power in the dud phase. Much of this experiment could be done in combination with an experiment on ECCD stabilization of the 2/1 NTM. The final step of applying MGI might need some dedicated shots, though the gas species and quantity could be selected for minimum after-effects. This experiment would work well in oordination with other experiments in the "Rapid Shutdown Schemes for ITER" task force to implement what is learned in those experiments into the PCS, or alternatively, to use the PCS logic developed under this proposal to perform control in other experiments under this task force.
Background:	An integrated system for stability control, disruption avoidance, and disruption mitigation is crucial for ITER, and is an important element of the next DIII-D five-year plan. This experiment combines several of the building blocks of such an integrated system, and should provide a foundation on which additional features can be added. Rapid shutdown schemes for ITER will require development of effective response scenarios as well as development of algorithmic methods by which such scenarios can be executed in real-time. Although the DIII-D PCS appears to contains the infrastructure necessary to support such integrated response scenarios, multi-layer responses of the complexity anticipated for ITER have not been attempted. An important part of this experiment is to push the capabilities of the PCS and to expand them if necessary to support such integrate response scenarios.
Resource Requirements:	Machine Time: 4 hour experiment + piggybacks Number of neutral beam sources: >=4 PCS programming support for implementation of detection and triggering logic.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	98: Determination of stochastic heat transport by RMP in a non-rotating plasma
Name:	Hudson bfhudson@ucsd.edu	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Evans	ITPA Joint Experiment :	No
Description:	Characterize nature of edge transport during RMP	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use balanced NBI plus ECH to achieve beta > 1.4, which is a typical minimum for ELM suppression. Vary ECH upward to increase the pedestal electron temperature at a constant density. Apply RMP and observe Te in ELM suppressed state vs. initial Te.
Background:	The leading theory of why RMP is effective at ELM mitigation is that the plasma edge is rendered stochastic and the resulting heat loss brings the pedestal stable to peeling-ballooning modes. However, traditional understanding of stochastic heat transport says that rapid parallel heat conduction takes place with electrons coming to thermal equilibrium with the cold outer region of the stochastic layer. Convection is reduced by the presence of an ambipolar electric field that develops in response to the slow ion thermal velocities. In RMP plasmas, we typically see the opposite, a small change in electron temperature but a large density drop. This experiment aims to decouple the density drop from the change in temperature. With known temperature and density profiles, and source input power, we can use steady-state power balance to determine transport coefficients with the CORSICA code. These can be compared with diffusion coefficients from field line tracing to determine if the edge is responding stochastically, or if the transport is anomalous.
Resource Requirements:	ECE, NBI (210 RT for MSE, 30 LT for MSE, 300 RT for CER, plus any additional needed to balance torque), Machine time: 1 day for ECE power scan and repeat shots.
Diagnostic Requirements:	MSE, CER, Thomson
Analysis Requirements:	CORSICA, TRIP3D
Other Requirements:	--

Title	99: Effect of magnetic stocasticity on bootstrap current
Name:	Hudson bfhudson@ucsd.edu	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Evans	ITPA Joint Experiment :	No
Description:	Compare an RMP ELM suppressed H-mode to a ELM-free H-mode to see if there is an impact on the bootstrap current.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We would compare two plasmas, an ELM-free H-mode plasma and an ELM-suppressed RMP plasma with identical Bt, Ip, beta, density, temperature and shape. We will try to match q, at least on axis and at the edge. A target ELM suppressed RMP discharge will be selected from the database to be reproduced. We will then try to achieve the same parameters in an ELM-free H-mode plasma. The only major difference between the two states should be the presence of the perturbation. A long time window in the ELM-free H-mode and during ELM suppression for averaging the MSE data will be helpful in getting a single measurement of a steady-state edge current. We specifically do not want a stationary plasma for this case as the ELM-free H-mode plasma has rotation.
Background:	The effect of stochastic magnetic fields on the bootstrap current is unknown, and of direct importance in a general assessment of RMP control in ITER. We can speculate that current would flow differently on open field lines than on good flux surfaces. This will also be a best-case attempt at being able to measure the bootstrap current with the present diagnostic set.
Resource Requirements:	RMP, NBI (30 LT + 210 RT for MSE, 330 RT for CER, others to match RMP rotation), ECE (if needed to adjust beta), machine time of 1 day or less depending on success in matching the discharges.
Diagnostic Requirements:	Thomson, CER, MSE
Analysis Requirements:	TRIP3D to analyze the bootstrap current vs. field line stochasticity
Other Requirements:	--

Title	100: Effect of plasma rotation on RMP
Name:	Hudson bfhudson@ucsd.edu	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Evans	ITPA Joint Experiment :	No
Description:	Vary plasma rotation to determine effect on RMP effectiveness.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin with stationary ECH and balanced beam plasma that exhibits ELM suppression with RMP. Bring in another co-beam to vary the rotation, while adjusting the balanced beam modulation to keep the fast ion density constant. This could identify a threshold (or rotation dependence) on ELM suppression.
Background:	In the plasma frame, the n=3 applied field appears to oscillate at the 3x the plasma rotation frequency. This acts like a high-frequency perturbation on a conducting medium and thus currents form that would cancel out the perturbation. This screening is widely expected, yet we see macroscopic effects on the plasma suggesting that the role of rotation on RMP is poorly understood. The rotation in ITER is expected to be low, but without knowing the underlying physics of why RMP works it is difficult to make predictive models to guide coil design.
Resource Requirements:	ECE, NBI (210 RT for MSE, 30 LT for MSE, 330 RT for CER, others to scan torque), Machine time: 1 day to do the torque scan, with several repeat shots to check the reproducibility of the observed ELM suppression.
Diagnostic Requirements:	Thomson, CER, MSE
Analysis Requirements:	--
Other Requirements:	--

Title	101: Dependence of ITER shape on effectiveness of RMP ELM suppression.
Name:	Hudson bfhudson@ucsd.edu	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Evans	ITPA Joint Experiment :	No
Description:	Perform a scan of plasma shape close to the ISS to study the effect of shape on ITER relevant ELM suppressed plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin with stationary ECH and balanced beam plasma that exhibits ELM suppression with RMP in the ISS. Vary the upper triangularity, upper squareness, and elongation about the ISS. Keep the lower triangulary and squareness fixed to preserve diverter pumping.
Background:	There will likely be refinements to the baseline ITER shape required in order to optimize fusion performance. Knowing the amount of flexibility in adjusting the shape of the ITER discharge while still suppressing ELM�??s will be a key operational constraint. To assess this we will scan the parameters that affect the upper shape and measure the degree of ELM suppression achieved.
Resource Requirements:	ECE, NBI (210 RT for MSE, 30 LT for MSE, 330 RT for CER, others to balance torque). Machine time: 1 day for three, single-parameter scans. Only one parameter will be altered at a time because of limits on number of shots.
Diagnostic Requirements:	Thomson, CER, MSE
Analysis Requirements:	--
Other Requirements:	--

Title	102: Fuel Ion Mass Scaling of Turbulence and Transport
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Transport Model Validation	Presentation time:	Requested
Co-Author(s):	G. McKee, C. Holland, T. Rhodes, L. Schmitz, A. White	ITPA Joint Experiment :	No
Description:	Measure the dependence of turbulence and transport characteristics and energy confinement on working ion mass. Compare scaling results with TGLF modeling and GYRO simulations.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish discharges similar to well-characterized discharges in a dimensionless scaling experiment (co-current L-modes). Perform steady state L-mode edge inner wall limited discharges very similar to previous rho* scan discharges, but in Hydrogen. The aim to run discharges in an operationally similar manner to a baseline in deuterium (101381/101391), but with dimensionless parameters (rho*, q, nu*, Mach #, Beta) matched, and the ion Mass, A, being the varying dimensionless quantity. There are several components to this experiment which may need to be prioritized: 1) L-mode ion mass scaling at near-balanced rotation 2) L-mode ion mass scaling with co-current rotation (matching previous experiments) 3) H-mode ion mass scaling at near-balanced rotation 4) H-mode ion mass scaling with co-current rotation Perform an ion mass scan at each condition, including a mix-species scan: 1) A=1 (Hydrogen) 2) A=2 (Deuterium) 3) A=1.5 (mixed species, H+D) 4) A=3 (Tritium) - Not! Measure turbulence in each condition as a function of position, wavenumber and field (n, T). All profiles will be measured for transport studies (may require beam swapping). Also, perform particle transport studies with Helium gas puffs. This experiment would necessitate operation in Hydrogen. Operationally, this could be performed in conjunction with the ITER Hydrogen & Helium plasma campaign. Incidentally, a Helium-Deuterium matched scan (same charge-to-mass ratio, factor of 2 difference in ion mass) would also be very fruitful and make for a compelling scan. This experiment would ideally be performed in conjunction with a deuterium rho* scan which might be performed elsewhere in the Transport Model Validation Task Force.
Background:	Previous experiments on various tokamaks have demonstrated increased confinement with higher atomic-mass fuel ions (e.g., DT: Scott, PoP, 1995). This issue has obvious implications for ITER and other future experiments. A preliminary hydrogen experiment on DIII-D in 2008 demonstrated a similar effect: normalized energy confinement was significantly lower in hydrogen plasmas compared with previous (c.a., 2000) dimensionlessly matched deuterium discharges (i..e., rho*, q, nu*, Mach #, Beta were matched). Preliminary turbulence measurements showed a consistent effect of higher normalized fluctuation magnitudes. Due to time constraints, this preliminary experiment was not completed, but offered intriguing initial observations. Several parameters were not well matched with previous the rho* experiment due to operational limitations. Because of the close connection between ion mass scaling and rho* scaling effects, this experiment would ideally be performed in conjunction with rho* scaling experiment. Basic arguments from gyrokinetic scaling considerations would suggest that the transport scaling with isotope mass should exhibit the opposite scaling to this experimentally observed trend, due primarily to gyroradius scaling with mass. GYRO simulations of transport scaling (Estrada-Mila, PoP, 2005) are consistent with this expected trend, but observe more complex behavior, with a stronger particle pinch for the higher mass isotope in mixed-species plasmas. Laboratory experiments have showed intriguing changes in the nonlinear dynamics (wave-wave coupling) of turbulence (dominant driven modes and energy cascades) as a function of ion mass (Sen, Sokolov). This may allow for interesting basic turbulence physics studies as well, since 2D time-resolved turbulence measurements will be obtained with BES. Resolving the ion mass scaling of particle and energy transport with both single ion and mixed species ions is of significant importance to future D-T experiments. Results, in terms of transport and turbulence, will be compared with TGLF and GYRO simulations.
Resource Requirements:	Hydrogen plasmas (performed in conjunction with the ITER Hydrogen & Helium campaign)
Diagnostic Requirements:	Fluctuation Diagnostics: BES, FIR (low & medium k), high-k backscattering, Doppler Backscattering, PCI, Langmuir probes Profile diagnostics: CER, MSE, Thomson scattering,...
Analysis Requirements:	The profile and fluctuation data from this experiment will be primarily geared towards testing the TGLF model and GYRO simulation.
Other Requirements:

Title	103: Pure ECH divertor power widths
Name:	Pitts richard.pitts@iter.org	Affiliation:	ITER Organization
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	C. Lasnier, J. Watkins	ITPA Joint Experiment :	No
Description:	Provide some of the first ever divertor target heat flux profiles in pure ECR heated L-modes and ELMing H-modes. Assist in building divertor and edge physics basis for ECR heated ITER plasmas	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Establish lower divertor IRTV optimised discharges in deuterium L-mode and ELMing H-mode with pure ECR heating. Investigate to the extent possible with the ECH system the profiles of power and particle deposition for a parameter range which would be a subset of previous similar measurements performed on DIII-D in NBI heated discharges. One or two key reference NBI discharges to be repeated to compare with previous larger database and to provide fiducial points for the ECH plasmas. L-mode plasmas are of equal interest given the lack of available tokamak divertor profiles. Here there may be more scope for parameter variation (e.g. density, plasma current) with ECH than in H-mode.
Background:	ECRH is one of the major heating systems planned for ITER and is likely to be the first ready for deployment. And yet the tokamak database of SOL and divertor measurements in the presence of pure ECR heating is extremely limited. Density pump-out effects on the SOL and divertor have yet to be properly characterised, even in L-mode. There is little or no comparison of the divertor heat flux widths during ECH with equivalent beam heated discharges.
Resource Requirements:	To begin with probably one half day of machine time provided target plasmas can be easily established. Both beams and ECH power required on the day. Max. ECH power required.
Diagnostic Requirements:	Target Langmuir probes can be used, but the key diagnostic is IRTV, viewing the lower divertor. There are some issues associated with potential damage to the camera due to ECH beams being refracted into the diagnostic port. Adequate counter-measures need to be installed before the experiments would be possible. Good pedestal profiles required for the H-mode discharges.
Analysis Requirements:
Other Requirements:

Title	104: Physics of NRMF torque
Name:	Garofalo garofalo@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Callen, Cole, Joseph, Park, Reimerdes, Solomon	ITPA Joint Experiment :	Yes
Description:	Test neoclassical paradigm of nonresonant magnetic field (NRMF) enhanced collisional transport as the basic process originating the torque exerted on a toroidal plasma by a NRMF. Neoclassical theory predicts that enhanced collisional transport from NRMFs leads to ion particle radial fluxes and return radial currents in the plasma. Cross product of these currents with the poloidal field produces the NRMF toroidal torque. This experiment aims at investigating the particle transport during application of large NRMFs under various plasma conditions. Particularly useful for identifying the NRMF-enhanced particle transport should be the sign change in the particle flux (from outward to inward) that is expected to accompany the sign change of the toroidal torque (from counter-Ip to co-Ip) when the plasma rotation goes from larger (i.e. more positive) to smaller (i.e. more negative) than the neoclassical "offset" rotation. Here, the sign of the plasma rotation is intended positive (negative) in the co-Ip (counter-Ip) direction.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The proposal is to carry out a detailed experimental study of two types of discharges from the 2007 campaign. These are a normal-Ip and a reversed-Ip discharges which exhibited a different behavior of the plasma density in response to a slowly increasing amplitude of an n=3, odd-parity I-coil field. For each type of discharge, we plan to carry out systematic scans of betan and of the I-coil amplitude for fixed initial toroidal rotation. These scans will allow to assess the correlation (if extant) between the density pump-out/pump-in and the n=3 external field or the n=3 plasma response. A rotation scan at fixed betan is also desirable. Profile measurements will provide input for TRANSP simulations of the discharge evolution (TRANSP), necessary to characterize the particle fluxes.
Background:	Recent DIII-D experiments have demonstrated the existence of the predicted neoclassical "offset" rotation, associated with the toroidal torque driven by a nonresonant magnetic field (NRMF) applied to a tokamak [Garofalo et al., PRL (2008)]. The existence of this counter-Ip directed offset rotation is clearly shown by a change in the sign of the measured NRMF torque as the initial (i.e. before NRMF application) plasma toroidal rotation is scanned (shot-to-shot) from a slow counter-Ip rate to a fast counter-Ip rate. During this scan, the NRMF torque is observed to change from an accelerating torque (for slow initial rotation) to a braking torque (for fast initial rotation). The inferred offset rotation profile is consistent with the neoclassical predictions for a plasma with ion collisionality between the asymptotic nu and 1/nu regimes. However, according to neoclassical theory, the sign change in the NRMF torque also implies a change in the enhanced particle transport from an outward transport (for the counter-Ip directed torque) to an inward transport (for the co-Ip directed torque). Hints of this sign change for the particle transport may be contained in the observations that the plasma density is usually reduced by the application of a NRMF which drives a torque in the counter-Ip direction, while the density is often increased by the application of a NRMF which drives a torque in the co-Ip direction. These observations were first made during the 2007 experiments on "Resonant vs. nonresonant braking" (D3DMP #2007-04-02 and #2007-04-04 by Garofalo et al.).
Resource Requirements:	Same resources as used for 127744 and 127908 (FY07 experimental campaign), except use a 7 kA capable I-coil hook-up. 127744 is a normal-Ip discharge. 127908 is a reversed-Ip discharge.
Diagnostic Requirements:	All standard magnetics and internal profile diagnostics, including reflectometer for density profile measurements. FIR and BES fluctuation measurements should also be acquired.
Analysis Requirements:	Analysis of the magnetics for extraction of the n=3 plasma response. Kinetic equilibrium reconstruction for accurate n=3 stability modeling. Time-dependent profile fitting for TRANSP modeling of the discharge evolution.
Other Requirements:	--

Title	105: Enhanced erosion from deuterium saturated materials during ELMs
Name:	Umstadter karl@ucsd.edu	Affiliation:	UCSD
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Requested
Co-Author(s):	Clement Wong, Dmitry Rudakov, Phil West, George Tynan, Russ Doerner	ITPA Joint Experiment :	Yes
Description:	All prior heat pulse testing of PFCs have been completed in vacuum environments without the presence of background plasma. ELMs will not be this kind of isolated event and one should know the effect of a plasma background during these transients. The retention of gas in PFCs may lead to enhanced erosion of material that will behave differently if ionized near the surface. Current models may underestimate the damage caused by ELMs due to these phenomena.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	DiMES samples of graphite and tungsten will be saturated by bombardment with deuterium ions (Eion~125eV) in the PISCES-A device. Saturated and unsaturated samples will be loaded into the DiMES system. During ELMing H2 experiments, with the strike point will be placed on the DiMES system. Disruptions, if they should occur will not adversely affect the experiment. The DiMES holder should be lowered between shots so that it is not exposed to He glow discharges.
Background:	Heat-pulse experiments have begun in the PISCES-A device utilizing laser heating in a divertor-like plasma background. Initial results indicate that the erosion of PFCs is enhanced as compared to heat pulse or plasma only tests. This enhanced erosion may be caused by trapped gases released during the heat pulse. Also self-sputtering of material that is ejected during the transient, ionized by the plasma near the surface and subsequently driven back to the surface may occur. Gas retention in PFCs and ELM energy are currently indicated as the cause. Current machines around the world don't see the damage witnessed in the lab because the sample (divertor) has a lower fluence before seeing an ELM, disruption, or glow cleaning. This will not be the case for ITER as one can expect >1E25 D/m2 between ELMs (2 Hz).
Resource Requirements:	The system should be operating for as many shots as possible during H2 experiments, ideally when the strikepoint can be moved onto the DiMES System. DiMES operator raise/lower before/after shot to avoid He glow.
Diagnostic Requirements:	-High speed video of the DIMES System. -DiMES Video -IR Cameras looking at DiMES -MDS â?? looking at Hg and W i & ii Center at 4320Ã? for W i 4294.6Ã?, W i 4302.1Ã?, H ï?§ 4340.5Ã?, W ii 4348.1Ã? -Fast Filterscopes -Divertor Thomson measurements of ne and Te in the divertor
Analysis Requirements:	NRA to be completed by SNL, SEM imaging, TDS and mass loss analysis will be completed at UCSD PISCES lab.
Other Requirements:

Title	106: ECCD within magnetic islands
Name:	Prater prater@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	C. Petty, F. Volpe, R. La Haye, T. Strait	ITPA Joint Experiment :	No
Description:	This experiment is designed to measure the Electron Cyclotron Current Drive within a slowly rotating magnetic island. The experiment will show whether the ECCD within an island is correctly evaluated by a code like TORAY, which assumes poloidal uniformity. The result could reduce the calculated ECCD power requirement for NTM stabilization in ITER.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Entrainment of a saturated 2/1 NTM in a rotating externally-applied magnetic field can produce a slowly rotating mode, say at 50 Hz, which allows the ECCD to be modulated synchronously with the O-point. If the magnitude of the ECCD is significant but too small to stabilize the island, the current driven in the island will be measureable through the modulation of the MSE pitch angles at 50 Hz. This will support determination of the driven current density.
Background:	The control of neoclassical tearing modes by ECCD within the islands has been shown to be highly robust, maybe more so than expected from the Rutherford equation. One reason may be that the poloidal bunching of the ECCD within the islands provides an un-included improvement in the ECCD efficiency compared to the poloidally uniform case. Alternatively, if the current-carrying electrons diffuse out of the islands before they isotropize through collisions, the ECCD will be diminished. These effects should be understood to project correctly to the requirements for ITER.
Resource Requirements:	At least 4 gyrotrons. The I-coil is an essential element.
Diagnostic Requirements:	MSE is crucial.
Analysis Requirements:
Other Requirements:

Title	107: Introduction of pre-characterized dust in divertor and SOL
Name:	Rudakov rudakov@fusion.gat.com	Affiliation:	UCSD
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Requested
Co-Author(s):	P. West, J. Yu, C. Wong, M. Groth, B. Bray, N. Brooks, M. Fenstermacher, S. Krasheninnikov, C. Lasnier, A. Litnovsky, A. Pigarov, R. Pitts, C. Skinner, R. Smirnov, W. Solomon	ITPA Joint Experiment :	Yes
Description:	Introduction of pre-characterized dust of varying size and chemical composition (carbon, boron, possibly aluminum and/or tungsten) in the lower divertor of DIII-D using DiMES and in the outboard SOL using reciprocating probe/MiMES. The objectives are: 1) characterization of core penetration efficiency and impact of dust of varying size and chemical composition on the core plasma performance in different conditions and geometries; 2) benchmarking of DustT modeling of dust transport and dynamics; 3) comparing results with other machines â?? TEXTOR, MAST, NSTX, LHD	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Dust will be introduced into the lower divertor using DiMES. Injection will be performed in LSN configuration by sweeping strike points over DiMES holder loaded with dust. Introduction of dust in the outboard SOL will be performed using mid-plane reciprocating probe by applying dust to the probe and plunging in into the SOL.
Background:	Dust penetration of the core plasma in ITER can cause unacceptably high impurity concentration and degrade performance. Therefore, knowledge of the dust transport and dynamics is important. Studies on the contemporary machines are needed to benchmark modeling for extrapolations to ITER. Introduction of pre-characterized dust from a known location offers a way to benchmark modeling of dust dynamics and transport. Dust can be either actively injected or launched off a surface by plasma contact. Initial studies of the dust launch by plasma contact have been performed in DIII-D and TEXTOR. Contrary to expectations, core penetration efficiency of the dust in limiter configuration (TEXTOR) proved much lower than in the divertor configuration (DIII-D). Joint experiments with participation from other machines are proposed to improve understanding and provide the basis for modeling.
Resource Requirements:	2-3 shot experiments: 1 setup shot, 1 reference shot (could be same shot as setup) + 1 injection shot. LSN patch panel, strike point sweeps for injection from DiMES.
Diagnostic Requirements:	DiMES, DiMES TV, lower divertor tangential TVs, UCSD fast camera, CER, Thomson (divertor and core), filterscopes, MDS, lower divertor Langmuir probes, SPRED. For SOL injection: mid-plane reciprocating probe and fast camera viewing probe/MiMES from 135T0 port.
Analysis Requirements:
Other Requirements:

Title	108: RMP Particle Sources and Sinks with Helium Discharges
Name:	Unterberg unterbergea@ornl.gov	Affiliation:	ORNL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Oliver Schmitz	ITPA Joint Experiment :	Yes
Description:	Recent analysis has shown that wall pumping (deuterium retention in graphite walls) could play a major role in RMP particle transport (i.e. providing a significant sink for particles expelled during the RMP). Helium (HE), as the main species of the plasma, on the other hand has a very low retention time in graphite and therefore has the possibility of testing the significance of wall pumping on the transport properties during the RMP. This experiment will aim to determine if there are significant changed in the sources and sinks during RMP operation due to HE as a main species. With the significant change to edge transport conditions (e.g. ionizations lengths, increased mass, etc) with HE as a main species manifest changes in aspects of particle transport physics could be determined. This would also help verify ELM suppression with RMPs as a viable option in ITER's first plasma campaign if it is in HE.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Ideally for direct comparisons with D2 experiments, ELM suppressed RMP plasmas like those of shot 134162 would be desired. But, the general development of a "standard" RMP recipe might be necessary in the new HE regime. By "standard recipe", it is meant that the RMP parameters that give reliable ELM suppression in HE regardless of other details (e.g. Icoil current, q95, density pumpout magnitude, βN values, shape, etc.).
Background:	The use of HE as the main species in stochastic boundary experiments has been tried before with limited success (see Ghendrih et al, PPCF 1996). TEXTOR will be conducting HE experiments with its DED system early in 2009. Therefore we aim on a cross comparison of HE stochastic boundary experiments in L-mode (from TEXTOR) and H-mode (from DIII-D) within ITPA task PEP19. Here the most desirable data is the efficiency of external pumping when a significant sink (the graphite walls) is not available and what effect this has on the resulting particle transport. Also valuable experience working with RMP in HE discharges will be gained from the TEXTOR work and here we want to compare in particular the change of transport in relation to the perturbed boundary as identified so far (i.e., the perturbed separatrix and laminar edge layer). This experiment could be conducted in conjunction with Gohil's proposal #67.
Resource Requirements:	Cryo-system "frosted" in Argon for HE plasmas. 5 co-NBI capability Full Icoil capability (i.e. up to 7kA)
Diagnostic Requirements:	Usual RMP edge diagnostics with relevant diagnostics calibrated for HE operations.
Analysis Requirements:	Single reservoir particle balances, profile analysis, perfect for EMC3-EIRENE/SOLPS analysis.
Other Requirements:	Probably requires a full run day.

Title	109: Improving long-distance FW coupling to H-mode plasmas with gas puffing
Name:	Pinsker pinsker@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Requested
Co-Author(s):	M.-L. Mayoral, J.C. Hosea, V. Bobkov, T. Petrie, M. Goniche, S. Wukitch, S. Moriyama	ITPA Joint Experiment :	Yes
Description:	It has been widely recognized that the required outer gap in ITER is probably not consistent with coupling adequate FW power levels to ITER, without doing something to increase the coupling. Several experiments have investigated gas puffing during the ICRF pulse in an attempt to increase the far SOL density and hence increase the FW coupling. Results obtained in various machines (JET, AUG, TS, JT-60U, etc.) are not entirely consistent. In some cases on JET, substantial improvement in the coupling has been obtained at plasma/antenna gaps of 0.15 m, while results in AUG have not been positive so far. It is thought that details of the geometry around the antennas - in JET the antennas are in ports, while on AUG they hang on the wall - are important. On DIII-D the antennas have a similar geometry to JET, so we might expect similarly positive results. In a piggyback experiment at the end of the FY08 run, we found that the effect of puff-and-pump (puffing from GASD) on the 285-300 antenna's loading was extremely sensitive to the value of DRSEP, and that the antenna loading correlated strongly with the measured VPLOWS (neutral pressure). In the autumn 2008 vent, we added two gas puff valves near the 285-300 antenna, one quite close to the antenna, the other farther away, but both being magnetically connected with the antenna for the usual helicity for DIII-D. We need to evaluate the effect of puffing on antenna loading as a function of the relative location of puffing and the antenna, the degree of magnetic connectivity between the puff location and the antenna, on details of the field line topology in the far SOL, on the puff rate, and possibly on the FW power level. Equally important, it is necessary to investigate possible deleterious effects of the puffing on confinement, on FW edge losses, and on power handling of the antenna. This is an important ITER task, and is being carried out in the framework of an ITPA joint experiment (IOS 5.2, 'Maintaining ICRH Coupling in Expected ITER Regime').	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Main idea is to establish a reproducible ELMing H-mode with a medium-to-large outer gap and investigate the effect of puffing from the various locations, both magnetically connected to the powered antenna or not, on antenna loading, on power handling (voltage standoff), on edge losses (by measuring the power coupled to the core and subtracting that value from the power leaving the antenna), and generally on the confinement. It is necessary to determine the importance of relative location of the puffing and the antenna, on the level of puffing, and on details of the field line geometry in the SOL (various DRSEP values, for example).
Background:	To quote from IOS 5.2 Joint Proposal: For effective and reliable operation of the ICRF system in ITER, good coupling conditions should be maintained. Gas fueling at the launcher or from a location magnetically connected to the launcher could raise the density in the SOL and increase the coupling resistance by moving the cut-off layer closer to the antenna. This technique has potentially deleterious effects on: wave propagation/absorption, confinement (pedestal height), antenna voltage stand-off (neutral pressure). Most of the devices are now equipped with gas injection systems possibly connected to the ICRH antenna. JET has shown substantial improvement of the coupling in ELMy H-mode with an antenna-separatrix distance as large as 0.15 m.
Resource Requirements:	Gas puffing from new injectors near the 285-300 antenna as well as from the usual locations (for comparison). At least the 285/300 FW system, but preferably the other two as well.
Diagnostic Requirements:	This experiment requires that the edge reflectometer adjacent to the 285-300 antenna be instrumented and working, in addition to all of the usual diagnostics. All available Langmuir probes should also be employed.
Analysis Requirements:
Other Requirements:

Title	110: Dependence of high-k Turbulence on ExB Shear
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	It is generally predicted that ExB shear should have little effect on ETG-scale turbulence, that is, turbulence at high wavenumbers. For this reason is it predicted that only high-k modes should remain in plasmas with strong ExB shear, such as co-injection hybrid discharges. The strong diagnostic set on DIII-D for measuring intermediate-k and high-k turbulence is well suited to experimentally testing this proposition. The high power ECH system can help destabilize ETG-scale turbulence, and the co/counter NBI system can systematically vary the ExB shear in the plasma. Therefore, DIII-D is an ideal device to determine the dependence (or lack thereof) of short wavelength turbulence on the ExB shear.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A decision first has to be made regarding the best target plasma. Such a plasma should have ample high-k turbulence. This could either be (1) a low density, balanced-NBI L-mode plasma with strong ECH, or(2) a co-NBI, rapidly rotating hybrid plasma. Next the NBI should be altered to vary the ExB shear as much as possible, either increasing the toroidal rotation in case (1) or reducing the rotating in case (1). To the extent possible, the plasma beta and Ti/Te ratio should be kept fixed during this ExB shear scan. The critical diagnostics are the sensitive sensitive to intermediate-k and high-k fluctuations, such as the microwave backscattering system and the Doppler backscattering system. The low-k diagnostics (BES, CECE) should also attempt to acquire data for completeness.
Background:
Resource Requirements:	NBI: Both co and counter sources are required. ECH: All 6 gyrotrons are desired.
Diagnostic Requirements:	Full complement of turbulence diagnostics.
Analysis Requirements:	GYRO modeling will be needed with synthetic diagnostics.
Other Requirements:

Title	111: Develop hybrid QH-mode
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Create a plasma which has the combined characteristics of a hybrid and a co-NBI QH-mode in order to simultaneously exploit the beneficial effects of small core tearing modes and the EHO	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Combine the hybrid recipe using beta feedback to induce 4/3 or 3/2 tearing modes early in the discharge with the co-NBI QH-mode recipe of low density operation with beta feedback.
Background:	Hybrid discharges with q95 > 4 simultaneously exhibit excellent energy confinement time a complete lack of sawteeth. The improved energy confinement time is associated with E x B shear stabilization for cases with rapid rotation. The mechanism for suppressing the sawteeth appears to be to effect of the 4/3 or 3/2 tearing mode which pumps magnetic flux out of the plasma core. Recent work by C. Petty has shown that the tearing mode interaction with the ELMs plays a role in this. QH-mode discharges also benefit from the presence of a continuous MHD oscillation, the EHO, which increases the edge particle transport and allows the discharge edge to reach a transport steady state under ELM-free conditions. Now that co-NBI QH-mode is a reality, it appears possible to simultaneously meet the conditions needed for hybrid operation and QH-mode operation. A key question is whether the EHO interaction with the core tearing modes could still allow operation with q(0) > 1. If a hybrid QH-mode operation is possible, it would be very elegant demonstration of plasma control, since it would utilize continuously oscillating MHD modes in the core and edge to eliminate MHD instabilities which grow explosively and then crash.
Resource Requirements:	6 NBI sources--all co-beams plus 210LT
Diagnostic Requirements:	Profile diagnostics, edge fluctuation diagnostics for EHO studies
Analysis Requirements:	--
Other Requirements:	--

Title	112: How do Zonal Flow-Induced Shear layers affect electron transport?
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Requested
Co-Author(s):	J.C. Hillesheim, T.L. Rhodes, A.E. White, W.A. Peebles	ITPA Joint Experiment :	No
Description:	Determine how quasi-stationary Zonal Flows affect electron transport in the limit of closely spaced and well separated low-order rational surfaces	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use early neutral beam heating to set up an L-mode plasma with reversed central magnetic shear. Obtain radial scans of the ExB flow velocity and the intermediate-scale fluctuation level by Doppler backscattering. The radial spacing of low-order rational surfaces changes with magnetic shear as a function of time. Radial DBS scans are taken at different times to obtain poloidal flows and fluctuation data for closely spaced and well separated Zonal Flow layers. Time-dependent transport analysis will be performed to determine the effect of Zonal Flows on electron (and ion) transport.
Background:	Zonal flow-induced shear layers at rational q-surfaces have been predicted by gyrokinetic modeling (R. Waltz et al., PoP), and electron temperature profile corrugations and transient electron transport barriers have been previously observed in D-III-D as the q=2 surface enters the plasma (M. Austin et al., PoP). Recently we have observed a quasi-stationary Zonal Flow-induced shear layer in a sustained electron transport barrier at the q=2 surface (using Doppler Backscattering). Substantially reduced intermediate-scale density fluctuations (k rho_s ~ 3.5) were observed in the shear layer, which coincides with the electron transport barrier.
Resource Requirements:	Beams
Diagnostic Requirements:	DBS, BES, FIR-scattering, PCI, MSE
Analysis Requirements:	GYRO simulation of L-mode reversed shear plasmas for the two limiting cases considered here (closely spaced and well separated rational q-surfaces).
Other Requirements:	--

Title	113: Optimized Plasma Shape for RMP ELM Suppression
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	B. Hudson	ITPA Joint Experiment :	No
Description:	Use the lessons learned from recent QH-mode studies to optimize the plasma shape for RMP ELM suppression. The most favorable shape for QH-mode was found to be high triangularity and balanced DND shape. The experiment should test whether such a plasma shape allows ELM suppression in standard H-mode plasmas with less non-axisymmetric magnetic fields than the plasma shape used previously in RMP studies.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Reproduce the optimal plasma shape from recent QH-mode experiment, but with co-NBI. (2) Adjust BT so that q_95 = 3.6 satisfy the RMP resonance condition. (3) For beta_N=2, increase the I-coil current to determine the amount needed to suppress ELMS. (4) Attempt to increase beta_N. Is a larger I-coil current needed to maintain the ELM suppression?
Background:	--
Resource Requirements:	RMP I-coil configuration required.
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	114: Imaging the RMP Stochastic Boundary
Name:	Unterberg unterbergea@ornl.gov	Affiliation:	ORNL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Oliver Schmitz, M. Fenstermacher, M. Groth	ITPA Joint Experiment :	Yes
Description:	Direct imaging of the stochastic boundary using visible and/or SXR frequencies was identified as a high priority set of experimental data for RMP physics understanding by the ITER IO and the RMP research community. This data would mostly be used to find evidence for perturbed magnetic structures induced by the n=3 I-coil field and by that validating field screening/amplification theories that have been proposed to explain the ELM suppression with the DIII-D RMP fields. This proposal will use helium gas puff imaging first with existing cameras (LLNL's tangential, lower X-point CID TV system with and w/o image intensifier) on DIII-D and then with a higher spatial resolution CCD camera from TEXTOR. The main purpose will be to measure the lobe structure of the stochastic x-point region and compare with modeled/predicted structures. Helium will also be used to obtain the (R,Z) electron density and temperature fields around the perturbed X-point region.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Development of the "standard" RMP recipe would be the only desire. By "standard recipe", it is meant that the RMP parameters that give reliable ELM suppression regardless of other details (e.g. Icoil current, q95, density pumpout magnitude, βN values, shape, etc.). It would be beneficial to change RMP phase of the perturbation between shots and time oscillate the perturbation during the shot. This will allow us to assess relative changes in the signal and even do frame-by-frame subtractions if necessary to improve the image contrast. The poloidal and toroidal location for an optimal helium gas puff would need to be determined by 3D modeling (EMC3/EIRENE coupled to a specialized collisional radiative model is ongoing) and field line tracing by TRIP3D.
Background:	Imaging of the edge islands using carbon impurity filters on CCD cameras has successfully imaged the islands produced by the RMP in L-mode TEXTOR discharges. A similar technique using helium gas puffs to image the islands is proposed here. Initially, three specific HE-I filters (667.8nm, 706.5 nm and 728.4 nm) will be placed in the existing 240deg lower x-point viewing CID camera in three consecutively repeated discharges. This will give a full view of the RMP perturbed x-point region, which has been modeled to have a wide spatial extent of stochasticity (as modeled by the vacuum code TRIP3d) and will allow us to deduce the (R,Z) electron density and temperature fields for the first time. This shall allow the resolution of structures and transport effects in the stochastic boundary consisting of laminar and stochastic field lines. However, resolution of the predicted magnetic islands deeper inside will need higher T_e emitting lines such as SXR frequencies planned for future application. The data obtained with this experiment will provide information for optimizing this approach. The lower view where we will attempt for this measurements is more preferential than a midplane view because of the poloidal flux expansion ( ~ 10X) and the subsequent spatial widening of the structure (e.g. island chains and the separation of the separatrix manifolds). But the 240deg camera may be limited by: 1) the spatial resolution, 2) contrast and/or 3) limited viewing of the stochastic region due to alignment. Therefore for 2010 a new high spatial resolution, MCP intensified camera is being proposed to resolve most of the issues mentioned above. This experiment will directly transfer techniques applied at TEXTOR-DED to DIII-D and is therefore part of the ITPA PEP19 joint experiments. It tackles field penetration and structural formation topics and is therefore addressing urgent ITER IO requests.
Resource Requirements:	Typical RMP neutral beam capability Full Icoil capability (i.e. up to 7kA) For 2010 campaign: port access for new CCD camera
Diagnostic Requirements:	Usual RMP edge diagnostics. LSN tangential CID cameras with He-I filter sets, DiMES TV.
Analysis Requirements:	Field line tracing using TRIP3d for magnetic topology for comparison with measurements, EMC3/EIRENE modeling including He puff location as source element (in preparation).
Other Requirements:	--

Title	115: RMP ELM Suppression at the NTV Offset Rotation
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	B. Hudson	ITPA Joint Experiment :	No
Description:	Establish RMP ELM suppression in a plasma with counter rotation. Allow the rotation to "lock" to the offset rotation given by NTV. 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. 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.
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	116: Quanitification of the requirements for ELM suppression by RMP from off mid-plane coils
Name:	Kirk andrew.kirk@ukaea.uk	Affiliation:	CCFE
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Evans, M. Fenstermacher (GA), Andrew Kirk (MAST), Yunfeng Liang (JET), Alberto Loarte (ITER) Rajesh Maingi (NSTX), Wolfgang Suttrop (ASDEX Upgrade)	ITPA Joint Experiment :	Yes
Description:	The goal is to verify the suppression requirement derived from DIII-D by achieving ELM suppression on other devices that are equipped with off-midplane coils.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Perform a similarity experiment with MAST. Start off by trying to get the best match to a shot on DIII-D in which ELM suppression has been established (i.e. shot 126006 at 3600 ms q95=3.6). Attempt to achieve ELM suppression in the MAST for this shot, if possible perform at high and low collisionality. Perform shots on DIII-D in which the pedestal quantities are matched to the MAST discharge (this has been successfully achieved previously as part of PEP-9, however, in order to maintain good discharges the magnitude of the toroidal field should be > 1T) and investigate the application of the I-coils. Perform a scan in DIII-D starting from the parameters of shot 126006 towards the MAST similarity shot to see the effect on ELM suppression, this will possibly involve a scan in beta_N or beta_pol. A magnetic configuration scan should be performed to investigate ELM suppression closer to connected double null in order to determine which parameter in separatrix separation, if any, is the important quantity. By comparing these discharges more information on the requirements for ELM suppression would be obtained which would increase the confidence that extrapolations could be made to ITER.
Background:	The goal is to verify the suppression requirement derived from DIII-D by achieving ELM suppression on other devices that are equipped with off-midplane coils. At present the only other machine equipped with off-midplane coils is MAST, which will be followed by ASDEX Upgrade in 2010. Using the expertise obtained on other devices the aim is to start by obtaining complete ELM suppression in a Lower SND type I ELM-ing discharge on MAST with internal off-midplane coils. If suppression is successful a similarity discharge would be performed on DIII-D.
Resource Requirements:	I-coil maximum current. May need BT = -0.6 T shots.
Diagnostic Requirements:	All pedestal and lower divertor diagnostics
Analysis Requirements:	All pedestal and lower divertor diagnostics
Other Requirements:	This is a 1 day experiment

Title	117: Study transient neutral particle burst transport in RMP perturbed boundary
Name:	Schmitz oschmitz@wisc.edu	Affiliation:	U of Wisconsin
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	L. Baylor (ORNL), T. Jernigan (GA), N. Comeaux (ORNL), T. Evans (GA), E. Unterberg (ORNL)	ITPA Joint Experiment :	No
Description:	In this experiment both, HFS and LFS pellet injection will be used to study the pellet ablation while it penetrates and the transport of the ablated, ionizing particles in an RMP perturbed boundary. Observation of the ablation signal and of the edge n_e and T_e profile manipulation in comparison to not suppressed cases shall allow to resolve the structure of the perturbed magnetic boundary and changes in the effective particle confinement time. This experiment will accompany the heat pulse transport suggested in ROF prop. ID6 but will focus also on the functional meaning of the disturbance caused by fueling pellets in an ELM suppressed H-mode pedestal. With this proposal we want to use pellet ablation as diagnostic foe these islands, locate the dominant pellet ablation spot (radially) on the pedestal and monitor the impact on the pedestal profile. In combination with ID6 we want to balance the T_e caused by the pellet with a simultaneous ECRH injection into the perturbed edge layer.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Inject D pellets from HFS and LFS with different injection speeds into RMP ELM suppressed H-modes. Monitor the ablation signal and try to resolve the location of (a) rational surfaces and (b) islands. In case we see rational surfaces in a no-RMP discharge they should vanish with RMP applied in the edge. Therefore we want to compare discharges with no RMP applied, with ELM suppression with marginal ELM suppression (detune q_95 a bit). We plan to change the toroidal phase of the I-coil field as this will move possible islands/structure poloidally at the pellet injection port leading to different ablation signals. In the last third of the experiment we want to balance the energy loss caused by the pellet due to radiation and ionization by a simultaneous ECRH pulse.
Background:	The application of pellets for edge fueling and pellet pacing is discussed for application in ITER. The implications of pellets injected for such purposes in RMP ELM suppressed H-modes is not clear and previous measurements have shown that ELM like activity potentially can be induced. In this experiment we employ a proposal/technique described in DIII-D physics memo D3DPM No. 9702, February 4, 1997 by L.R. Baylor and T. C. Jernigan. Here it was demonstrated that observation of the ablation signal of a penetrating pellet obeys local heat sinks on resonant surfaces caused by the ionized particles. This heat loss can not be balanced due to finite field line connection length and therefore drops in the ablation signals are seen on these resonance surfaces and in particular in magnetic islands, in this proposal induced as natural tearing modes. For an RMP perturbed boundary magnetic islands caused by the RMP field are predicted and shall be visible too in the ablation signal.
Resource Requirements:	pellet gun from LFS and HFS prepared, D pellets as fast as possible (> 400 m/s), acquisition of ablation signal, ECH with all 5 gyrotrons ready, standard ISS plasmas used for ELM suppression experiments (e.g. #132741), SPA supplies for I-coil, C-coil in standard n=1 EFC setting (same as #132741), 1.0ITER04 f-coil patch panel
Diagnostic Requirements:	pellet ablation signal, visible and IR divertor cameras, target Langmuir probes, Thomson Scattering (core, tangential, divertor desirable) in fast mode at each pellet injection time, MSE and fast Li beam desirable for kinetic EFIT, fast Li beam for density profile, reflectometers
Analysis Requirements:	TRIP3D runs for island location, kinetic EFIT, profiles.py runs
Other Requirements:	modeling of ECH deposition before experiment

Title	118: Stabilization of ICRF-induced giant sawteeth by suppressing core-localized TAE activity.
Name:	Kramer none	Affiliation:	PPPL
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	G.J. Kramer (PPPL), M.F.F.Nave (IST), A. Turnbull (GA)	ITPA Joint Experiment :	No
Description:	Goals: - Measure the fast-ion pressure evolution inside the q=1 surface during the Giant Sawtooth (GST) cycle. - Investigate the role of core-localized TAEs on GST stability. - Stabilize the ICRH-generated GST completely by stopping the current diffusion to the core with ECCD. And in a related proposal by M.F.F. Nave: - investigate the ICRH induced rotation during the GST cycle and its correlation with fast ions.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	As a target plasma we propose to use a sawtoothing plasma with balanced beam injection so that the plasma rotation as close to zero as possible (as required by related intrinsic rotation proposal). Then we will apply various levels of ICRF to document the sawtooth period and C-TAE activity as function of ICRF power. ICRF might be applied as 4th harmonic majority heating and/or 4th and 6th harmonic majority heating, depending on which heating scheme creates the best stabilization. When the GST regime is established with core-localized TAEs we want to apply ECCD at an off-axis location to stop the current diffusion to the core and keep q on axis to a value between 0.9 and one to avoid the onset of low-n TAEs.
Background:	On several large tokamaks (JET, JT-60, TFTR, DIIID) it was found that sawteeth can be stabilized for one second or more when ICRH is applied in the plasma center. The sawtooth free period is usually ended by a large (or monster) sawtooth crash. The GST is stabilized by the fast-ion pressure inside the q=1 surface. Up to 0.5 s before this GST core-localized Alfven eigenmodes are observed, starting with high toroidal mode numbers (n=10 and higher) and decreasing to low toroidal mode numbers i(n=5 and lower) just before the crash. The sequence of TAEs can be explained by a slow decrease of the magnetic safety factor in the core of the plasma whereby the resonant condition for successive C-TAEs is met. While high-n TAEs are highly localized inside the q=1 surface and are thought not to contribute to the transport of fast ions from inside to the outside of the q=1 surface, low-n TAEs have a much broader radial extend to well outside the q=1 surface. These low-n TAEs can redistribute the fast ions, which are responsible for the sawtooth stabilization, from inside to outside the q-1 surface. When the fast-ion pressure then drops below the value to stabilize the sawtooth a GST is triggered.
Resource Requirements:	A sawtoothing plasma with balanced beam injection so that the plasma rotation as close to zero as possible. NBI, ICRF, and ECCD heating.
Diagnostic Requirements:	MSE, fast fluctuation diagnostics, diagnostics hat can measure fluctuations locally in the plasma core, rotation diagnostics, FIDA, BES.
Analysis Requirements:	The Equilibrium reconstruction will be done with with EFIT and the Ideal kink instability analysis with GATO. The sawtooth trigger analysis will be done using the Porcelli model and NOVA-K. NOVA-K will also be use for The TAE analysis and stability while the fast ion transport modeling will be done with the SPIRAL and ORBIT/ORBIT-RF codes.
Other Requirements:

Title	119: Exploration of He gas puf imaging for measurement of 2D electron density and temperature fields
Name:	Schmitz oschmitz@wisc.edu	Affiliation:	U of Wisconsin
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	E.A. Unterberg (ORNL), T. Evans (GA), M.E. Fenstermacher, M. Groth (LLNL), R. Maingi (ORNL), P. West (GA)	ITPA Joint Experiment :	No
Description:	In this proposal we want to test the capability of He gas injection and spectroscopic line emission measurement for measurement of 2D electron density and temperature fields at the divertor X-point region. This is in preparation of ROF proposal ID114 and can be performed piggy-back in any discharge with 500 ms free H-mode time accepting, short (order of 100 ms) He puffs with a gas flow of appr. 10^18 atoms per second. Three He-I lines will be measured in three similar discharges and evaluated with a collisional radiative model emloyed with good success at TEXTOR. This shall allow to measure directly n_e and T_e in the complete field of view.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	puff He from different injection capabilities (DiMES, midplane, pellet tubes) first into L-mode plasmas then into H-mode. Repeat discharges twice for different filters and monitor lines at different input powers. Try to acquire strong lines without image intensifier. Second step: commission and apply TEXTOR camera system and establish technique as standard.
Background:	Emission spectroscopy on thermal He injected for diagnostic purposes was used successfully at TEXTOR [O. Schmitz et al. 50 PPCF (2008) 115004] and represents actually the working horse of plasma edge diagnostics. With this attempt we want to transfer this technique to DIII-D in general as a new method to measure SOL and separatrix electron density and temperature fields in 2D. The ultimate goal is to apply this technique to resolve structures in the RMP perturbed boundary as described in ROF proposal ID114. However, establishing this technique needs dedicated tests which need to be performed backing up on existing programs. We need to test the observation system (we plan on the first stage using the LLNL tangential camera systems and in a second stage a new CCD system as applied at TEXTOR) and in particular which of the existing gas valve is optimal to apply. The injected He flow will not exceed a level of 10^18 He atoms s^-1 and will be applied for a short time only (~100 ms) during these tests. For measurement of electron density and temperature three He-I lines are used: 31D -> 21P (667.8 nm) (with nomenclature Nml (N:=main quantum no., m:=spin quantum no., l:=angular momentum quantum no.)), 31S -> 21P (728.4 nm) and 33S -> 23P (706.5 nm). To acquire them in these tests we need three discharge as similar as possible which represents the limitation for piggy back operation. It would be beneficial to start the tests in L-mode plasmas and extend successively to more challenging H-mode plasmas.
Resource Requirements:	L- and H-mode plasmas in LSN, different power level would be helpful as soon as the observation is commissioned
Diagnostic Requirements:	tangential camera system with standard and intensified camera. He I filters mounted and prepared to be exchanged repetitively, DiMES TV, spred spectrometers to monitor He lines, Thomson scattering to exclude edge profile manipulation by puffing
Analysis Requirements:	He collisional radiative model and EMC3/EIRENE coupled to specialized He data set (TEXTOR), python.py runs
Other Requirements:	Forst attempts can be done mostly piggy back, 2x 1/3 of a run day would be desirable after first tests and initial commissioning

Title	120: Disruption Statistics and Prediction
Name:	Wesley wesley@fusion.gat.com	Affiliation:	GA
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):	Al Hyatt, Pete Taylor, Dave Humphreys, Mike Walker and Sean Flanagan	ITPA Joint Experiment :	No
Description:	Effect automated search and analysis of disruption-related plasma and tokamak data. Compile 'statistics' re disruption causes, correlation of disruptivity with operation mode, proximity to operation limits, nature of experimental activity. Identify disruption pending indicators.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Off-line/background development of data acquisition and automated data analysis, on past and daily tokamak and plasma operations data. Intent is to minimize eventual staff requirements/effort re monitoring and assessment of data. May eventually result in coupling to disruption avoidance and/or mitigation methods and test of benefits in reducing â??disruptivityâ?? and/or failed shots owing to early disruption.
Background:	Important issue/need re ITER and DEMO
Resource Requirements:	Staff time and participation for method development, daily review and refinement. Interaction with scenario development and exploration campaigns; perhaps eventually runtime to test avoidance strategies and results.
Diagnostic Requirements:	May require disruption and/or VDE specific diagnostic settings, gains, time base, etc. to detect and record data on precursor signals, during-event plasma data and machine fault logic.
Analysis Requirements:	Automated or semi-automated analysis methods development and validation. Assessment of results and correlation with operations program. Eventual development of 'real-time' capable output capable of effecting real-time avoidance or repair or 'soft-landing' action(s).
Other Requirements:	Practical implementations should focus on providing use 'feedback' to physics operator and session leaders and in improving 'good shot' yield for routine and exploratory plasma ops.

Title	121: Optimized Low-Z and Mixed Gas MGI
Name:	Wesley wesley@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	Team MGI/DM	ITPA Joint Experiment :	Yes
Description:	Optimize low-Z (H2/D2/He) and low-fraction mixed gas massive gas injection (MGI) using the six-valve MEDUSA injector. Find for each/various pure and mixed species, find optimal 6-valve pulse duration to optimize added free and total electron density, assimilation fraction (minimize exhaust gas) and/or Rosenbluth density ratio. Also assess thermal energy effect (see #TBD). For mixed species, determine optimal low-Z carrier gases (H2, D2 or He) and fraction of Ne and/or Ar admixture. Provide data for integrated model validations of species, impurity radiation and impurity delivery (impurity entrainment) attributes	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Vary valve-open duration (6 valves the same) to assess pulse duration and species (pure and mixed) effect on fast shutdown attributes, especially added electron content, Rosenbluth density fraction (ne/nRB) and in-plasma gas assimilation. Focus on the high-quantity regime, > 2000 torr-l added in < 2 ms most relevant to ITER application. Compare with similar quantity + species injection via solid/shatter pellet. Quantify trade-off among gas utilization efficiency, attainment of max ne/nRB and applicable quantity scaling behavior. Also effect of increased Wth (see ROF #TBD)
Background:	Experiments in 2007 and 2008 show promising FPS and RE mitigation attributes for low-Z and low-fraction (eg D2 + 2% Ne) mixed gases in the short-pulse injection regime. Typical experiments to date have utilized only ca 1/2 the short-pulse capability of six-valve MEDUSA. Extrapolation of present data suggests 6-valves x 2-ms can deliver 2000-4000 torr-l of H2, D2 He or weak mixed gas. FPS and ne/nRB attributes in this 2000 torr-l plus regime can be tested/optimized. Ambiguities in present 1000-2000 torr-l regime can be resolved and scaling data extended to higher quantity (2000-4000 torr-l) and added electron content, hopefully also with increased assimilation. Results can also be compared to planned similar-quantity solid-pellet injection tests. Timely results will be relevant to on-going ITER DM system concept evaluations and hardware development
Resource Requirements:	Standard MGI target plasma (2 beams), MEDUSA injector, supplies of candidate gas mixtures, between shot gas change capability. Option to increase NBI input to full 7-beam capability, plus EC and/or IC power as available. Adequate runtime to allow for adverse impact (on NBI) of massive helium injection.
Diagnostic Requirements:	Standard + MGI/disruption-specific diagnostics, fast camera, etc. CO2 interferometer is critical (high n*l)., Urgent need for during-CQ data on [cold/high-density] plasma temperature, free and total e- density (Zeff) and in plasma n, T and neutral profiles. RE production is not expected
Analysis Requirements:	Per existing MGI experiments, supplemented with improved during CQ configuration
Other Requirements:	Helium injection adversely impacts NBI cryostability and resulting multi-beam to next-shot recovery time. EC or IC heating may be adequate for NBI-less tests at standard Wth. But pace of high-Wth many-NBI shots will necessarily be slow. Use of H2 or H2 mix may impact subsequent IC operations. Need to consider scheduling aspects.

Title	122: Dust generation from deposited layers and leading edges
Name:	Rudakov rudakov@fusion.gat.com	Affiliation:	UCSD
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Requested
Co-Author(s):	P. West, C. Wong, J. Yu, M. Groth, B. Bray, N. Brooks, M. Fenstermacher, S. Krasheninnikov, C. Lasnier, A. Pigarov, R. Pitts, R. Smirnov, W. Solomon	ITPA Joint Experiment :	No
Description:	Characterize dust generation from DiMES samples with pre-deposited hydrocarbon films and specially machined leading edges.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	DiMES samples with pre-deposited hydrocarbon films and specially machined leading edges will be exposed to known particle/heat fluxes at the strike point in LSN configuration. Dust generation will be characterized by available diagnostics (visible cameras, IR TV, MDS) and postmortem analysis of the samples.
Background:	Dust production and accumulation present potential safety and operational issues for ITER by contributing to tritium inventory rise and leading to radiological and explosion hazards. In addition, dust penetration of the core plasma can cause undesirably high impurity concentration and degrade performance. Projections of dust pro¬duction rates based on experience from existing devices are needed.
Resource Requirements:	Two ½ day experiments, one with pre-deposited layers, one with a leading edge. LSN patch panel, OSP on DiMES.
Diagnostic Requirements:	DiMES, DiMES TV, lower divertor tangential TVs, UCSD fast camera, CER, Thomson (divertor and core), filterscopes, MDS, lower divertor Langmuir probes, SPRED. IR TV and fast visible camera with view of DiMES are highly desirable.
Analysis Requirements:
Other Requirements:

Title	123: Thermal Energy Scan for Low-Z MGI
Name:	Wesley wesley@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	Team MGI/DM	ITPA Joint Experiment :	Yes
Description:	Previous experiments (2006 and before) with argon MGI using the Mark IV 'directed jet' injection system demonstrate a significant increase in assimilated ion and electron content with increasing plasma thermal energy (W_th), albeit with a co-linear quantity (Q) dependence. Experiments in 2007-2008 with helium MGI using a 5-valve MEDUSA injector showed high assimilation, ~30%, with W_th = 0.7 MJ and experiments with H2/D2 and weak mixed gases showed similar and still-increasing with quantity assimilation, plus possible (but ambiguous) indication of a Wth dependence for OH vs. L-mode vs. ELMy H-mode plasmas. Increase in W_th is expected to increase assimilation. A 2-2.5 MJ target plasma will be more 'ITER-like' with regard to reaching a high Rosenbluth density fraction in terms of the ratio of W_th to added He and electron content. The basic concept will be to do 6-valve short-pulse He MGI into a maximum W_th 'ITER-like' ELMy H-mode plasma with 7-source NBI heating. Supplemental ECH and/or ICH is an option. Need to assess whether incremental Wth is significant for DIII-D.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Make a reproducible 'ITER-like' ELMy H-mode plasma with W_th (diamagnetic) approaching 2 MJ. Do 6-valve x 2 ms He MGI (2500 torr-l). Evaluate thermal collapse, added density (n*l with fast C02) and current quench attributes. If time permits, attempt W_th 'scan' 0.5 -> 2+ MJ (1-7 sources + supplemental rf heating as available). Maybe with OH. One 2+ MJ example will indicate proof-of-principal effect, but a controlled 'W_th scan' series is needed for publication and development of Wth scaling basis for ITER.
Background:	Assimilation of injected neutral gas in MGI depends on plasma thermal energy. Past data and experience + theories indicate positive W_th dependence of assimilation fraction. With regard to achieving total electron densities = Rosenbluth density (~5e22 m-3), all present tokamaks, DIII-D included, have lower ratio, at 'maximum performance', of W_th/n_RB than ITER will have. Hence experiments with maximum W_th are needed to better approach 'ITER-like' conditions and elucidate W_th scaling in the otherwise most efficient short-pulse MGI (MEDUSA) regime. Past DIIi-D and other tokamaks also show species dependence of assimilation, etc., so a multi-species, multi-W_th data set is ultimately needed
Resource Requirements:	q ~ 3 'ITER-like' ELMy H-mode target with 1->7 source NBI, plus supplemental ECH or ICH if warranted and available. Option: 'advanced baseline' hybrid target rather than standard sawtoothing H-mode. MEDUSA six-valve injector, helium and D2 and mixed gas. Standard MGI/disruption diagnostics, especially fast magnetics, 4-chord fast C02 interferometer, DISRAD + TS etc. for before-injection characterization. Shot plan will have to accommodate need (~ 30 min) to recycle NBI cryopumping after each shot. Consider scenarios for a dedicated experiment day (or half day) with ca 4 MGI shots vs. four end-of-day experiments
Diagnostic Requirements:	Standard MGI diagnostics (per #121). Also spectroscopic and/or probe measurements of current quench plasma attributes.
Analysis Requirements:	Characterization of pre-injection target; eventual modelling with 2-D codes and diagnostic simulations to interpret C02 data, etc.
Other Requirements:	See discussion of pure and mixed gas optimization in #121. Proposed massive pellet (shatter pellet) experiments with solid pellets will explore similar/more challenging neutral dissociation and atomic ionization requirements (> 0.25 MJ for D2) and hence may comprise an extension to the W_th requirement scans planned here with MGI. Combination of gas and solid pellet results may provide a better empirical and/or model basis scaling for ITER

Title	124: 2/1 NTM Stab by ECCD in ITER Demo Discharges
Name:	La Haye rbrtlahaye@gmail.com	Affiliation:	Retired from GA
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	R. Prater, A. Isayama, A. Welander	ITPA Joint Experiment :	Yes
Description:	DIII-D ITER demo discharges at q95=3.1 and betaN~1.8 in 2008 usually developed and were terminated by m/n=2/1 NTMs. ITER relies on ECCD stabilization of such modes. However, as q95 is low and rho21 is large where Te is low, ECCD per MW injected is low. Thus no existing device has demonstrated this stabilization in an ITER matched discharge yet. With 6 good high power gyrotrons, DIII-D will have enough injected power to drive enough jeccd at rho=0.8 (q=2) to match the jboot of the ITER demo discharges in 2008 that developed 2/1 modes. Provided that the density can be reduced, this will take about 4.7 MW. TORAY-GA calculations show that with this power and reduced density, the conditions of jeccd=jboot and FWHM=wmarg will be met at q=2 for successful stabilization.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Repeat #131497 which developed a 2/1 NTM at 4300 msec before flattop ended. But turn off gas and optimize cryopumping to lower density to about 4.4E13 for effective ECCD. See if the 2/1 mode still appears (earlier possible?) Use the "dud" detector to turn on ECCD on the start of a mode and "search and suppress" with deltaZ followed by active tracking (as the absorption will be at the top of the q=2 flux surface). If the mode should not appear at lower density (unlikely?) raise betaN to reliably get the mode. If the ECCD proves inadequate with the available gyrotron power (perhaps because density still too high?), develop a target at lower BT and IP keeping shape and q95 the same. One first wants to demonstrate at the ITER q95=3.1 that one can successfully detect the onset of the 2/1 mode, turn on ECCD, align, and suppress it completely. Second, one wants to demonstrate that pre-emptive ECCD turned on with active tracking of alignment to q=2 can avoid the mode otherwise destabilizing.
Background:	ITER relies on ECCD suppression of the 2/1 NTM but this has not been demonstrated in existing devices with q=2 far out on the profile at low q95. A control strategy is proposed for ITER by La Haye et al IAEA 2008 and to be submitted to Nuclear Fusion. An experimental confirmation of this strategy would give confidence for success in ITER. This was proposed as idea 195 for 2008 but received no time. This is included in the ITPA MHD proposals for 2009 under MDC-8 "Current drive prevention/stabilzation of NTMs.
Resource Requirements:	1 DAY, ITER shape, -1.9T, 1.5 MA, cryopumps, as in #131497, 5 co beams, 6 gyrotrons
Diagnostic Requirements:	Standard particularly Thomson, MSE and ECE.
Analysis Requirements:	EFIT with MSE for equilibria, TORAY-GA for ECCD, etc are all standard.
Other Requirements:	None.

Title	125: Dependence of C deposition and D co-deposition rates on the surface temperature
Name:	Rudakov rudakov@fusion.gat.com	Affiliation:	UCSD
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	A. Litnovsky (FZJ), V. Philipps (FZJ), W. West, C. Wong, R. Boivin, W. Wampler	ITPA Joint Experiment :	Yes
Description:	Expose differentially heated metal strip mounted on DiMES holder to detached ELMing H-mode discharges. Because of the surface temperature variation, net deposition rates of C and D, D/C ratios, and deposited layer thickness will vary along the strip.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Insert the DIMES holder with metal strip into the private flux zone of high-density LSN ELMing H-mode plasmas. Expose for ~8 plasma discharges. Remove and analyze deposit thickness and C and D areal density by various techniques (ellipsometry, IBA, SIMS, XPS).
Background:	During the 2004-08 experimental campaigns it was shown that a moderate increase of the surface temperature has a dramatic effect on C deposition and D co-deposition at recessed surfaces in a tokamak divertor under detachment. C deposition was observed on molybdenum mirrors recessed below the divertor floor at room temperature, and was fully suppressed at elevated temperature between 90 - 180ºC. At 200ºC carbon deposition down a simulated tile gap was reduced by about a factor of 2-4 and D co-deposition by an order of magnitude compared to those at room temperature. Here we are proposing to measure a continuous dependence of the deposition/co-deposition rates on the surface temperature. This is also relevant for prevention of deposition on diagnostic mirrors, joint ITPA activity DIAG-2.
Resource Requirements:	½ day experiment; high density ELMing H-mode, NBI
Diagnostic Requirements:	DIMES with a specially built head, all available lower divertor diagnostics
Analysis Requirements:	Various surface analyses: IBA, ellipsometry, SIMS, XPS
Other Requirements:

Title	126: Transient divertor reattachment and detachment control
Name:	Pitts richard.pitts@iter.org	Affiliation:	ITER Organization
Research Area:	ITER Demonstration Discharges	Presentation time:	Requested
Co-Author(s):	T. Petrie, C. Lasnier, P. West, J. Watkins, P. C. Stangeby, A. Leonard	ITPA Joint Experiment :	No
Description:	Characterise divertor response to sudden loss of impurity seeding or fuelling and across confinement transitions in the presence of seeding. Study the combination of extrinsic seeding with local fuel gas injection as a degree of detachment controller.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The aim of these experiments is twofold: first to investigate the transient evolution of heat, particle flux and divertor radiation following sudden loss of extrinsic impurity seeding maintaining detachment and second to study the effectiveness of combining extrinsic seeding with local (fuel) deuterium injection to control the degree of detachment in high power H-mode discharges. In both cases the best possible measurements of divertor target heat and particle flux profiles and time evolution are required, along with the divertor radiated fraction and time dependence. This constrains the experiment to be run in LSN with optimum strike positioning for IRTV coverage of both strike points, whilst arranging for the optimum gas injection locations for both impurity and deuterium injection with respect to divertor pumping capability. Experiments should be performed in ELMing H-mode with high input power and favourable ion B�?��??B drift direction, as will be the case in ITER. The latter has relatively weak pumping, primarily through the private flux region. Ideally, both deuterium and the seeding impurities should be injected into the lower divertor, but the former could be introduced into the main chamber if this is not possible. The approach would be to first establish partially detached divertor operation with noble gas injection (neon and argon and if possible also nitrogen seeding) and then abruptly cut the injection whilst following the response of target particle and heat flux profiles and divertor radiation fraction and distribution. The same applies for the behaviour during and following an H-L back transition in the presence of seeding, provoked by cut or reduction in heating power (avoiding disruption). For experiments combining extrinsic seeding and local D fuelling, high power H-mode shots are required in which the level of local deuterium injection is varied from zero to some (to be determined) maximum level to investigate the degree to which target detachment can be influenced by the addition of gas fuelling. The effect of this local D injection on pedestal profile erosion should be studied.
Background:	The ITER divertor high heat flux components cannot tolerate the elevated power fluxes that will occur during fusion burn if the divertor plasma falls out of the partially detached state in which the machine must operate at full power. Thermal load calculations show that the time to react if plasma reattachment occurs must be on the order of 1-3 seconds at most. In this sense, reattachment is defined as a case in which divertor radiated fraction falls from the high (~60%) required in the partially detached regime to values in the region of 20%. Even somewhat higher values (e.g. 30%) extend the time to react, but not significantly. In the case of a full W divertor, foreseen for the DT phase of ITER, this radiation will have to be almost entirely supplied by extrinsic seeding impurities, most likely Ne and or Ar. However, there is not yet a quantitative physics basis on which to assess how fast and under which circumstances the reattachment might occur. Experiments and mining of previous data are thus being sought on a maximum of operating devices to construct the experimental basis, upon which modeling efforts can build (such efforts are already underway within the IO). The issue of divertor reattachment is the subject of the newly launched ITPA Divsol task DSOL-20, being led by the main author of the current proposal. ITER is currently assessing the provision for divertor injection capability and proposals have been made to increase flexibility in the injection locations, in particular allowing for seeding or fuelling into both the inner and outer divertors separately whilst allowing for reasonable toroidal uniformity. One possible gain would be to allow for a higher degree of detachment control by local deuterium fuelling in combination with the baseline impurity seeding (fuelling of the core plasma via divertor gas injection is not expected to be significant on ITER, but local effects might be more important). There have been few experiments performed on this in the past on tokamaks �?? DIII-D has the tools and the diagnostics to make an important contribution.
Resource Requirements:	One half day provided adequate reference discharges can be set up quickly. This should be possible on the basis of extensive seeding experience already gained on DIII-D. High power ELMing H-mode required, dRsep <= -2 cm (ITER simulation experiment).
Diagnostic Requirements:	All possible diagnostics for the lower divertor to construct the best possible data set for subsequent modeling, but especially IRTV, Langmuir probes and bolometry. Neutral pressure measurements in the lower sub-divertor if possible for seeding fractions and neutral fuel atom density. Pedestal diagnostics for measurement of effect of D injection. CER for core impurity density.
Analysis Requirements:	UEDGE, MIST (for edge modeling and core impurity accumulation). Some SOLPS5 modelling within the IO also possible in support of these experiments.
Other Requirements:	--

Title	127: Effect of Islands on ECCD
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	R. Prater	ITPA Joint Experiment :	No
Description:	ECCD is an important tool to control MHD, such as tearing modes. While DIII-D has done detailed studies of ECCD, these have been for an axisymmetric plasma. The helical perturbations from tearing modes may significantly change the ECCD profile, which in term could affect its application to MHD control. This experiment will examine two facets of the effect of islands on ECCD. First, the flux-surface-average parallel current density will be compared for deposition at the island O-point or X-point. Second, the ECCD profile will be decomposed into separate toroidal and helical components. This second case requires a slowly rotating island, which can be achieved using entrainment with the I-coil.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Part I: Effect of islands on flux-surface-average parallel EC current density. (1) Target plasma is to be taken from successful modulated ECCD experiment to stabilize the 2/1 NTM. Probably a mixture of co/counter NBI will be used to slow the island rotation frequency to <5 kHz. (2) During the ECCD measurement phase, the 30LT and 210RT beams should be on continuously for MSE data acquisition, (3) With EC deposited at the island O-point, compare co/radial/counter ECCD injection. For the co-ECCD case, the power should be limited so that the island is NOT stabilized. (4) Repeat last step for ECCD deposition at the island X-point. (5) Repeat last step with continuous ECCD (i.e. not modulated). Part II: Helical current from ECCD (1) The target plasma should be taken from a successful entrainment experiment where the I-coil is used to force a 2/1 tearing mode to rotate at a frequency <1 kHz. (2) Apply co/counter/radial ECCD at the q=2 location continuously (i.e. not modulated). (3) Compare co/radial/counter ECCD injection. (4) Compare modulated ECCD at island O-point or X-point for co/radial/counter injection.
Background:	Experiments on DIII-D over the last 10 years have made detailed comparisons between ECCD theory and experiments on the local level. However, these experiments specifically avoided MHD such as sawteeth and tearing modes. Thus, the ECCD studies were done in a axisymmetric plasma configuration. The highly localized region of ECCD led to the development of methods for direct analysis of the MSE signals without equilibrium reconstruction. This direct analysis method was able to determine the ECCD profile with spatial resolution limited only by the MSE diagnostic itself. Later, this methodology was extended to include the helical perturbations from tearing modes. The helically perturbed current for a m/n=2/1 "quasi-stationary" mode was successfully determined using MSE data and was reported at the 2006 EPS meeting.
Resource Requirements:	NBI: Both co and counter beams are required. EC: 6 gyrotrons are required. I-coil: Entrainment of rotating 2/1 mode required for Part II of this experiment.
Diagnostic Requirements:	MSE is critical, with highest time resolution possible.
Analysis Requirements:
Other Requirements:

Title	128: ELM filament propagation through SOL and interaction with main chamber wall
Name:	Rudakov rudakov@fusion.gat.com	Affiliation:	UCSD
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Requested
Co-Author(s):	J. Boedo, J. Yu, N. Brooks, M. Groth, T. Evans, M. Fenstermacher, E. Hollmann, C. Lasnier, A. Leonard, R. Moyer, R. Pitts, P. Stangeby, J. Watkins, P. West, C. Wong	ITPA Joint Experiment :	No
Description:	Characterize ELM filament propagation through SOL and interactions with the main chamber wall under varying density/collisionality. Collect data on the filament decay rates for comparison with analytical models (Fundamenski, Stangeby). Measure material erosion rates at the outer wall using MiMES.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Synchronize as many fast edge/pedestal diagnostics as possible to get reliable relative timing measurements. Perform measurements of the radial and poloidal propagation of ELM filaments and radial decay lengths of the filament density and temperature at 3-4 different normalized densities n/n_GW of 0.4-0.9. Perform material sample exposure with MiMES. Vary outer wall gap to increase/decrease plasma interaction with the outer wall. Turn on I-coil during the last second of the shot to attempt ELM reduction.
Background:	Plasma interaction with the main chamber wall is one of the critical issues for ITER. Interaction caused by transient events such as ELMs and disruptions are of particular concern. A series of ELM-characterization experiments during CY2004 experimental campaign at varying discharge densities has yielded excellent results on ELM dynamics. However, diagnostics of plasma interaction with the main chamber wall back then relied almost solely on the mid-plane reciprocation probe. In 2006 two filterscope channels and three camera views of the outboard wall were added. A fast framing camera proved to be particularly useful for diagnostics of the PMI caused by transient phenomena such as blobs and ELMs. A material study module â?? MiMES â?? was installed at the end of 2007 campaign. We propose repeating 2004 study with these new diagnostic capabilities. In addition, I-coils can be turned on at the end of each shot to attempt ELM control. Collected data will also be used for comparison with ELM filament models (Fundamenski, Stangeby).
Resource Requirements:	1 day experiment; LSN H-mode, 4 sources of NBI, I-coils
Diagnostic Requirements:	All available fast edge and pedestal diagnostics; profile diagnostics.
Analysis Requirements:
Other Requirements:

Title	129: Testing the sensitivity of GYRO-calculated turbulence to non-Maxwellian distribution functions
Name:	White whitea@mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	General Integrated Modeling	Presentation time:	Not requested
Co-Author(s):	M. E. Austin, J. C. DeBoo, R. W. Harvey, C. Holland, G. R. McKee, R. Prater, and T. L. Rhodes	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to obtain the first measurements of multi-field, multi-scale fluctuations in tokamak plasmas where the electron distribution function, fe(v), is intentionally distorted in a controlled manner. We will create two simple plasma conditions using the newly available 6 ECH gyrotrons at DIII-D: a first condition where little to no perturbation in the electron distribution function is expected and a second condition where large perturbations are expected. The effects of the ECH in the second condition will be a large non-thermal electron population. The fe(v) will be modeled using CQL3D to produce model distribution functions, fm(v), for the experimental conditions. The turbulence will be monitored with the BES (local, low-k n-tilde), CECE (local, low-k Te-tilde) and Doppler backscattering (DBS) (local, intermediate-k n-tilde) systems and will be compared in the two cases. The ultimate goal of this experiment will be a new dataset that can be used for direct comparisons with GYRO calculations of the turbulence and transport using the CQL3D modeled form for fm(v) as input to GYRO. Integrated modeling efforts would take place during 2009, with the experimental run-time occurring during the 2010 DIII-D run-period.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Recreate a discharge similar to 117940 which utilized varying deposition angles for ECH/ECCD and attained perturbations to fe(v) that were successfully modeled with CQL3D [1]. Perturbations to fe(v) can be optimized by adding adding counter-current ECH/ECCD deposited either near r/a ~ 0.15 or r/a ~ 0.4. An effective single neutral beam source (30L/330L 10/10 duty cycle) will be used to obtain CER/MSE data during reference shots and (150 L) will be used for BES data on different shots. Diagnostic beam blips (10-20 ms) may be used instead of full sources. Measure fluctuations using CECE, BES and DBS at two radial locations, (1) near the ECH/ECCD deposition radius, e.g. r/a ~ 0.4 and (2) far away from it, e.g. r/a ~ 0.7. In cases where large fe(v) perturbations are expected, we will monitor carefully the changes in the turbulence to determine what, if any, noticeable effects the perturbation has on the turbulence in terms of fluctuation level, spectrum, or correlation length. Case 1: Produce standard conditions using ECH to modify plasma profiles and change the turbulence and transport drives without producing any large perturbations in fe(v). The ECH will change the profiles which will then change the ITG/TEM drives and fluctuation drives. These indirect effects of the ECH on Te-tilde and n-tilde will be monitored using between shot autoonetwo/TGLF runs to analyze profile changes and ITG/TEM growth rate changes. Case 2: Produce non-standard conditions using ECH to drive a strongly non-Maxwellian fe(v) using counter-injected ECH/ECCD to generate a significant population of non-thermal electrons. In this condition direct effects of the ECH/ECCD on the EC emission will lead to enhanced Te measured with the 40-channel ECE radiometer and the Michelson interferometer. Enhanced Te-tilde (CECE) are expected, but any direct effects on n-tilde (BES/DBS) are not known.
Background:	A dedicated experiment to investigate the effects of a non-Maxwellian distribution function on turbulence is important for two reasons. First, ECH is often used as a tool for turbulence experiments and Transport Modeling Validation experiments at DIII-D. Being able to model these conditions regularly with integrated Fokker-Planck (FP) codes and turbulence simulation codes and will provide strong evidence that the primary effect of ECH in these conditions is local heating and profile modification rather than distribution function distortion, which is a key assumption in these works. This new experimental/integrated modeling study is therefore important for validation efforts as it will help eliminate any uncertainties regarding interpretation for fluctuation measurements and design of synthetic diagnostics in these types of conditions - this is of particular interest for electron temperature fluctuation measurements and new CECE synthetic diagnostic design. Second, and more generally, the effect of non-Maxwellian distribution functions on transport and turbulence is of interest for understanding reactor-grade tokamak plasmas. In a reactor, the slowing down distribution function for alpha particles may affect alpha transport and the high energy tails in the electron energy distribution function may impact electron transport and current-drive efficiency. Drift-wave transport models based on nonlinear gyrokinetic (NLGK) theory solve for the fluctuations and transport in conditions where the distribution function is Maxwellian or very nearly a Maxwellian. Solving for the turbulence and transport using non-Maxwellian distribution functions is more complicated but has been studied recently by several researchers [2,3]. Despite the challenges, the process of integrating model distribution functions from CQL3D into GYRO or other NLGK equation solvers is important for ITER and beyond as eventually these types of codes will be used to predict current-drive and heating efficiency (FP codes) and transport (NLGK codes) in reactor-grade conditions where the distribution functions may be strongly non-Maxwellian. References: [1] C. C. Petty et al. GA Report A25804 (2007) [2] C. Angioni et al. 13th EU-US TTF Workshop (2008) [3] E. Gravier, et al. Phys. Plasmas, 15, 122103 (2008)
Resource Requirements:	1 day experiment, 5-6 gyrotrons, 30L/330L/150L beams.
Diagnostic Requirements:	Thomson scattering, Michelson interferometer, 40-channel ECE radiometer, CER and MSE, fast magnetics, all available fluctuation diagnostics.
Analysis Requirements:	EFIT, ONETWO/autoonetwo, GENRAY, TORAY, and CQL3D, TGLF, GYRO.
Other Requirements:	--

Title	130: Investigate Disagreements Between Thomson Scattering and ECE Measurements in High Te Discharges
Name:	White whitea@mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	Heating & Current Drive	Presentation time:	Requested
Co-Author(s):	M. E. Austin, B. Bray, R. Pinsker, R. Prater	ITPA Joint Experiment :	Yes
Description:	The goal of this experiment is to search for a discrepancy between Thomson scattering (TS) and ECE measurements of Te on DIII-D in discharges with high electron temperature. To carry out the experiment, L-mode discharges with core transport barriers will be created with neutral beam injection (NBI) plus fast wave heating (FWH) to attain central electron temperatures of Te(0) = 7-8 keV or higher. Multiple repeat discharges can be used to obtain statistically significant data, which can include plasma jogs to reduce scatter in profile measurements. Core electron temperature measurements from TS and the absolutely calibrated Michelson interferometer would be compared to look for any discrepancy between the two diagnostics. As an ITPA joint experiment, either a positive or a negative result on this topic from DIII-D can significantly impact international efforts to understand the past discrepancies that have been reported on TFTR and JET [1,2]. For example, a positive result, the observation of the discrepancy between TS and ECE on DIII-D, would verify the discrepancy on an additional machine and would therefore motivate a new and detailed investigation of the phenomenon. However, a negative result, the observation of no discrepancy under a variety of conditions with high electron temperature produced with NBI and FWH, would be equally beneficial as it would show that agreement between TS and ECE can be obtained in high temperature tokamak discharges. In both cases, the experimental results and associated modeling will improve the understanding of heating and diagnostic techniques in high temperature plasmas relevant for ITER and reactor-grade tokamak experiments.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Recreate a discharge similar to 87311 (Bt=1.9T, Ip=1.2 MA) which attained a central electron temperature of Te(0) = 8 keV using 2.5-5 MW of NBI plus 2.5-3.0 MW of FWH. In this new experiment we will adjust the timing of NBI and FWH to maximize the effect and create the most robust improved core confinement without creating large MHD modes in order to access the highest electron temperatures possible without using ECH. Also, increasing the B-field can be used to further improve electron energy confinement. Short (0.5 sec) pulses of ECH can be added to the discharges to increase FW absorption. This experiment will employ the unique set of turbulence diagnostics at DIII-D (BES, CECE, reflectometry, scattering) to monitor turbulence changes as well as any low-amplitude, core localized MHD activity or other oscillations that may exist at times when the TS/ECE measurements disagree. In addition to allowing access to high electron temperatures via NBI and FWH in order to study TS/ECE discrepancies, this type of discharge would also be useful to demonstrate and characterize increased FW absorption with electron temperature, which would be a natural joint experiment.
Background:	Note that in the core of typical tokamak plasmas with Te(0) < 7 keV the TS and ECE measurements of electron temperature are in very good agreement. Also note that TS and ECE measurements of electron temperature often disagree in high Te(0) discharges that are strongly heated via ECH, but in these cases the disagreement can be explained by a well understood perturbation of the electron energy distribution function caused by the ECH [3]. In contrast to these cases, the cause of the TS/ECE discrepancy in discharges heated with only NBI and FWH where Te(0) > 7 keV is not known. Such a discrepancy has been observed in NBI discharges and discharges heated with both NBI and Ion Cyclotron Resonance Heating (ICRH) discharges in TFTR and JET [1,2]. In these cases, the central electron temperature Te(0) measured with ECE diagnostics is 10-20% higher than the TS measurement of Te(0) The discrepancy starts at Te(0) ~ 7 keV and increases approximately linearly with electron temperature. Theoretically, a non-Maxwellian electron distribution f(v) with distortion near the thermal velocity may create such a measurement discrepancy between TS and ECE measurements [4], however, no known mechanism can sustain that type of distribution with finite heating power. References: [1] E. de la Luna, et al., Rev. Sci. Instrum. 74, 1414 (2003) [2] G. Taylor, PPPL report 4202 (2006) [3] C. C. Petty et al. GA Report A25804 (2007) [4] V. Krivenski et al. 29th EPS Conference on Plasma Phys. and Contr. Fusion Montreux, 17-21 June 2002 ECA Vol. 26B, O-1.03 (2002)
Resource Requirements:	1.5 experimental days 1/2 day for discharge development and search for discrepancy followed by 1 day to fully characterize any discrepancy in a variety of heating scenarios. 2 gyrotrons FW heating systems All available NB sources
Diagnostic Requirements:	Thomson scattering, Michelson interferometer, 40-channel ECE radiometer, CER and MSE, fast magnetics, all available fluctuation diagnostics.
Analysis Requirements:	EFIT, ONETWO/autoonetwo, GENRAY, TORAY,and CQL3D.
Other Requirements:	--

Title	131: DiMES Erosion Measurements with Detached Plasmas Induced by Argon Injection
Name:	Rudakov rudakov@fusion.gat.com	Affiliation:	UCSD
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	W. Wampler, W. West, T. Petrie, D. Whyte, C. Wong, R. Moyer, S. Allen, N. Brooks, P. Stangeby, J. Boedo	ITPA Joint Experiment :	No
Description:	DiMES will be used to measure the rate of carbon and tungsten erosion at the outer strike point (OSP) of H-mode plasmas detached by argon injection. This experiment should determine whether argon-detached plasmas can be used with a carbon/tungsten divertor in a tritium fueled next step device. Results will be compared to those from a previous DiMES exposures to plasmas detached by neon injection and semi-detached plasmas with argon injection.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	This is a continuation of previous DiMES experiments with noble gas injection. A DiMES probe containing depth-marked graphite and tungsten (surface-deposited films or solid buttons) samples will be exposed to the OSP with LSN H-mode plasmas detached by argon injection. Plasma and machine parameters should be close to those used in the previous experiments with neon and argon injection. The OSP should be moved onto the DiMES probe during periods of detached H-mode for a total exposure time of 10-12 seconds (3-4 discharges). Heat flux reduction during Ar injection should be about a factor of 4 compared to the attached phase. Divertor plasma conditions should be characterized to allow comparison of experimental results with erosion/deposition models. Pellet ELM pacing may be used to prevent density runaway in ELM-free H-mode.
Background:	Carbon and tungsten are the materials currently in the ITER divertor design. A possibility of switching to all-tungsten divertor for D-T phase is being discussed. Divertor detachment is required to reduce heat flux and erosion rates of the targets. With all-tungsten divertor impurity injection will be required to reach the detachment. Previous studies using DiMES showed that net carbon erosion rate at the OSP with neon injection can be rather high even with detachment. Attempts have been made to repeat this experiment with argon injection, but due to various problems the desired plasma conditions were not achieved. During the latest attempt argon injection rate had to be limited because of the density runaway at detachment. As a result, OSP stayed attached through most of the exposure. Preliminary results indicate that tungsten from a surface-deposited strip has eroded. An experiment with better controlled conditions is needed for better extrapolation to ITER.
Resource Requirements:	½ day experiment; high density ELMing H-mode, NBI, pellet injection for ELM pacing desirable
Diagnostic Requirements:	DiMES, all available lower divertor diagnostics
Analysis Requirements:	IBA analysis of the exposed samples
Other Requirements:

Title	132: New Optimal Plasma Shape for AT Scenario?
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	For the high q_min, steady-state AT scenario, switch the plasma shape from the standard unbalanced DND shape to the lower SND shape in shot 129323. The new plasma shape is proved to have high beta limits and low electron heat transport for the low q_min hybrid scenario. If these properties are present in the q_min>1.5 AT scenario, the result will be (1) higher electron temperature (and higher confinement), and (2) higher noninductive current fraction.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The main objective of this experiment is to repeat the high-beta, steady-state AT scenario but with the plasma shape given by shot 129323. The heating waveforms during the current ramp up phase will been to be optimized to raise q_min above 2 at the beginning of the flat top phase. If stronger cryopumping is desired to reduce the plasma density, than reverse BT direction may be required.
Background:	During an ECCD stabilization experiment in 2007, it was recognized that the discharges developed had some interesting properties (example: shot 129323). Although RWM feedback stabilization was not being used, the plasma beta exceeded the ideal no-wall limit with beta_N reaching 3.5 before the beam power topped out. Even more interesting was the fact that the core electron temperature was ~1 keV higher than normal for the hybrid scenario. This was a result of a much lower than typical electron heat transport. Usually for the hybrid scenario in the standard AT plasma shape, heat loss through the electron channel is dominant. This is attributed to ETG-scale turbulence. However, for the lower SND shaped used in this ECCD experiment, the electron heat loss was much lower than the ion heat loss. This plasma shape was used for high-beta, steady-state hybrid experiments in 2008. Here it was found that even with 3.0 MW of ECCD and Te=Ti except near the axis, the confinement time remained high with H_98=1.4. This is a much better transport result than for ECH hybrid experiments in the standard AT plasma shape where H_98 normally drops below 1.1.
Resource Requirements:	NBI: All co beams required (5 in 2009, 6 in 2010). 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	133: Comparison of Phase Between Density and Electron Temperature Fluctuations with GYRO Predictions
Name:	White whitea@mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	Transport Model Validation	Presentation time:	Requested
Co-Author(s):	T. L. Rhodes, W. A. Peebles, J. C. DeBoo, J. H. Hillesheim, C. Holland, G. R. McKee, G. Staebler, R. E. Waltz	ITPA Joint Experiment :	No
Description:	The primary goal of this experiment is to make direct comparisons of the measured phase angle between electron temperature and density fluctuations with both existing, and future, GYRO and TGLF predictions. During 2008 a diplexed reflectometer/CECE system demonstrated the required cross phase measurement capability - for the first time in a tokamak. The measurement was performed in diverted Ohmic and ECH heated plasmas. In addition, during 2007, nonlinear GYRO calculations were performed at r/a = 0.5 and r/a = 0.7 for NBI heated diverted plasmas (t =1500 ms in 128913; see White POP 2007). Analysis indicated that density and temperature fluctuations were partially out-of-phase, and that the temperature fluctuations were responsible for ~80% of the electron thermal transport. Direct comparison of measured and predicted phase relationships is a critical test of code validity. These phase relationships are fundamental to the physics embedded in the codes and ultimately govern the predicted transport. The method proposed for this experiment and model comparison is two-fold. First, we will directly test the above existing predictions from GYRO in 128913 (White POP 2007). These comparisons will be performed at a variety of normalized radii. Second, based on the work of Waltz presented at the APS 2008 Conference, it is possible to use TGLF to make predictions for changes in the phase angle from quasilinear theory under new experimental conditions - ahead of time - and then test these predictions experimentally. The proposed experiment represents an important and unique contribution to transport model validation while also adding to our knowledge of the fundamental physics of turbulence and transport in core tokamak plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce as a baseline a simple, sawtooth-free L-mode plasma such as 128913 t = 1300 ms. The baseline beam heated case will be perturbed with ECH deposited in the core at t= 1500 ms. Later in time, the beam heating will be stopped except for diagnostic beam blips for CER/MSE measurements. The phase angle between density and electron temperature fluctuations will be measured with diplexed reflectometer/CECE systems in an L-mode plasma with varying levels of ECH power. The discharge will be repeated to scan the reflectometer/CECE systems spatially to 1) optimize spatial overlap and 2) measure the phase at different radial locations ( 0.6 < r/a < 0.8). Control over the experimental conditions (shape, density, # of shots in scans, etc.) is crucial to ensure optimization of the phase measurement. This experiment will allow for development of improved synthetic diagnostics and other validation analysis techniques as we introduce a new parameter, the n-tilde Te-tilde phase angle, to the measurable set of turbulence characteristics. After this experiment is performed and comparisons with GYRO and TGLF are completed, natural extensions of this work will be to study the phase angle between the two fluctuating fields under conditions where the transport is expected to change significantly. For example, future scans of kappa, shear, beta, Te/Ti, collisionality, etc. could all include phase angle measurements and comparisons with GYRO/TGLF predicted trends for those scanned conditions.
Background:	This new experiment is an extension of successful modeling and experimental results from 2007 and 2008. In 2007, electron temperature and density fluctuations were measured for the first time at DIII-D in neutral beam heated L-mode plasmas, and the results were modeled with TGLF and GYRO. GYRO was used to calculate the electron temperature-density phase angle in this simple, L-mode plasma 128913, with the result that electron temperature and density fluctuations are predicted to be out of phase. In 2008, the electron temperature-density phase angle was measured in ECH/OH plasmas using the new diplexed reflectometer/CECE systems. The results from 2008 show that the electron temperature-density phase angle can be measured reliably using the diplexed reflectometer/CECE systems. It should be noted that the measured phase rela-tionship was observed to change incrementally with increasing ECH power launched into an Ohmic target plasma, but the existing GYRO prediction is for only one condition is a beam-heated L-mode plasma with no ECH. The next step is to extend and connect the past modeling and experimental results using a new experiment with two parts. In the first part, we will measure the phase angle in a simple L-mode discharge similar to 128913 in order to immediately and directly test the GYRO predictions for that experi-mental condition that are already in-hand. In the second part, we will measure the phase angle in a modified L-mode case using ECH to modify the profiles and turbulence drives. To plan for this second part, experimental conditions where the phase angle is expected to change significantly can be identified ahead of the experiment using TGLF modeling. Work by Waltz APS 2008 showed that the phase angle between fluctuating quantities from quasilinear theory is in good agreement with full nonlinear theory in many conditions. Also, TGLF modeling of the phase angle in 128913 will be performed and cross-checked against the GYRO predictions to help validate the new TGLF predictions that will be made ahead of time for the new experiment.
Resource Requirements:	1 to 1 1/2 experimental days At least 5 gyrotrons, 6 would be better. All beam sources
Diagnostic Requirements:	CECE, multi-channel tunable reflectometer, FIR intermediate- and low-k systems, BES, perhaps the new PCI system. Thomson scattering, 40-channel ECE radiometer, CER and MSE, magnetics, CO2 interferometer, profile reflectometer.
Analysis Requirements:	Standard analysis for GYRO/TGLF input file generation: EFIT, ONETWO, and autoonetwo are essential. General k-space averaging capabilities (i.e. synthetic diagnostic modules) for phase angle and fluctuation level estimates from the Quasilinear Theory calculations (TGLF) and Nonlinear Theory calculations (GYRO) are highly desired.
Other Requirements:

Title	134: Fast pellet mass drift physics experiment
Name:	COMMAUX commaux@fusion.gat.com	Affiliation:	ORNL
Research Area:	General ITER Physics	Presentation time:	Requested
Co-Author(s):	L. Baylor, T. Jernigan, P. Parks, B. Pegourie	ITPA Joint Experiment :	No
Description:	Pellet injection is planned to be the main fueling method on ITER. The very high density and temperature of the plasma will not allow a deep penetration of the pellet. But this shallow penetration is expected to be compensated a strong curvature and gradB drift which will deposit the particles much deeper in the plasma for the pellets injected from the inner wall. A relation between the pellet deposition profile and the q profile has been observed on several machines. This effect could have important consequences on the particle deposition profile on ITER. In order to evaluate the consequences and the efficiency of the fueling on ITER, a good understanding of this drift effect is important.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Inject HFS pellets in ELMing H modes and changing slowly the penetration of the pellet with respect to integer q surfaces (penetration inside or outside of q=3). This can be done by changing the pellet speed (from 80 m/s up to 200 m/s and changing the q profile using Ip, Bt and NBI current drive to change the position of q=3 in the plasma. The analysis of the pellet deposition profile will be used to determine how the q profile affects the fast drift of the pellet mass.
Background:	Using the pellet injection data from DIII-D, a relation has been found between the pellet deposition profile and the integer q surfaces. But there is a strong variation of plasma and pellet injection conditions for these shots (NBI power, Ip, Bt, H or L mode...). A systematic study with controlled changes in the plasma parameters is still required to prove this effect and to evaluate its consequences on the fueling of ITER.
Resource Requirements:	1 day experiment with Co/Cn NBI, fast reflectometry, 30 NBI line (for MSE), pellet injector
Diagnostic Requirements:	Thomson burst mode (if available), fast reflectometers, MSE data
Analysis Requirements:	--
Other Requirements:	--

Title	135: Influence of the density on the RMP ELM suppression
Name:	COMMAUX commaux@fusion.gat.com	Affiliation:	ORNL
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	L. Baylor, T. Jernigan, E. Unterberg, O. Schmitz, A. Polevoi, W. Houlberg, T. Evans	ITPA Joint Experiment :	Yes
Description:	The heat load from ELMs on the plasma facing components is an important issue for the design of ITER. A new technique using non axisymetric resonant magnetic perturbations to suppress the ELMs has been successfully tested on DIII-D and could be applied on ITER. ITER is to be operated at high Greenwald fraction (0.85 for the baseline scenario and 0.95 for the 500MW inductive scenario at Q=10). The RMP suppresses the ELM activity at low density on DIII-D. But several experiments at higher densities did not allow a good ELM. It is therefore important to test the effect of the density on the RMP ELM suppression	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Apply the n=3 "usual" ELM suppression RMP field on ELMing H modes to obtain high quality ELM suppression. Add HFS pellets injection and try to change the density in order to to measure a possible density threshold on the ELM suppression below which the suppression is achieved. Then lower the toroidal field and the plasma current in order to maintain the q95 value in the resonant window and try again to get a threshold value on the density. It would allow determining if this value can be expressed in term of Greenwald fraction or absolute density
Background:	All the successful ELM suppression attempts have been achieved at low Greenwald fraction (ne<0.45 neGr). But most of these shots have attempted at the same Greenwald density value (i.e the same plasma current). So the current RMP database is not sufficient to determine if this possible density effect is an absolute density effect or a Greenwald fraction effect.
Resource Requirements:	1 day experiment with Co/Cn NBI, pellet injector, I coils with current capacity up to 7kA
Diagnostic Requirements:	fast reflectometry, fast magnetic, IR camera, CER diagnostic, UCSD fast camera
Analysis Requirements:	CER rotation analysis
Other Requirements:	--

Title	136: Pellet fueling in high density ITER-like discharges
Name:	COMMAUX commaux@fusion.gat.com	Affiliation:	ORNL
Research Area:	General ITER Physics	Presentation time:	Requested
Co-Author(s):	L. Baylor, T. Jernigan, E. Doyle	ITPA Joint Experiment :	No
Description:	The fusion performance expected on ITER has been calculated assuming an energy confinement in agreement with the H98 empirical confinement scaling. ITER plasmas are planned to be operated at significant Greenwald fraction (0.85 for the 400 MW inductive baseline scenario and 0.95 for the 500MW inductive scenario) with H98=1 in order to achieve high Q. But it has been proven on several tokamaks that such high density conditions can be difficult to obtain using gas puffing and can show confinement degradation when compared to the H98 scaling. DIII-D is uniquely capable of producing high performance discharges with the ITER shape. It is therefore important to test the consequences of such high density operation on the plasma performance of the ITER scenario using this unique characteristic.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Obtain the ITER baseline scenario plasmas. Compare the behavior of the plasma when using gas puffing or pellet injection (HFS shallow injection) in order to increase the density to a Greenwald fraction of ~0.9. If possible, try to improve the pumping efficiency by small adjustments of the strike points. Use the pellets also as a probe to determine the particle transport coefficients. Use a combination of NBI and RF heating to maintain a strong central heat deposition at high density. NBI alone at high density will lead to off axis heating, which will result in reduced stored energy and lower confinement than the H98 scaling.
Background:	The 2008 campaign was successful at demonstrating several ITER operational scenarios on DIII-D. It was possible to reproduce the shaping, the beta, the collisionality... but the density was not at the Greenwald fraction expected on ITER (it is difficult to achieve both ITER relevant collisionality and Greenwald fraction on DIII-D). It is therefore important to extend these studies to the high density regime to understand the consequences of high density on the confinement.
Resource Requirements:	1 day experiment with Co/Cn NBI, pellet injector, fast wave heating, ECH heating
Diagnostic Requirements:	fast magnetics
Analysis Requirements:	--
Other Requirements:	--

Title	137: Determination of particle transport using pellet injection
Name:	COMMAUX commaux@fusion.gat.com	Affiliation:	ORNL
Research Area:	Transport	Presentation time:	Requested
Co-Author(s):	L. Baylor, T. Jernigan, T. Rhodes	ITPA Joint Experiment :	No
Description:	The particle transport is an important topic for the achievement of the good fusion performance on ITER. The density profiles have been assumed rather flat in the fusion performance assessments guiding the design of the machine. But recent simulations have shown that this profile could be significantly peaked at the center thus improving the performance. It is then important to determine accurately the particle transport coefficients in ELMing H mode. The new density profile diagnostics available in DIII-D coupled with perturbation methods like pellet injection can allow a precise determination of the transport coefficients in different scenarios.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Obtain basic ELMing H mode plasmas. Inject HFS pellets to obtain a perturbation of the density profile and diagnose the density profile during the relaxation with all the possible density diagnostics available (Thomson scattering, fast reflectometry...). Modify characteristics of the plasma in order to investigate the influence of these parameters on the transport coefficient (q95, Greenwald fraction, I coil current, NBI power, shaping...)
Background:	Transport coefficient measurement on basic H mode scenario have already been attempted on DIII-D but these experiments did not use pellet as a perturbation method, did not benefit from the fast reflectometry diagnostic which can increase the temporal resolution of the measurements and were carried out using the former PFC configuration of DIII-D. It is then important to carry out new experiments in order to improve the understanding the particle transport (which has been much less studied than the heat transport because of the measurement difficulties)
Resource Requirements:	1 day experiment with Co/Cn NBI, pellet injector
Diagnostic Requirements:	fast reflectometry, Thomson scattering
Analysis Requirements:	a comparison of these experimental transport coefficients with theoretical coefficients (GLF23 model...)
Other Requirements:	--

Title	138: Validation of an Atomic Model for the Motional Stark Effect Spectrum.
Name:	Antoniuk antoniuk@fusion.gat.com	Affiliation:	PPPL
Research Area:	Heating & Current Drive	Presentation time:	Requested
Co-Author(s):	K. Burrell, B. Den Hartog, D. Den Hartog	ITPA Joint Experiment :	No
Description:	We propose to to investigate and validate a code that does atomic modeling of the MSE spectra. This would be a collaboration between DIII-D and MST at UW-Madison, and would validate a code that is part of ADAS. The B-Stark diagnostic has the unique ability to directly measure many of the predictions of this model. Once the model is validated it would have wide applicability to MSE based diagnostics. To do this validation, scans of the plasma density, temperature and beam voltage are required.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This experiment would be done in conjunction with the B-Stark Validation experiment. Produce and maintain plasmas during which density, temperature and beam voltage scans can be performed. Density scan to be done over the range 2x10^19-6x10^19. Temperature and beam voltage ranges to be determined. For these scans either L or H mode shots may be used. Several Beam Into Gas with toroidal field shots at different gas pressures will also be taken.
Background:	Diagnostics based on the motional Stark effect are a standard way of measuring the internal magnetic field in a variety of magnetic confinement fusion devices. The ability to accurately model the MSE spectra is important to the operation and calibration of these diagnostics. Several codes have been developed for the purposes of this modeling. The accuracy of these codes in plasma relevant conditions have not been well tested. The use of these codes is essential for interpretation of the MSE spectrum in cases where the level populations of the beam neutrals are not in statistical equilibrium. This is certainly true for beam into gas shots, which are needed for calibration of MSE and B-Stark. In plasma shots, deviations from the statistical levels are also seen, particularly at lower densities (< 10^14 cm^-3). This is a significant issue at MST which generally has densities of 0.5-2.5 x 10^13 cm^-3. For any MSE system in ITER it will be necessary to perform in-situ calibrations. One way of doing this is to use Beam Into Gas. Accurate modeling of the level populations in beam into gas will be necessary for this to be successful. The B-Stark diagnostic installed here on DIII-D is ideally suited for validation of these codes. With the B-Stark diagnostic the population levels of the beam neutrals can be measured. These measurements can be compared with the predictions of the codes. Validation of an atomic model will also allow an major improvement in the analysis routines for the B-Stark diagnostic. By using the model to constrain the population levels in the spectral fit, the sensitivity of the diagnostic can be greatly improved.
Resource Requirements:	330L, 330R, 30L & 210R beams
Diagnostic Requirements:	MSE, CER, Thompson.
Analysis Requirements:	--
Other Requirements:	1/2 day runtime. To be combined with "B-Stark Validation" (#139)

Title	139: B-Stark Validation
Name:	Antoniuk antoniuk@fusion.gat.com	Affiliation:	PPPL
Research Area:	Heating & Current Drive	Presentation time:	Requested
Co-Author(s):	K. Burrell	ITPA Joint Experiment :	No
Description:	In order to determine the effectiveness of the B-Stark diagnostic in measuring the magnetic field line pitch, detailed comparisons need to be made with MSE over a range of fields and currents. A diagnostic of similar design to B-Stark is being considered for ITER. This type of system has not yet been validated.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This experiment would be done in conjunction with the Validation of an Atomic Model for the Motional Stark Effect Spectrum experiment. Produce and maintain plasmas during which current and field, and density scans can be performed. In addition to the scans we will also use current diffusion during the start up to scan q. Field to be scanned between 1.40 - 2.16 Tesla. Current scan to be performed over the maximum possible range consistent with q95<3, <2.5 preferred. Density scan over the range 2x10^19-6x10^19. For these scans either L or H mode shots may be used.
Background:	The magnetic field line pitch is an essential input for magnetic equilibrium reconstruction codes for many magnetic confinement fusion devices. On DIII-D, as in most tokamaks, the field line pitch is measured though motional Stark effect (MSE) polarimetry. This technique is based on measuring the polarization angle of D-alpha light emitted from the neutral beams. In machines with high temperatures and densities, such as ITER, coatings on the plasma facing mirrors are expected to be a major issue for MSE polarimetry based diagnostics. The coatings can cause changes in the polarization direction of light incident upon them. An alternative method to using the polarization direction is to measure the intensities of the individual lines in the Stark spectra. This method is insensitive to the polarization direction. While it is still sensitive to polarization dependent transmission, it may be easier to calibrate for this effect. The B-Stark diagnostic is based on this method of measuring the intensities of the Stark lines. So far no thorough studies of the performance and accuracy of this diagnostic have been performed. DIII-D has the advantage of having an excellent MSE Polarimetry system with which the B-Stark results can be compared.
Resource Requirements:	330L, 330R, 30L & 210R neutral beams.
Diagnostic Requirements:	MSE, CER, Thomson
Analysis Requirements:	--
Other Requirements:	1/2 Day runtime. To be combined with "Validation of an Atomic Model for the Motional Stark Effect Spectrum" (#138)

Title	140: ELM pacing using the new pellet dropper configuration
Name:	Baylor baylorlr@ornl.gov	Affiliation:	ORNL
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	N. Commaux, T. Jernigan, J. Yu	ITPA Joint Experiment :	Yes
Description:	The ELM heat loads could damage the plasma facing components on ITER. An ELM mitigation system is required to minimize this damage. The injected fueling pellets trigger ELMs very reliably. Using high frequency pellet injection to increase the natural ELM frequency, and thus decrease the energy loss from individual ELMs could lower the consequences of the ELM activity. This method has only been tested on ASDEX-U with a low frequency (2 times the natural ELM frequency). The purpose of this experiment on DIII-D is to demonstrate that the pacing is possible with ITER relevant pellet injection (reaching at most the top of the pedestal) at higher frequencies (up to 5 times the natural ELM frequency).	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Use the ITER shaping to obtain plasmas with a low natural ELM frequency (down to 10Hz or less). Then inject high frequency pellets (30-40 Hz) in this plasma with the pellet dropper and measure the ELM frequency during the injection phase and compare it with the natural ELM frequency to measure the improvement. Compare the plasma energy loss from â??naturalâ?? ELMs with the triggered ones to confirm whether or not the energy loss content scales approximately as the inverse of the pacing frequency. Several NBI power levels, density levelsâ?¦ would be tested as well to test the effect of the pedestal parameters on the pacing efficiency and ELM behavior
Background:	The pellet pacing on DIII-D using the pellet dropper has been briefly tested during the 2008 campaign without success. It appeared that the pellets were bouncing out of the plasma due to the fast ions in the SOL and did not penetrate. This was probably due to the low speed of the pellet normal to the plasma (~3m/s). A bouncing plate has been added to the guiding tube which will make the pellet bounce toward the plasma, thus inducing an almost normal trajectory and increasing the normal speed up to 10m/s. This will hopefully improve the results of the dropper pellet pacing.
Resource Requirements:	1 day experiment with Co/Cn NBI, the pellet dropper after proven effective in piggyback.
Diagnostic Requirements:	UCSD fast framing camera, fast reflectometers, fast magnetics
Analysis Requirements:
Other Requirements:

Title	141: Pellet ELM triggering physics study
Name:	Baylor baylorlr@ornl.gov	Affiliation:	ORNL
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	N. Commaux, T. Jernigan	ITPA Joint Experiment :	No
Description:	The heat loads from ELMs could damage the plasma facing components on ITER. An ELM mitigation system is therefore required to minimize this damage. The injection of fueling pellets from all locations triggers ELMs very reliably and is thus pellet pacing of ELMs is being considered for an ELM mitigation system. But the ELM triggering physics is still not well understood. The ITER requirements for the pellet pacing system are based mainly on assumptions. It is therefore necessary to understand the mechanism of the pellet triggering in order to predict the consequences of the pellet pacing on ITER and to check if the ITER requirements are correct. Some experimental data show that the peeling/ballooning model used to predict the onset of an ELM is not sufficient to explain what happens when a pellet triggers an ELM since the ELM appears to be triggered before any change in the global pressure gradient or current profile in the pedestal.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Use the ITER shaping to obtain plasmas with a low natural ELM frequency (down to 10Hz or less). Then use the pellet injector to inject pellets from various locations (LFS, HFS, vertical injections) with different speed and sizes in order to test the influence of the poloidal angle and the penetration of the pellet on the ELM properties and on the onset of the ELM. Several pedestal parameters should be tested using NBI power, triangularity, Ip, densityâ?¦ to change the pedestal and the ELM parameters
Background:	Several examples of pellets triggering ELMs have allowed the study of some characteristics of the ELM triggering. But the available data lack a number of key plasma and pellet injection measurements such as fast magnetics. No systematic study has been carried out in order to understand the ELM triggering physics
Resource Requirements:	1 day experiment with Co/Cn NBI, the pellet injector
Diagnostic Requirements:	UCSD fast framing camera, fast reflectometers, fast magnetics, Thomson scattering burst mode (if possible)
Analysis Requirements:
Other Requirements:

Title	142: RMP-pellet compatibility
Name:	Baylor baylorlr@ornl.gov	Affiliation:	ORNL
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	N. Commaux, T. Jernigan, E. Unterberg, O. Schmitz, A. Polevoi, W. Houlberg, T. Evans	ITPA Joint Experiment :	Yes
Description:	The heat load from ELMs on the plasma facing components is an important issue for the design of ITER. A new technique using non axisymetric resonant magnetic perturbations to suppress the ELMs has been successfully tested on DIII-D. But some experiments combining pellet fueling and RMP on DIII-D have shown that pellet injection can temporarily disturb the ELM suppression. Some cases of individual pellets show no sign of immediate ELM triggering. Pellet injection is planned to be the main fueling method on ITER. Therefore the combination of RMP ELM suppression and pellet fueling is very likely in ITER plasmas. It is then important to understand the physics of the interaction in order to maintain reliably the ELM suppression during the pellet fueled high density plasmas. This is a high priority ITPA and ITER sanctioned research area for RMP ELM suppression.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Apply the n=3 â??usualâ?? ELM suppression RMP field on ELMing H modes to obtain high quality ELM suppression. Add individual HFS pellets injection (low frequency ~1Hz) and try to change the plasma parameters to suppress the ELM bursts that pellet injection can trigger (injected torque, shaping, strike point position to increase the pumping efficiencyâ?¦). Then increase the pellet injection frequency to test the influence of the pellet injection frequency. One possibility is modifying the injected torque since the rotation damping induced by the pellet injection could be an effect explaining the loss of the ELM suppression in some cases. The influence of the q95 should be tested (through a q95 scan) also inside the resonant window. It will be also interesting to change the pellet parameters to test the influence of the pellet penetration and injection location on the occurrence of ELM bursts after a pellet injection.
Background:	Several shots with non-optimized RMP ELM suppression have been proven sensitive to pellet fueling when the density is significantly increased. In these shots (for example 131466, 131467) the first pellet injected triggered a bifurcation back to ELMing H mode. It is important to understand what triggers this bifurcation and a recipe to avoid it. Cases exist with single pellet injection where ELMs are not triggered and the plasma maintains an ELM suppressed state.
Resource Requirements:	1 day experiment with Co/Cn NBI, Bt at 2.15T (for ECE), pellet injector, I coils with current capacity up to 7kA
Diagnostic Requirements:	fast reflectometry, fast magnetic, IR camera, CER diagnostic, fast camera, ECE
Analysis Requirements:	CER rotation analysis
Other Requirements:

Title	143: Beta Scaling of Multi-Field, Multi-Scale Turbulence on DIII-D
Name:	White whitea@mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	T. A. Carter, C. Holland, J. Kinsey, G. R. McKee, W. A. Peebles, C. C. Petty, T. L. Rhodes, L. Schmitz, G. Staebler	ITPA Joint Experiment :	No
Description:	The purpose of this experiment is to investigate beta scaling of long-wavelength density and electron temperature fluctuations and multi-scale density fluctuations on DIII-D. As pointed out by Candy POP 2005, high-beta regimes present an important challenge for electromagnetic drift-wave transport models used in codes such as GYRO. These high-beta regimes are therefore important to study as part of ongoing verification and validation efforts aimed at developing predictive transport codes for ITER and beyond. This proposed beta scaling experiment will produce a unique fluctuation data set that can be used for quantitative comparisons with GYRO, and will therefore provide valuable information about the limitations of both the turbulence models and the experimental measurements. TGLF will be used extensively to evaluate target discharge conditions before and after the experiment -- before GYRO simulations are performed -- in order to contribute to the Transport Model Validation Task Force at DIII-D during the experimental campaigns of 2009-2010. This would be an exciting and timely experiment because we will combine targeted testing of the TGLF and GYRO transport models with a fundamental study of turbulence and transport by making use of advanced turbulence diagnostics at DIII-D that are capable of monitoring density fluctuations from the low- to high-k regions of k-space, as well as low-k electron temperature and density fluctuations simultaneously.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Based on a TGLF survey of L-mode beta scans at DIII-D performed in the past, target experimental conditions where the fluctuations and transport are expected to change with beta most strongly will be identified. How much the new conditions must differ from the past reference discharges will depend on the particular needs of the fluctuation diagnostics that will be employed. Doppler Backscattering and high-k backscattering measurements can be optimized with a careful selection of plasma shape and vertical positioning. The BES diagnostic must utilize the 150L beam source, and the beta-scan beam mix must be versatile enough to obtain the highest quality CER/MSE data and BES data throughout the experiment efficiently. The CECE diagnostic, like other ECE diagnostics, will have a radial range of accessibility determined largely by the magnetic field for a given set of measured frequencies. The range of beta that can be accessed experimentally will likely be limited by the fluctuation diagnostic requirements. However, despite these anticipated limitations, it should be noted that the exact range of beta expected in the real experimental conditions can be studied directly with TGLF scans before the experiment is performed. If the given beta range is large enough to produce measurable changes in the measured fluctuation fields and wavenumber ranges, then those theory predictions can then be directly compared with new and unique fluctuation measurements.
Background:	Most beta scans that are performed focus on the scaling of transport with beta, and in order to access values of beta approaching the beta limit, H-mode plasmas are typically studied. Recent results from several different machines indicate that is some cases the scaling in H-mode is favorable (DIII-D, JET, NSTX), but in others unfavorable (JT60-U, ASDEX) [Petty POP 2008]. Few experiments have reported fluctuation measurements during beta scans, but those that have involved only density fluctuations [Petty POP 2008]. There has been no targeted experimental study of multi-field, multi-scale turbulence during a beta can. Past theory work regarding beta scaling of transport was performed using GYRO but using standard base cases for the code rather than actual experimental conditions [Candy POP 2005]. It is now possible to measure high-k density fluctuations with backscattering, intermediate-k density fluctuations with Doppler Backscattering, and low-k density fluctuations with the upgraded BES diagnostic and low-k electron temperature fluctuations with the CECE diagnostic at DIII-D. It is desirable to monitor the beta scaling of both turbulence and transport in a controlled dimensionless parameter beta-scan where the fluctuations can be measured. The fluctuation measurements can be used to constrain and evaluate the quality of the drift-wave model predictions for transport scaling with beta. Also, new information about the fundamental nature of the turbulence and transport will be uncovered via the comparison between theory and many different fluctuation measurements.
Resource Requirements:	1 experimental day, beam sources 30L, 330L, 210R essential for CER and MSE, 150L for BES
Diagnostic Requirements:	Core and tangential Thomson, ECE Michaelson Interferometer, ECE radiometer, Density profile reflectometer, CER, CO2 interferometers, Magnetics, MSE, Doppler backscattering reflectometer, High-k backscattering, BES and CECE
Analysis Requirements:	TGLF ahead of time. After the experiment, EFIT, ONETWO/TRANSP/autoonetwo, TGLF and GYRO
Other Requirements:

Title	144: Can the RMP coils eliminate ELMs from SNs with B x gradB out of the divertor?
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment is the first necessary step f or determining whether the ELM suppression method developed here at DIII-D is compatible with radiating divertor scenarios. Previous DIII-D studies focused on the effect that particle drifts in the SOL/divertor had on fueling, particle pumping, and radiating divertor behavior. We concluded that the most promising (only?) way to successfully employ a radiating divertor in order to reduce heat flux at the divertor targets with a minimal cost to plasma core H-mode properties was to use a SN plasma characterized by having the gradB ion drift directed OUT of the divertor. Presently, however, it is unclear whether ELM suppression in SNs using the RMP coils is attainable, if the gradB ion drift is directed out of the divertor. In this experiment, we investigate if it is possible to suppress ELMs of a SN plasma with the gradB drift direction out of the divertor. Once demonstrating the feasibility of eliminating ELMs under these conditions, we are then ready to examine the behavior of trace injected impurities in an RMP ELM-suppressed environment and ultimately to demonstrate the feasibility of RMP ELM suppression in a radiating divertor environment in subsequent experiments.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The upper SN plasma is maintained in a standard ELMing H-mode regime (i.e., Ip =1.2 MA, Bt = -1.75 T, dRsep = +1.0 cm, q95=4.2, and Pinj = 6 MW). These parameters yielded the best of the radiating divertor results, but the resulting q95 may (or may not) be optimal for ELM suppression with the I-coil. To identify the range in q95 that yields the best prospects for ELM suppression, q95 is reduced during the shot by reducing Bt while the I-coil current is set to maximum. Once this q95 range is identified, choose value of q95 in the middle of that range and run successive shots with increasingly lower I-coil current. This is done to identify the minimum coil current, so as to minimize the perturbing effect of the RMP on the pedestal region.
Background:	Eliminating ELMs from H-mode plasmas using the I-coil approach presents an interesting possibility for resolving the ELM-issue in ITER and future highly powered tokamaks. Yet, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, whereby an impurity gas is injected into a pumped divertor with simultaneous deuterium gas puffing upstream of the divertor, has shown promise as a way to reduce peak power loading at the divertor targets without concomitant degradation of the ELMing H-mode plasma properties [IAEA 2006, PSI2006]. However, in combining the I-coil approach with such puff and pump scenarios while still maintaining favorable H-mode operation, the injected impurity must still be prevented from escaping the divertor and contaminating the main plasma. The most promising radiating divertor scenario involves using a SN divertor with the gradB directed out of the divertor. However, it has not been demonstrated that a SN with the gradB out of the divertor is itself compatible to ELM suppression with the RMP coils. We suspect it is, because SN plasmas run at typically lower density than corresponding plasmas with the gradB drift directed into the divertor. This should result in lower collisionality in the pedestal in the gradB OUT case and thus better ELM suppression. Lower collisionality in the pedestal is helpful in ELM suppression with the RMP coils. As a result, the collisionality in the ion gradB drift direction out of the divertor cases will be lower than generally found in the more standard ion gradB drift direction into the divertor cases, and the former would be expected to tolerate the higher gas puff rates needed to impede the escape of the impurities from the divertor.
Resource Requirements:	Machine time 0.5 day (in forward Bt), I-coil, dome- and upper baffle cryo-pumps cold, minimum 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, dome and upper baffle locations), core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER.
Analysis Requirements:	UEDGE, with ONETWO
Other Requirements:	--

Title	145: Comparison of impurity screening between ELMing and ELM-suppressed plasmas
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment presents the second step in determining whether the ELM suppression method developed here at DIII-D is compatible with radiating divertor scenarios. This experiment, which uses non-perturbing (trace) argon under puff-and-pump scenarios, provides a side-by-side comparison of how well the injected impurity is screened in ELMing H-mode plasmas and in ELM-free H-mode plasma (with I-coil). DIII-D IS UNIQUELY CONFIGURED TO DO THIS EXPERIMENT. We focus on addressing the following questions: (1) Is there a significant difference in the argon accumulation in the plasma core under I-coil operation? (2) How does the exhaust enrichment change between the ELMing- and the ELM-free (I-coil) cases?	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The plasma is maintained in a standard ELMing H-mode regime under the conditions established in the lead-in experiment (e.g., Ip =1.2 MA, Bt = -1.75 T, dRsep = +1.0 cm, and upper SN). A trace amount of argon is injected into the private flux region of the upper divertor, while deuterium plasma flow toward the divertor is enhanced by a combination of deuterium gas injected from the bottom of the vessel and active cryo-pumping from both upper divertor locations. Trace argon is injected at a steady but trace level from t = 3.0 s to 7.0 s. A standard ELMing H-mode regime is established previous to t = 4.5 s of the discharge. At t = 4.5 s, the I-coil is activated and the ELMs are eliminated. This provides a direct comparison of the trace argon behavior between ELMing H-mode and ELM-free H-mode (I-coil) under similar conditions. [Note that the q95 and the coil current selected have been determined from the lead-in experiment: Can the RMP coils eliminate ELMs from SNs with the gradB out of the divertor?]
Background:	Eliminating ELMs from H-mode plasmas using the I-coil approach presents an interesting possibility for resolving the ELM-issue in ITER. Yet, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, whereby an impurity gas is injected into a pumped divertor with simultaneous deuterium gas puffing upstream of the divertor, has shown promise as a way to reduce peak power loading at the divertor targets without concomitant degradation of the ELMing H-mode plasma properties [IAEA 2006, IAEA2008, NF2008]. However, in combining the I-coil approach with such puff and pump scenarios while maintaining favorable H-mode operation, the injected impurity must still be prevented from contaminating the main plasma. This is by no means assured, due to the ergodic nature of the pedestal. In this experiment, we compare the dynamics of impurity screening between ELMing H-mode plasmas and ELM-free H-mode (plus I-coil) plasmas.
Resource Requirements:	Machine time 0.5 day (forward Bt), I-coil, dome- and upper baffle cryo-pumps cold, 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, dome and upper baffle locations), Penning gauge, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER.
Analysis Requirements:	UEDGE, with ONETWO and MIST
Other Requirements:	--

Title	146: Is the Radiating Divertor Scenario Compatible With ELM Suppression?
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment is the third in a series experiments for determining whether the I-coil method of ELM suppression is compatible with radiating divertor (i.e., puff and pump) scenarios. This experiment provides a side-by-side comparison of how a standard ELMing plasma and an ELM-free plasma (with I-coil) respond to a puff-and-pump scenario with a perturbing amount of argon as the injected impurity. DIII-D IS UNIQUELY CONFIGURED TO DO THIS EXPERIMENT. We focus on addressing the following questions: (1) Is there a significant change in the argon accumulation in the plasma core under I-coil operation? (2) How does argon entrainment in the divertor change when the I-coil is activated? (3) How does the ratio of divertor-to-core radiated power change when the I-coil is activated?	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The plasma is maintained in a standard ELMing H-mode regime (e.g., Ip =1.2 MA, Bt = +1.7 T, dRsep = -1.0 cm, and lower SN). This setup is different from the previous two proposed experiments in that we use lower SN plasmas, as opposed to upper SN. This is due to need to have an operating IR camera measuring the changes in heat flux between RMP and non-RMP conditions. Significant deuterium gas puffing, which is needed to raise the SOL plasma flow into the divertor, will raise the density. A previous experiment will have shown that ELMs are suppressed under these conditions by a pre-selected I-coil current: Comparison of impurity screening between ELMing and ELM suppressed plasmas?. Hence, it is possible that adjustments to the I-coil settings may be necessary if the D2 injection rate is changed, as it is in this experiment. After this prep work is done, argon is injected into the private flux region of the lower divertor, while deuterium plasma flow toward the divertor is enhanced by a combination of deuterium gas injected from the top of the vessel and active cryo-pumping from the lower outer divertor location. A radiating divertor plasma in an ELMing H-mode regime is established previous to t = 4.5 s of the discharge. At t = 4.5 s, the I-coil is activated and the ELMs are eliminated. This provides a direct comparison between ELMing H-mode and ELM-free H-mode under similar plasma conditions. The argon injection rate with a fixed D2 injection rate is scanned; then the D2 injection rate with a fixed argon injection rate is scanned. Important measurables are the changes in the radiated power distribution and heat flux values, the accumulation of argon in the core and divertor plasmas, and the density and temperature conditions at both divertor targets.
Background:	Eliminating ELMs from H-mode plasmas using the I-coil approach presents an interesting possibility for resolving the ELM-issue in ITER. Yet, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, whereby an impurity gas is injected into a pumped divertor with simultaneous deuterium gas puffing upstream of the divertor, has shown promise as a way to reduce the peak power loading at the divertor targets without concomitant degradation of the ELMing H-mode plasma properties [IAEA2006, NF2008, and IAEA2008]. However, in combining the I-coil approach with such puff and pump scenarios, it is by no means clear that the injected impurities can be prevented from building up the main plasma as effectively as in the ELMing H-mode cases.
Resource Requirements:	Machine time 0.5 day, I-coil, lower baffle cryo-pumps cold, 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, in the dome and upper baffle locations), Penning gauge, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, lower divertor IR camera, and CER.
Analysis Requirements:	UEDGE, ONETWO, MIST.
Other Requirements:	--

Title	147: Compatibility of ELM suppression with the radiating divertor in the hybrids
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment is the fourth in a series of experiments for determining if the ELM suppression method using the I-coil is compatible with radiating divertor scenarios in the hybrid regime. This experiment provides a side-by-side comparison of how an ELMing hybrid plasma and an ELM-free hybrid plasma with I-coil perform under puff-and-pump scenarios with argon as the injected impurity. DIII-D WOULD BE UNIQUELY CONFIGURED TO DO THIS EXPERIMENT. We focus on addressing the following questions: * Is there a significant change in the argon accumulation in the plasma core under I-coil operation? * How does argon entrainment in the divertor change when the I-coil is activated? * How does the ratio of divertor-to-core radiated power change when the I-coil is activated? This experiment will be attempted only after it has been demonstrated that RMP ELM suppression has been demonstrated in the hybrid regime, where the gradB ion drift is directed OUT of the divertor.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The plasma is maintained in a ELMing hybrid H-mode regime (e.g., Ip =1.2 MA, Bt = +1.7 T, dRsep = -1.0 cm). Argon is injected into the private flux region of the lower divertor, while deuterium plasma flow toward the divertor is enhanced by a combination of D2 gas injected from upstream of the lower divertor targets and active cryo-pumping from the lower divertor location. A radiating divertor plasma in the ELMing hybrid regime is established previous to t = 4.5 s of the discharge. At t = 4.5 s, the I-coil is activated and the ELMs are eliminated. This provides a side-by-side comparison between ELMing hybrid and ELM-free hybrid. The argon injection rate and the deuterium injection rate are separately scanned. Key measurables are the changes in the radiated power distribution and heat flux values, the accumulation of argon in the core and divertor plasmas, and the density and temperature conditions at both divertor targets.
Background:	Plasma operation in the hybrid regime has been put forward as a paradigm for ITER, and would become particularly attractive if ELMing could be eliminated. Suppressing ELMs from hybrid H-mode plasmas using the I-coil approach is an intriguing possibility for resolving the ELM-issue in ITER. However, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, in this case, puff and pump, has shown promise as effective way of reducing peak power loading in ELMing hybrid plasmas, while at the same time maintaining good hybrid properties, e.g., energy confinement time. Our experiment will address whether the puff and pump radiating divertor concept is also as effective for hybrid ELM-free plasmas during I-coil operation. At this juncture, the I-coil method has not been completely successful in eliminating ELMs for plasmas in the hybrid regime. Clearly, this experiment should be attempted after ELM suppression in hybrid is successfully demonstrated.
Resource Requirements:	Machine time 0.5-1.0 day, I-coil, lower baffle cryo-pumps cold, 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, dome and upper baffle locations), Penning gauge, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, lower divertor IR camera, and CER.
Analysis Requirements:	UEDGE, ONETWO, and MIST
Other Requirements:	--

Title	148: What is the nature of the heat flux outside a slot divertor and are particle drifts important?
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	N. Brooks	ITPA Joint Experiment :	No
Description:	We investigate the changes in the heat flux width on particle drifts in the SOL and divertor. In the lower single-null (SN) cases addressed in this experiment, we have found greater interaction of the plasma with the top of the lower baffle extension for the grad-Bt direction out of rather than into the lower divertor, if these particle drifts are important. We determine the differences in the scrapeoff widths in density, temperature, and heat flux for each Bt direction, in attached and detached cases. If the divertor scrape-off widths in ne, Te, and Qp are significantly different in the forward and reverse Bt cases, this will require two different divertor slot widths. This piece of information bears directly on an acceptable design of a slot divertor for ITER and beyond.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The ELMing H-mode plasmas considered for this experiment are a high triangularity, lower SN, with dRsep = 3.0 cm. The toroidal field direction is forward, i.e., the direction of the gradB ion drift is downward. The outer lower cryopump is cold. (1) Use the D2 gas injector from locations upstream of the lower divertor (e.g., GasA). Scan line-averaged density up to and including the H-L back-transition. This range in density will include the partial detachment of the outer divertor separatrix. Reverse the Bt direction and repeat. Important measurables from this experiment are the changes in the heat flux (IR) and particle flux (Langmuir probes) profiles in the slot and along the top of the lower baffle. Upstream electron density, electron temperature, and ion temperature measurements in the pedestal and near-scrapeoff layer (Thomson scattering, midplane movable probe and CER) are essential. Radial sweeps of the outer strike point inward are likely to be needed for a full radial profile in ne, Te, and Qp.
Background:	During the plasma shaping experiments from 2000 we noted that the heat flux on the horizontal section of the upper outer baffle shelf was NOT reduced during high density, radiating upper SN divertor operation, even though the heat flux inside the slot (i.e., between the dome and baffle pump opening) was reduced 3-5 times. Even under high density, detached conditions inside the slot, there was no reduction of heat flux outside the lip of the divertor slot. Formation of a strong radiating region on the horizontal upper shelf with significant heating on the baffle top were observed. Energetic ions far out in the SOL, largely decoupled from the radiatively-cooled electrons, were thought to be responsible for this undiminished heating on the horizontal shelf of the upper outer baffle. Subsequent analysis with the UEDGE code has suggested that particle drifts may have been responsible for the spillover interaction the plasma with the baffle. Reversing the direction of the grad-Bt drift reduced this plasma-baffle interaction. The same kind of interaction was observed again in recent radiating divertor experiments. UEDGE analysis done in 2008 indicates that drifts in the SOL and divertor are playing key roles. Engineering implications of this are significant: If the electrons and ions are decoupled in the SOL outside the slot, then radiative divertor operation has little effect on heat flux outside the slot. Hence, the width of the slot must be widened when the grad-Bt drift direction is out of the divertor. If our supposition about drifts is correct, then a narrower slot can be used if the grad-Bt drift is into the divertor. Should the "slot" divertor concept be applied to future generation, high power tokamaks, our understanding of this result would be particularly important.
Resource Requirements:	Machine time 0.25 day (forward Bt) + 0.25 day (reverse Bt), lower divertor cryo-pumps cold, minimum 5 co-beam sources.
Diagnostic Requirements:	Asdex gauges inside the lower outer baffle, IRTV monitoring the lower divertor, core Thomson scattering, all of the lower divertor fixed Langmuir probes, movable midplane probe, filterscopes, and CER.
Analysis Requirements:	UEDGE with full drifts option turned on.
Other Requirements:	--

Title	149: Realistic test of the compatibility of radiative divertor with AT plasma operation with RMP
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	SSI group	ITPA Joint Experiment :	No
Description:	This study will combine ALL the essential elements for making the first real test of a radiating divertor concept in an AT/hybrid DN (or near-DN) plasma, using realistic high triangular shape and particle exhaust configuration anticipated for high performance tokamaks. PRESENTLY, ONLY DIII-D HAS THE CAPABILITY TO MAKE THIS TEST. Argon is injected into the private flux region near the upper outer divertor separatrix target. Enhanced deuterium plasma flow toward the divertor in the low field SOL is enhanced by a combination of deuterium gas injected upstream of both outer divertor targets and active cryo-pumping from both outer divertor locations; advanced tokamaks will likely not use an inner pump for a variety of reasons that will not be discussed here. Previous experiments have shown that setting the ion gradB drift direction toward the lower divertor and taking dRsep = +0.5 cm will yield the best chances of optimizing the benefits of (near) double-null shape with maintaining a high performance relatively clean of impurity accumulation. An additional benefit of having the ion gradB drift direction out of the divertor is that we are running at a significantly lower core density than if the ion gradB drift direction reversed, so we expect to include RF in our advanced scenarios. Depending on how successful the ELM suppression experiments in ELMing H-mode plasmas have been, we would also attempt to run AT-class plasmas with RMP under radiating divertor conditions.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The plasmas are near-DN AT, and both outer divertor cryo-pumps are at liquid helium temperature. The gradB-ion drift direction is downward. dRsep = +0.5 cm. RF heating is used. This experiment is probably best done in as follows: * First establish the sensitivity of AT plasmas to deuterium gas injection. Scan the deuterium gas puff rate, i.e., establish operational limit to how much D2 gas injection the AT plasma can accommodate before plasma degradation results. Trace argon is injected into the private flux region of the upper divertor. * Scan of the argon injection rate at a reasonable D2 injection rate, using a deuterium gas puff rate that maintains good AT properties as established above. * Repeat under under ELM suppressed conditions. Important measurables from this experiment are the changes in the (poloidal) radiated power distribution and heat flux values, changes in the density and electron temperature at the divertor targets, and the accumulation of argon in the core and divertor plasmas.
Background:	High performance AT in the DN and near-DN configurations are attractive for future power reactor operation due to their high toroidal beta and energy confinement properties. However, for futuristic AT-machines (like ARIES-AT), there can be severe divertor power loading problems. A possible way of reducing excessive power loading at the divertor target(s) is to radiate significant power outside the main plasma, mainly in the divertor (hence, radiative divertor). But the resulting divertor cooling may also lead to a cooling of the upstream (core) plasma, which, in turn, may result in a marked degradation in AT-edge properties (e.g., bootstrap current). The expected increase in the argon presence in the pedestal can also be expected to affect the AT-pedestal adversely. Previous work with radiating divertor H-mode DN plasmas has shown that the balanced DN results in overly rapid accumulation of the seeded impurity (argon) in the core plasma. Two important reasons for this are (1) the relatively easy penetration of an impurity specie from the high field side into the core plasma of the DN and (2) the particle drifts in the scrape-off layer plasma in one of the divertors that always assist in the escape of injected impurities from the divertor region to the vulnerable high field side SOL [NF2008, PSI2008]. 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 shown 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. If previous experiments with ELM suppression with RMP are successful for standard H-mode plasmas, we think that it would be very important to attempt applying the radiating divertor for AT plasmas under ELM-suppressed conditions.
Resource Requirements:	Machine time less than or equal to 1 day, both outer baffle cryo-pumps cold, minimum of five co-beam sources. ECH
Diagnostic Requirements:	Asdex gauges inside the dome and inside both outer baffles, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, upper divertor IRTV, and CER.
Analysis Requirements:	UEDGE, MIST, ONETWO
Other Requirements:	--

Title	150: Further development of co-NBI QH-mode
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	P. Gohil, T.H. Osborne, P.B. Snyder, W.M. Solomon, W.P. West	ITPA Joint Experiment :	No
Description:	Broaden operating range for the co-NBI QH-mode discovered in 2008 through systematic, theory-guided parameter scans	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	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. In order to properly carry out all these scans, three experimental days are requested.
Background:	QH-mode with all co-injection was discovered during serendipitously during the 2008 campaign and a dedicated experiment was performed for one day. We have just barely begun the investigation of the co-NBI QH-mode. The goal of the present proposal is to use our knowledge of counter-NBI QH-mode to find ways to broaden the co-NBI QH-mode operating space so that co-NBI QH-mode can be used more routinely. The parameter scans listed in the experimental approach are based on empirical results from counter-NBI QH-mode combined with theoretical understanding of the QH-mode operating boundaries based on peeling-ballooning mode stability analysis. All QH-mode experiments to date indicate that lowering the target density is beneficial for QH-mode. Theory tells us that more strongly shaped plasmas and increased rotational shear are both beneficial for QH-mode. In addition, edge stability depends on safety factor. Finally, the theory of the EHO says there is a range of outer gaps over which the EHO will exist and modify the particle transport.
Resource Requirements:	This is a multi-day experiment. 6 NBI sources (5 co plus 210LT).
Diagnostic Requirements:	All profile and edge fluctuation diagnostics
Analysis Requirements:
Other Requirements:

Title	151: Optimal Location For Fueling Pumped DN Plasmas
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	SOL Main Ion and Impurity Flows	Presentation time:	Not requested
Co-Author(s):	N. Brooks	ITPA Joint Experiment :	No
Description:	We compare the deuterium fueling efficiency in DN plasmas among low-field side, high-field side, and divertor fueling locations. With simultaneous pumping on both outer divertor legs of a magnetically balanced high-triangularity DN now possible, DIII-D IS UNIQUELY CONFIGURED TO DETERMINE DEFINITIVELY whether it is more effective to fuel a DN plasma from the high field side or from a divertor versus from the standard low-field side location. Related to this question, it is also important to assess any positive or negative effects to fueling from these locations (e.g., how is tauE affected). This experiment can be done in either forward and reverse toroidal field.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A high triangularity, symmetric DN shape is maintained throughout the shot. The upper- and lower outboard cryo-pumps are cold and the inboard (dome) cryo-pump is warm. Conceptually speaking, this approach is made up of four parts: (1) A steady gas puff is injected from the inboard centerpost location into a standard DN H-mode plasma. (2) On the subsequent shot, the same steady gas puff is injected directly into the private flux region of the divertor in the direction of the gradB ion drift (e.g., pfx1 for ion gradB up cases). (3) In the third shot, the same steady gas puff is injected from an outboard location (e.g., GasA). We expect that gas injection from the first two locations to yield higher core densities than from GasA. (4) GasA is now programmed to reach the densities achieved in the centerpost puff and divertor puff cases. We expect that more gas will be required to reach these densities using GasA than was used by the gas injectors from the other two locations. It is best to do this experiment in both forward and reverse toroidal field directions in order to account for any effect that differences in the structure of the upper and lower divertors.
Background:	It is important to determine the most efficient way of fueling DN divertors in order to minimize the amount of deuterium (or tritium) needed to maintain a set density value. We have previously found that unpumped DN H-mode plasmas fuel much faster than comparable SN plasmas (PSI1998). In those experiments fueling was done from the outboard vessel side. When gas puffing was used to fuel DNs, an almost immediate detachment of the inboard divertor legs permitted recycled gas from the outboard legs to escape to the inboard side of the core plasma, suggesting after some analysis (UEDGE) that fueling from the inboard side would be more effective than fueling from the outboard side. Experimentally, IR camera and Langmuir probe data from DN cases indicate that both ne and Te may be significantly lower along the inner SOL than along the outboard SOL and that the scrapeoff widths are also narrower on the inboard SOL (PSI2002). We thus expect easier neutrals penetration of the core from the inboard side and thus more effective fueling. An absence of ELMs along the inboard SOL (NF2003) is another reason to expect effective fueling from the inboard side. More recent data from radiating divertor experiments (2007) indicated that injecting impurity (argon) directly into the divertor pointed to by the gradB ion drift direction was very efficient in fueling the core plasma with impurity ions (argon). Due to the very quick detachment of the inner divertor leg of that divertor, one might expect that deuterium gas injected into the private flux region would also fuel the core plasma just as efficiently as from the high field side.
Resource Requirements:	Machine time 0.3 (forward Bt), both upper and lower outer divertor baffle cryo-pumps cold, minimum 5 co-beam sources.
Diagnostic Requirements:	Asdex gauges inside both outer baffles, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER.
Analysis Requirements:	UEDGE, ONETWO
Other Requirements:	--

Title	152: Effect of particle drifts on deuteron and impurity exhaust in an ITER-like configuration
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	N. Brooks, R. Pitts	ITPA Joint Experiment :	No
Description:	We compare the effectiveness of the puff-and-pump technique for concentrating impurities in the divertor in an ITER-like configuration with that found from the standard DIII-D puff-and-pump approach. ITER has chosen the pumping of neutrals from the private flux region as its mode of particle control. To-date, modeling of the effectiveness of this approach has been done but this modeling presently does not include the effect of particle drifts in the divertor, such as those involving ExB, might have on these results. It is the intention of this experiment to evaluate whether or not neglecting these drifts has an appreciable effect on particle exhaust and detachment, when particle control is based on pumping on the private flux region. Note that detachment is crucial to ITER in maintaining power loading to acceptable values and it also would be affected divertor particle drifts. Hence, these are rather important issues that need to be resolved sooner rather than later.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	An upper SN plasma (dRsep= +3 cm) has its outer strike point on the side of the outer baffle just above the entrance to the outer baffle. The inner strike point is on the centerpost adjacent to the inner pump plenum. Simulate the ITER plasma shape as much as practical. Only the upper inner and upper outer cryo-pumps are at liquid helium temperature. Hence, neither the inner nor the outer strike points are pumped directly; only neutrals in the private flux region are pumped. This configuration is similar to the pumping setup describe in the ITER Technical Basis [G A0 FDR 1 01-07-13 R1.0], except for absence of a semi-permeable dome located between the pump and the private flux region. The experiment is straightforward. The main parameter scan is core density and (trace) impurity accumulation in the core vs deuterium gas puff rate (from the bottom of the vessel). Argon injection is from the private flux region. The range in core density for which the plasmas are attached and detached and the neutral pressure values in the private flux region and inside the upper outer plenum are determined in both forward and reverse Bt cases.
Background:	The geometry of the ITER divertor has been based on simulations obtained using the B2-EIRENE Monte Carlo code and by extrapolation from results of tokamak experiments. The reference configuration for the ITER divertor is a vertical target/baffle with an open private flux region and a dome below the Xpoint. Neutrals that have entered the dome (which is partially open to neutrals from the inner and outer divertor legs) can then be pumped out of the vessel. Also, the magnetic field lines intercept the vertical target at an acute angle for power spreading on a larger wetted area and for easier partial detachment. The calculations used to evaluate pumping expectations were based on models that do not include the cross-field particle drifts that studies at DIII-D have indicated play a very large roles in particle behavior in the divertor. We can simulate the ITER pumping scenario well enough to determine whether or not neglecting drifts are important for ITER. In addition to experimental verification, the progress that UEDGE has made in modeling these types of plasmas with drifts, particularly the Etheta x B and the Er x B drifts, should provide a theoretical underpinning to our measurements. Etheta is the poloidal component of the electric field on an SOL flux surface, arises principally from the poloidal gradient in Te on that flux surface, and is directed toward the divertor target.
Resource Requirements:	Machine time 0.25 (forward Bt) + 0.25 days (reverse-Bt), upper inner- and upper outer divertor cryo-pumps are at liquid helium temperature, minimum 5 co-beam sources.
Diagnostic Requirements:	Asdex gauges inside both outer baffles, Asdex gauge in the upper PFR, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, modified Penning gauge, upper divertor IRTV, and CER.
Analysis Requirements:	UEDGE, ONETWO
Other Requirements:	--

Title	153: Heat flux reduction in double-null plasmas
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	N. Brooks	ITPA Joint Experiment :	No
Description:	We compare the possibility of cooling both targets simultaneously during puff-and-pump operation. We ask the question: In reducing power loading at both outer targets of double-null plasmas, is it sufficient only to inject a seeded impurity into the divertor which has been determined to be better at preventing the escape of these impurities (thereby cooling that divertor) and rely on parallel electron transport to reduce the heat flux at the other divertor? The existence of strong particle drifts in the SOL and divertor have shown that this may not be the case, based on the results of previous studies [IAEA2006, IAEA2008, NF2008, NF1008(2)]. From previous experiments, this better divertor has been determined to be the one for which the ion gradB drift is directed away from. This is an important issue in understanding the viability for applying a puff-and-pump scenario to the DEMO double-null concept. If the results of this experiment and the subsequent UEDGE modeling indicates that impurity injection need only occur in one divertor, this significantly simplifies double-null divertor design. With simultaneous pumping on both outer divertor legs of a magnetically balanced high-triangularity DN now possible, DIII-D IS UNIQUELY CONFIGURED TO MAKE A DEFINITIVE DETERMINATION.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The experiment is straightforward. The argon impurity is injected from only into the private flux region of the divertor that has the ion gradB drift directed away from it. The main parameter scan is argon impurity injection rate versus (1) the electron temperature in the divertor from where the argon is injected and (2) the divertor that has no argon injection. Argon injection is from the private flux region. With Bt in the forward direction (ion gradB drift down), inject argon into the upper divertor, first at trace levels, and then at more perturbing levels on subsequent shots; the deuterium gas injection rate is held constant at 80 torr l/s. Use a radial sweep of the outer strike points to obtain Te and ne profiles under both outer divertor legs. Monitor surface temperature behavior with the IR camera monitoring the lower divertor. On a different day, repeat this sequence, except not that the Bt direction is now reverse and the argon is injected into the private flux region of the lower divertor. Again, monitor surface temperature behavior with the IR camera monitoring the lower divertor.
Background:	Our puff-and-pump studies have shown that it is virtually impossible to balance the radiated power between upper and lower divertors so the heat flux reduction can be balanced between them. We have found that no matter where we inject the argon, it always ends up predominantly going to the divertor that is opposite the ion gradB drift direction; call this divertor A. This divertor is also the stronger radiator of the two divertors. Our studies comparing experiment with UEDGE modeling have shown that particle drifts in the SOL and divertor are responsible for these asymmetries. This begs the question: Is the divertor toward which the ion gradB drift is directed (call this divertor B) cooled as efficiently as the other with the larger radiated power activity? Is electron conduction in the low field side SOL strong enough to cool down divertor B as divertor A is cooled? This experiment will provide the data needed for our UEDGE modeling to resolve this question and provide a basis for predicting future behavior in DIII-D and other tokamaks under puff-and-pump conditions.
Resource Requirements:	Machine time 0.25 (forward Bt) + 0.25 days (reverse-Bt), only the upper outer divertor and lower outer cryo-pumps are at liquid helium temperature, minimum 5 co-beam sources.
Diagnostic Requirements:	Asdex gauges inside both outer baffles, Asdex gauge in the upper PFR, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, upper divertor IRTV, and CER.
Analysis Requirements:	UEDGE, ONETWO
Other Requirements:	--

Title	154: Active impurity removal from the core plasma
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	N. Brooks	ITPA Joint Experiment :	No
Description:	This experiment examines the feasibility of using changes in the magnetic balance to remove impurities from the core of DN plasmas. Changing the magnetic balance from dRsep = 0 to dRsep = + 0.5 cm (with the ion gradB drift downward) reduces pedestal (and line-averaged) density by about one-third. 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 downward. The issue we want to examine here is that do we get preferential loss of core impurities with respect to deuterium when dRsep is changed from 0 to +0.5 cm and then returned to DN. With simultaneous pumping on both outer divertor legs of a magnetically balanced high-triangularity DN now possible, DIII-D IS UNIQUELY CONFIGURED TO MAKE A DEFINITIVE DETERMINATION.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The experiment is straightforward. Start with a DN shape and inject 70 torr l/s of deuterium gas from gasA, starting at t = 2.0 s. The direction of the toroidal field is forward. The argon impurity is injected at 0.3 torr l/s into the private flux region of both divertors. Wait for steady conditions; this should take the discharge out to about t = 4.0 s. Between t = 4.0 s and t = 4.4 s, change dRsep from 0 to +0.5 cm. Hold dRsep = +0.5 cm from t = 4.5 s to 4.7 s. Then return dRsep to 0, starting at t = 4.7 s and finishing up at 5.2 s. Compare argon impurity density before dRsep is changed with the impurity density after dRsep is restored to dRsep = 0. How long does it take the argon density to return to its original value?
Background:	Previous experiments [JNM2003, JNM2008] have demonstrated that plasma density can be regulated by altering magnetic balance. We also have a limited set of data that suggests that impurities already in the core plasma can be preferentially exhausted by using this same regulating method [NF2008]. If indeed we can demonstrate conclusively that we can (actively) exhaust impurities from the core plasma of double-null and near DN plasmas, we have a powerful tool that can significantly improve the prospects of futuristic double-null tokamaks, like DEMO, which may have a serious problem with impurity accumulation in the core, including helium.
Resource Requirements:	Machine time 0.25 (forward Bt), only the upper outer divertor and lower outer cryo-pumps are at liquid helium temperature, minimum 5 co-beam sources.
Diagnostic Requirements:	Asdex gauges inside both outer baffles, Asdex gauge in the upper PFR, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER.
Analysis Requirements:	UEDGE, ONETWO, MIST
Other Requirements:	--

Title	155: Investigate QH-mode without EHO
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Investigate the mechanism which allowed some reverse current shots in the 2008 campaign to operate ELM-free without the EHO	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Goal is to determine whether wall conditions contributed to the edge transport increase which allowed ELM-free operation without an EHO. Start with a reverse current QH-mode plasma using the standard upper single null shape. Utilize cryopumping as is standard for the QH-mode. Fire sufficient number of shots to be sure that the upper portion of the vacuum vessel is clean. Move to balanced double null conditions with all counter-injection and clean the vessel floor. Next, repeat the beam timing and co-counter beam mix used in shots in the 131889-898 range to see if the ELM-free operation without EHO occurs again under clean wall conditions. If it does, then the wall conditions are not the issue. Detailed miniproposal will contain list of other items to investigate if this result occurs. If the ELM-free operation without EHO does not occur, then stop using cryopumping for several shots to deliberately load walls and see if the ELM-free operation without EHO comes back.
Background:	On three days during the 2008 campaign where we operated with reverse current, we saw some discharges which were ELM-free but which didn't have the EHO. These occurred during shots on February 27 and 28, 2008 and June 17, 2008. Only the work on February 28 was part of dedicated QH-mode experiments. Besides reverse current, the only obvious common features of these shots is strong interaction with the lower divertor and some indication of gassy wall conditions. These shots were typically low power (2 to 3) sources, often with a mix of co and counter NBI. These shots remain ELM-free apparently because, even though they are H-mode, the confinement is relatively poor; the edge pedestal seems to be transport limited below the peeling-ballooning limit even without the EHO. The goal of this experiment is to investigate whether the wall conditions are the key which controls the edge transport. If we can find out the mechanism for edge transport increase in these shots and if we can find a way to control this, we will then have another way to control the edge pedestal and control ELMs.
Resource Requirements:	All 7 NBI sources requested.
Diagnostic Requirements:	Profile diagnostics, edge fluctuation diagnostics for EHO studies
Analysis Requirements:
Other Requirements:

Title	156: Joint NSTX/DIII-D poloidal rotation experiment
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	R.E. Bell, W.M. Solomon	ITPA Joint Experiment :	No
Description:	Test aspect ratio dependence of neoclassical theory of poloidal rotation. Determine if deviation from neoclassical could be due to effects of turbulence.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Run matched pairs of discharges on NSTX and D III-D and measure the poloidal rotation of the carbon impurities.
Background:	A predictive understanding of poloidal rotation is needed as part of an overall predictive understanding of transport because the poloidal rotation can contribute to the E x B shear stabilization of turbulence. In addition, the poloidal rotation speed enters in to the calculation of the offset velocity in the theory of neoclassical toroidal viscosity (NTV). The NTV effects are important in determining the effects of nonresonant error fields. Neoclassical theory predicts the poloidal rotation of the plasma ions and electrons. The theory has had some success; the bootstrap current comes from the same order of the theory as the poloidal rotation calculation and the predicted and measured bootstrap currents agree. In addition, recent poloidal rotation measurements on NSTX showagreement with theory. On the other hand, poloidal rotation measurements in standard aspect ratio tokamaks at higher toroidal field (D III-D, JET and TFTR) show significant disagreement with the theory. The goal of the present experiment is to see whether the agreement on NSTX is due to aspect ratio. Because the magnetic pumping effect on poloidal rotation is much stronger at low aspect ratio, it may be that this dominates in NSTX while turbulence effects are strong enough to compete with magnetic pumping at larger aspect ratio. At present, our plan is to match toroidal field on axis; plasma current, plasma shape; and density, temperature and toroidal rotation speed profiles. This will match rho* and beta but will not match collisionality because of the shape-dependent factors which enter in to its definition. Our concept of the best match may evolve as we work on the detailed miniproposal. As part of this development, we will consider the previous NSTX/DIII-D joint experiments. This experiment is the D III-D portion of a joint experiment with NSTX. The NSTX portion has been given one run day in the 2009 campaign on NSTX.
Resource Requirements:	Requires good vacuum conditions in order to operate plasmas at 0.55 T
Diagnostic Requirements:	CER for poloidal rotation measurements. All other profile diagnostics.
Analysis Requirements:	Compare results with NEO and GTC-NEO code predictions.
Other Requirements:	--

Title	157: Measuring the Structure of Tearing Modes
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	Rob La Hay, Troy Carter, Jon Hillesheim	ITPA Joint Experiment :	No
Description:	Measure the localized structure of slowly rotating tearing modes to determine the helical perturbations in density, temperature, and magnetic field. Compare these with theoretical predictions from codes such as NIMROD. This should be done for 2/1 tearing modes and possibly for 3/2 tearing modes as well. The key is to reduce the tearing mode rotation frequency to less than ~500 Hz using a mix of co and counter NBI.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use plasmas in the hybrid regime to produce long lasting discharges with a large saturated 2/1 tearing mode (and 3/2 tearing mode as well). Trade off counter-injection beams for co-injection beams to bring the 2/1 mode rotation frequency to ~500 Hz. Use feedback control of the 2/1 mode frequency if the PCS has been upgraded to that feature. Use continuous 30LT and 330 beams to collect MSE and CER data at high time resolution. May need repeat shots to collect all of the CER data, and also may need repeat shots for the BES data.
Background:	Previously the co-injection of beams into DIII-D resulted in a rotation frequency of ~10 kHz for the 2/1 tearing mode. At this high frequency, only localized data from ECE and BES has been found to be useful in measuring the tearing mode structure. (Attempts have been made to collect fast MSE data, but to date no quantitative information has resulted from this.) If the tearing mode rotation frequency can be reduced to ~500 Hz using a more balanced beam injection, then localized data also can be collected by the CER diagnostic (~4 kHz rate) and MSE diagnostic (2 kHz rate, perhaps this could be increased to 4 kHz). This would allow the helical perturbations in electron density, electron temperature, ion temperature, ion toroidal and poloidal velocities, and local magnetic fields to be obtained. Possibly other diagnostic could be added to this mix. This experiment should result in an excellent dataset for which to compare with theoretical predictions to improve (or verify) our understanding of tearing modes.
Resource Requirements:	NBI: All 7 sources needed (not simultaneously).
Diagnostic Requirements:	MSE, BES, CER, ECE, fast camera, DBS.
Analysis Requirements:
Other Requirements:

Title	158: FW coupling and electron heating in QH mode
Name:	Pinsker pinsker@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	C.C. Petty, F.W. Baity, A. Nagy, J.C. Hosea	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 either with Resonant Magnetic Perturbations (RMPs) or in intrinsically ELM-free regimes such as the QH mode. So far, there have not been any antenna loading measurements in QH-mode, and certainly there have been no attempts to measure FW core heating efficiency in this regime. Here we propose to investigate those issues. At minimum, we need to measure the antenna loading in QH-mode by powering the FW antennas at low, non-pertubative levels (a few tens of kilowatts each) - this can be done, and should be done, in piggyback on ongoing QH-mode experiments. A new feature that is possible is to study the antenna loading at extremely high time resolution - sampling rate of up to 1 GHz - to look for coherent features that would correlate with the EHO. Again, this can be done at very low (non-perturbative) power levels. Next, if the ongoing attempts to expand the QH-mode operating space are able to produce a QH-mode at reduced beam power, so that the maximum plausible ratio of FW power to beam power is not negligible compared to unity, we should attempt to measure the incremental core heating efficiency of the FW in this regime. This line of experiment is analogous to the proposed continuation of the FW experiments in the other DIII-D ELM-free regime (the RMP-stabilized regime).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In the first part of the experiment, which should be piggybacked on the ongoing QH-mode work, antenna loading is measured at low (non-perturbative) power levels. In the dedicated portion of the proposed work,which probably should be attempted only if the QH-mode regime is successfully enlarged to permit robust QH-mode operation at reduced NBI power, the FW power from all three antenna systems is raised to the maximum level, limited by the peak system voltage (inversely proportional to the antenna loading obtained). FW power modulation can be employed to improve the signal-to-noise ratio of the incremental core heating efficiency measurement. Clearly reduced density QH-modes would also be greatly beneficial for observing the incremental FW heating; this is also a direction in which the QH-mode work is going.
Background:	The proposed work would be the first attempt to study ICRF heating in the QH-mode regime. In order to gain understanding of the edge losses in ICRF heating regimes, it is important to study FW heating under as many distinct edge conditions as possible. In view of the possible unacceptability of ELMs in ITER, the most ITER-relevant regimes to study are the ELM-free ones.
Resource Requirements:	Piggyback time is requested during QH-mode experiments; one dedicated day is requested if a QH-mode regime with reduced beam power and/or with reduced density is found in those experiments.
Diagnostic Requirements:	In addition to the usual complement of profile diagnostics, the edge reflectometer adjacent to the 285/300 antenna would be highly desirable.
Analysis Requirements:	--
Other Requirements:	--

Title	159: Expand the high li, betaN >4 operating regime through instability avoidance and higher heating power
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	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 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. Comparison with ONETWO indicate significant anomalous fast ion loss, possibly a result of semicontinuous 1/1 mode activity.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	160: Maintaining high li at high betaN using RMP and near-axis current drive
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Make high li, betaN >4 discharges more stationary by reducing the rate of decrease of li through replacement of ohmic current near the axis with ECCD and by using RMP to reduce the H-mode pedestal density in order to reduce the edge bootstrap current. Results would be used to determine what would be required to produce a true, steady-state, high li, high betaT discharge.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	There are two primary parts to this experiment. First, the deposition profile of the ECCD would be varied in order to determine its effectiveness in replacing the core ohmic current. A accompanying goal would be to determine if q(0) can be raised slightly in order to avoid 1/1 activity. It is unlikely that, even with six gyrotrons, the ohmic current can be completely replaced in the core. However, we can obtain scaling information in order to compare with models and determine how much external current drive would be necessary to make a stationary high li discharge. In the other part of the experiment, the RMP fields would be used to reduce the H-mode pedestal pressure and, particularly, the density. Supposedly there is a window for ELM stabilization with q just above 7 which matches the best high li discharge from 2008. However, ELM stabilization isn't necessarily needed, and the RMP fields are reported to have an effect on the pedestal height even away from the resonance required to stabilize ELMs. The best discharges in 2008 were strongly overdriven with noninductive current which may have contributed to the rate of decrease of li. Some exploration of the balance between total plasma current and the noninductive current would be done in order to test the effect on the li decrease.
Background:	In the fNI >1 discharges produced in 2008, li decreases slowly as the trapped ohmic current decays in the discharge core and the bootstrap current builds in the H-mode pedestal. In the best performance discharge, because Ip was relatively low (q95 about 7), and the noninductively driven current exceeded the total programmed current, the edge surface voltage was driven negative.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	161: Cross-cutting experiment to address LH transition physics and formation of edge shear layer
Name:	Rhodes rhodes@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Cross-cutting experiment addressing: ITER scenario access, Hydrogen and Helium Plasmas, Pedestal Structure, Transport The formation of the edge velocity shear layer and transport reduction layer during the H-mode formation as well as that leading up to and during ELMs would be addressed using high time and space resolution diagnostics. This will consist of a multi-scale and multi-field turbulence approach utilizing the unique array of DIII-D diagnostics. In addition, fast profile (10 microsecond) reflectometer system as well as the heterodyne ECE will be used to follow fast profile changes. This is a cross-cutting experiment that would provide new and unique information on the transition behavior in a variety of plasmas and would address important issues regarding ITER H-mode access; edge transport and barrier formation, and ELM physics.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	NBI and ECH L-H transitions would be utilized to vary how the transition is formed. ELM initiation and behavior will also be be a focus of this effort (note that RMP suppression of ELMs can also be studied). Plasmas will be designed for optimum diagnostic access (a crucial step). Deuterium (and hydrodgen/helium if available) working gases will be used. Hydrogen/helium plasmas provide a unique opportunity to provide ITER relevant/priority information on the H-mode transition physics. Even if hydrogen/helium are not available new and unique information can still be obtained. This data includes high time/space resolution fluctuation flow velocity; low, intermediate and high-k density fluctuations; low k temperature fluctuations; radial correlation lengths, and potentially information on zero-mean frequency zonal flows.
Background:	Due to the wide array of fluctuation and profile diagnostic systems DIII-D is now in a unique position to more fully probe the physics associated with H-mode transitions and ELM behavior. A recent paper [Happel, et al, EPL, 84 (2008) 65001] proposes a two-step process in the edge shear formation: 1) a seeding mechanism linked to gradients and 2) an amplification process in which edge shearing rates and fluctuations self-organize near marginal stability. These observations can be verified and extended.
Resource Requirements:	Machine Time: TBD Number of Neutral Beam Sources: required ECH: required
Diagnostic Requirements:	All fluctuation and profile diagnostics.
Analysis Requirements:	GYRO, TGLF, XPTOR, ELITE, plus others.
Other Requirements:

Title	162: Dependence of confinement and stability on toroidal rotation in high li discharges
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Produce a high li discharge that runs without beta collapse at significant betaN (for example, about 4). Evaluate the dependence of confinement on toroidal rotation. Evaluate the effect of the toroidal rotation velocity on the stability limit, both the maximum attainable betaN and the no-wall limit as measured with MHD spectroscopy.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce a high betaN discharge similar to those produced in 2008. Add counter injection beams. In order to reduce the rotation to low values, it will probably be necessary to operate at less than the maximum betaN.
Background:	The normalized confinement in high li discharges seems to increase as the beam power increases, possibly indicating a dependence of confinement on toroidal rotation velocity. On the other hand, experiments in the 1990s also indicated that the enhanced confinement at higher li depends on the poloidal field strength profile. It is essential to understand the confinement at high li under low rotation conditions as might be expected in a reactor. Also, a motivation for the high li scenario is that high betaN can be obtained in the absence of wall stabilization. The stability at low rotation in DIII-D discharges should be tested to determine the role of the wall in stabilization. Discharges in 2008 had phases with betaN below the no-wall limit and phases with betaN above the limit.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	163: Extend the fNI = 1 phase in the steady-state scenario and maintain q_min above 1.5
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	In 2010, use the improved pulse length capability of the tokamak and the neutral beams to extend the 100% noninductive fraction duration in steady-state scenario discharges. Possibly operate at higher betaN with the increased neutral beam power available with the return of the 30 right source. Place emphasis on maintaining q_min at a constant value above 1.5. Determine the best ECCD aiming, combining the needs for 2/1 stability and maintaining the q profile.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	--
Background:	The best steady-state, q_min >1.5, scenario discharges produced in 2008 had the fNI = 1 phase limited by the available toroidal field duration and neutral beam energy. Additional capability to extend the pulse is expected in 2010. Also, q_min tended to drift below 1.5 in these discharges.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	164: Routinely use feedback control of q_min during the Ip ramp up
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Begin using feedback control of q_min during the formation phase of steady-state scenario discharges. This should improve reproducibility of the target q profile. To accomplish this, we will need to become comfortable with using a new discharge start up scheme rather than exactly reproducing the startup of previous discharges.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	--
Background:	Feedback control of q_min during the Ip ramp up has been developed to the point that it is ready to test in routine use. We need to get some experience to determine whether the feedback scheme is robust enough. If it is, we should have improved reproducibility of the target q profile for the high beta, fNI = 1 phase of the discharge.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	165: Addressing the physics of ECH density pump-out via multi-scale/field turbulence measurements
Name:	Rhodes rhodes@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Cross-cutting experiment addressing ITER scenario, ITER demo, heating and current drive, and transport. The reduction of density during ECH will be examined using DIII-Dâ??s full array of multi-scale / multi-field turbulence diagnostics as well as its profile diagnostics. The observed increase in temperature fluctuations, reduction of rotation, and apparent constant low k density fluctuation levels are thought to be keys to this phenomenon.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Utilize ECH in Ohmic, L, and H-mode plasmas to modify the local drive terms. Select target plasmas based upon previous runs that showed either a strong or no effect.
Background:	ECH heating often leads to so-called density pump-out where the plasma density is reduced while the electron temperature is increased. Experiments on ASDEX-U attribute this to the excitation of TEM and associated thermodiffusion. On DIII-D multiple examples are known where the effect of ECH on turbulence during density pump-out is strong. For example, a QDB discharge where both ne and Ti drop, Te increases during ECH (Casper, APS â??05) and there is a concurrent increase in fluctuations at 2 cm-1 (possibly ITG type range). A second example is a H-mode where ne drops, Te increases, low k ( 1 cm-1 ) decreases, and intermediate k ( 7 cm-1) increases during the ECH (Wong APS â??05). Examples also exist with no density pumpout for comparison. We will use ECH as a tool to probe the turbulence response to changes in local drive terms, Te/Ti, Ln, L_T, etc. Results will be compared to gyrokinetic turbulence simulations (e.g. GYRO, GS2, ..).
Resource Requirements:	ECH and NBI required
Diagnostic Requirements:	all fluctuation and profile diagnostics
Analysis Requirements:	GYRO, TGLF, XPTOR, plus others.
Other Requirements:

Title	166: Clarification of minimum EC-driven current for complete NTM stabilization
Name:	Isayama isayama.akihiko@qst.go.jp	Affiliation:	QST
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	R.J. La Haye, E. Strait	ITPA Joint Experiment :	Yes
Description:	Clarify the minimum EC wave power to completely stabilize an m/n=2/1 NTM and obtain data with different beta value, different ECCD scheme (with/without modulation)	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	First stabilize 2/1 NTM and then decrease EC power shot by shot until the mode cannot be stabilized completely. In the experiments, real-time control with the 'search and suppress' scheme will be useful. Once a set of data is obtained, similar data set is taken at different beta regime by changing NB power. First priority of this experiment is to obtain data using unmodulated ECCD. If EC wave power can be modulated in synchronization with NTM rotation, similar data using the modulated ECCD will be taken.
Background:	Minimum EC wave power (~ EC-driven current) for complete stabilization of an m/n=2/1 NTM was investigated in JT-60U. The ratio of EC-driven current density to bootstrap current density at the mode rational surface was about 0.4. In DIII-D, on the other hand, the ratio was about 2.7 (C.C. Petty et al., Nucl. Fusion 44 (2004) 243). One reason for the difference may be attributed to narrow ECCD deposition width with respect to the marginal island width below which an NTM spontaneous decays. Another reason is the difference in the marginal island width: the width is twice the ion banana width in DIII-D while it is ~5 times in JT-60U, which may make stabilization in JT-60U easier. Difference in beta value is another candidate. The aim of this experiment is to compare the JT-60U result with DIII-D experiments. Similar data in a different beta regime will clarify the dependence of the threshold EC-driven current on beta value (if exists). In addition, since the investigation on the threshold current for modulated ECCD has not been done in any other devices, results from this experiment will be a unique one.
Resource Requirements:	ECRF with enough power to stabilize a 2/1 NTM. NB power enough to destabilize a 2/1 NTM. Discharge scenario to obtain a stationary 2/1 NTM is needed; first follow the previous experiments
Diagnostic Requirements:	ECE & Thomson scattering for Te(r); CER for Ti(r); Thomson scattering for ne(r); MSE for q(r); Mirnov & saddle coils for dB/dt
Analysis Requirements:	Codes to calculate EC-driven current and bootstrap current, REVIEW, NEWSPEC
Other Requirements:	'Search and Suppress' feedback system, Modulated ECCD (if possible; not necessary)

Title	167: Assess the bootstrap current fraction as a function of the q profile in steady-state discharges
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	Assess Steady-State Current Profiles for Optimum Performance	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	This experiment is directed toward DIII-D milestone 170. In this experiment, q_min and q95 will be varied in discharges similar to the steady-state scenario in order to determine the parameter regime that really produces the most bootstrap current. The density profile will be varied to the extent possible. BetaN will be varied to assess the changes in bootstrap current density. BetaT (relevant to fusion power density) will be compared. Single null and double null divertor shapes will be compared. Changes in stability and the ability to maintain a stationary current profile will also be documented. Confinement will be monitored in order to determine the parameter range that produces values of H98 >1 and H89 >2.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Data exist for a variety of discharge conditions, but not for a systematic scan. For instance, at low q_min we have the hybrid and high li scenarios, at q_min near 1.5 is the standard steady-state scenario. There is some limited data at higher q_min in steady-state type discharges. These previous discharges will be used as a guide for choosing discharge parameters. In addition, there are many discharges operated in a close-to-double-null shape and there are ITER shape steady-state scenario discharges from 2008 which look promising for high bootstrap current fraction. So, the two discharge shapes should both be studied to look for the impact on density profiles, in particular. Since the density profile has a strong effect on the bootstrap current density, the line average density should be varied and the match to the cryopump should also be changed in order to get a scan in density profiles.
Background:	Because the bootstrap current density is inversely proportional to the local poloidal field strength, the most common approach to producing high bootstrap current density is to operate with elevated values of q (both q_min and q95). Beta is maximized in order to increase the bootstrap current through increased pressure gradient. However, because the confinement depends on the q profile and confinement tends to be better at lower values of q, it isn't obvious that the highest bootstrap current fraction will necessarily be produced with higher values of q. It may be that at lower q, with a given heating power higher pressure gradient can be obtained. In addition, the achievable bootstrap current density will depend on the achieved temperature and density profiles which will likely depend on the q profile, the discharge shape, and the amount of divertor cryopumping. A uniform density profile will possibly result in lower bootstrap current density even at high beta.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	168: Detailed measurement of NTM island structure
Name:	Isayama isayama.akihiko@qst.go.jp	Affiliation:	QST
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	R.J. La Haye, M. Austin	ITPA Joint Experiment :	No
Description:	Measure structure of magnetic island associated wtih NTMs using ECE (and CER)	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Measure the evolution of island structure with ECE (and CER if possible) during NTM stabilization with ECCD. If data were taken in past experiments, no additional data are needed.
Background:	Structure of rotating magnetic islands associated with 3/2 and 2/1 NTMs during ECCD was measured in JT-60U using ECE radiometer. In general, profile of electron temperature perturbation in the island region becomes an M-shaped one, in which the two peaks correspond to the island separatrix, and the local minimum point corresponds to the island center. During NTM stabilization with ECCD, the mode amplitude at the inner half of the island quickly decays, and that at the outer half of the island decays slowly. The distance between the two peaks, which corresponds to the full island width, continuously decays like magnetic perturbation evolution (A. Isayama et al., Phys. Plasmas 12 (2005) 056117). While the behavior is not fully understood, such evolution was observed in both 3/2 and 2/1 NTMs. Detailed island measurement in DIII-D will clarify island evolution physics such as whether such asymmetry is observed also in DIII-D, and how much is the time scale of the decay etc. Measurement of the island structure of ion temperature is also informative.
Resource Requirements:	NB and ECRF to obtain and stabilize NTMs
Diagnostic Requirements:	ECE, CER, Mirnov/saddle coils
Analysis Requirements:	REVIEW, NEWSPEC, etc
Other Requirements:	--

Title	169: Multi-scale, multi-field turbulence in reversed shear and ITB plasmas
Name:	Rhodes rhodes@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Measurements utilizing DIII-Dâ??s unique array of multi-scale and multi-field fluctuation diagnostics will be carried out in the core of reversed shear and ITB plasmas. New and unique information will be obtained on the behavior of the fields in normal/reversed magnetic shear and ITB conditions. Importantly, information on potential sources of electron thermal transport in these conditions will be obtained.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Early beam injection will be used to control the magnetic shear profile. The formation of ITB plasmas will be of particular interest.
Background:	Reversed shear and internal transport barrier (ITB) plasmas often exhibit core regions of flat temperature and density profiles where the turbulent drive is expected to be low and yet the transport is high. We will probe the turbulence and transport response in these regions utilizing DIII-Dâ??s unique measurement capability for profiles of low, intermediate, and high k density fluctuations, low k temperature, and fast turbulence flow velocity measurments. A particular focus will be on radial distribution and response of intermediate and high-k turbulence and low k temperature turbulence and their possible contribution to electron thermal transport. The effect of magnetic shear will also be investigated since in these plasmas q naturally varies in time. Since transport remains anomalous in the core of these plasmas these measurements will provide a broad wavenumber range with which to challenge non-linear turbulence simulations. All available fluctuation diagnostics together with CER and MSE will be utilized.
Resource Requirements:	ECH and NBI required. 1 day expt.
Diagnostic Requirements:	all fluctuation and profile diagnostics.
Analysis Requirements:	GYRO, TGLF, XPTOR, ELITE, plus others. Use of TGLF/XPTOR/similar is expected in order to facilitate the design of the experiment
Other Requirements:

Title	170: Beta limit and bootstrap current fraction in ITER steady-state scenario discharges
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	Yes
Description:	Study the performance of steady-state scenario discharges in the ITER discharge shape in order to establish the physics basis and optimum operating scenario for the ITER steady-state mission. Determine the beta limit and bootstrap current density as a function of q_min. Increase the plasma current over what has been used previously to push q95 down to 5. 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. A portion of this experiment addresses ITPA IOS group high-priority experiment 3.1.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce the 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
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. 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 operated at betaN = 3.7 without a 2/1 NTM. The discharge shape that was used doesn't quite match the intended ITER scaled shape
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	171: Improve the ability to produce the exact, scaled ITER discharge shape in DIII-D
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Find a new patch panel configuration that will allow the lower, outer squareness in the DIII-D "scaled ITER shape" to better match the actual squareness in the ITER shape.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Try a different patch panel configuration, probably with return current in the F4B coil. This will probably require more than just a couple of shots because it is hard to predict the effect of changes on the patch panel on the currents during the full discharge evolution. Typically there are unexpected excursions in current on various coils that need to be dealt with. During 2008, a fair amount of time was spent trying to overcome this type of problem. If possible, modeling using TokSys would be useful before the experiment in order to test the effect of various patch panel configurations.
Background:	During 2008, the ITER demonstration discharge task force attempted to produce a discharge shape in DIII-D that was an exact match to the ITER shape scaled by a factor of 3.7. The shapes produced were a good match except in the lower, outer squareness. The shape mismatch was a consequence of the VFI constraint on the DIII-D patch panel configuration. Two different patch panel configurations were tried. Both had too much current in the F7B coil, resulting in lower outer squareness which was too small. At least one alternative patch panel configuration exists which might improve the squareness match.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	172: Establish the requirements for 2/1 tearing mode stabilization by broadly deposited ECCD
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine the operating conditions that allow the 2/1 tearing mode to be avoided in high betaN discharges through the use of EC current driven over a broad radial region. Issues to be addressed: required EC power, deposition profile, relation to the location of the q = 2 surface, range of q profiles for which the ECCD is effective, relationship to the pressure gradient profile. Determine the physics basis for the 2/1 stability at high betaN.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce a discharge from 2007/2008 which was stable to 2/1 with broadly deposited ECCD and demonstrate clearly that the effect is a result of the ECCD. Perturb parameters about this discharge (as listed in "description") to establish the conditions for 2/1 stability. Use these results to design an experiment to establish the physics mechanism.
Background:	During 2007 and 2008 experiments in steady-state scenario discharges, it was found that betaN up to 3.7 could be maintained for 2 s without a 2/1 mode occurring if sufficient ECCD power was deposited approximately in the region 0.25
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	173: Turbulence and transport dependence on rho_* utilizing working gas species
Name:	Rhodes rhodes@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Use helium and hydrogen plasmas with matched Te, Ti, etc. to compare fluctuation and transport properties. Compare to predictions of GYRO/TGLF/other.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use helium and hydrogen plasmas with matched Te, Ti values. Target discharge will be an upper single null plasma to allow higher auxiliary power without H-mode. Te, Ti ratios and profiles will be adjusted using ECH, co and counter neutral beams. The rotation and radial electric fields will also be monitored and matched as closely as possible. Use will be made of all the available turbulence and other appropriate diagnostics.
Background:	There is a well-known dependence of the plasma confinement on the mass of the working gas, varying approximately as (Mass)^0.5. This variation will be addressed by comparing helium and hydrogen plasmas with Ti, Te matched as closely as possible. This should give approximately the same rho_i,e but of course different working gas masses. Theoretical turbulence expectations appear to depend upon the value of rho_i,e rather than the mass itself so the naïve expectation is that the plasmas should be very similar. Comparison of low, intermediate, high k turbulence, transport analysis, etc. will be made between the plasmas and theoretical expectations (analytic, numerical �??GYRO, GS2,�?�) .
Resource Requirements:	ECH and NBI required
Diagnostic Requirements:	all fluctuation and profile diagnostics.
Analysis Requirements:	GYRO, TGLF, XPTOR, plus others. Use of TGLF/XPTOR/similar is expected in order to facilitate the design of the experiment.
Other Requirements:	--

Title	174: Maximize the high noninductive fraction duration in the ITER shape discharges
Name:	Ferron ferron@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Make use of increased pulse length capability in DIII-D (longer toroidal field duration, more neutral beam energy) to push the duration of fNI = 1 discharges in the ITER shape toward 2 tauR.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In 2010, DIII-D will have an additional neutral beam available, some of the beams will be able to operate with longer pulse length, the toroidal field pulse length capability will be increased, and there will be six gyrotrons available. This provides increased hardware capability that should allow the extension of the pulse length duration of discharges with moderate values of betaN. Take advantage of this capability and the good performance of the 2008 ITER steady-state scenario discharges to study longer duration fNI = 1.
Background:	In the steady-state scenario ITER demonstration discharges produced in 2008, fNI = 1 was obtained with relatively low betaN = 3.1. This discharge is a good candidate for the study of longer duration with fNI = 1.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	175: High-power modulated ECH to assess power deposition and heat transport models
Name:	Austin austin@fusion.gat.com	Affiliation:	U of Texas, Austin
Research Area:	Transport	Presentation time:	Requested
Co-Author(s):	Ken Gentle, Dmitry Meyerson, Marco Zerbini	ITPA Joint Experiment :	No
Description:	High-power modulated ECH experiments are needed for two purposes. The first is to resolve the issue of the power deposition profile. Recent analysis [1] has shown a discrepancy in the power deposition determined from time changes of ECE-measured Te for which the simplest explanation is that TORAY is wrong. Higher ECH power that would give more accurate values of dT/dt at switch-on would clarify the issue. However, if the calculated power deposition is wrong, more work will be required to distinguish between different possible explanations, for example lack of mode purity from the antenna or unexpected mode conversion in the plasma. Regardless, a 50% uncertainty in the power deposited at resonance would have serious repercussions for future applications of ECH such as ITER and should be resolved. The second purpose is to measure the heat flux as a function of gradient [Q(dT/dr)] over a broader range of gradient and for different operating regimes. This would provide better data for evaluating heat transport models.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply high-power modulated ECH into sawtooth-free L-mode discharges at several beam powers with scans of ECH power and deposition radius. Repeat for other suitable conditions: RMP, quiescent H-mode, etc. Use a variety of modulating frequencies and waveforms (sinusoidal, triangular, square) chosen based on the focus of the experiment (for example, high frequencies are better for measuring the deposition profile while slower modulation is more revealing for transport model comparison). Post-experiment, different techniques for analyzing the data would be used, such as standard FFT, wavelet, and cross-correlation [2].
Background:	The high accuracy, spatial, and time resolution of electron temperature measurements by ECE has enabled a conceptual advance in our characterization of electron thermal transport. Coupled with modulated ECH, experiments can move beyond approximating transport with an effective thermal conductivity to a direct measurement of the thermal flux as a function of temperature gradient, at least over a certain range of gradients. This advance has numerous advantages. It is provides a more natural basis for comparison with theoretical simulations, which compute the flux for a fixed gradient, and it can provide a uniquely sensitive test of the simulations by comparing the Q(dT/dr) functions from theory and experiment. Simple application of the method, however, has produced paradoxical results. The Q(dT/dr) trajectory around a cycle of ECH modulation was double-valued â??strong hysteresis. That could be interpreted as nonlocality (cf. â??cold-pulsedâ?? experiments), but of a special form. The hysteresis arises entirely from jumps at ECH switching, exactly as if the power deposition were wrong. In fact, the hysteresis disappears if the local power deposition is inferred from the initial dT/dt instead of from calculations of absorption profiles. The resulting ECH power deposition profile remains strong peaked at the resonance position, but it has a broad, weak tail that accounts for half the total input power. The revised analysis accounts for the full measured ECH power input and gives well-behaved Q(dT/dr) for all radii. [1] K.W. Gentle, et al, Phys. Plasma 13, 012311 (2006). [2] M. Zerbini, et al, Plasma Phys. Control. Fusion 41 931 (1999).
Resource Requirements:	1-day experiment Minimum 4 gyrotrons, 2 neutral beams
Diagnostic Requirements:	Essential: ECE 40-chan radiometer Fluctuation diagnostic set highly desirable
Analysis Requirements:	TORAY
Other Requirements:

Title	176: Turbulence Reduction via Nonresonant Magnetic Field-Driven Velocity Shear
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	A. Garofalo, General Atomics	ITPA Joint Experiment :	No
Description:	Exploit the nonresonant magnetic field-driven offset velocity and inherent velocity shear to control, reduce and suppress turbulence to optimize the energy confinement improvement in otherwise non-rotating discharges. Characterizing the confinement changes through direct measurement of turbulence will aid in understanding the effect and its extrapolability to large scale experiments like ITER. This may be especially important since large non-axisymmetric fields are envisioned to suppress ELMs, and the beam-driven torque will be relatively small.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Counter Ip (to maximize available counter-NBI to offset intrinsic co-current rotation and achieve near zero initial rotation), counter-BT high-beta discharges. Reversing BT as well maintains the standard field helicity that will allow for turbulence measurements with BES. q95~5 to insure that n=3 will be non-resonant. The NTV is predicted to scale with dTi/dr, which suggests that the offset rotation effect will be maximized at high-beta. Increasing the I-coil to 7 kA should further increase the NTV torque. Scan the initial rotation velocity to discern changes in rotation with NTV. Document fluctuation characteristics in all cases.
Background:	Neoclassical Toroidal Viscosity offset rotation, predicted by A. Cole (PRL, 2007), has been experimentally observed in DIII-D (Garofalo, PRL, 2008). Application of non-resonant nonaxisymmetric static magnetic fields via I-coils configured to apply an n=3 perturbation results in a counter-current rotation with a magnitude comparable to but opposite in direction to the ion diamagnetic velocity. In an initially low-rotation high-beta discharge, this results in a counter-current acceleration that is observed to increase global energy confinement, suggesting that turbulence suppression may be active. The magnitude of the offset rotation is relatively large (~100 krad/s on axis) and, more importantly, the resulting Er shear of the offset rotation profile appears large enough to affect and suppress turbulence.
Resource Requirements:	I-coils with up to 7KA DC current
Diagnostic Requirements:	All fluctuation and usual profile diagnostics required.
Analysis Requirements:
Other Requirements:

Title	177: Integrated Diagnostics Tool for Non-Axisymmetric MHD Mode Internal Structure
Name:	Bogatu nbogatu@far-tech.com	Affiliation:	FAR-TECH, Inc.
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Y. In (FAR-TECH, Inc.), J.S. Kim (FAR-TECH, Inc.), M. Okabayashi (PPPL), E.J. Strait (GA), M.J. Lanctot (Columbia University)	ITPA Joint Experiment :	No
Description:	There is a serious difficulty in distinguishing n=1 resistive wall modes (RWM) and neoclassical tearing mode (NTM), both being performance-limiting instabilities. Regardless of the subtleties between those MHDs, the identification of the internal structures of non-axisymmetric (n=1) MHDs based on multiple diagnostics will greatly help us to understand the physics of each separate MHD mode. The multi-faceted physics involved motivates us to mobilize and compare several diagnostics, rather than one particular diagnostic that is most sensitive to the relevant MHD, and which may be finally used for accurate early detection for feedback and control. Moreover, by revealing the phenomena inter-correlations, the analysis with the integrated multi-diagnostic tool could lead to new and more efficient feedback and control techniques. To understand the multi-physics of non-axisymmetric MHDs we need several diagnostics able to measure inside the plasma column. Finally, for comprehensive feedback control, only the most sensitive one might be used. The proposed diagnostics tool, integrating the non-magnetic diagnostics SXR arrays, ECE, CER, and MSE, should be able to provide real-time correlated information on the internal structure of the non-axisymmetric MHD modes, such as magnetic flux surfaces displacement, perturbations of electron and ion temperatures, rotation velocity, and internal magnetic field. The most important global challenge in achieving a comprehensive feedback control of non-axisymmetric MHD modes is to find the essential links between several parameters of very different nature (e.g. magnetic field perturbation, in Gauss, and mode displacement, in cm) for which the corresponding diagnostics have the accuracy of spatial and temporal resolution required for real-time identification. More specific challenges are the following: 1. Build a coherent understanding of the multi-faceted physics by multi-diagnostic analysis of the internal structure of MHDs (e.g. surfaces displacement from SXR) and external structure of magnetic field from magnetic sensors 2. Develop or improve the diagnostics data analysis methods/techniques able to provide an early and accurate identification of MHDs (e.g. n=1 RWM, by enhancing the �??contrast�?�, so that low amplitude structures are detected) 3. Establish the necessary and sufficient parameter sets and the link(s) among them 4. Test the identification capability of the integrated diagnostic in critical cases (e.g., onset of the global 2/1 NTM, recently found in DIII-D to depend on plasma rotation near the no-wall limit, which may add to the complexity of identification and controlling of plasma rotation-dependent RWM).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The physics on which the integrated diagnostics tool for the internal structure should be based requires the determination and understanding the correlation of the following different-nature parameters: 1. Internal structure of SXR relative displacement for n=1 MHD mode evolution from the three 12-chord SXR toroidal arrays (SXR TAs), at 45, 165, and 195 deg., w/o contrast enhancing technique (CET); in addition to the SXR TAs, to increase the accuracy, the SXR poloidal arrays (PAs) camera (2x16 viewing chords) at 90 R+1 can by used together with DISRAD at 210 deg., as both have similar components. 2. Internal m-structure from the SXR PAs at 90 deg. (two cameras, each with two 16 viewing chords at one poloidal location) with spatial Fourier transform 3. Overlay the EFIT-reconstructed positions of qmin, q~2, and q~3 surfaces to correlate SXR displacement profile with q-surface position 4. Te perturbation from ECE to see if magnetic islands appear and to determine the displacement of Te profile convected by the MHD mode; determine also the �??absolute�?� displacement from SXR data 5. Internal magnetic field perturbation from direct measurement by MSE diagnostic with 120 deg. toroidal separation to estimate the flux surface displacement 6. Ion temperature profile perturbation from CER to infer the displacement of Ti profile convected by MHD mode 7. Toroidal velocity profile evolution from CERFIT with atomic physics correction 8. Magnetic field perturbation outside plasma (at mid-plane) from magnetic sensors
Background:	We have been developing the CET for SXR measurements and demonstrated it for detection of the low-amplitude RWM on DIII-D tokamak [1]. We tested the potential of SXR CET to provide the details of the internal structure information on different types of discharges. For an ELM-free high-rotation non-resonant braking discharge [2] the SXR relative displacement, from chord-by-chord signal difference and radial gradient, indicated an early internal structure of significantly large amplitude of non-axisymmetric n=1 component, while mid-plane poloidal magnetic sensors do not detect it. Later on we observed a strong correlation between the magnetically detected mode amplitude and the SXR-based plasma displacement. It indicated that RWM is a global mode whose structure extends from the plasma core to the wall and even beyond it. Another observation was that when RWM is weakly damped, SXR relative displacement of the surface q=2 is negative (inward pushing) while that of the qmin surface is positive (outward pulling). All the results showed that m/n=2/1 structure appears to play the dominant role. For the discharges where the balanced neutral beam injectors (NBI) were used (2006 DIII-D run-campaign) our analysis [3] showed that there was strong non-axisymmetric n=1 core MHD activity, which correlates with and also even with the radiated power in the lower divertor, as measured by the bolometers, indicating a multi-faceted physics of the global mode. In qmin >1 discharges, we frequently found that there is an early non-axisymmetric SXR n=1 precursor within the core which appears before RWM starts to grow. An analyzed discharge from 2008 DIII-D campaign has a non-rotating n=1 RWM during ECCD with a very slow growth (tg~ 100 ms). We compared the amplitude and phase from magnetic sensors and the SXR TAs. An early development of an internal n=1 structure is detected by SXR TAs on the central viewing chord which is propagating outwardly and synchronizes with the amplitude and phase from magnetic sensors. The correlation between magnetic and SXR signals indicate a close link between the evolutions of two different nature parameters: magnetic field outside plasma and internal (spatial) displacement within the core. As SXR shows always the capability of early detection of RWM, understanding its basic relationship with magnetic signals is crucial for incorporating SXR into any feedback system for MHD mode control and suppression. [1] I. N. Bogatu et al., Resistive Wall Mode Identification by Contrast Enhancing Technique of Soft X-Ray Measurements on DIII-D, Rev. Sci. Instrum. 75 (9), 2832 (2004) [2] I. N. Bogatu et al., Resistive Wall Mode Internal Structure Identification by Soft X-Ray Contrast Enhancing Technique, 47th APS DPP: APS Meeting: Bull. Am. Phys. Soc. 50, 79 (2005) [3] I.N. Bogatu et al., Investigation of Resistive Wall Mode Internal Structure, 48th APS DPP: APS Meeting: Bull. Am. Phys. Soc. 51, 347 (2006)
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Piggy-back on discharges for �??Physics of Non-Axisymmetric Field Effects in Support of ITER�?�

Title	178: Test neutral penetration model for pedestal density width
Name:	Groebner groebner@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Requested
Co-Author(s):	A. Leonard, T. Osborne, J. Callen, L. Owen, T. Rognlien, W. Stacey, J. Canik	ITPA Joint Experiment :	No
Description:	Obtain a complete set of SOL and divertor data for use in testing several coupled plasma fluid/neutral transport models. Obtain these data in plasmas with fairly wide density pedestal profiles. Perform analysis with edge plasma/neutral transport codes to determine if modeled neutral penetration length is comparable to density width.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Develop a long steady-state ELMing H-mode discharge with a wide pedestal density. Do this by operating at moderately high triangularity and beta. Optimize shape to improve data from edge diagnostics used in fuelling studies. Perform edge sweeps to increase spatial resolution of these diagnostics, as appropriate. Obtain data at conditions of low and high density.
Background:	A simple analytic model, based on coupled particle and neutral transport equations, predicts that the pedestal density width is about equal to a characteristic neutral penetration depth. This model has had mixed success in comparisons to experiment. For instance, one study from a coupled plasma/neutral transport model found that the neutral penetration was significantly smaller than the density width. This is evidence that the simple neutral penetration idea is insufficient to explain all of the physics. This proposal is to obtain a complete data set for edge fuelling studies that would be used in a number of edge models, which have different capabilities. This data set would both allow for comparisons of the models and for testing the density width physics. This work would be follow on work for a group of modelers who have been benchmarking codes from data obtained from previous DIII-D discharges. These codes include UEDGE, SOLPS, GTNEUT coupled to an 2-point plasma model and ONETWO. For this work to proceed, much improved sets of SOL and divertor data are required in the benchmarking.
Resource Requirements:	DIII-D tokamak. 5-6 neutral beams. LSN discharges, probably in hybrid regime. Pumping. Ip ~ 1-1.5 MA, Bt ~ 2.0 T.
Diagnostic Requirements:	TS, CER, CO2, bolometers, filterscopes, pressure gauges, divertor probes, IR divertor camera, profile reflectometer, MDS system, divertor TS
Analysis Requirements:	Reduce data to obtain profiles of Te, ne, Ti and Zeff through pedestal into SOL. Obtain divertor heat and particle fluxes, te and ne. Obtain line integrated D_alpha data. Use codes noted in background section to obtain best solutions of edge plasma and neutral transport, consistent with all of the data. Compare the results from the models and compare the neutral penetration lengths to density pedestal width.
Other Requirements:	--

Title	179: Test EPED1 pedestal model at high density
Name:	Groebner groebner@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Requested
Co-Author(s):	C. Maggi, T. Leonard, T. Osborne, P. Snyder	ITPA Joint Experiment :	No
Description:	Perform discharges at high density, comparable to typical densities in ASDEX-Upgrade, ideally in the AUG shape. Perform beta scans by increase of power. Determine if EPED1 model predicts observed variations of pedestal width and height. Also, determine of density width increases with beta.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Develop hybrid operation in AUG shape, similar to work performed in 2006. Develop steady operation at densities comparable to those used in AUG. Perform beta scan on shot by shot basis by varying heating power. Obtain high quality pedestal data by running long steady state phases and performing separatrix sweeps. Obtain similar data at same density in standard DIII-D hybrid shape.
Background:	The EPED1 model has successfully described DIII-D pedestal widths and heights over an order of magnitude variation in pedestal height. These data were obtained for moderate to low pedestal densities with nped < 0.5 of the Greenwald density. In these experiments, both the Te and ne widths were observed to increase with the pedestal poloidal beta. As a further test of this model, we propose to obtain the required data at a significantly higher density, comparable to the densities run in AUG. An additional motivation for this test is the observation in AUG that the density width is independent of pedestal beta, whereas the Te width increases with pedestal beta. The observations of ne width behavior are different than in DIII-D. We need to determine if DIII-D shows the same behavior as AUG at high density or the difference between the machines is related to something else.
Resource Requirements:	DIII-D tokamak. 6-7 NB sources. LSN operation at moderate and high triangularity. Ip ~ 1-1.5 MA, Bt~ 1.5-2.1 T.
Diagnostic Requirements:	TS, CER, CO2, bolometers, profile reflectometer, ECE, MSE
Analysis Requirements:	Obtain best fits to edge profiles to obtain te and ne pedestal widths and pedestal height for each discharge. Perform EPED1 analysis to predict pedestal width and height. Compare EPED1 predictions with measurements. Examine behavior of pedestal newid and tewid with beta.
Other Requirements:	--

Title	180: ELM Pacing with n=0 I-coil Configuration
Name:	West west@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	Al Hyatt	ITPA Joint Experiment :	No
Description:	The axisymmetric, n=0 configuration would used to pulse the squareness of the plasma and provide a sudden reduction of the pedestal stability, leading to the inducement of an ELM. We know from previous experiments by John Ferron followed by more experiments by Tony Leonard, as well as the theoretical studies using ELITE by Phil Snyder that squareness can be a very sensitive knob for tuning edge stability. Both upper and lower rows of the I-coil, or either row alone, could be useful and should be tried (first modeling, then experiment). The n=0 configuration implies that the sense of the current is the same for each window frame in a particular row. If both rows are used, an â??odd parityâ?? configuration, i.e. the sense of the current in the upper row is opposite to that in the lower row, is most likely the best for pulsing the squareness. Reversing the sense of the currents would change the direction of change (increase or decrease) of the squareness.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Both upper and lower rows of the I-coil, or either row alone, could be useful and should be tried (first modeling, then experiment). The n=0 configuration implies that the sense of the current is the same for each window frame in a particular row. If both rows are used, an â??odd parityâ?? configuration, i.e. the sense of the current in the upper row is opposite to that in the lower row, is most likely the best for pulsing the squareness. Reversing the sense of the currents would change the direction of change (increase or decrease) of the squareness. Repetitive pulses would provide ELM pacing
Background:	ELM control is essential for ITER. The RMP ELM suppression is the leading technique for ITER, but seems to have narrow windows of resonance with edge q. Pacing with pellet injection is also promising, but inherently requires fueling, which may put excessive loads on ITER pumping.
Resource Requirements:	H-mode with nbi and proper shaping,
Diagnostic Requirements:	high resolution pedestal profile diagnostics, fast magnetics, edge reflectometry
Analysis Requirements:	Fast Wdia, pedestal profiles,
Other Requirements:	I-coil in n=0 configuration, spas on i-coil for fast pulsing, may want to trigger n=0 pulse with feedback control on beta.

Title	181: Turbulence and Transport Scaling with Te/Ti in Low Rotation L & H Modes
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	A. White, C. Holland. T. Rhodes, L. Schmitz, G. Wang	ITPA Joint Experiment :	No
Description:	Examine the dependence of turbulence over a wide wavenumber range, spatial range and multiple fields, as well as particle, momentum and thermal transport on the electron to ion temperature ratio, Te/Ti, shown theoretically and experimentally to be a critical dimensionless parameter for transport. The electron temperature profile will be self-similar as Te/Ti is varied, while other dimensionless parameters (rho*, nu*, q, beta) are kept nearly constant.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish low power L-mode target plasma, with 2 balanced injection sources (minimize rotation), providing diagnosing beam for BES and CER, MSE (150R, 1/2-30L,330L), inner wall limited discharges to maintain L-mode conditions. - Scan Te at constant Ti w/off-axis ECH heating, maintaining self-similar electron temperature profile shapes by varying ECH deposition location as necessary -Maintain constant Ti temperature at low rotation through adjustment of co/counter beam sources and power (previously in co-rotation discharges, rotation dropped significantly as Te increased during previous experiment) - Scan Ti at constant Te (via balanced NBI injection) - Scan Te and Ti at constant Beta (trade off ECH and NBI) as feasible - Increase density if necessary to further equilibrate temperatures - Obtain fluctuation data with all fluctuation diagnostics (expanded high-sensitivity BES, multi-wavenumber FIR, ECE, microwave back-scattering, correlation reflectometry, PCI, probes) - Examine particle diffusivity with helium puffs - Repeat experiment in H-mode plasmas. Contrasting L-mode and H-mode turbulence response to Te/Ti will be a central aspect of this experiment.Establish low power L-mode target plasma, with 2 balanced injection sources (minimize rotation), providing diagnosing beam for BES and CER, MSE (150R, 1/2-30L,330L), inner wall limited discharges to maintain L-mode conditions. - Scan Te at constant Ti w/off-axis ECH heating, maintaining self-similar electron temperature profile shapes by varying ECH deposition location as necessary -Maintain constant Ti temperature at low rotation through adjustment of co/counter beam sources and power (previously in co-rotation discharges, rotation dropped significantly as Te increased during previous experiment) - Scan Ti at constant Te (via balanced NBI injection) - Scan Te and Ti at constant Beta (trade off ECH and NBI) as feasible - Increase density if necessary to further equilibrate temperatures - Obtain fluctuation data with all fluctuation diagnostics (expanded high-sensitivity BES, multi-wavenumber FIR, ECE, microwave back-scattering, correlation reflectometry, PCI, probes) - Examine particle diffusivity with helium puffs - Repeat experiment in H-mode plasmas
Background:	Numerous experiments have demonstrated that as Te/Ti, which is typically less than one in beam-heated experiments, increases towards unity or above, transport increases, in general agreement with ITG-based turbulent transport theory, through modification of the critical gradient. This experiment will seek to quantify the underlying turbulence mechanism giving rise to this dependence. Previous experiments showed that there was little or no increase in low-wavenumber turbulence as Te is increased at constant Ti in co-current rotating L-mode discharges, however the experiments were not conclusive because of uncontrolled variation in the rotation and resulting ExB shear. This experiment will seek to investigate this issue in low-rotation discharges, now feasible with the balanced beam-injection, and also to investigate behavior in both L-mode and H-mode plasmas, as well as to employ the wide array of new and upgraded diagnostics. H-modes (hybrid) showed a more significant increase in broadband low-k density fluctuations with increasing Te/Ti, consistent with the increase in thermal transport. These experiments will be simulated with TGLF and GYRO as reasonable to contribute to the effort to compare and challenges the codes and ultimately validate turbulence and transport simulations.
Resource Requirements:	Maximum ECH power (5 or more gyrotrons) to increase Te/Ti and balanced NBI
Diagnostic Requirements:	BES, FIR, DBS, CECE, PCI, high-k backscatter
Analysis Requirements:	TGLF/GYRO
Other Requirements:

Title	182: Dynamic Error Field Correction for ITER
Name:	Strait strait@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The goal of this experiment is to develop the use of feedback-controlled dynamic error field correction in low beta plasmas. Such plasmas can develop non-rotating n=1 modes at low density, driven by error fields. As the mode becomes marginally stable it should amplify the error field, just as RWMs do at high beta, providing input for the feedback system. The experiment will also provide data for comparison to IPEC modeling of the plasma response.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Use a low-density ohmic plasma with all error correction turned off. Ramp the density down in the current flattop phase until a locked mode appears. On the next shot, turn on "slow" feedback (time constant 10-50 ms) early in the discharge, using C-coils and SPAs. We expect to see that the feedback calls for something close to the "standard" error correction currents, and thereby avoids the locked mode. Repeat the process in an ELMing H-mode plasma, but well below the no-wall beta limit.
Background:	Error field measurement and correction techniques are a key physics R&D need, as communicated by the ITER IO. ITER is likely to have little or no capability to measure field errors directly. The field errors may also change as the coils and their support structures are cooled in the cryostat. Therefore "dynamic error field correction" in real time is likely to be required for all scenarios in ITER. However, this technique has so far been applied only in plasmas at or above the no-wall limit. The proposed experiment would provide the first demonstration of the technique in a wider range of operating conditions. Recent results from IPEC modeling show that plasma response is important even at low beta, suggesting that feedback-controlled â??dynamic error field correctionâ?? may be a viable approach.
Resource Requirements:	C-coils and SPAs
Diagnostic Requirements:
Analysis Requirements:	A good compensation matrix (C-matrix) in the control system to remove direct coupling of the sensors to the toroidal field, F-coils, etc. is a prerequisite. This is likely to be more critical than for dynamic error field correction at high beta, since the plasma response may be smaller.
Other Requirements:

Title	183: Comparison of resonant and non-resonant n=1 error fields
Name:	Strait strait@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to distinguish the effects of resonant and non-resonant error fields. The results are intended to confirm the hypothesis that a single helical component of error field dominates the plasma response - and hence that error correction in ITER need not be perfectly matched to the poloidal structure of the error field. The results will also be used to benchmark IPEC and MARS-F models for the plasma response to varying error field configurations. In this experiment, the mixture of resonant and non-resonant n=1 fields is varied by changing the upper-lower I-coil phase difference. Resonant and non-resonant behavior will be distinguished by the amplitude of the plasma response and the dependence of the braking torque on plasma rotation. This is an extension of the 2008 experiment on resonant n=1 braking.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply a rotating n=1 field of increasing amplitude, and measure the plasma response and braking of rotation. Repeat with two or more values of neutral beam torque, in order to obtain the braking torque over a wide range of plasma rotation frequency. Carry out these measurements for upper-lower I-coil phase differences varying from 60 to 300 degrees. *** A second possible approach is to vary the I-coil phase difference continuously, i.e., apply a fixed n=1 field with one row of I-coils while slowly rotating the toroidal phase of the other row. This technique can go through the full 360 degree range of I-coil phasing several times in a single shot. The variation of measured plasma response and magnetic braking during this cycle can be separated to obtain the resonant and non-resonant contributions. Then scan the NBI torque shot-to-shot in order to make the measurement at different values of plasma rotation. *** The beam torque scan at different I-coil phasing is the most significant feature that was not present in the 2008 experiment. The wider range of I-coil phasing will also allow more complete comparison with IPEC and MARS-F.
Background:	Two key results from the 2008 experiment were: (1) Preliminary indication of resonant and non-resonant torque behavior at different values of rotation (in a comparison of shots from different days). (2) Strongly varying plasma response for upper-lower I-coil phase differences of 120, 180, 240 degrees, in good agreement with MARS-F calculation. - A primary aim of the proposed experiment is to confirm the behavior of braking torque using several combinations of resonant/non-resonant fields, under the same plasma conditions. - The proposed experiment will extend the MARS-F comparison to a wider range of I-coil phasing.
Resource Requirements:	Both 210 sources for counter-injection. I-coils and SPAs.
Diagnostic Requirements:	--
Analysis Requirements:	MARS-F and IPEC calculations of plasma response.
Other Requirements:	--

Title	184: Dependence of halo currents on plasma current and q95
Name:	Strait strait@fusion.gat.com	Affiliation:	GA
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The goal of this experiment is to measure the scaling of halo currents and vessel forces with plasma current and q95. In particular, it will test â?? under consistent plasma conditions â?? the hypothesis that vessel forces scale as Ip-squared, and the empirical observation that vessel forces peak at intermediate values of q95. The results will provide data for empirical scalings and for comparison to modeling of halo currents and vessel forces.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	In a low beta, lower single-null discharge, develop a reproducible disruption with VDE (perhaps by low-intensity gas puffing from the Medusa valve). Perform a shot-to-shot scan of q95 from ~6 to ~2.2, by increasing plasma current at constant toroidal field. If the vessel displacement reaches a predetermined limit, reduce the plasma current and toroidal field together at constant q95. Then resume the upward plasma current scan. At q95~3.2, vary the plasma current and toroidal field together. This scan should cover a factor of 2 in plasma current, subject to the limits of acceptable vessel motion. The â??safeâ?? amplitude of vessel motion must be determined ahead of time, in consultation with the operations group.
Background:	Prediction of vertical and horizontal vessel forces is a key physics R&D need, as communicated by the ITER IO. The upper envelope of the DIII-D disruption data may support the hypothesis that vessel forces scale as Ip-squared. It also seems to show that vessel forces peak at intermediate values of q95, not the lowest values as might be expected from Ip-squared scaling. However, these discharges represent widely varying plasma conditions. The proposed experiment will test these dependences in a more systematic way.
Resource Requirements:	The Medusa gas valve is required.
Diagnostic Requirements:	Measurements of halo currents, and vertical and horizontal vessel forces, are essential for this experiment.
Analysis Requirements:	A â??safeâ?? amplitude of vessel motion must be determined ahead of time, in consultation with the operations group. The experiment will be designed not to exceed this level.
Other Requirements:

Title	185: Study of edge plasma turbulence with RMP
Name:	Krasheninnikov none	Affiliation:	UCSD
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	ELM suppression is important for ITER operation. One of the promising mechanism of ELM suppression is RMP. After ELMs are suppressed by RMP, the power-load on the target and plasma-main chamber wall interactions are determined by remaining edge plasma turbulence (which actually plays crucial role in reduction pedestal parameters and ELM suppression). However, there is practically no experimental data even of the main features of edge plasma turbulence in divertor tokamak with RMP (the results from MAST are just coming). Meanwhile there are some theoretical results (e.g. B. Scott and UCSD/Lodestar) which may be compared with experimental data, which may help the understanding of the physics of ELM suppression.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	186: Physics of sawtooth suppression in hybrid discharge
Name:	Suzuki none	Affiliation:	JAEA
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):	C. Petty	ITPA Joint Experiment :	No
Description:	To investigate what mechanism (anomalous resistivity or some clamping mechanism) suppresses the sawtooth in hybrid discharge	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Toroidal field ramp down, in order to change edge safety factor q_95 from ~6 to ~4, is performed in hybrid discharge with or without m/n=3/2 NTM (q_min,target~1.1-1.2). With the decrease in q_95 from 6 to 4, it is expected that qmin becomes below 1 if 3/2 NTM does not exist. We want to check how the q (or current profile) changes if 3/2 NTM exists. Does the q_min still stays above unity or the q_min becomes below 1 and sawtooth appears, after the Bt ramp down. If q_min stays above unity, there may be some (unknown) mechanism to clamp q_min > 1 in hybrid discharge, or the anomalous resistivity is a strong function of q around q~1. If q_min becomes below unity, this result may reject the clamping mechanism (and may suggest anomalous resistivity, such as hyper-resistivity).
Background:
Resource Requirements:	NBI: up to 7 sources required to produce hybrid
Diagnostic Requirements:	MSE and diagnostics for current profile calculation
Analysis Requirements:
Other Requirements:

Title	187: SOL and divertor behaviour in hydrogen
Name:	Pitts richard.pitts@iter.org	Affiliation:	ITER Organization
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):	C. Lasnier, J. A. Boedo, D. Rudakov, P. West, J. Watkins, P. C. Stangeby, A. Leonard	ITPA Joint Experiment :	No
Description:	Improve confidence in the prediction of deuterium plasma exhaust properties in ITER, based on an initial Hydrogen plasma phase.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Repeat, the experiments performed in the 2008 power width scaling sessions of 2008 (07/23 and 08/05) in high purity hydrogen discharges. Emphasis on heat flux profiles, for which very little information is available in hydrogen, and on divertor detachment behaviour. Experiments in both L and H-mode required to provide scope for scaling studies and to allow high quality turbulence studies (deep probe plunges at lower power). A particularly important aspect, addressed by this proposal, is the physics of near-SOL heat transport, reflected strongly in the divertor target power e-folding widths. Available scalings contain virtually no information from hydrogen plasmas (for example the recent JET scalings for ï?¬p which are based almost entirely on deuterium plasmas, with some helium data). Discharges should be performed in LSN for IRTV and with forward BT. These experiments also provide an ideal opportunity to study pedestal density width sensitivity to isotopic mass. They might also be used, if the experiments are performed carefully, to investigate the unrelated, but important for ITER question of isotopic tailoring by using short D phases at the end of some pulses and monitoring the uptake of hydrogen fuelling in subsequent pulses.
Background:	The early, non-active phases of ITER operation will be run in hydrogen and/or helium. Although all tokamaks have run hydrogen plasmas at one time or another, we have very little systematic information on divertor and SOL characteristics in pure hydrogen (the situation is somewhat better, though far from sufficient in helium). Whilst this state of affairs prevails, confidence in the prediction of operation in deuterium in ITER will not be improved by an early phase in hydrogen. This is a consequence of the different mass dependences involved in neutral atom dynamics, impurity production rates and ion transport timescales. By running at lower current (7.5 MA), ITER anticipates H-mode operation in hydrogen, even if the heating power is limited. Studying the behaviour of H-mode power exhaust in hydrogen is thus equally important. The physics of near SOL heat transport is still far from understood, with a variety of possible scalings proposed for divertor target power flux widths. Good quality measurements in hydrogen, repeating those in deuterium, are currently missing in the tokamak database.
Resource Requirements:	A maximum of two days of experiments (in L and then H-mode) to match the set of reference discharges performed in deuterium as part of the power width experiments performed on 07/23 and 08/05. Those deuterium discharges are presently being analysed. Completion of this analysis will help to decide on the key discharges to aim for in hydrogen to provide matched pairs.
Diagnostic Requirements:	All lower divertor particle and heat flux diagnostics, all SOL profile and turbulence measurements. Best possible pedestal coverage. Monitoring of degree of purity of hydrogen discharges
Analysis Requirements:
Other Requirements:

Title	188: Rotation effect on high beta_p small ELM regimes
Name:	Oyama oyama.naoyuki@jaea.go.jp	Affiliation:	Japan Atomic Energy Agency
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	A.Leonard, T. Osborne, Y. Kamada, H. Urano, K. Kamiya	ITPA Joint Experiment :	Yes
Description:	Final goals of this inter-machine experiment are to establish H-mode plasma operation with small ELMs and extension of operational regimes for these small ELM regimes to ITER relevant plasmas. Especially, the effect of the toroidal rotation should be confirmed by using the unique capability of rotation control with co- and counter- NBs in DIII-D and JT-60U. (ITPA inter-machine experiment PEP-17)	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A search through the extensive DIII-D database of high beta_p discharges did not reveal grassy ELM regime so far. In JT-60U, on the other hand, grassy ELMs with ELM frequency of ~400 Hz was obtained with zero plasma rotation in high q (q95>6) and high delta(d>0.5) plasmas. Therefore, following experiments will be proposed in DIII-D. 1) The first attempt will be a beta_p scan to reproduce small ELMs in DIII-D following the grassy ELM recipe with balanced NB injection (balanced NB injection + more co NBIs to increase beta) in high q and high delta plasmas. 2) Once we obtain small ELM similar to grassy ELM in DIII-D, parametric dependence of important parameters (delta , q95, beta_p and VT) to enter the grassy ELM regime will be compared between DIII-D and JT-60U.
Background:	Small ELM regimes have been intensely studied in several devices. The grassy ELM regime discovered by JT-60U is a candidate for a small ELM operation in ITER to combine tolerable ELM energy losses at low pedestal collisionality, and no degradation of pedestal pressure. Since the grassy-like ELMs has been observed in AUG and JET following the grassy ELM prescription developed in JT-60U, we expect that similar small ELM regime can be established also in DIII-D.
Resource Requirements:	1 day experiment NBI: co, bal, counter injections
Diagnostic Requirements:	Standard diagnostics for edge pedestal and ELM study
Analysis Requirements:
Other Requirements:

Title	189: Controllability of pedestal and ELM characteristics by edge ECH/ECCD
Name:	Oyama oyama.naoyuki@jaea.go.jp	Affiliation:	Japan Atomic Energy Agency
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T. Osborne, A.Leonard, Y. Kamada, H. Urano, K. Kamiya	ITPA Joint Experiment :	Yes
Description:	In order to establish new ELM control tool for ITER, the controllability of ELM and pedestal characteristics will be investigated with physics understanding of the mechanism of ELM control by edge ECH/ECCD.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This inter-machine experiment is aimed to confirm whether edge ECH/ECCD/LHCD can be a possible ELM control tool in ITER or not. To achieve it, following experiments will be proposed. (1) Reproduce similar heating condition (power density, radial profile of heating power) used in JT-60U to decrease the ELM size in type I ELMy H-mode with low pedestal collisionality. (2) Survey the required condition such as the dependence of heating power, the location of heating (HFS/LFS) and heating/CD methods (ECH/ECCD). (3) Evaluate capabilities of ECH/ECCD to modify ELM size (frequency), pedestal structure and plasma confinement. Based on experimental results obtained in this inter-machine experiment, we will discuss recommendation for the specification of EC system in ITER with respect to the effective ELM control.
Background:	In ITPA-CC meeting held in June, one specific heating and current drive issue related to the ELM control was explained. The steering range of upper launcher in ITER can be adjusted, if experiments indicate that this is worthwhile. In ASDEX-Upgrade, ELM frequency was locked at ~100 Hz using modulated ECH/ECCD in H-mode plasma with fELM~150 Hz. On C-Mod a variable-phase LH launcher can be used for off-axis current drive and electron heating, with the capability of changing both the radial deposition of LH power and the relative degree of resulting LHCD and electron heating. LHCD has already been used to modify the pedestal structure in EDA H-modes, and will be used in future experiments in ELMy H-modes. In JT-60U, resent experimental results show that fELM can be increased by localized edge ECH/ECCD within the pedestal. It is noted that edge ECH/ECCD near the top of the plasma at high-field side is effective, while no clear effect has been observed in the case of edge ECH/ECCD near the top of the plasma at low-field side [2]. This observation might suggest that the effect of trapped electron for effective ECCD, although the evaluated ECCD was only 5% of the evaluated bootstrap current at the pedestal. These experimental observations of edge ECH/ECCD may suggest a possibility of ELM control.
Resource Requirements:	1 day experiment NBI+ECH(~3MW for 2-3sec)
Diagnostic Requirements:	Standard diagnostics for edge pedestal and ELM study
Analysis Requirements:
Other Requirements:

Title	190: Edge FW power loss versus edge density
Name:	Hosea jhosea@pppl.gov	Affiliation:	PPPL
Research Area:	Heating & Current Drive	Presentation time:	Requested
Co-Author(s):	F.W. Baity, B. LeBlanc, C. Petty, C.K. Phillips, R. Pinsker, P.M. Ryan, G. Taylor, et al.	ITPA Joint Experiment :	No
Description:	Measurements of FW heating of a relatively high density L-mode plasma at 60 MHz and 116 MHz on DIII-D (R. Pinsker et al, Nuclear Fusion 2006) indicate that the heating efficiency for the 60 MHz case is substantially reduced from that obtained at 60 MHz for the low density L-mode plasma case. Furthermore, the 116 MHz heating efficiency is further reduced relative to the 60 MHz case for the higher density plasma. (The reduced effective power provided to the core plasma must be used to calculate the FW power going into energetic beam ions, electron heating, and current drive.) These results point to edge loss enhancement at the higher density due to the onset density for perpendicular propagation being moved toward the antenna/wall as is observed on NSTX (J. Hosea et al., Phys. Plasmas 2008). We are proposing an experiment on DIII-D to determine edge power deposition and possible causes vs edge density, especially for power lost to SOL in vicinity of the antenna and subsequently deposited outside the plasma. In particular, we would like to focus in on the possible loss process associated with surface fast wave propagation, which is not a multi-pass process, to help determine if excessive FW power is being lost to the SOL in the vicinity of the antenna region and ultimately to the outer divertor SOL region. This experiment is needed to set the limits on edge density for efficient DIII-D heating and to provide guidelines on the acceptable maximum density that can be used for enhancing coupling with gas puffing near the antenna. Scans of edge density in L-mode and H-mode, via increasing the core plasma density and via gas puffing, will serve to highlight the characteristic of the loss enhancement and potentially the properties of the loss process itself. In particular, it is proposed that measurements of heating at the outer SOL region of the divertor be investigated and compared to the surface power loss to determine the importance of the SOL loss mechanism.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	A density scan of heating efficiency will be made first in L-mode discharges by increasing the overall plasma density in steps of order 1 X 10^19 m^-3, starting at a core density of ~ 1.5 X 10^19 m^-3, for plasma conditions similar to those used in the paper by Pinsker et al. referenced in the Description section. Both 90 MHz sources and the 60 MHz source will be employed at the maximum powers available. The scan will be made for 90° antenna phasing (although if run time were to be available, it would be very informative to perform a second L-mode scan at 180° antenna phasing and /or the 60 MHz source connected to the 0° or 180° antenna). It is very important that we diagnose the possible linking of the antenna SOL regions to possible divertor region heating with visible and IR cameras, divertor region thermocouples and probes, and any other applicable diagnostics available. A second density scan of heating efficiency will be made for a suitable H-mode regime with the gap desired for the H-mode gas puffing experiments planned for the experiment proposed by R. Pinsker et al. to enhance coupling. Edge density will be increased by increasing the edge density and also by edge gas injection. Both of these scans will be used to set experimental limits for edge density that permit enhanced coupling with gas puffing while maintaining efficient core heating. It is important for these experiments that we have a reasonably accurate measurement of edge density in the scrape off layer in the mid-plane, and thus, the installation of the ORNL reflectometer would be beneficial for these experiments.
Background:	See Description. Also, RF power injected at the antenna that does not reach the core plasma has been under investigation for some time even for the fundamental and second harmonic heating regimes planned for ITER. In fact, outer divertor plate erosion on ASDEX (Noterdame et al., FED, 1990) and Alcator C-Mod (Wukitch et al., AIP CP933, 2007) point to significant RF power causing bombardment at the intersection of the SOL with the divertor. The study we propose is important for supporting DIII-D and ITER with regard to minimizing edge power losses and localized erosion.
Resource Requirements:	2 run days 3 FW sources (60 MHz + 2 @ 90 MHz) NB sources
Diagnostic Requirements:	Complete set of diagnostics required for determining heating efficiency and edge properties, as well as effect of FW on energetic ion heating and current drive: edge density from TS, edge reflectometer, etc.; stored energy; kinetic profiles; spectroscopy; etc. Also, diagnostics for linking edge power loss to heating in the divertor: fast visible and IR camera views of antennas and divertor region, as available; Dimes probe to measure RF interaction zone in the divertor, if possible; etc.
Analysis Requirements:	Includes analysis for determining efficiency of FW heating (edge losses), data analysis for other diagnostics, and RF code analysis for predicted power deposition
Other Requirements:	--

Title	191: DIII-D/JET steady-state scenario comparison
Name:	Challis clive.challis@ukaea.uk	Affiliation:	CCFE
Research Area:	Core Integration (Steady State)	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	The goal is to compare the behaviour of candidate scenarios for steady-state operation of ITER in DIII-D and JET. This would be done by, firstly, matching key conditions in the two devices (i.e. plasma shape, q-profile, thermal betaN and normalised rotation), and then varying conditions around the match point to establish the most critical parameters for performance optimisation. This approach has two potential deliverables. The first would be an additional insight into the key tools for performance optimisation (in terms of both stability and confinement) compared with the analysis of the accessible domain of operation of each device independently. For example, JET has the ability to scan the q-profile in a slightly different way to DIII-D due to the large ratio of the resistive time to the energy confinement time, and is able to access low rho*. DIII-D has the ability to operate with high plasma shaping and vary the level of torque applied at high NBI heating power. By connecting the regimes being developed on the two devices via a common operating point (i.e. with similar plasma shape, q-profile, thermal betaN and normalised rotation), the different domains accessible on the machines could be connected. The second deliverable would be a frame within which to refine the extrapolation of the present regimes to future devices, such as ITER. The extrapolation of the performance potential of present advanced tokamak scenarios to ITER has a greater degree of uncertainty compared with inductive H-mode scenarios, for which specific confinement scalings have been developed. The development of a match point between DIII-D and JET for scenarios being developed for steady-state application to ITER would allow performance extrapolation techniques to be validated for machines of different size.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	There are a variety of parameters that have been identified as being important for the performance of advanced tokamak scenarios. The following are considered most crucial for this proposed comparison experiment: plasma shape; q-profile; thermal betaN; and normalised rotation. It is, therefore, not proposed to attempt a strict identity experiment in terms of rho* and nu*, although the envisaged comparison domain of the two devices should result in these parameters being of a similar order. The proposed approach is the following. The DIII-D experiments would be based on the recent ITER demonstration steady-state discharges, which are already relatively similar to the JET high triangularity plasma shape currently used for the development of advanced tokamak scenarios. The first step would be to modify the divertor configuration slightly reduce the elongation to more closely match the JET reference configuration. It is possible that some scenario optimisation will be necessary in terms of the time evolution of the fuelling and heating power (or beta control) to compensate for any deleterious effects of the configuration change. The target would be a plasma at high total betaN (around 3) without ECCD (to match JET conditions) and q-min below 2. The second step would be to vary the toroidal rotation velocity by varying the mix of co-injected and counter-injected NBI power. The object of this is twofold: firstly, to establish a range of rotation speed that could be matched to typical JET conditions in terms of either mach speed of normalised rotation frequency (depending on the physics of interest in the comparison). Secondly, it would provide a basis to assess the importance of toroidal rotation for the plasma performance in this domain. The third step would be to vary thermal betaN (from 2 up to the maximum achievable). The object is again twofold. To form an overlap with the domain achieved on JET and provide a basis to assess the importance of this parameter for the performance of the scenario. It is estimated that a reasonable progress could be made on the above in two run days. Further scans in plasma shape and q-profile are not envisaged in this proposal as they already exist in one or other or both device databases, which can be connected via this proposed experiment.
Background:	Development of the JET high betaN domain with q95 close to 5 has provided an opportunity for comparison with the existing DIII-D steady-state scenarios. This work (noted under the then ITPA SSO-1) has resulted in the establishment of a significant database of pulses at JET with an extensive variation in the q-profile [1-2]. In the meantime, DIII-D experiments to produce ITER demonstration discharges [3] have resulted in a version of the plasmas being developed for steady-state application with a plasma shape much closer to the JET pulses mentioned above. Together these developments provide an opportunity to significantly improve the comparison of advanced tokamak plasmas in these devices. [1] C Challis presentation at the DIII-D Science Meeting on 4-Apr-08 [2] C Challis presentation at IEA W68 at GA on 24-Jun-08 [3] E J Doyle et al 2008 22nd IAEA Fusion Energy Conference, Geneva, Switzerland EX/1-3
Resource Requirements:	Full NBI power (co- and counter-NBI) Real-time betaN control using NBI power
Diagnostic Requirements:	Plasma pressure and rotation diagnostics (including charge-exchange, Thomson scattering, electron cyclotron emission, interferometer). Visible spectroscopy. q-profile diagnostics (motional Stark effect). Magnetics. Neutrons.
Analysis Requirements:	Analysis is desirable using similar tools on the two devices (e.g. TRANSP for transport and current drive interpretative analysis)
Other Requirements:

Title	192: Comparison of Rotation Effects on Type I ELMing H-mode in JT-60U and DIII-D
Name:	Kamada none	Affiliation:	JAEA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	A. Leonard, T. Osborne, N.Oyama, H.Urano,M.Yoshida	ITPA Joint Experiment :	Yes
Description:	In order to improve predictive capability for ITER H-mode operation and control, the effects of toroidal rotation on the Pedestal structure and ELMs and core transport are investigated systematically over a wide range of the plasma shape based on DIIID and JT-60 data. In addition, the effects of rotation and ripple loss on the pedestal structure and ELMs are separated. DIII-D and JT-60 have quite unique capability to study plasma rotation with co and counter NBs. On the other hand, the plasma shape, thus the ELM stability, is different between DIII-D and JT-60. By combining these two conditions, this work can clarify the universal effects of rotation and dependence of rotation effects on plasma shape. (ITPA PEP-18)	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	By Utilizing the unique capability of rotation control with co- and counter- NBs in DIII-D and JT-60U and by utilizing the difference in the plasma shape and edge stability between the two tokamaks, we propose to conduct the inter-machine experiments on the rotation effects on the pedestal structure and type I ELMs. Based on the ITPA pedestal database, the pedestal structure in JT-60U and DIII-D are quite different: DIII-D has large pressure gradient and narrow pedestal width compared with JT-60U. This difference seems to be due to the plasma shape. In order to clarify the effects of rotation at different pedestal situation. In 2008, as the first step of the study, in JT-60U, effects of the toroidal rotation have been clarified at medium triangularity ~0.3 with CO, BAL, and CTR NB injected dischsrges at Ip=0.9,1.1,1.6 and 1.8MA and the ELM crash and inter-ELM dynamics were measured with fast diagnostics. In 2009, we propose rotation scan experiments at higher triangularity in DIII-D and take the following data for comparison with the JT-60U data taken in 2008: 1) Frequency and energy loss (incl. ELM affected area) of type I ELMs, and Pedestal width and inter-ELM transport at the same beta-p-ped and q95 with JT-60U, 2) Frequency and energy loss of type I ELMs, and Pedestal width and inter-ELM transport at the same pedestal collisionality and q95 with JT-60U, and 3) Core thermal confinement of the plasmas in 1) and 2). Related experiments reflecting the DIIID results will be proposed to JT-60U.
Background:	Recent tokamak experiments have revealed that the pedestal and core transport of the H-mode plasmas are determined under the linkage among pressure, current and rotation profiles. The goal of this research is to understand this complex system in order to improve predictive capability for ITER, and to develop control schemes for the pedestal parameters and ELMs and core transport. Concerning the parameter linkage, plasma rotation and its radial profile seem to play critical roles. Recent JT-60U experiment has demonstrated a shift of toroidal plasma rotation into co-direction reduces the inter-ELM transport loss and increase the pedestal height and width. In addition, type I ELM energy loss normalized to the pedestal stored energy (DWELM/Wped) increases with increasing co-directed rotation. The critical importance is to clarify the rotation effects on the pedestal structure and ELMs over a wide range of the plasma shape. Is is also important to separate the effects of rotation and ripple loss on the pedestal structure and ELMs. As for the core confinement of H-mode plasmas, both DIII-D and JT-60U have shown improved performance with co-directed rotation compared with counter rotation. The purpose of this study is to clarify the roles of plasma rotation systematically by utilizing the unique capability of rotation control with co- and counter- NBs in DIII-D and JT-60U and by utilizing the difference in the plasma shape and edge stability between the two tokamaks.
Resource Requirements:	1 day Experiment. CO & Counter NBs
Diagnostic Requirements:	Standard set. Pedestal Diagnostics. In particular fast measurement of CER.
Analysis Requirements:
Other Requirements:

Title	193: Rotation and beta effects on ELM suppression/control by RMPs
Name:	Loarte none	Affiliation:	ITER
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to perform experiments to suppress ELMs at various levels of plasma rotation and determine the influence of plasma rotation on the magnitude of the RMP perturbation that needs to be applied	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The proposed experiments are as follows : a) In order to avoid other effects affecting the experiments, the experiments on rotation will be done at a constant value of beta. For this experiment, I would propose to take a high delta (0.5) plasma with Ip ~ 1.1 MA and q95 = 3.6 with Pinp ~ 10 MW by NBI (co and counter). One would start from a plasma with pure co NBI and increase the fraction of counter NBI until problems with mode locking, etc. appear. b) The studies of beta effects, on the contrary, would be done at constant plasma rotation. This would be done by starting at low beta with Pinp ~ 4 MW by NBI with the minimum value of rotation that avoids problems with mode locks, etc., and increase the input power and plasma beta while maintaining as far as possible constant plasma rotation
Background:	Shielding of the fields by RMP by plasma rotation and its consequences for ELMs suppression remains one open issue regarding the application of this scheme to ITER conditions. ITER is expected to have low rotation but its resistivity is low and therefore the effects of plasma shielding of RMP perturbations in the pedestal region remain relatively uncertain. Similarly effects of plasma beta (thorugh shafranov shift or by changes in the pedestal structure) remain unclear. Demonstration of the sensitive or insensitive of ELM suppression by RMP to plasma rotation and beta would be required to evaluate consequences for the application of this method to ITER
Resource Requirements:	DIII-D with cryo-pumping and NBI heating and I-coils
Diagnostic Requirements:	Core, pedestal and ELM measurements to determine changes in pedestal and ELMs. Measurements of power fluxes to divertor
Analysis Requirements:	Analysis of plasma rotation, edge parameters and comparsion with appropiate models
Other Requirements:

Title	194: Requirements of resonance window for ELM suppression with consant Ip
Name:	Loarte none	Affiliation:	ITER
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to perform one experiment in which ELM suppression is achieved with the standard DIII-D method and the width of the resonance window is determined	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The idea is to study resonance window for ELm suppression by RMP (usually done by plasma current changes) by carrying out discharges with constant plasma current at various levels of magnetic field. The aim would be to demonstrate that ELM suppression can also be achieved in this way (i.e. the resonance condition is a physics requirement) and that the width of the window is the same in q95 regardless of the way that q95 is varied (field or current). The experiment could be done taking as basis a typical high delta (0.5) plasma with Ip ~ 1.1 MA and q95 = 3.6 with Pinp ~ 8 MW by NBI (co) where ELM suppression by RMPs is well characterised already
Background:	Experiments in DIII-D show that the perturbation by I-coils needs to be resonant within a relatively narrow window in order to achieve full suppression. The DIII-D experiments always explore this window by changing the plasma current, which is know to affect edge currents and pedestal instability. A demonstration of the requirement for resonance and of its width for ELM suppression with RMP with constant plasma current would be desirable to firm up the basis for the resonance condition for ITER
Resource Requirements:	DIII-D with cryo-pumping. NBI heating, I-coils
Diagnostic Requirements:	Core, edge and ELM diagnostics for pedestal and core measurements
Analysis Requirements:	Analysis of pedestal measurements and edge stability
Other Requirements:

Title	195: Fast-Ion Driven MHD Instabilities and Fast-Ion Transport in ASDEX Upgrade Similar Plasmas
Name:	Garcia-Munoz none	Affiliation:	IPP
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	W.W. Heidbrink and M.A. Van Zeeland	ITPA Joint Experiment :	No
Description:	It is the main goal of this proposal to link the internal redistribution of fast-ions with their losses, the global stability properties of fast-ion driven MHD instabilities and their impact on fusion performance. For this purpose, we plan to: - Trigger Alfven Eigenmodes (AEs) during the plasma current ramp-up phase with fast-ions of ICRH origin. The main goal is to reproduce the best diagnosed AE discharges performed in ASDEX Upgrade (AUG). - 1 or 2 NBI sources will be applied by the end of the plasma current ramp up phase to trigger Energetic Particle Modes (EPMs), making use of the low densities and the very high energetic ions generated by the ICRF-NBI synergy. - Document the AE and EPM internal structure by means of the internal fluctuation diagnostics (fast SXR, ECE radiometry,â?¦) - Investigate the EPM stability dependence on q-profile and fast-ion distribution function. - Study the fast-ion channeling process in phase space due to a chain of fast-ion driven MHD instabilities e.g. BAAEs+TAEs, RSAEs+TAEs, TAEs+EPMs, etc. The fast-ion redistribution in phase-space and possible loss will be documented using the FIDA, FIDA Imaging (FIDAi), and (Fast-Ion Loss Detector) FILD systems.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin with discharge #122116 as a reference. This discharge uses one equivalent co- neutral beam source and contains relatively low amplitude RSAE, TAE, and BAAE activity. Low power 60 MHz ICRF heating will then be applied to drive AEs with the very energetic ions generated by the fast waves 4th harmonic. The ICRH power will then be stepped up to increase the AEs driving force and so the fast-ion transport. 1 or 2 extra NBI sources might be applied at the end of the plasma current ramp up phase to modify the fast-ion population. Next, the discharge with the clearest AE/EPM activity and induced fast-ion transport will be chosen and a combination of scans will be made with the goal of best diagnosing the fast-ion transport and MHD instability internal structure, studying the impact of various instabilities with different amplitudes (and possible internal structure overlapping) on the fast-ion transport. A ne and Te scan (ECRF needed) will be made to evaluate the critical β_fast for the onset of EPMs as well as to modify the instability amplitude and the subsequent fast-ion transport.
Background:	Alfven Eigenmodes (AEs) and Energetic Particle Modes (EPMs), are common in most present large fusion devices. They have the potential to cause important redistribution and loss of fast-ions. Therefore, a full understanding of the interplay between fast-ions and fast-ion driven MHD instabilities is mandatory to avoid grave consequences in burning plasma devices. In ASDEX Upgrade (AUG), fast-ion driven MHD instabilities are common in low density plasmas during ICRF heating. While the losses of fast-ions induced by these instabilities are well diagnosed by the AUG Fast-Ion Loss Detector (FILD) [1, 2], the internal MHD induced fast-ion redistribution is not. The time-resolved energy and pitch angle measurements of fast-ion losses obtained with FILD on AUG have allowed identifying the main fast-ion loss mechanisms. In DIII-D, AEs structures are very well documented by its unique internal fluctuation diagnostic system and modelling [3-5]. Recently, a large reduction in the fast-ion pressure during AE activity has been observed with the FIDA diagnostic [6, 7]. However, the underlying fast-ion transport mechanism remains to be conclusively identified. For this purpose a FILD system is being installed on DIII-D. Time-resolved FILD measurements of lost ions energy and pitch angles will allow the identification of the fast-ion loss mechanisms. Furthermore, the new fast-ion diagnostic set on DIII-D, FIDA, FIDAi and FILD, will have the capability to conclusively connect the redistribution and loss of fast-ions due to multiple MHD instabilities. As mentioned, while the primary goal of this experiment is to utilize the unique fast-ion diagnostic set on DIII-D to establish a conclusive connection between the observed MHD activity and fast-ion redistribution and loss (channelling) in AUG similar plasmas, a secondary result of these experiments will be a set of discharges useful for fast-ion physics modelling validation. Moreover, these discharges will be used to investigate the global EPM stability properties based on the internal fast-ion distribution function.
Resource Requirements:	Machine Time: 1 day NBI = 3 sources ICRH = 60 MHZ and maximal heating power ECRH = 4 gyrotrons
Diagnostic Requirements:	FIDA, FIDAi, FILD, neutron fluxes and internal fluctuation diagnostics
Analysis Requirements:
Other Requirements:

Title	196: FILD commissioning
Name:	Garcia-Munoz none	Affiliation:	IPP
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	R. Fisher, W. W. Heidbrink and M. A. Van Zeeland	ITPA Joint Experiment :	No
Description:	The main goal of this proposal is to validate the FILD design on DIII-D as well as to set its operational limits based on some critical variables. For this purpose, we, basically, plan to make a prompt-loss study in discharges with: - A plasma current scan of 0.4, 0.6 and 0.8 MA to vary the amount of prompt losses. - An NBI scan using all possible NBI sources co- and counter and modifying (if it is possible) their voltage. - A Bt scan to check the energy-pitch angle grid resolution. - R_FILD scan; change in FILD radial position to get enough fast-ion signals but without overheating the probe. - Maximal ICRF heating, on- and off-axis will be applied to learn about the general FILD capabilities during FW heating.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The plan is to establish a quiet shot, then cycle through all four angles of injections repeatedly in the same shot. 60 MHz ICRH will be applied for one or two cycles of the four sources to accelerate deuterium beam ions with the 4th harmonic. We would like to change the injection energy from the usual 75-81 keV down to 50 keV on one of the shots to benchmark the energy-pitch angle grid (on the lowest plasma current shot?). An Ip scan (3-4) points, a Bt scan (2-3 points) and a RFILD scan will complete the commissioning of the FILD. If time permits, an outer gap scan for one of the conditions could also be attempted.
Background:	A new diagnostic for fast-ion losses has been installed on DIII-D. The DIII-D fast-ion loss detector (FILD) design is based on the concept of a similar system installed in ASDEX Upgrade [1, 2]. The detector acts as a magnetic spectrometer, dispersing fast-ions onto a scintillator with the hit point depending on their gyroradii (energy) and pitch angle. The radial position of the FILD head can be varied through a radial manipulator in order to find out the best location within the limiter shadow, a compromise between fast-ion loss signal and heat load. [1] GARCIA-MUNOZ, M., et al., Nucl. Fusion 47, L10 (2007) [2] GARCIA-MUNOZ, M., et al., Phys. Rev. Lett. 100, 055055 (2008)
Resource Requirements:	Machine Time: ½ day NBI = all sources ICRH = 60 MHz with maximal heating power
Diagnostic Requirements:	FILD
Analysis Requirements:
Other Requirements:

Title	197: Fast-Ion Driven MHD Instabilities and Fast-Ion Transport in ASDEX Upgrade Similar Plasmas
Name:	Garcia-Munoz none	Affiliation:	IPP
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	W.W. Heidbrink and M.A. Van Zeeland	ITPA Joint Experiment :	No
Description:	It is the main goal of this proposal to link the internal redistribution of fast-ions with their losses, the global stability properties of fast-ion driven MHD instabilities and their impact on fusion performance. For this purpose, we plan to: - Trigger Alfven Eigenmodes (AEs) during the plasma current ramp-up phase with fast-ions of ICRH origin. The main goal is to reproduce the best diagnosed AE discharges performed in ASDEX Upgrade (AUG). - 1 or 2 NBI sources will be applied by the end of the plasma current ramp up phase to trigger Energetic Particle Modes (EPMs), making use of the low densities and the very high energetic ions generated by the ICRF-NBI synergy. - Document the AE and EPM internal structure by means of the internal fluctuation diagnostics (fast SXR, ECE radiometry,â?¦) - Investigate the EPM stability dependence on q-profile and fast-ion distribution function. - Study the fast-ion channeling process in phase space due to a chain of fast-ion driven MHD instabilities e.g. BAAEs+TAEs, RSAEs+TAEs, TAEs+EPMs, etc. The fast-ion redistribution in phase-space and possible loss will be documented using the FIDA, FIDA Imaging (FIDAi), and (Fast-Ion Loss Detector) FILD systems.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin with discharge #122116 as a reference. This discharge uses one equivalent co- neutral beam source and contains relatively low amplitude RSAE, TAE, and BAAE activity. Low power 60 MHz ICRF heating will then be applied to drive AEs with the very energetic ions generated by the fast waves 4th harmonic. The ICRH power will then be stepped up to increase the AEs driving force and so the fast-ion transport. 1 or 2 extra NBI sources might be applied at the end of the plasma current ramp up phase to modify the fast-ion population. Next, the discharge with the clearest AE/EPM activity and induced fast-ion transport will be chosen and a combination of scans will be made with the goal of best diagnosing the fast-ion transport and MHD instability internal structure, studying the impact of various instabilities with different amplitudes (and possible internal structure overlapping) on the fast-ion transport. A ne and Te scan (ECRF needed) will be made to evaluate the critical βfast for the onset of EPMs as well as to modify the instability amplitude and the subsequent fast-ion transport.
Background:	Alfven Eigenmodes (AEs) and Energetic Particle Modes (EPMs), are common in most present large fusion devices. They have the potential to cause important redistribution and loss of fast-ions. Therefore, a full understanding of the interplay between fast-ions and fast-ion driven MHD instabilities is mandatory to avoid grave consequences in burning plasma devices. In ASDEX Upgrade (AUG), fast-ion driven MHD instabilities are common in low density plasmas during ICRF heating. While the losses of fast-ions induced by these instabilities are well diagnosed by the AUG Fast-Ion Loss Detector (FILD) [1, 2], the internal MHD induced fast-ion redistribution is not. The time-resolved energy and pitch angle measurements of fast-ion losses obtained with FILD on AUG have allowed identifying the main fast-ion loss mechanisms. In DIII-D, AEs structures are very well documented by its unique internal fluctuation diagnostic system and modelling [3-5]. Recently, a large reduction in the fast-ion pressure during AE activity has been observed with the FIDA diagnostic [6, 7]. However, the underlying fast-ion transport mechanism remains to be conclusively identified. For this purpose a FILD system is being installed on DIII-D. Time-resolved FILD measurements of lost ions energy and pitch angles will allow the identification of the fast-ion loss mechanisms. Furthermore, the new fast-ion diagnostic set on DIII-D, FIDA, FIDAi and FILD, will have the capability to conclusively connect the redistribution and loss of fast-ions due to multiple MHD instabilities. As mentioned, while the primary goal of this experiment is to utilize the unique fast-ion diagnostic set on DIII-D to establish a conclusive connection between the observed MHD activity and fast-ion redistribution and loss (channelling) in AUG similar plasmas, a secondary result of these experiments will be a set of discharges useful for fast-ion physics modelling validation. Moreover, these discharges will be used to investigate the global EPM stability properties based on the internal fast-ion distribution function. [1] GARCIA-MUNOZ, M., et al., Nucl. Fusion 47, L10 (2007) [2] GARCIA-MUNOZ, M., et al., Phys. Rev. Lett. 100, 055055 (2008) [3] VAN ZEELAND, M.A., et al., Phys. Rev. Lett. 97, 135001-1 (2006) [4] VAN ZEELAND, M.A., et al., Phys. Plasmas 14, 056102-1 (2007) [5] NAZIKIAN, R., et al., Phys. Plasmas 15, 056107 (2008) [6] HEIDBRINK, W.W., et al., Phys. Rev. Lett. 99, 245002-1 (2007) [7] HEIDBRINK, W.W., et al., Nucl. Fusion 48, 084001 (2008)
Resource Requirements:	Machine Time: 1 day NBI = 3 sources ICRH = 60 MHZ and maximal heating power ECRH = 4 gyrotrons
Diagnostic Requirements:	FIDA, FIDAi, FILD, neutron fluxes and internal fluctuation diagnostics
Analysis Requirements:
Other Requirements:

Title	198: Stability and electron thermal transport effects of high-n modes in QH plasmas
Name:	Nazikian none	Affiliation:	PPPL
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	Mike Van Zeeland, W. Heidbrink, N.N. Gorelenkov, G. Y. Fu, ...	ITPA Joint Experiment :	No
Description:	The QH regime exhibits a broad spectrum of high-n modes that appear to be energetic particle driven but are poorly understood. This experiment will make use of new diagnostic capability to explore the density and temperature structure of the modes and to assess the role of these modes on electron thermal transport.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Generate QH plasma. Obtain strong high-n modes. Obtain ne, te measurements on linear ECE and BES array and on high resolution measurements from UCLA. Add ECH to qmin and on-axis, assess mode stability. Add pulse of ECH in unstable and stable regime and assess effect on thermal transport.
Background:	The QH regime is characterized by a multitude of high-n modes. The role of these modes on core transport is not understood. The possibility that these modes are thermally driven and that they strongly affect electron thermal transport will be explored.
Resource Requirements:	BES, ECE, ECH.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	199: Effect of low-n RSAEs, TAEs on electron thermal transport
Name:	Nazikian none	Affiliation:	PPPL
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	M. Van Zeeland, W. Heidbrink, G.J. Kramer, ...	ITPA Joint Experiment :	No
Description:	Multiple low-n modes are often observed in the preheat phase of DIII-D plasmas. Nothing is known of the effect of these modes on electron transport. This experiment will vary the stability of these modes to determine their effect on electron thermal transport.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Regenerate 122117 or equivalent. Obtain ECE, BES radial measurements. Scan beam voltage and keep power constant to vary strength of the mode activity. Add ECH pulse train and see propagation characteristics of the pulses.
Background:	We know nothing about the effects of these low-n modes on electron thermal transport. This experiment will address this issue.
Resource Requirements:	BES, ECE, ECH.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	200: stability and structure of the E-GAM
Name:	Nazikian none	Affiliation:	PPPL
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	M. Van Zeeland, W. Heidbrink, G. Y. Fu, ...	ITPA Joint Experiment :	No
Description:	This experiment aims to resolve the reason why the E-GAM induces such large loss/redistribution of the beam ions by measuring the mode amplitude well off the midplane and measure the redistribution of the beam ions and losses correlated with the mode bursts. In addition the stability and dispersion of the mode will be explored using a beam energy scan.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce 134505 or equivalent. Scan beam voltage keeping power constant. Move plasma well off midplane and observe fluctuation level.
Background:	A mode called the E-GAMs was recently discovered on DIII-D that produces intense bursting of the neutrons indicative of rapid fast ion redistribution. The mechanism for the loss/redistribution needs to be resolved. This experiment aims at resolving this issue by measuring the peak mode amplitude off the midplane and to measure the distribution properties of the losses.
Resource Requirements:	Voltage scan of beams.
Diagnostic Requirements:	BES, ECE, FIDA, scintillator detectors.
Analysis Requirements:	M3D
Other Requirements:

Title	201: Validation of Integrated Modeling
Name:	Budny none	Affiliation:	PPPL
Research Area:	General Integrated Modeling	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Measure changes in the H-mode characteristics of ITER-like plasmas when the heating shifts from NBI-only to include simulated alpha heating, and test the ability of PTRANSP to simulate the plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Form ITER-demo H-mode plasmas simulating external heating with near balanced NBI, then substitute central ECH to simulate alpha heating. Try various combinations of P_NB / P_EC, and try feedback control of P_EC on n(0)*T(0). H-mode characteristics to be measured include 1) pedestal temperatures, 2) core temperatures / stiffness, 3) energy confinement, and 4) back transition power.
Background:	Experiments were done in JET (2000) simulating burn control with feedback control of central ICRH power simulating alpha heating. Recent studies using the PTRANSP code [R.Budny, et al., Nuc. Fus <48> 075005] have predicted high Q_DT and ignition in ITER H-mode plasmas. One of the assumptions used was that the pedestal temperatures remain constant as the heating changes from external (NNBI, ICRH, and ECH) to alpha heating. This experiment will test the pedestal and stiffness assumptions and thus challenge the predictions. The data will be used for PTRANSP time-dependent, self-consistent simulations. TORAY-GA (and GENRAY if available) will be used to compute the ECH. GLF23 (and TGLF if available) will be used to simulate the temperatures, momentum, and density evolutions. Recently PTRANSP has been upgraded to improve the momentum and density predictions. Much progress has been made verifying the numerical solutions of PTRANSP. PTRANSP-TORIC simulations of ICRH in ITER agree with AORSA-CQL3D to first order. Also PTRANSP-TORAY simulations are in excellent agreement with GENRAY. The validations from this experiment will assess the reliability of PTRANSP for simulating ITER and also for scenario development in present experiments. Simulations of the JET burn control plasmas have been done using PTRANSP-TORIC-GLF23. Also PTRANSP simulations have been done of two of the DIII-D ITERDEMO H-mode plasmas (131498 and 131499). A complementary experiment, similar to this one is being proposed for JET. The DIII-D and JET experiments might be suitable for an ITPA joint experiment.
Resource Requirements:	one-day experiment with near-balanced beams and at least 3 MW of ECH power.
Diagnostic Requirements:	Accurate Thomson, CER, ECE, and MSE. Fluctuation measurements would be useful for indicating changes in turbulence as the nature of the heating changes
Analysis Requirements:	TRANSP analysis and PTRANSP predictions. Perhaps later simulations with GYRO / TGYRO.
Other Requirements:

Title	202: Validation of Integrated Modeling
Name:	Budny none	Affiliation:	PPPL
Research Area:	General Integrated Modeling	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Measure changes in the H-mode characteristics of ITER-like plasmas when the heating shifts from NBI-only to include simulated alpha heating, and test the ability of PTRANSP to simulate the plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Form ITER-demo H-mode plasmas simulating external heating with near balanced NBI, then substitute central ECH to simulate alpha heating. Try various combinations of P_NB / P_EC, and try feedback control of P_EC on n(0)*T(0). H-mode characteristics to be measured include 1) pedestal temperatures, 2) core temperatures / stiffness, 3) energy confinement, and 4) back transition power.
Background:	Experiments were done in JET (2000) simulating burn control with feedback control of central ICRH power simulating alpha heating. Recent studies using the PTRANSP code [R.Budny, et al., Nuc. Fus <48> 075005] have predicted high Q_DT and ignition in ITER H-mode plasmas. One of the assumptions used was that the pedestal temperatures remain constant as the heating changes from external (NNBI, ICRH, and ECH) to alpha heating. This experiment will test the pedestal and stiffness assumptions and thus challenge the predictions. The data will be used for PTRANSP time-dependent, self-consistent simulations. TORAY-GA (and GENRAY if available) will be used to compute the ECH. GLF23 (and TGLF if available) will be used to simulate the temperatures, momentum, and density evolutions. Recently PTRANSP has been upgraded to improve the momentum and density predictions. Much progress has been made verifying the numerical solutions of PTRANSP. PTRANSP-TORIC simulations of ICRH in ITER agree with AORSA-CQL3D to first order. Also PTRANSP-TORAY simulations are in excellent agreement with GENRAY. The validations from this experiment will assess the reliability of PTRANSP for simulating ITER and also for scenario development in present experiments. Simulations of the JET burn control plasmas have been done using PTRANSP-TORIC-GLF23. Also PTRANSP simulations have been done of two of the DIII-D ITERDEMO H-mode plasmas (131498 and 131499). A complementary experiment, similar to this one is being proposed for JET. The DIII-D and JET experiments might be suitable for an ITPA joint experiment.
Resource Requirements:	one-day experiment with near-balanced beams and at least 3 MW of ECH power.
Diagnostic Requirements:	Accurate Thomson, CER, ECE, and MSE. Fluctuation measurements would be useful for indicating changes in turbulence as the nature of the heating changes
Analysis Requirements:	TRANSP analysis and PTRANSP predictions. Perhaps later simulations with GYRO / TGYRO.
Other Requirements:

Title	203: Input power requirements for access to H-mode, Type III ELMy H-mode and high confinement H-modes
Name:	Alberto none	Affiliation:	ITER
Research Area:	General ITER Physics	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to investigate the required power levels in DIII-D to access H-mode, Type III ELMy H-mode and Type I ELMy H-mode, under the assumptions that this will be required to get H ~ 1 in DIII-D, in D, H and He for the flat top and transient current ramp-up/down phases with NBI heating	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	For this initial experiment, I would propose to take some typical DIII-D discharges with Ip ~ 1.2 MA and q95 = 3 in the flat top and low triangularity ~ 0.2 and high triangularity ~0.5 and investigate three levels of density in the L-mode phase plus three levels of fuelling in the H-mode phase aiming to (for high delta) ne/nGW ~ 0.7 (or unfuelled density), 0.9 and 1.1 (or highest density in Type I ELMy H-mode). This should be done for the flat top phase and also for Ip ~ 1.0 MA both in flat top and current ramp-up/down phases with various levels of dIp/dt
Background:	For determination of the level of additional heating required for ITER to meet its performance expectations during the various operational phases we assume that we will need to exceed the H-mode threshold by a margin in terms of required additional heating. The specification of this margin is done on empirical evidence from some tokamaks which may not be of general application.
Resource Requirements:	DIII-D tokamak with cryo-pumping and NBI heating
Diagnostic Requirements:	Usual diagnostics with emphasis on pedestal parameters to determine H-mode onset and characteristics
Analysis Requirements:	Experimental analysis of measurements obtained
Other Requirements:

Title	204: Pedestal scaling for discharges dominated by pellet fuelling
Name:	Loarte none	Affiliation:	ITER
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to investigate the role of core versus edge fuelling in determining the pedestal density width	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	For this experiment, I would propose to take some typical DIII-D discharges with Ip ~ 1.2 MA and q95 = 3 with medium triangularity with Pinput/Pl_H ~2 and strike point positions providing maximum pumping and perform a fuelling scan with pellet injection and a similar one with gas fuelling, Ideally one would like to have matching points in terms of pedestal density by pellet fuelling and gas puffing from low to high densities in Type I ELMy H-mode and possibly a combination of both
Background:	The physics processes that determine the scaling of the pedestal width need to be characterised experimentally in order to provide a firm physics basis for our expectations in ITER. Comparison of pedestal measurements among devices shows largest discrepancies for the pedestal density width. This is believed to be due to the role of particle sources in influencing the pedestal width. Core particle sources in ITER are likely to be very different in ITER because fuelling by recycling neutrals is expected to be very ineffective and core fuelling will be dominated by pellets. This raises some issues about the extrapolability of present pedestal scalings to ITER conditions
Resource Requirements:	DIII-D tokamak with cryo-pump, NBI heating and pellet fuelling system,
Diagnostic Requirements:	Core and edge diagnostics, including pellet diagnostics
Analysis Requirements:	Analysis of experimental measurements and comparsion with edge modelling
Other Requirements:

Title	205: Effect of TF ripple on pedestal plasma studied by radial shifts
Name:	Loarte none	Affiliation:	ITER
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to investigate the effect of TF ripple on pedestal characteristics and plasma performance in DIII-D by changing the ripple level at the plasma by radial shifts. The normal level of ripple at the separatrix in DIII-D is 0.4%, which is at a level for which noticeable effects can be measured at JET for low collisionality conditions. I estimate that a radial shift of the plasma by 10 cm this level can be reduced by almost one order of magnitude more or less. In this way we can cover the range from basically no ripple to the level that is probably achievable in ITER (with ferritic inserts) with this technique.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	For this experiment, I would propose to take a medium delta very high clearance configuration (i.e. small plasma) with Ip ~ 0.8-1.0 MA and q95 = 3 with Pinput/Pl_H ~2 and strike point positions providing maximum pumping and perform a radial scan of the distance to the outer wall from the standard position (which gives 0.4% ripple level at the separatrix) to that + 10 cm. This scan should probably be done in subsequent shots to be sure that we are not changing too many things (such as wall recycling) transiently during the discharge. Due to the collisionality effects identified in JET this experiment should be done at the lowest level of density achievable and at medium and high density in terms of greenwald fraction (while staying in Type I ELMy H-mode)
Background:	Experiments in some tokamaks (JET, JT-60U, etc.) have shown that TF ripple can have a significant effect on the thermal and particle confinement in H-mode plasmas through its effects on the pedestal plasma characteristics. The physics processes behind this behaviour are not clear and their consequences are shown to change with pedestal plasma collisionality. This makes the extrapolation of such results in terms of ripple correction requirements for ITER uncertain
Resource Requirements:	DIII-D tokamak with cryo-pumping and NBI heating
Diagnostic Requirements:	Obviously a key measurement in this case is the pedestal parameters. It is important to check the compatibility of the required displacements to change the ripple and the required measurements.
Analysis Requirements:	Analysis of experimental pedestal and core measurements
Other Requirements:

Title	206: H-mode/pedestal physics and particle transport ECRH heated Type I ELMy H-modes
Name:	Loarte none	Affiliation:	ITER
Research Area:	General ITER Physics	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to compare pedestal physics, ELM and overall plasma behaviour for plasmas exclusively heated with NBI and ECRH	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	For this experiment, I would propose to take a medium (0.3) or high delta (0.5) plasma with Ip ~ 1.2 MA and q95 = 3 with Pinput/Pl_H ~2 with both heating systems (NBI co-injection) and strike points providing good pumping and perform a fuelling scan from zero (i.e., natural density) to the level at which Type III ELMy H-mode is achieved. If ECRH system allows (i.e. restriction on absolute power and cut-offs, etc.) a comparison between gas fuelling and pellet fuelling could be carried out to study the effect of ECRH on core fuelled H-modes
Background:	All heating systems in ITER will predominantly deposit power in electrons and will input relatively low torque into the plasma. The understanding of pedestal behaviour and H-mode physics in such conditions is based on a very restricted experimental database of discharges meeting ITER requirements for QDT=10, which is dominated by high torque input/ion heated plasmas
Resource Requirements:	DIII-D tokamak with NBI and ECRH heating, cryopumping and pellet fuelling
Diagnostic Requirements:	Core, edge, ELM and pellet diagnostics. Some NBI blips in the ECRH-only discharges will be needed to measure ion parameters and plasma rotation
Analysis Requirements:	Analysis of measurements obtained and comparison with stability and edge modelling
Other Requirements:

Title	207: Comparison of rotation in ECCD plasmas to C-Mod LHCD
Name:	Rice none	Affiliation:	MIT PSFC
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	Nat Fisch Matt Reinke Ron Parker John Wright John deGrassie Wayne Solomon	ITPA Joint Experiment :	Yes
Description:	The purpose of this experiment is to compare toroidal rotation velocity profiles in DIII-D ECCD discharges to those in C-Mod with LHCD. Parameter scans of electron density, ECCD power and deposition location will allow a comparison of C-Mod results, and to a model based on an inward pinch of energetic trapped electrons.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The approach will be to vary ECCD power and deposition location under different operating conditions, and to measure the toroidal rotation velocity profiles.
Background:	Rotation velocity profile control without neutral beam injection is desirable for ITER. In C-Mod LHCD plasmas, strong counter-current toroidal rotation has been observed, which is in contrast to the co-current rotation in ICRF heated discharges. Parameter scans of electron density and waveguide phasing have revealed a strong correlation of the rotation to the plasma internal inductance. Modeling indicates that this may be due to an inward pinch of energetic trapped electrons. Previous DIII-D results have shown a variation of rotation velocity profiles with ECCD deposition location. A comparison between the two methods will help reveal the underlying physics and lead to a powerful rotation velocity profile control technique.
Resource Requirements:	ECCD power and deposition location control.
Diagnostic Requirements:	CXRS and MSE from beam blips.
Analysis Requirements:	Transport and wave codes.
Other Requirements:

Title	208: Generation of current hole by ECCD alone and its conversion to AT plasma
Name:	Shiraiwa none	Affiliation:	PSFC, MIT
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	A. Garofalo, M. Makowski	ITPA Joint Experiment :	No
Description:	Current hole is a phenomenon observed at the extreme limit of reversed shear profile, and is reported to have peculiar features such as current clamping. In the past, current hole was produced by intense heating during plasma current ramp-up, delaying the Ohmic current penetration. We propose to produce current hole by early ECCD injection without plasma current ramp-up (or only overdrive by ECCD). By applying ECCD from the very initial phase, plasma is expected to evolve towards current hole formation. Such an operation allows us to sustain current hole in steady state, providing a good basis to study the current hole.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The discharge scenario used in this experiment resembles the scenario used on JT-60U [1], and is planned in the upcoming experimental campaign on Alcator C-mod [2]. First, we produce a plasma by induction. The initial plasma current must be small enough (~150kA) so that it can be sustained by ECCD alone, and a lower density is favored to ensure a sufficiently high EC driven current. Then, we start ECCD. It is important to start off-axis ECCD immediately after break down before the plasma current penetrates to the core, which would occur in a rather short time scale due to low temperature. By using the increased ECCD power available in the next campaign, we hope to drive up to ~200 kA at minor radius rho~0.4, with density of ~2x1019m-3. Since there is no Ohmic loop voltage at the plasma edge, the current profile will evolve towards the EC driven current profile, which has a current hole. We might spend some discharges to find the best combination of initial plasma current, ECCD power, ECCD location, and timing. Once this discharge scenario is established, we will try off-axis NBCD in the later phase of discharge to increase the pressure, and to convert the target plasma to a high f_BS AT plasma.
Background:	urrent hole has been observed in several tokamaks including JET[3], JT-60U[4], DIII-D[5], and ASDEX-U. Although 10 years have passed since it was first reported, its nature is still not well understood. Generally, it is produced using a technique to produce a reversed shear plasma, combining intense heating and plasm current ramp-up to delay the current penetration. In plasmas produced in such a way, several ingredients such as bootstrap current, internal transport barrier, and loop voltage are tightly linked together (so-called "self-organized" state) and make it difficult to discuss the roles of each ingredient. But, what if the extent of self-organization is reduced? The current hole plasma produced by this experiment is in a less self-organized state. Current profile is well controlled by the external source and the fraction of bootstrap current is small because of low density. Once such a target plasma is established, by adding NBCD (off-axis or even on-axis, if current clamping works) we can produce plasmas with an arbitrary extent of self-organization. This research also relates to the concept of solenoid-less tokamak reactor. It is envisioned that the Ohmic solenoid will play a relatively small role in the steady-state phase of a tokamak reactor. Reducing, and eventually eliminating the Ohmic solenoid has been considered as an extended concept of advance tokamak reactor. In fact, there are design works of such a reactor and its economical benefit has been discussed [6]. In solenoid-less operation, the technique mentioned above is not possible. For example, in the previous nearly solenoid-less experiment on JT-60U [1], the off-axis LHCD overdrive was used to produce a plasma to which NB injection became possible, and it was observed that the plasma current profile evolved towards one with current hole. To study the plasma driven by ECCD instead of LHCD is important for the sake of developing an operational scenario of a solenoid-free toakamk reactor. [1] S. Shiraiwa et. al., Phys. Rev. Lett. 92, 035001 (2004) [2] S. Shiraiwa et. al., Bull. Am. Phys. Soc. 53, (2008) [3] N. C. Hawkes, et. al., Phys. Rev. Lett. 87, 115001 (2001) [4] T. Fujita, et. al., Phys. Rev. Lett. 95, 075001 (2005) [5] R.J. Jayakumar, et. al, Nucl.Fusion 15004 (2008) [6] S. Nishio, et. al., FT-P1/21, Proc. 19th Int. Conf. on Fusion Research 2002 (Lyon, 2002)
Resource Requirements:	At least 5 gyrotrons for ECCD, 6 highly desirable
Diagnostic Requirements:	MSE
Analysis Requirements:
Other Requirements:	Research area and plasma shape preference is TBD

Title	209: Effects of mixed plasma exposure on D retention and surface damage of tungsten
Name:	Ueda none	Affiliation:	Osaka U, Japan
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):	Clement Wong	ITPA Joint Experiment :	No
Description:	We propose mixed plasma (D, He, and Ar (Ne)) exposure experiments to several tungsten samples in DIII-D with the DiMES probe to study D retention and surface damage (blistering etc.). The purpose and the aim are described below. Recently, the ion beam experiment (~10^20 m-2s-1, ~1 keV, Osaka University [presented in ICFRM13 (2007) and TOFE18(2008)]) and the high density plasma simulator experiment (~10^22 m-2s-1, ~ 60 eV, UCSD(PISCES)) showed simultaneous irradiation of D(H) and He significantly reduces D retention and blistering of tungsten. In addition, when the He ion energy was sufficiently reduced compared to H ion energy (He ion energy : 0.6 keV and H ion energy : 1.5 keV, Osaka Univ.[TOFE18(2008)]), blistering was enhanced, which means this effect is dependent on relative ion energies. The reason could be attributed to the formation of He bubble layers, which seem to be diffusion barrier of D(H). In actual tokamak devices, edge plasma ion irradiation conditions are different from those of ion beams and high density plasma simulators. For example, impinging ions have angular and energy dependence, which cannot be reproduced by ion beams and high density plasma simulators. In addition, plasma impurities such as carbon could affect this phenomena through the formation of mixing layers. Therefore, it is important to know whether mixed irradiation of D with He and/or Ar(Ne) reduces D retention and suppresses blistering in actual tokamak devices.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Several tungsten samples (Pure tungsten (ITER grade), recrystallized tungsten, VPS(Vacuum Plasma Sprayed) tungsten, and UFG (Ultra Fine Grained)-W) will be exposed to the DIII-D edge plasmas by using the DiMES probe. VPS tungsten is coated on F82H with the thickness of 0.5 mm. We want to expose samples to two types of plasmas. One is D and He mixed plasma with He concentration of 5-10% (simulation of burning plasma). The other is D, He and Ar(Ne) plasma for the simulation of the actively edge plasma cooling by Ar(Ne) puffing. Total exposure fluence of (5-10) x 10^23 m-2 and sample temperature of 200-300 °C are planned. The higher fluence is better. The D and He exposure experiment has higher priority than D, He and Ar(Ne) mixed exposure. After the experiment in DIII-D, the samples will be brought back to Japan to measure depth profile of D by an ion beam (NRA or ERDA), total D retention by TDS, surface morphology by SEM, surface mixed layer composition by XPS, and near surface microstructure by TEM. All of the samples were prepared in Japan. The estimated preparation time before the experiment is about 3 months.
Background:	In ITER, installation of tungsten divertor before the first DT plasma and tungsten first walls in the high duty DT phase are under consideration. Although in-vessel T retention in tungsten could be reduced compared with CFC and Be, it is necessary to estimate T retention. In tungsten, implanted hydrogen isotope atoms diffuse into the bulk ( at more than roughly 300 K ) and are trapped at either intrinsic or neutron-induced trapping sites ( bulk retention). In other words, T retention in tungsten is governed by T diffusion behavior and trap site characteristics (density and trapping energy). In addition, surface material mixing layers (ex. W and C mixing layers), He bubbles, and blisters generally affect hydrogen isotope diffusion and desorption. Therefore, T retention in tungsten depends on many parameters such as ion impinging energy (energy distribution), impurity ions in plasmas (wall impurity and He ash), temperature, tungsten material characteristics, neutron flux, and so on. So, in order to estimate T retention in tungsten, we need to make experiments in various plasma exposure conditions (laboratory plasmas and actual magnetic confinement devices) to comprehensively understand this phenomena.
Resource Requirements:	DiMES probe
Diagnostic Requirements:	Not special
Analysis Requirements:	Not special. Sample analysis will be done afterward in Japan.
Other Requirements:

Title	210: Comparison of Rotation Effects on Type I ELMing H-mode in JT-60U and DIII-D
Name:	Kamada none	Affiliation:	Japan Atomic Energy Agency
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	A. Leonard, T. Osborne, N.Oyama, H.Urano, M. Yoshida	ITPA Joint Experiment :	Yes
Description:	This proposal is the ITPA PEP-18. In order to improve predictive capability for ITER H-mode operation and control, the effects of toroidal rotation on the Pedestal structure and ELMs and core transport are investigated systematically over a wide range of the plasma shape based on DIIID and JT-60 data. In addition, the effects of rotation and ripple loss on the pedestal structure and ELMs are separated. DIII-D and JT-60 have quite unique capability to study plasma rotation with co and counter NBs. On the other hand, the plasma shape, thus the ELM stability, is different between DIII-D and JT-60. By combining these two conditions, this work can clarify the universal effects of rotation and dependence of rotation effects on plasma shape.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	By Utilizing the unique capability of rotation control with co- and counter- NBs in DIII-D and JT-60U and by utilizing the difference in the plasma shape and edge stability between the two tokamaks, we propose to conduct the inter-machine experiments on the rotation effects on the pedestal structure and type I ELMs. Based on the ITPA pedestal database, the pedestal structure in JT-60U and DIII-D are quite different: DIII-D has large pressure gradient and narrow pedestal width compared with JT-60U. This difference seems to be due to the plasma shape. In order to clarify the effects of rotation at different pedestal situation. In 2008, as the first step of the study, in JT-60U, effects of the toroidal rotation have been clarified at medium triangularity ~0.3 with CO, BAL, and CTR NB injected dischsrges at Ip=0.9,1.1,1.6 and 1.8MA and the ELM crash and inter-ELM dynamics were measured with fast diagnostics. In 2009, we propose rotation scan experiments at higher triangularity in DIII-D and take the following data for comparison with the JT-60U data taken in 2008: 1) Frequency and energy loss (incl. ELM affected area) of type I ELMs, and Pedestal width and inter-ELM transport at the same beta-p-ped and q95 with JT-60U, 2) Frequency and energy loss of type I ELMs, and Pedestal width and inter-ELM transport at the same pedestal collisionality and q95 with JT-60U, and 3) Core thermal confinement of the plasmas in 1) and 2). Related experiments reflecting the DIIID results will be proposed to JT-60U.
Background:	Recent tokamak experiments have revealed that the pedestal and core transport of the H-mode plasmas are determined under the linkage among pressure, current and rotation profiles. The goal of this research is to understand this complex system in order to improve predictive capability for ITER, and to develop control schemes for the pedestal parameters and ELMs and core transport. Concerning the parameter linkage, plasma rotation and its radial profile seem to play critical roles. Recent JT-60U experiment has demonstrated a shift of toroidal plasma rotation into co-direction reduces the inter-ELM transport loss and increase the pedestal height and width. In addition, type I ELM energy loss normalized to the pedestal stored energy (DWELM/Wped) increases with increasing co-directed rotation. The critical importance is to clarify the rotation effects on the pedestal structure and ELMs over a wide range of the plasma shape. Is is also important to separate the effects of rotation and ripple loss on the pedestal structure and ELMs. As for the core confinement of H-mode plasmas, both DIII-D and JT-60U have shown improved performance with co-directed rotation compared with counter rotation. The purpose of this study is to clarify the roles of plasma rotation systematically by utilizing the unique capability of rotation control with co- and counter- NBs in DIII-D and JT-60U and by utilizing the difference in the plasma shape and edge stability between the two tokamaks.
Resource Requirements:	1 day experiment, CO and Counter NB injection
Diagnostic Requirements:	Standard set & pedestal diagnostics. In particular Charge Exchange Recombination Spectroscopy.
Analysis Requirements:
Other Requirements:

Title	211: Role of ECRF on toroidal rotation profile
Name:	Yoshida none	Affiliation:	JAEA
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	Dr. P Gohil, Dr. J. DeGrassie, Dr. W. Solomon	ITPA Joint Experiment :	No
Description:	In order to understand the formation mechanism of the toroidal rotation velocity (Vt) profile with ECRF, the response of Vt profile and the momentum transport coefficients are investigated.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	To investigate the dependency of the EC deposition radius and the local behavior of Vt, EC deposition scan (for example r/a~0.3, 0.45, 0.6) is demonstrated. To separate the effects of the momentum transport and intrinsic rotation, Vt of the target plasmas is scanned as CO, BAL, CTR-rotation. To evaluate the momentum diffusivity and pinch term, magnetic perturbative experiments in both EC injected and non-injected plasmas are carried out.
Background:	In JT-60U, we obtained some results in H-mode plasmas using the fundamental O-mode EC wave as follows. ECRF drives the CO-intrinsic rotation inside the EC deposition and drives the CTR-intrinsic rotation outside the EC deposition. ECRF degrades both the momentum and thermal confinements. The core Vt and Ti reduce with ECRF (EC deposition r/a~0.3) in positive shear plasmas. However, for understanding the formation mechanism of Vt profile with ECRF, we need more experiments and analysis.
Resource Requirements:	1 day experiment. NBI: CO and BAL injections ECH: ~3 MW (~6 gyrotrons), ~3 second
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
Analysis Requirements:	magnetic perturbations for momentum transport study
Other Requirements:

Title	212: Transport during transients in ITER
Name:	Alberto none	Affiliation:	ITER
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to investigate the changes in plasma parameters (and associated power fluxes to PFCs, see associated proposal on divertor re-attachment by R. Pitts) during some of the confinement transients expected in ITER, namely : controlled and uncontrolled ELMs, L-H transitions and H-L transitions and collapses of regimes with confinement in excess of H-mode (ITBs and hybrids)	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Some of the objectives in this proposal do not actually require additional experiments while others do. The conditions to be explored would be of Ip ~ 1.0-1.2 MA q95 = 3-5 (from H-mode to ITB or advanced) with an input power over H-mode threshold of < 2 (for the standard scenario) or as required for the hybrid scenario, all NBI heated dominated.: a) The target ELM energy loss to emulate uncontrolled ELMs in ITER both for standard scenario and advanced/hybrid scenarios would be DWELM/Wped > 0.15 and for controlled ELMs DWELM/Wped < 0.05 (or lower if possible), I guess that the ELM part does not require any experiment in addition but to get good ELM/pedestal measurements and power flux measurements in other experiments already planned. b) L-H and H-L confinement transients. Here one would need to investigate the changes of plasma density and temperature (core and pedestal) for various rates of additional heating increases and fuelling rates (to emulate several rates of density increase/decrease (for L-H and H-L). Two starting points for the H-L transition should be studied a normal H-mode and a hybrid scenario with high beta. In the second phase of the experiments we would select some fuelling rates and try to emulate the expected change of alpha heating in ITER during these phases by scaling the NBI additional heating (possibly with wplasma^2 or ne^2) taking into account some delay to account the expected thermalisation of alphas in ITER c) Collapses of advanced confinement regimes. For this study we would need to take a hybrid scenario discharge (or ITB) with b as high as possible and force a transition back to â??normalâ?? H-mode operation or Type III ELMy H-mode with beta_p/beta_N ~ 0.7/2.0 or lower (for Type III). This should be probably done in two ways : a) by a sudden decrease of the input power or b) by changing the current profile (decreasing q95) to trigger this transition by growth of NTMs
Background:	The requirements for plasma position control in ITER are very stringent as contact of the plasma with some plasma facing components could damage them and/or lead to a disruption. A key ingredient to determining the plasma movements in ITER is the evolution of the plasma temperature and density (and thus plasma energy and alpha heating) following confinement transients. These transients are of common occurrence in tokamaks but in many cases good measurements are not available.
Resource Requirements:	DIII-D tokamak with cryo-pumping and NBI heating
Diagnostic Requirements:	Core and pedestal measurements and power/particle fluxes to PFCs
Analysis Requirements:	Analysis of experimental data and comparsion with transport models
Other Requirements:

Title	213: NSTX/DIII-D TAE avalanche and rsAE simularity experiment
Name:	Fredrickson none	Affiliation:	PPPL
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	Neal Crocker, Bill Heidbrink	ITPA Joint Experiment :	No
Description:	Revisit NSTX simularity discharges previously developed to compare TAE avalanche thresholds, Alfven Cascade beta scaling, and make polarization measurements of CAE.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce condition developed previously for TAE aspect ratio scaling experiment (e.g., 120190 etc). Lower density to reduce beta below stabilization threshold for Alfven Cascades (rsAE) as was done in NSTX experiments. Find condition where TAE present and do power scan to find threshold for TAE excitation using parameters from DIII-D shot 75346 as a guide. Attempt to reach threshold for TAE avalanching. CAE will likely be present in these shots and use new Mirnov coils to measure polarization of CAE, other modes.
Background:	Alfven Cascades seen in high field DIII-D plasmas and are suspected of fast ion redistribution. Operation at low field, and similar parameters to NSTX should put beta above threshold for "stabilization". First goal is to search for this beta-scaling of AC frequency sweeps. Second goal is to look for beta-fast threshold for onset of TAE, and then attempt to push beta-fast up to threshold for avalanch onset. Threshold might be higher in higher aspect ratio device. Finally, interesting results have been found regarding the polarization of TAE, AC and CAE on NSTX. Data from DIII-D could contribute to understanding the observed polarizations.
Resource Requirements:	Machine operation at 0.5 to 0.6 T and neutral beams at 65 - 80 kV. Possibly cryo-pumps to allow low density operation. Possibly need TF scan.
Diagnostic Requirements:	Fast Mirnov acquisition (5-10 MHz), reflectometers, BES, other fast internal mode diagnostics. MSE, CER, TS, neutrons and certainly FIDA.
Analysis Requirements:	EFIT and TRANSP. Higher level analysis by NOVA-k and/or M3D.
Other Requirements:

Title	214: Model-based Current Profile Control during the Ramp-up Phase in DIII-D
Name:	Schuster schuster@lehigh.edu	Affiliation:	Lehigh U
Research Area:	Integrated and Model-Based Control	Presentation time:	Requested
Co-Author(s):	Yongsheng Ou, Chao Xu - Lehigh University John Ferron, Tim Luce, Mike Walker, Dave Humphreys - General Atomics Tom Casper, Bill Meyer - Lawrence Livermore National Laboratory	ITPA Joint Experiment :	No
Description:	Establishing a suitable current profile has been demonstrated to be a key condition for the achievement of advanced scenarios with improved confinement and possible steady-state operation. The current approach at DIII-D focuses on creating the desired current profile during the plasma current ramp-up and early flattop phases with the aim of maintaining this target profile during the subsequent phases of the discharge. The controller used for the q evolution during the ramp-up phase is presently a simple PI (proportional-integral) algorithm with empirically determined gains. The q profile is obtained in real time from a complete equilibrium reconstruction using data from the Motional Stark Effect (MSE) diagnostic. The controller requests a power level to the actuator (electron cyclotron heating (ECH) or neutral beam heating (NBH)) which is equal to preprogrammed feed-forward value plus the error in q times a PI gain. Present limitations of this controller (oscillations and instability), the high dimensionality of the problem, and the strong coupling between the different variables describing the dynamics of the current profile of the plasma motivates the design of a model-based, multi-variable controller that takes into account the dynamics of the q response to the different actuators. The Advanced Scenario thrust is interested in developing a model based controller to be used in forming desirable current profiles during the plasma current ramp-up. Some characteristics of the problem that make it difficult are the limited actuator power, the need to avoid unstable MHD regimes, and the significant nonlinearity of the problem. The objective of this experiment is to implement model-based controllers developed for the regulation of the q profile evolution during the early phase of the discharge, including ramp-up and beginning of the flattop, with the ultimate goal of achieving a desired target profile at some time during the first part of the flattop phase. The goal of the experiment is twofold: 1- It will allow for performance evaluation of optimal open-loop controllers obtained by applying optimization techniques to a CORSICA-based full predictive model. The experimental results will also be used to validate a control-oriented reduced-order model which will be used for closed-loop control synthesis. 2- It will allow for performance evaluation of optimal closed-loop controllers. The evaluation of control feasibility through the use of open-loop actuator trajectories and the validation of the control-oriented model are both key prerequisites for the following step, which is to implement a feedback controller to drive the q profile to the desired target. It is expected that closed-loop controllers will add robustness to previously tested open-loop controllers.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The CORSICA-based open-loop optimal control laws will be expressed as time trajectories for the actuators: total plasma current, average plasma density, and non-inductive current drive (NBI, ECH) power. The closed-loop controller will regulate in real-time these three actuators based on real-time measurements of poloidal flux or current. Our goal is to carry out the open-loop experiment early in the experimental campaign and the closed-loop experiment late in the experimental campaign, after the model validation and open-loop control testing experiment. For the open-loop experiment special care will be put in reproducing those initial conditions for the poloidal magnetic flux considered for the synthesis of the open-loop optimal control laws. The evolution of the poloidal magnetic flux, plasma density and plasma temperature will be used for the validation of a control-oriented model which in turn will be employed for the synthesis of closed-loop controllers. We will assess the ability of the combined open-loop and closed-loop controllers to drive the current profile from an initial condition different from (but close to) the nominal one to a specific target profile. Different initial and target profiles will be considered mainly in L-mode but we also intend to carry out part of the experiment in H-mode.
Background:	The control group at Lehigh University (LU) headed by Prof. Eugenio Schuster has been working on this problem for more than two years. A preliminary first-principle control-oriented model of current profile evolution in response to auxiliary heating and current drive systems (NBI, EC) and electric field due to induction was developed for the plasma current ramp-up and early-flattop phases in 2006 [1]. In 2007, an optimal open-loop "extremum-seeking" control scheme was developed based on the simplified control-oriented model [2]. The extremum-seeking algorithm predicts 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 1/2-day experiment showed: 1- qualitative agreement with the q profile evolution predicted by the simplified model, 2- the implementation of the open-loop actuator trajectories obtained during the extremum seeking procedure was feasible, which indicates that the actuators constraints were correctly taken into account during the control synthesis, 3- success in achieving monotonic target profiles with positive and near-zero shear near the axis. Nevertheless, reversed shear target profiles could not be achieved. This motivated the use of CORSICA (full predictive model instead of simplified control-oriented model) for the development of extremum-seeking controllers and further refinement of the simplified control-oriented model which is currently being used for the development of closed-loop control techniques. A reduced order model is obtained from the original simplified control-oriented infinite-dimensional model and combined with Optimal Control theory to synthesize closed-loop controllers. Based on initial results obtained in simulation studies, it is anticipated that the scheme can play an important role in experiments at the DIII-D tokamak. The development of model-based current profile controllers aims at saving long trial-and-error periods of time currently spent by fusion experimentalists trying to manually adjust the time evolutions of the actuators to achieve the desired current profile at some pre-specified time during the early flattop phase. [1] Y. Ou, T.C. Luce, E. Schuster, J.R. Ferron, M.L. Walker, C. Xu, and D.A. Humphreys, Towards Model-based Current Profile Control at DIII-D, Fusion Engineering and Design 82 (2007) 1153-1160. [2] Y. Ou, C. Xu, E. Schuster, T.C. Luce, J.R. Ferron, M.L. Walker and D.A. Humphreys, Extremum-Seeking Open-Loop Optimal Control of Plasma Current Profile at the DIII-D Tokamak, Plasma Physics and Controlled Fusion, 50 (2008) 115001.
Resource Requirements:	Machine time: 2 day (1 day: open-loop experiment, 1 day: closed-loop experiment) Note: Some coordination with the Steady-State Scenario group might allow use of piggybacks or individual shots on their experimental days.
Diagnostic Requirements:	Core and tangential Thomson, CER, CO2, magnetics, MSE, ECH diagnostics, a reasonable set of fast ion instability diagnostics (UF interferometers, FIR scattering, ECE at 500 kHz, fast magnetics with fast delay set in the current ramp), FIDA. For closed-loop experiment real-time magnetic measurements and equilibrium reconstruction including the q-profile (EFIT and RTEFIT with MSE) are essential, and real-time measurements of the ion and electron temperature profiles, as well as line-averaged plasma density or density profile are required.
Analysis Requirements:	--
Other Requirements:	--

Title	215: Model-based Current and Kinetic Profile Control during the Flattop Phase in DIII-D
Name:	Schuster schuster@lehigh.edu	Affiliation:	Lehigh U
Research Area:	Integrated and Model-Based Control	Presentation time:	Requested
Co-Author(s):	Chao Xu, Yongsheng Ou - Lehigh University Didier Moreau, Didier Mazon - CEA, France John Ferron, Tim Luce, Mike Walker, Dave Humphreys - General Atomics Tom Casper, Bill Meyer - Lawrence Livermore National Laboratory	ITPA Joint Experiment :	No
Description:	Establishing a suitable current profile has been demonstrated to be a key condition for the achievement of advanced scenarios with improved confinement and possible steady-state operation. The current approach at DIII-D focuses on creating the desired current profile during the plasma current ramp-up and early flattop phases with the aim of maintaining this target profile during the subsequent phases of the discharge. A closed-loop controller is necessary to regulate the current and kinetic profiles around the target values during the flattop. The objective of this experiment is to implement model-based controllers developed for the regulation of the current profile and temperature profile evolutions during the flattop phase of the discharge. The goal of the experiment is twofold: 1- Identification from data of a control-oriented reduced-order model for the evolution of the q and Te profiles. The identified dynamic model will be used for closed-loop control synthesis. 2- Synthesis of a model-based controller for combined control of the current and temperature profiles. The multivariable, model-based controllers to be developed within this project differ from non-model-based, empirically-tuned, PID (proportional-integral-derivative) controllers, in two distinctive aspects: 1- knowledge of the system (identified model) is incorporated during the synthesis of the controller, ii- the relationships among all input and output variables are taken into account during the synthesis of the controller. These two distinctive aspects are indeed the reasons for which improved performance is expected from advanced multivariable model-based controllers. Indeed, the strong coupling between the different physical variables involved in the plasma transport phenomenon and the high complexity of its dynamics make unavoidable the use of information of the to-be-controlled system, i.e., dynamic models, during the synthesis of plasma profile controllers. It is important to emphasize at this point that the PCS (plasma control system) at DIII-D does have infrastructure for implementing such advanced controllers.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The first task will be the selection of feasible q and Te nominal profiles which are not too close to power and performance limits, and whose realizations require real-time closed-loop control. For the system identification of a dynamic model for the q and Te responses, the total plasma current, average plasma density, and non-inductive current drive and heating (NBI, ECH, etc.) power will be excited around the nominal trajectories. The dynamic response data will be used to identify a state-space model for the evolution of q and Te using subspace identification techniques. The identified model will be used for the synthesis of a reduced-order controller that will exploit the time-scale separation between kinetic and magnetic variables and will optimally regulate the q and Te profiles around the nominal values. The plan requires to carry out the open-loop system identification experiment early in the experimental campaign and the closed-loop control experiment late in the experimental campaign, after the dynamic model is obtained and the closed-loop controller is synthesized.
Background:	A group of researchers at JET, including Didier Moreau and Didier Mazon, have been working for more than five years now on the development of model-based controllers for the regulation of an equilibrium profile during the flattop phase of the discharge. Different current and temperature gradient target profiles have been reached and sustained for several seconds at JET during the flattop current phase. The control schemes rely on the experimental identification of linearized static and dynamic response models, using lower hybrid current drive (LHCD), ion cyclotron resonance heating (ICRH) and neutral beam injection (NBI) as actuators. The controller designed based on a static response model, which finally reduces to a proportional integral regulator incorporating information of the static response of the system, has been shown effective when rapid plasma events are absent. If the controller is expected to respond to rapid transients, such as MHD phenomena, which may displace the system on a short timescale during the slow evolution of the current density profile towards its desired shape, information of the dynamic response of the system must be incorporated into the controller synthesis. Exploiting the different time scales of kinetic and magnetic variables, a dynamic model has been recently identified at JET and used for the synthesis of a two-timescale controller [1]. 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 [2]. Further experiments are needed to identify a dynamic model for both q and Te profile evolutions. A reduced-order state-space model obtained from data using subspace identification techniques can be combined with Optimal and Robust Control theory to synthesize closed-loop controllers that optimally regulate both the q and Te profiles around target values. Based on initial results obtained in simulation studies, it is anticipated that the scheme can play an important role in experiments at the DIII-D tokamak. [1] D. Moreau et al., A two-time-scale dynamic-model approach for magnetic and kinetic profile control in advanced tokamak scenarios on JET, Nucl. Fus. 48 (2008) 106001. [2] C. Xu, Y. Ou, E. Schuster, J. Ferron, T.C. Luce, M.L. Walker, D.A. Humphreys, T.A. Casper, W.H. Meyer, Current Profile Evolution Modeling via Subspace Identification Algorithms, 50th Division of Plasma Physics (DPP) Annual Meeting of the American Physical Society (APS), Dallas, Texas, November 17-21, 2008.
Resource Requirements:	Machine time: 2 days (1 day: model identification, 1 day: closed-loop control experiment) Note: Some coordination with the Steady-State Scenario group might allow use of piggybacks or individual shots on their experimental days, particularly for the model identification phase.
Diagnostic Requirements:	Core and tangential Thomson, CER, CO2, magnetics, MSE, ECH diagnostics, a reasonable set of fast ion instability diagnostics (UF interferometers, FIR scattering, ECE at 500 kHz, fast magnetics with fast delay set in the current ramp), FIDA. For closed-loop experiment real-time magnetic measurements and equilibrium reconstruction including the q-profile (EFIT and RTEFIT with MSE) are essential, and real-time measurements of the ion and electron temperature profiles, as well as line-averaged plasma density or density profile are required.
Analysis Requirements:	--
Other Requirements:	--

Title	216: Shape effects on li during ITER-like current ramp down
Name:	Casper thomas.casper@iter.org	Affiliation:	ITER Organization
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):	G. Jackson, D. Humphreys, T. Luce	ITPA Joint Experiment :	No
Description:	Study the effect of shaping, primarily elongation, on the variation of li and the vertical growth rate during current ramp down in the ITER shape	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Run ITER ELMing H-mode demo discharges at scaled ITER nominal performance level and full scaled ITER current. During the plasma current ramp down, vary the elongation and possibly the triangularity to determine a range of variation of li and vertical instability growth rate that can be expected in ITER. Repeat for a few different stored energy settings to determine sensitivity.
Background:	ITER is in the process of developing scenarios to safely ramp down from 15MA without disruptions or loss of control. The main concern is that when coming out of H-mode the current narrows (loss of edge bootstrap current), li rises and the plasma may be difficult to control. One approach is to use shape variation to control li and maintain stability and controllability. ITER has requested data to develop rampdown techniques and controllers.
Resource Requirements:	Baseline ELMing H-mode ITER demo discharge operation. May expand to use ITER hybrid mode conditions as well. This proposal only needs access and control of the current ramp down phase of the shot. The ramp up and current flattop phases are not affected and can be used for other purposes. However, most of the discharges should be run at or near the nominal scaled performance level for ITER.
Diagnostic Requirements:	Density, temperature and MSE measurements to determine H-mode profiles including the current density profile.
Analysis Requirements:	Determination of li and vertical instability growth rate
Other Requirements:	--

Title	217: Heating, H-to-L transition and limited plasma effects on li during ITER-like current ramp down
Name:	Casper thomas.casper@iter.org	Affiliation:	ITER Organization
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):	G. Jackson, D. Humphreys, T. Luce	ITPA Joint Experiment :	No
Description:	Study the effects of heating, H-to-L transition at various plasma currents, and limiting the plasma on the variation of li and the vertical growth rate during current ramp down in the ITER shape	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Run ITER ELMing H-mode demo discharges at scaled ITER performance level and full scaled ITER current. During the plasma current ramp down, vary the heating power to change the time of the H-to-L transition and obtain a database of li and vertical instability growth rate changes with heating power and plasma current at the H-to-L transition. Repeat a sequence of shots to vary the time of onset of the H-to-L transition by forcing the plasma into L-mode by biasing the vertical position upwards for the ITER LSN shape. Repeat a sequence of shots to force the plasma into limited operation. The final result will be a database of variations of li and growth rate suitable for benchmarking ITER ramp down scenarios.
Background:	ITER is in the process of developing scenarios to ramp down from 15MA without disruptions or loss of control. The main concern is that when coming out of H-mode the current narrows (loss of edge bootstrap current), li rises and the plasma may be difficult to control. Control of the H-to-L transition via heating and/or limiting the plasma are techniques proposed to limit the rise in li and growth rate to maintain controllability. ITER has requested data to benchmark ramp down scenarios and develop controllers.
Resource Requirements:	Baseline ELMing H-mode ITER demo discharge operation. May expand this to include ITER hybrid-mode discharges. This proposal only needs access to and control of the current ramp down phase of the shot. The ramp up and current flattop phases are not affected and can be used for other purposes. However, most of the discharges should be run at or near the nominal scaled performance level for ITER.
Diagnostic Requirements:	Density, temperature and MSE measurements to determine H-mode profiles including the current density profile.
Analysis Requirements:	Determination of li and vertical instability growth rate
Other Requirements:	--

Title	218: Generation of ring of relativistic electrons
Name:	Prater prater@fusion.gat.com	Affiliation:	GA
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	C.C. Petty, M. Austin, R. Harvey, E. Hollman, G. Guest	ITPA Joint Experiment :	No
Description:	Generate a thin ring of very energetic (significantly relativistic) electrons using the ECH system. This may have a variety of practical applications: affecting plasma stability, diagnosing transport, creation of transport barriers, flow shear modification, modification of the radial electric field, and so on, but of course these are very speculative. Because these applications are so speculative, this proposal is being entered as a Torkil Jensen Award. Also, this experiment would be a nice test of the science of ECH and Fokker-Planck theory.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply second harmonic O-mode exactly perpendicular to the local magnetic field where the rays intersect the resonance. Then the only broadening is due to the relativistic mass shift. The X2 mode is still very strongly absorbed, so it is mostly on not-very-energetic electrons. But the O2 mode is more weakly absorbed, by a factor Te/mc2, so it will penetrate to more relativistic electrons. Being that there are fewer of them, there will be a large power per particle deposited in a small number of electrons. This should generate a runaway condition in energy for these electrons. These electrons should be visible on the SXR, HXR, and ECE diagnostics.
Background:	We have a lot of experience with the X2 mode but practically none with the O2 mode, particularly for perpendicular launch. Calculations using the TORAY code show that in a high performance-type discharge (122907, an NTM shot) the absorption of the O2 mode is nearly complete, so the experiment can be done safely without endangering the inner wall.
Resource Requirements:	Five gyrotrons should be available.
Diagnostic Requirements:	ECE, HECE, SXR, HXR, plus the usual TS and MSE diagnostics.
Analysis Requirements:	The principal analysis will be done using the CQL3D code.
Other Requirements:	--

Title	219: Rotation Dependence Of Hybrid Tearing Beta Limit & ECRH Control Requirements
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Holger Reimerdes, Ted Strait	ITPA Joint Experiment :	Yes
Description:	The principal beta limit for the hybrid scenario comes from the 2/1 NTM. Decreasing plasma rotation is known to lower the 2/1 NTM beta limit in ITER-like baseline scenarios from values of betan ~3 in co injected plasmas, to ~2 in balanced beam plasmas. Hybrid scenarios are reliant on high betan access to achieve long pulse and high performance goals, but ITER will have a relatively low Alfven-normalised plasma rotation. Thus it is important to assess how the 2/1 NTM betan limit changes with rotation in hybrid scenarios, in order to understand whether control is needed and how effective An important second step is to assess how this behaviour changes with ECCD mode control. This is important, because if the tearing mode destabilisation at low rotation originates from changes in classical tearing stability, then removing the bootstrap hole with ECCD will not be so effective (it will stop bootstrap amplification of island but leave residual delta prime driven island) �?? it is important to gauge the effect and the size of the residual island.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish robust regime at intermediate betan. Adjust to a target initial torque value. Ramp up heating power until 2/1 NTM. Complementary supportive data can also be obtained by ramping down torque at different high levels of betan. Extend scan by performing comparison pulses with ECCD (exploring pre-emptive application and/or responsive after mode struck) to see how this changes the threshold for mode and/or whether it leaves a residual island at low rotation. These studies may good synergy (and share shot development) with attempts to develop ITER relevant low rotation hybrid scenarios.
Background:	Basis for a rotation effect is well established, both theoretically and experimentally. Parameter dependencies from DIII-D show the effects to act in the parameter range of concern. Phenomenology of 2/1 NTM onset suggests it can often be related to a change in intrinsic tearing stability, implying ECCD control requirements may be less effective, or may need to act differently than previously expected. Performing these experiments is therefore important to understanding the control and limit issues for the scenario �?? particularly what residual islands can be tolerated, and whether modes can be completely removed. Learning about the physics mechanism in this way is also important for rho* extrapolating the NTM onset threshold.
Resource Requirements:	SND ITER like hybrid, normal Ip direction. 4 co beams and 2 counter beams. ECCD with real time adjustment (to plasma) systems.
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be self evident in intershot parameter, but further analysis will be executed to explore trends in rotation and local �??NTM�?? parameters.
Other Requirements:	--

Title	220: Rotation Dependence Of Hybrid Tearing Beta Limit & ECRH Control Requirements
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Holger Reimerdes, Ted Strait	ITPA Joint Experiment :	Yes
Description:	The principal beta limit for the hybrid scenario comes from the 2/1 NTM. Decreasing plasma rotation is known to lower the 2/1 NTM beta limit in ITER-like baseline scenarios from values of betan ~3 in co injected plasmas, to ~2 in balanced beam plasmas. Hybrid scenarios are reliant on high betan access to achieve long pulse and high performance goals, but ITER will have a relatively low Alfven-normalised plasma rotation. Thus it is important to assess how the 2/1 NTM betan limit changes with rotation in hybrid scenarios, in order to understand whether control is needed and how effective An important second step is to assess how this behaviour changes with ECCD mode control. This is important, because if the tearing mode destabilisation at low rotation originates from changes in classical tearing stability, then removing the bootstrap hole with ECCD will not be so effective (it will stop bootstrap amplification of island but leave residual delta prime driven island) �?? it is important to gauge the effect and the size of the residual island.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish robust regime at intermediate betan. Adjust to a target initial torque value. Ramp up heating power until 2/1 NTM. Complementary supportive data can also be obtained by ramping down torque at different high levels of betan. Extend scan by performing comparison pulses with ECCD (exploring pre-emptive application and/or responsive after mode struck) to see how this changes the threshold for mode and/or whether it leaves a residual island at low rotation. These studies may good synergy (and share shot development) with attempts to develop ITER relevant low rotation hybrid scenarios.
Background:	Basis for a rotation effect is well established, both theoretically and experimentally. Parameter dependencies from DIII-D show the effects to act in the parameter range of concern. Phenomenology of 2/1 NTM onset suggests it can often be related to a change in intrinsic tearing stability, implying ECCD control requirements may be less effective, or may need to act differently than previously expected. Performing these experiments is therefore important to understanding the control and limit issues for the scenario �?? particularly what residual islands can be tolerated, and whether modes can be completely removed. Learning about the physics mechanism in this way is also important for rho* extrapolating the NTM onset threshold.
Resource Requirements:	SND ITER like hybrid, normal Ip direction. 4 co beams and 2 counter beams. ECCD with real time adjustment (to plasma) systems.
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be self evident in intershot parameter, but further analysis will be executed tom explore trends in rotation and local �??NTM�?? parameters.
Other Requirements:	--

Title	221: Two Point q=3.1 "ITER Reference" Scans For Baseline Scenario Beta And Error Field Limits
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Demonstration Discharges	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Ted Strait, Holger Reimerdes	ITPA Joint Experiment :	Yes
Description:	Experiments on DIII-D have established a decreasing tearing mode betan limit and increasing error field sensitivity as rotation is reduced for ITER-like SND baseline scenarios at q95=4.4. This q95 was chosen to allow a subsequent mode study phase for other combined experiments. It is important to see what sort of limits this implies at the ITER operational point of q95=3.1. Two two-point experiments are proposed each with a shot at strong co-torque and near balanced beam injection to assess the range of the rotation effect. In the first power is to be ramped to determine the 2/1 NTM betan limit vs rotation. In the second, error fields are to be ramped at the ITER betan value (~1.9) to determine error field threshold	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish robust regime at intermediate betan. Adjust to a target initial torque value. Ramp up heating power until 2/1 NTM for first pair of discharges (at high and low torque). Repeat with fixed betan=1.9 and ramp I240 n=1 error field to measure mode.
Background:	The physics effects here are well established from previous DIII-D experiments - the aim is to provide some quantitative assessment at the ITER-like q95 value, which would be expected to influence tearing stability. This data will also make a nice comparison set for q95 dependence with previous data (q95=4.4) and 2009 proposal for NSTX NTM comparison discharges at q95=7)
Resource Requirements:	SND ITER like hybrid, normal Ip direction. 4 co beams and 2 counter beams. I coils 240 configuration for two error field points.
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be self evident in intershot parameter, but further analysis will be executed tom explore trends in rotation and local �??NTM�?? parameters.
Other Requirements:	--

Title	222: Two Point q=3.1 "ITER Reference" Scans For Baseline Scenario Beta And Error Field Limits
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Ted Strait, Holger Reimerdes	ITPA Joint Experiment :	Yes
Description:	Experiments on DIII-D have established a decreasing tearing mode betan limit and increasing error field sensitivity as rotation is reduced for ITER-like SND baseline scenarios at q95=4.4. This q95 was chosen to allow a subsequent mode study phase for other combined experiments. It is important to see what sort of limits this implies at the ITER operational point of q95=3.1. Two two-point experiments are proposed each with a shot at strong co-torque and near balanced beam injection to assess the range of the rotation effect. In the first power is to be ramped to determine the 2/1 NTM betan limit vs rotation. In the second, error fields are to be ramped at the ITER betan value (~1.9) to determine error field threshold	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish robust regime at intermediate betan. Adjust to a target initial torque value. Ramp up heating power until 2/1 NTM for first pair of discharges (at high and low torque). Repeat with fixed betan=1.9 and ramp I240 n=1 error field to measure mode.
Background:	The physics effects here are well established from previous DIII-D experiments - the aim is to provide some quantitative assessment at the ITER-like q95 value, which would be expected to influence tearing stability. This data will also make a nice comparison set for q95 dependence with previous data (q95=4.4) and 2009 proposal for NSTX NTM comparison discharges at q95=7)
Resource Requirements:	SND ITER like hybrid, normal Ip direction. 4 co beams and 2 counter beams. I coils 240 configuration for two error field points.
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be self evident in intershot parameter, but further analysis will be executed tom explore trends in rotation and local �??NTM�?? parameters.
Other Requirements:	--

Title	223: Two Point q=3.1 "ITER Reference" Scans For Baseline Scenario Beta And Error Field Limits
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Ted Strait, Holger Reimerdes	ITPA Joint Experiment :	Yes
Description:	Experiments on DIII-D have established a decreasing tearing mode betan limit and increasing error field sensitivity as rotation is reduced for ITER-like SND baseline scenarios at q95=4.4. This q95 was chosen to allow a subsequent mode study phase for other combined experiments. It is important to see what sort of limits this implies at the ITER operational point of q95=3.1. Two two-point experiments are proposed each with a shot at strong co-torque and near balanced beam injection to assess the range of the rotation effect. In the first power is to be ramped to determine the 2/1 NTM betan limit vs rotation. In the second, error fields are to be ramped at the ITER betan value (~1.9) to determine error field threshold	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish robust regime at intermediate betan. Adjust to a target initial torque value. Ramp up heating power until 2/1 NTM for first pair of discharges (at high and low torque). Repeat with fixed betan=1.9 and ramp I240 n=1 error field to measure mode.
Background:	The physics effects here are well established from previous DIII-D experiments - the aim is to provide some quantitative assessment at the ITER-like q95 value, which would be expected to influence tearing stability. This data will also make a nice comparison set for q95 dependence with previous data (q95=4.4) and 2009 proposal for NSTX NTM comparison discharges at q95=7)
Resource Requirements:	SND ITER like hybrid, normal Ip direction. 4 co beams and 2 counter beams. I coils 240 configuration for two error field points.
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be self evident in intershot parameter, but further analysis will be executed to explore trends in rotation and local �??NTM�?? parameters.
Other Requirements:	--

Title	224: ITB torque scan
Name:	Greenfield greenfieldcm@ornl.gov	Affiliation:	ORNL
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Scan ITB discharges from full co-NBI to full counter-NBI, including balanced.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Two half days, one co and one counter IP. Scan ITB discharges (preferably L-mode edge) from 4 co/0 ctr to 2 co/2 ctr and 3 co/0 ctr to 1 co/2 ctr. The same scans should be done in reverse with counter-NBI.
Background:	In 1999 we established that counter-NBI driven ITBs (with pressure gradient dominated ExB shear) could be broader than with co-NBI (rotation dominated). Our analysis at that time suggested that balanced NBI would have similar characteristics. This experiment will also test the differences in balanced NBI discharges with forward and reversed IP. Although the physics to be studied here was one of the motivations for reversing the 210 beamline, the experiment has yet to be done.
Resource Requirements:	2x0.5 days of experimental time, with both plasma current polarities. Could be paired with proposal 226.
Diagnostic Requirements:	All of the usual transport diagnostics
Analysis Requirements:	Both toroidal and poloidal rotation; TRANSP
Other Requirements:	--

Title	225: Error Field Sensitivity Of Low Torque Intermediate Betan Plasmas
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Rob la Haye, Ted Strait, Holger Reimerdes	ITPA Joint Experiment :	Yes
Description:	The error field sensitivity of ITER-like baseline scenarios has already been explored as a function of rotation at betan=1.9 in co-injected plasmas, and at lower betan values with counter rotation. The data suggests a dependence on rotation and beta value, although the exact form of this is not well understood. In particular the beta dependence is not clear (do error field thresholds fall with betan rises in this range?), and it is important to understand the action of the error fields (do the error fields act to change stability by modifying rotation? Or are there increased error field amplification effects with proximity to ideal or tearing mode beta limits?). Simple experiments are proposed to get at this question by inducing modes with error fields at modest betan and co-rotation values.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Perform error field ramps to strike a mode at various values of betan between 1 and 2 in plasmas with a net low but positive torque injection. Also explore varying torque value modestly between shots.
Background:	This experiment represents a simple extension of previous studies to determine the key controlling parameters. This is important to understand , so that the requirements for ITER can be interpreted �?? does it need to maintain rotation? How much does it need to minimise error fields? Is operation well below tearing or ideal beta limit important?
Resource Requirements:	2 co and counter beams. Standard ITER-like plasma scenarios matched to previous reference configuration from 2008 studies. I coils 240 configuration.
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be self evident in intershot parameter, but further analysis will be executed to explore trends in rotation and local NTM parameters.
Other Requirements:	--

Title	226: ITB dynamics while changing between pressure gradient and rotation dominated ExB shear
Name:	Greenfield greenfieldcm@ornl.gov	Affiliation:	ORNL
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Reproduce the TFTR experiments reported in Synakowski's PRL: Start from an ITB with balanced-NBI, and rapidly shift to fully co-NBI.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Form an ITB with balanced NBI. Turn off the two counter NBI sources and turn on two co-NBI sources and fully document the dynamics. Repeat the same procedure in reverse.
Background:	Synakowski's PRL showed some very interesting dynamics as the balanced NBI ITB was lost when the pressure gradient and rotation terms of the ExB shear cancelled and then was regained as a rotation dominated ITB. The diagnostic set now available on DIII-D will allow much more complete study of these dynamics. Although the physics to be studied here was one of the motivations for reversing the 210 beamline, the experiment has yet to be done.
Resource Requirements:	2x0.5 days of experimental time, with both plasma current polarities Could be paired with proposal 224.
Diagnostic Requirements:	All of the usual transport diagnostics
Analysis Requirements:	Both toroidal and poloidal rotation; TRANSP; fluctuation diagnostics
Other Requirements:	--

Title	227: Scaling Of Baseline Scenario Error Field Sensitivity Towards ITER
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Rob la Haye, Ted Strait, Holger Reimerdes	ITPA Joint Experiment :	Yes
Description:	To extrapolate error field sensitivity to ITER it is important to measure density and toroidal field dependencies. Previous studies have used a dimensional constraint [Connor] applicable to Ohmic regimes to infer cross machine scalings. The constraint will change in heated discharge, which have another degree of independence, and so the exponents in the extrapolation might be expected to change. Given recent observations of high error field sensitivity in low rotation ITER-like baseline scenarios, it is important to understand how this will extrapolate to ITER Bt, ne and R values. Therefore it is proposed to measure error field threshold dependence vs ne and Bt at an otherwise fixed operational point (fixed q95, shape, betan, and normalised rotation), while using a modified Connor invariance to infer machine size scaling.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish betan=1.9 modest rotation discharge. Ramp up error field to induce a mode. Repeat at different densities, adjusting torque to obtain same (Alfven) normalised rotation. Repeat at different Ip and Bt values (same q95).
Background:	Scaling of error field thresholds towards ITER in heated plasmas has never been established.
Resource Requirements:	2 co and counter beams. Standard ITER-like plasma scenarios matched to previous reference configuration from 2008 studies. I coils 240 configuration.
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be clear from parameter scalings
Other Requirements:	--

Title	228: One Shot 2/1 Island Size Dependence On Rotation
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Ted Strait	ITPA Joint Experiment :	Yes
Description:	A rotation dependence in 2/1 NTM onset has been observed. It has been suggested this is due to changes in underlying tearing stability (rather than, say triggering physics). An easy way to test this is to observe behaviour of a saturated mode as rotation is changed at constant betan. This is readily achievable as a one shot studying DIII-D, but has not been done previously due to desire to piggy back mode control and beta ramp-down studies. The results will readily tell what sort of percentage effect rotation has on the island drive.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1 shot: Strike high rotation 2/1 mode at 'safe' q95 (high value to avoiding locking). Vary beam torque in steps to see effect on mode size. (This may well be piggy back to NSXT comparison NTM beta ramp-down studies or ECCD mode control studies.)
Background:	Easy way to get a key physics question.
Resource Requirements:	4 beams and an NTM inducing shot at highish q95.
Diagnostic Requirements:	CER, MSE, Magnetics, TS are important
Analysis Requirements:	Trivial, though detailed analysis of profile evolution may prove interesting
Other Requirements:	--

Title	229: Disruption forces
Name:	Greenfield greenfieldcm@ornl.gov	Affiliation:	ORNL
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Measure vessel forces during a disruption	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	I'm not sure whether we have the capability to do this now, or might for 2010. I am putting this in because it's an important issue for ITER and if we can't do it now, we should think about how to do it.
Background:	Identified as urgent ITER issue.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	230: Looking For The Upturn Of NTM Thresholds With Strong Counter Rotation To Test NTM Physics
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Requested
Co-Author(s):	Ted Strait, Rob La Haye, Gary Jackson, John De Grassie	ITPA Joint Experiment :	Yes
Description:	In previous DIII-D studies a fall in 2/1 NTM betan thresholds was established first from co to balanced beam injection, and then a further fall with increasing counter injection. This latter aspect is novel and interesting, with theoretical models not accounting for the co/counter asymmetry of the apparently robust trend. A key element in resolving the physics is whether there will be an 'upturn' with beta thresholds rising at higher counter rotation. Such a regime was not successfully tested in previous experiments due to changes in plasma profiles at strong counter rotation. Therefore it is proposed to revisit this issue, deploying further techniques (machine conditioning, q95 value, changes to fuelling, Ip and heating ramp) to maintain good quality H modes. This is useful in resolving the rotation and NTM onset mechanisms, and so extrapolation of mode drives towards ITER.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish target regime at intermediate betan and counter rotation. Optimise configuration for low density good H modes, raise counter torque to maximum while slowly raising total heating power. Repeats adjust torque vs betan trajectory.
Background:	This is an extension of previous scans to help understand a key physics effect.
Resource Requirements:	*Reverse Ip* 5 "co" beams, 1 "counter"
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be clear from analysis of profiles
Other Requirements:	--

Title	231: ELM suppression by RMPs at ITER-like additional heating level
Name:	Loarte none	Affiliation:	ITER
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to perform one experiment in which ELM suppression is achieved at two levels of additional heating one providing Pinp/PL-H ~1.5 and one with Pinp/PL-H ~3.0 and compare both the required perturbation with RMPs and the width of the resonance window in both cases	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The experiment should be done taking as basis a typical high delta (0.5) plasma with Ip ~ 1.1 MA and q95 = 3.6 and employ two levels of NBI (co) which meet the requirements above. It is understood that the associated changes in ELM suppression by RMPs caused by changes in rotation and plasma beta are already assessed in a different proposal and this one would concentrate on the effects of proximity to the H-L threshold. In principle these effects could be studied in this proposal as well by adequate combination of co and counter NBI for the two power levels proposed
Background:	ITER will operate at power levels which will not exceed the H-mode threshold by a large fraction, typically Pinp/PL-H < 2. Most experiments on ELM suppression by RMP in DIII-D are carried out at levels of additional heating much higher (in terms of L-H threshold) than those required in ITER. ELM suppression in these conditions may present different behaviour both in terms of the required size of the RMP perturbation as well as of its consequences for the pedestal plasma and overall plasma confinement, which need to be assessed for ITER
Resource Requirements:	DIII-D with cryopump. NBI. I-coils
Diagnostic Requirements:	Core, pedestal and ELM diagnostics
Analysis Requirements:	Analysis of pedestal measurements and edge stability
Other Requirements:

Title	232: ELM suppression by RMPs at low and high densities and associated pedestal/divertor behaviour
Name:	Loarte none	Affiliation:	ITER
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	R. Pitts	ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to perform experiments to determine the effects of ELM suppression in both low density and high density conditions (by gas fuelling) for otherwise similar plasmas	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The experiment could be done taking as basis a typical high delta (0.5) plasma with Ip ~ 1.1 MA and q95 = 3.6 with the power/rotation conditions that allow easiest (in terms of I-coil current level) access to ELM suppression in DIII-D and then increase plasma density with gas puffing and pellets and I-coil current while avoiding Type I ELMs. The effects on pedestal plasma parameters and gradients, overall plasma confinement, scrape-off layer and divertor plasma parameters and power fluxes to the divertor must be measured. This may impose some restrictions in the configuration to be used so that divertor measurements are optimised
Background:	The physics of ELM suppression by RMP remains uncertain, in particular the role of plasma density and or collisionality on the observed results remains to be assessed for ITER as well as the overall effects on plasma confinement in both conditions. This is particularly important as ITER will operate in conditions of low collisionality and high density that cannot be simultaneously met in present experiments. Similarly the observations of power fluxes to the divertor in low density/high density conditions in DIII-D are strikingly different and it is not clear what would be the expectations for ITER
Resource Requirements:	DIII-D with cryopumping, NBI, I-coils and pellets
Diagnostic Requirements:	Core, edge (pedestal and SOL) and divertor measurements to determine parameters in pedestal, SOL and divertor
Analysis Requirements:	Meassurement analysis, edge stability analysis and edge modelling
Other Requirements:

Title	233: ELM suppression by RMPs during transient phases of discharges
Name:	Loarte none	Affiliation:	ITER
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to perform experiments that demonstrate that ELM suppression can be achieved in transient phases of DIII-D discharges within the limitations of the I-coil system	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The experiment could be done taking as basis a typical high delta (0.5) plasma with Ip ~ 1.1 MA and q95 = 3.6 and do a relatively slow NBI power ramp-up and down to have a long ELM free phase. During this phase the RMP current level would be increased/decreased with the aim to achieve a transition from H~1 H-mode (without ELMs) from/to L mode in which all ELMs are avoided or lead to very small energy losses. Another experiment would demonstrate continuous ELM suppression during various phases of the discharge in which plasma conditions are varied at constant Ip by varying input power and/or plasma density. This could be done by pre-programming the currents in the I coils or by feedback on the basis of the results obtained in the experiments on beta and rotation effects above
Background:	ELM control and/or suppression in ITER maybe required for current levels much lower than 15 MA and during transient phases : ramp-up/down and H-mode conditions at 15 MA flat top away from full performance. The present design of the ELM control system in ITER is very flexible to allow this control but a demonstration of the feasibility of this technique in these conditions is still outstanding
Resource Requirements:	DIII-D with cryopumping, NBI heating and I-coils
Diagnostic Requirements:	Pedestal, core and ELM measurements. Ideally also power fluxes should be measured
Analysis Requirements:	Analysis of edge measurements, ELM parameters, power fluxes to PFCs and edge stability analysis
Other Requirements:

Title	234: Localization of ripple-field induced fast-ion losses.
Name:	Kramer none	Affiliation:	PPPL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	G.J. Kramer (PPPL), R. Nazikian, (PPPL), R.B. White (PPPL)	ITPA Joint Experiment :	No
Description:	Investigate the ripple-lost fast-ion deposition on the first wall in the presence of a toroidally localized magnetic disturbance to bench-mark particle loss codes that are used for studying the impact of ripple fields generated by testblanket modules in ITER.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In our experiment we want to use the error field coils in different configurations (n=1, n=3, and if possible with only one coil energized to create a single localized perturbation) to determine beam-ion losses with the scintillator probes that will become available in FY-10. We like to use off-axis NBI to maximize the trapped particle fraction which is most sensitive to the field ripple.
Background:	In ITER testblanket modules will be installed to study tritium breeding. These testblankets contain a substantial amount of ferritic steel and therefore, they create large toroidally localized disturbances of the toroidal field. Such a toroidally localized ripple can affect amongst others the alpha particle confinement negatively. From alpha particle orbit-following simulations for ITER it was found that the main effect of the testblanket modules on the fusion-born alpha particle confinement is the localization of the lost particles in front of the testblanket modules. We want to verify those predictions as well as possible on a present device.
Resource Requirements:	Quiescent plasma with low plasma current or reversed shear. Off-axis NBI heating. Error field coils in various configurations.
Diagnostic Requirements:	Scintillator probes. Neutron detectors. FIDA. BES.
Analysis Requirements:	The Equilibrium reconstruction will be done with with EFIT and the fast ion transport modeling will be done with the SPIRAL and ORBIT codes.
Other Requirements:

Title	235: Impact of a Mock-up Ferromagnetic Test Blanket Module on Plasma Operations in DIII-D
Name:	Snipes none	Affiliation:	ITER Organization
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	J A Snipes, M Schaffer, L Giancarli, A Loarte	ITPA Joint Experiment :	No
Description:	A series of experiments are proposed to study the impact of a large ferromagnetic mock-up Test Blanket Module (TBM) on plasma operations in DIII-D. A large piece of ferromagnetic material covering most of a horizontal port will be mounted on rails so that it can be moved to within perhaps 10 cm of the plasma boundary or be pulled back well away from the plasma to avoid perturbing the plasma without breaking vacuum when other experiments are being performed. The experiments are aimed at measuring and quantifying the effects of the asymmetric magnetic field perturbation produced by such a mock-up TBM on plasma startup, plasma equilibrium, error field correction and locked modes, H-mode threshold, L-mode and H-mode confinement, plasma rotation, and energetic particle losses. These experiments are urgently required to make decisions for designing the TBMs on ITER. To properly mock-up the magnetic field perturbations on ITER, the mock-up TBM on DIII-D should be large enough to increase the local toroidal field ripple at the plasma boundary in front of the TBM by at least 1% when placed close to the plasma and then moved back from the plasma to vary the perturbation by a factor of two or more. The experiments should be performed over a broad range of collisionality in both L-mode and especially H-mode to determine how the effects of the TBM on confinement and plasma rotation change with collisionality. If the TBM magnetic field perturbation can be at least partially suppressed with the I-coils or C-coils, then experiments should also be performed to attempt to correct the TBM perturbation to determine how effective such corrections can be across the range of effects on plasma operations. Experiments should also be performed with input power near the H-mode threshold to determine if the TBM magnetic field perturbation has an impact on the H-mode threshold. Experiments should be performed with the ITER plasma shape and near q95 = 3, but a range of plasma currents for the same toroidal field should also be run to measure energetic particle losses and determine if there are enhanced losses due to the TBM magnetic field perturbation versus energetic particle orbit radius.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	This experiment requires construction and installation of a mock-up ferromagnetic TBM inside a horizontal port. The TBM needs to be movable from shot-to-shot under vacuum to within about 10 cm of the plasma boundary and as far back as possible so that the magnetic field perturbation can be minimized when TBM experiments are not being performed. Since the construction and installation of a mock-up TBM will require a large amount of resources and there are a large number of effects of TBMs on plasma operations, a number of experiments should be performed over several run days, though they need not be consecutive, assuming that the TBM perturbation can be made negligibly small when pulled back away from the plasma. The experiments should first be performed with the TBM up close to the plasma perhaps 5 - 10 cm from the plasma boundary to ensure that the local toroidal field ripple in front of the TBM should increase by more than 1%, where the effects on plasma operations should be measureable. Assuming the effects are easily measureable, then subsequent experiments should be performed recessing the TBM further from the plasma to reduce the perturbation by a factor of 2 to determine how the effects scale with the size of the local magnetic field perturbation at the plasma boundary.
Background:	Experiments have been performed on JET (e.g., Saibene, et al, IAEA 2008) showing that H-mode confinement is reduced in proportion to the increase in TF ripple at low collisionality, whereas at higher collisionality, the effect disappears. Non-axisymmetric fields have substantial effects on plasma rotation (e.g., Garofalo, et al., PRL 2008). Energetic particle losses are enhanced by TF ripple (e.g., Tsuzuki, et al, NF 2003). These effects are due to periodic magnetic perturbations, but the effects of a non-periodic low toroidal mode number perturbation like that of a TBM could be even larger since low n perturbations will penetrate deeper into the plasma than higher n perturbations.
Resource Requirements:	The construction and installation of a mock-up TBM on DIII-D will require substantial resources and effort. The ITER Organization intends to help support such experiments and a detailed cost sharing arrangement needs to be agreed upon.
Diagnostic Requirements:	Thermal energy and particle confinement, plasma rotation, error fields, and energetic particle confinement all need to be well diagnosed.
Analysis Requirements:	Energy and particle confinement, equilibrium, plasma rotation, error fields, and energetic particle confinement all need to be analyzed in detail. Disruption forces acting on the mock-up TBM need to be assessed in the design of the support structure for the TBM.
Other Requirements:	A moveable mock-up TBM and support structure needs to be designed, built, and installed in DIII-D.

Title	236: Characterization of H-to-L and L-to-H transitions and control in ITER shape
Name:	Casper thomas.casper@iter.org	Affiliation:	ITER Organization
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):	Characterization of H-to-L and L-to-H transitions and control in ITER shape	ITPA Joint Experiment :	No
Description:	Study the details of H-to-L and L-to-H transitions in the ITER shape during current ramps up and down and at full ITER-scaled performance and current.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Run ITER ELMing H-mode demo discharges at scaled ITER performance level and full scaled ITER current. Vary heating power to stimulate both slow and rapid H-to-L transitions and L-to-H back-transitions to study current profile evolution and controller response in the ITER shape.
Background:	There have been some differences in the controller modeling response to simulated H-to-L transitions in ITER scenario-2 conditions. Some codes indicate control and others lose control with the plasma moving upwards and inwards to hit the wall and disrupt. This appears to be dependent on the details of the current profile and, therefore, li and possibly to the effects of impurities when the plasma moves near the wall. This experiment would provide detail profile evolution for benchmarking the various models and time-dependent simulations in use to predict ITER controller and PF system performance. This data will also provide a benchmark for the different power threshold models for H-to-L and L-to-H transitions in the ITER shape.
Resource Requirements:	Baseline ELMing H-mode ITER demo discharge operation. May expand this to include ITER hybrid-mode discharges. Need adequate neutral beam and ECH power to have ITER-shape H-mode conditions and then control the heating power to explore details of the transitions. Both NBI and ECH H-modes condtions would be studied. If time permits, we would explore sensitivity of recovery (L-to-H back-transitions) dependent on the heating source available.
Diagnostic Requirements:	Density, temperature and MSE measurements to determine H-mode profiles including the current density profile. Fluctuation measurements to diagnose details of the transitions.
Analysis Requirements:	Determination of the profile evolution, equilibrium and controller response.
Other Requirements:

Title	237: Destabilisation Of Fast Particle Stabilised Sawteeth In ITER-Like Baselines For NTM Avoidance
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	Ian Chapman, Rob La Haye, Olivier Sauter	ITPA Joint Experiment :	Yes
Description:	ITER will need to deploy ECCD sawtooth destabilisation to avoid large sawteeth triggering the onset of multiple low betan NTMs with potentially large size at mode onset. Previous attempts at demonstrating this have either not been in right regime or not used the right tool. In particular, recent analysis of q=1 ICRH on JET has shown that it act as a kinetic effect rather than changing the sawtooth via magnetic shear modification (as ECCD must do in ITER). Other studies have not been in relevant ITER-like baseline with significant heating power and beta, and/or have not acted on fast particle stabilised sawteeth. Therefore a demonstration of the technique proposed for ITER (with ECCD to change local magnetic shear) is required, to ascertain whether other strategies (eg off axis beams or a move to hybrid scenario) will be required. This demonstration should start with testing the principal of whether the sawtooth can be controlled in the right regime, and extend to tracking the q=1 radius to provide a viable demonstration for ITER and to ascertain impact on NTM threshold.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A sawtoothing ELM y H mode must be established. ICRH should be applied to lengthen sawtooth periods (or possibly low density beam heated plasma may suffice). ECCD should be applied to reduce sawtooth size and increase frequency. Heating ramps should be applied to measure 3/2 and 2/1 mode onset thresholds comparing ECCD and no ECCD cases. Real time systems will need to be deployed to keep ECCD tracking q=1 surface - this may require some shot to shot optimisation.
Background:	Discussed above. See Chapman IAEA.
Resource Requirements:	4 co beams, ECRH, ICRH, real time mode targeting systems
Diagnostic Requirements:	CER, MSE, Magnetics, TS, SXR are important
Analysis Requirements:	Results should be self evident from comparison of discharges. Cutting edge modelling will be executed to test sawtooth behaviour against latest models for mode triggering and ITER implications.
Other Requirements:	--

Title	238: Quantify effects of test blanket module on performance
Name:	Greenfield greenfieldcm@ornl.gov	Affiliation:	ORNL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine potential effects of ITER TBM on fast ion confinement specifically as well as performance in a more general sense.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	I am putting this in because it's an important issue for ITER and if we can't do it now, we should think about how to do it. Two obvious approaches are (a) install a new coil or change the wiring of a segment of the I- or C-coil to impose a local magnetic field perturbation, or (b) place a piece of ferromagnetic material close to the vessel. Not sure if either is a realistic possibility in 2009-10.
Background:	Identified as urgent ITER issue.
Resource Requirements:
Diagnostic Requirements:	Fast ion diagnostics, all of the usual transport and confinement diagnostics.
Analysis Requirements:
Other Requirements:

Title	239: Quantify effects of test blanket module on performance
Name:	Greenfield greenfieldcm@ornl.gov	Affiliation:	ORNL
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine potential effects of ITER TBM on fast ion confinement specifically as well as performance in a more general sense.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	I am putting this in because it's an important issue for ITER and if we can't do it now, we should think about how to do it. Two obvious approaches are (a) install a new coil or change the wiring of a segment of the I- or C-coil to impose a local magnetic field perturbation, or (b) place a piece of ferromagnetic material close to the vessel. Not sure if either is a realistic possibility in 2009-10.
Background:	Identified as urgent ITER issue.
Resource Requirements:
Diagnostic Requirements:	Fast ion diagnostics, all of the usual transport and confinement diagnostics.
Analysis Requirements:
Other Requirements:

Title	240: Disruption Mitigation By Real Time Control Of Locked Modes
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Anders Welander	ITPA Joint Experiment :	Yes
Description:	Many disruption types involve the occurrence of a q=2 locked mode, which is instrumental in the process leading to plasma termination. It is proposed to develop, in a step wise manner, a technique to target ECCD control on such modes in real time to delay the plasma termination, thereby providing time for further systems and techniques (deshaping, Ip rampdown, gas, massive gas, etc) to act to mitigate the event, terminate the plasma more safely, or even recover to avoid the event. This should use real time systems to identify locked mode formation and location. Resonant I coil fields should then be applied to orient the q=2 island for ECCD injection, deploying real time control to target the mode radius (adjust plasma radius, Bt, and/or gyrotron deposition radius). ECCD should be applied to try to arrest the mode's growth by driving currents down the island core. This will provide time for various mitigation actions to be explored. If the island shrinks, it may even allow recovery to a spin-up mode, if the plasma can be got back to a safe state. The advantage of this technique is that as island becomes bigger, it becomes easier to see, diagnose location, and hit with the ECCD.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	In a first step a reliable disruptive process involving locked modes at low q95 (where they are likely to be disruptive) should be established, using error field ramps, or inducing rotating 2/1 NTMs and allowing them to lock. A pre-programmed event (with know radial location) is easiest to use to test whether ECCD can delay or prevent the disruption in such a case. A refinement of this should then be attempted with real time detection of the locked mode to trigger gyrotrons. At a more advanced level, one might also go on to target mode using real time MSE (not starting with gyrotrons locked on q=2). It would also be important to try this for other forms of disruption, such as density limit, low q disruption, or possibly RWM disruption processes that involve locked modes.
Background:	ITER must develop the operational techniques to minimise high force disruptions. DEMO must be disruption free.
Resource Requirements:	4 beams. ECRH. I coils. Real time systems and mode tracking. Prep work for ECRH real time targeting.
Diagnostic Requirements:	MSE, CER, TS, ECE,
Analysis Requirements:	Results should be self-evident from increased disruption timings. Analysis for ECH prior to studies.
Other Requirements:	--

Title	242: Disruption Mitigation By Real Time Control Of Locked Modes
Name:	Buttery buttery@fusion.gat.com	Affiliation:	GA
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	Rob La Haye, Anders Welander	ITPA Joint Experiment :	Yes
Description:	Many disruption types involve the occurrence of a q=2 locked mode, which is instrumental in the process leading to plasma termination. It is proposed to develop, in a step wise manner, a technique to target ECCD control on such modes in real time to delay the plasma termination, thereby providing time for further systems and techniques (deshaping, Ip rampdown, gas, massive gas, etc) to act to mitigate the event, terminate the plasma more safely, or even recover to avoid the event. This should use real time systems to identify locked mode formation and location. Resonant I coil fields should then be applied to orient the q=2 island for ECCD injection, deploying real time control to target the mode radius (adjust plasma radius, Bt, and/or gyrotron deposition radius). ECCD should be applied to try to arrest the mode's growth by driving currents down the island core. This will provide time for various mitigation actions to be explored. If the island shrinks, it may even allow recovery to a spin-up mode, if the plasma can be got back to a safe state. The advantage of this technique is that as island becomes bigger, it becomes easier to see, diagnose location, and hit with the ECCD.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	In a first step a reliable disruptive process involving locked modes at low q95 (where they are likely to be disruptive) should be established, using error field ramps, or inducing rotating 2/1 NTMs and allowing them to lock. A pre-programmed event (with know radial location) is easiest to use to test whether ECCD can delay or prevent the disruption in such a case. A refinement of this should then be attempted with real time detection of the locked mode to trigger gyrotrons. At a more advanced level, one might also go on to target mode using real time MSE (not starting with gyrotrons locked on q=2). It would also be important to try this for other forms of disruption, such as density limit, low q disruption, or possibly RWM disruption processes that involve locked modes.
Background:	ITER must develop the operational techniques to minimise high force disruptions. DEMO must be disruption free.
Resource Requirements:	4 beams. ECRH. I coils. Real time systems and mode tracking. Prep work for ECRH real time targeting.
Diagnostic Requirements:	MSE, CER, TS, ECE,
Analysis Requirements:	Results should be self-evident from increased disruption timings. Analysis for ECH prior to studies.
Other Requirements:	--

Title	243: Evaluation of Diamond as a Plasma Facing Material
Name:	Lisgo steve.lisgo@iter.org	Affiliation:	ITER Organization
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Requested
Co-Author(s):	G. De Temmerman, D. Rudakov	ITPA Joint Experiment :	No
Description:	Expose polycrystalline diamond to neon seeded detached L-mode plasma via DiMES (buttons), in order to quantify the chemical erosion rate.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Expose diamond DiMES samples, heated to 300 degrees Celsius, to the outer strike-point of a weakly detached L-mode plasma with neon seeding, for which high levels of erosion have been observed for graphite. Repeat to maximise the total fluence.
Background:	Recent advances in low cost chemical vapour deposition techniques (CVD) for the production of polycrystalline diamond have made possible its use as an engineering material. For example, diamond windows are now readily available. With respect to PWI, diamond is attractive due to its very high thermal conductivity; resistance to neutron damage; and reduced chemical erosion as compared to graphite or CFC. CVD could also, in principle, be carried out via remote handling, raising the possibility of in situ repair. Unfortunately, pure diamond is an electrical insulator, which raises questions about its suitability for tokamak applications. A single diamond button sample was exposed for a total of 24 s to a deuterium detached L-mode plasma in 2008 using heated DiMES, with no evidence of substantial erosion (as with graphite), arc damage or graphitization. The proposal for 2009-2010 is to expose a range of button samples to detached plasmas with neon seeding, for which high levels of net erosion have been observed for graphite [Whyte, Fusion Sci. and Tech., 2006], to see how diamond compares. Included in the samples will be boron doped diamond, which increases the electrical conductivity to semiconductor levels, potentially maximising performance in large area applications (full strike-points) where arcing may be more of a problem.
Resource Requirements:
Diagnostic Requirements:	divertor probes, DiMES TV, divertor cameras, MDS, DTS
Analysis Requirements:
Other Requirements:

Title	244: Characterization of heat loads due to Type-I ELMs
Name:	Jakubowski marcin.jakubowski@ipp.mpg.de	Affiliation:	Max-Planck Institute for Plasma Physics
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	Charles Lasnier, Todd Evans, Max Fenstermacher, Mathias Groth, Oliver Schmitz, Jonathan Watkins	ITPA Joint Experiment :	No
Description:	Recent experiments have shown that the filamentary nature of Type-I ELMs is also reflected in their heat deposition patterns. This has been also confirmed on DIII-D. Use of DIII-D and TEXTOR infrared cameras gives unique opportunity to study power loads due to Type-I ELMs with very good time resolution at two different toroidal locations. Characterization of power loads and power deposition patterns at different plasma parameters, like collisionality or direction of magnetic field with ability to resolve toroidal asymmetries is the main idea behind this experiment.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Perform discharges at different collsionalities and with different plasma shapes. It should be performed in two experimental units with different direction of Bt/Ip
Background:	Type-I Edge Localized Modes (ELMs) are a significant concern in tokamak plasmas. They appear as a series of rotating filamentary structures due to pedestal pressure gradients found at the edge of H-mode plasmas. It has been reported from ASDEX-Upgrade, that Type-I ELMs create helical footprint patterns of heat flux on the divertor surface. Several strike lines were detected outside the original strike point of the outer leg albeit at very low amplitude. They form helically aligned structures, which are clearly related to the topology of the magnetic field. At DIII-D the same observation have been made recently. Additionally, it has been measured that the wetted area of power deposition due to Type-I ELMs depends lineary on the ELM size, which is consistent with the hypothesis of separatrix stochastization due to filaments.
Resource Requirements:
Diagnostic Requirements:	TEXTOR and DIII-D infrared cameras running simultaneously, fast measurements of diamagnetic energy
Analysis Requirements:
Other Requirements:

Title	245: Aspect ratio scaling of Alfven eigenmode structure and avalanche transport
Name:	Crocker ncrocker@physics.ucla.edu	Affiliation:	UC, Los Angeles
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	E. D. Fredrickson, N. N. Gorelenkov, W. W. Heidbrink, G. J. Kramer, R. M. Nazikian, M. Van Zeeland, W. A. Peebles, S. Kubota, T. Rhodes,	ITPA Joint Experiment :	Yes
Description:	The structure of Alfvén eigenmodes is sensitive to the structure of the Alfvén continuum, which is affected aspect ratio. The structure of Alfvén eigenmodes, including, in particular, both the their radial extent and poloidal spectral width, influences their effect on fast-ion orbits. Similar plasmas will be created in NSTX and DIII-D, differing primarily in aspect ratio. The plasma conditions will be chosen so that the NSTX plasma exhibits avalanches. Measurements of mode structure will be compared to mode structure predicted by NOVA-K. F ast-ion transport, including incidence and severity of avalanche transport, will studied and compared with prediction using ORBIT.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A model NSTX shot with avalanches and conditions achievable in DIII-D will be identified. This shot will reproduced in NSTX using the full suite of available fast-ion population diagnostics and mode structure diagnostics. A similar shot, except for aspect ratio, will be produced in DIII-D using the full suite of available fast-ion population diagnostics and mode structure diagnostics.
Background:	Calculations of Alfvén eigenmode structure using NOVA-K indicate that NSTX eigenmodes typically have a broad spectrum of poloidal components and a broad radial extent. The calculated eigenmode structures may be usefully understood as resulting from linear coupling between "local" extremum modes of the types such as Alfvén Cascade modes and TAEs that theory predicts in the limit of low inverse aspect ratio, low beta and high toroidal mode number. Such coupling is expected be stronger at higher inverse aspect ratio, leading to the expectation that for plasma conditions in DIID-D and NSTX that are similar except for aspect ratio, the eigenmodes in NSTX will have a broader spectrum of poloidal components and a broader radial extent. Both the broad spectrum of poloidal components and broad radial extent of NSTX eigenmodes may be expected to make them more effective at inducing fast-ion transport and creating avalanches than the eigenmodes in the similar DIII-D plasmas. The broader poloidal spectrum leads to many more significant phase space resonances and increases the likelihood of resonance overlap, which is the phenomenon responsible for avalanches. The broader radial extent of the eigenmodes leads to more significant fast-ion orbit perturbations in the plasma edge, which increases their effectiveness at causing fast-ion loss.
Resource Requirements:	Beams, low magnetic field (B ~ 0.6 T), cryo pumping (?)
Diagnostic Requirements:	MSE, BES, Thomson scattering, PCI, reflectometers, FLIP, FIDA, Neutron Emission
Analysis Requirements:	NOVA-K, ORBIT, EFIT, TRANSP
Other Requirements:

Title	246: Validation of Code Predictions for the Sensitivity of Turbulence and Transport to Te/Ti Ratio
Name:	White whitea@mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	T. A. Carter, J. C. DeBoo, J. H. Hillesheim, C. Holland, J. Kinsey, G. McKee, W. A. Peebles, T. Rhodes, L. Schmitz, G. Staebler	ITPA Joint Experiment :	No
Description:	The use of linear and nonlinear gyrokinetic turbulence codes to analyze and model experimental data with the goal of ultimately validating the models and codes is becoming increasingly important as ITER draws near. This experiment will use the new transport model TGLF and the nonlinear gyrokinetic simulation code GYRO to motivate and guide the design of experiments that maximize the predicted response of turbulence to changes in the ratio Te/Ti, while minimizing sources of input error to the code. Predictions for turbulence characteristics obtained prior to an experiment will be compared with the experimental results. This in contrast to most comparisons done now, where nonlinear gyrokinetic simulations that calculate transport coefficients and fluctuation characteristics are run only after an experiment has been performed. The experiments in this experiment will be designed to optimize access for novel turbulence diagnostics that have the potential to reveal new physics regarding the scaling of transport with Te/Ti, e.g. measurements of high-k density turbulence and low-k electron temperature turbulence.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The experimental approach uses primary and secondary parameter scans to investigate the sensitivity of turbulence and transport to the ratio Te/Ti. The ratio Te/Ti will be the parameter varied in the primary scan and it is anticipated that rotation or Er and collisionality will be parameters used in the secondary scan. TGLF will be used to model past experiments [1,2,3] in order to investigate the changes in the linear growth rates of the ITG, TEM and ETG modes with changes in Te/Ti. Then, exploratory primary and secondary TGLF scans about the baseline parameters will be used to identify regions of parameter space that maximize the predicted response of the turbulence to changes in Te/Ti. This highlights a goal of this experiment: to use predicted changes in turbulence characteristics to design target experimental conditions. Preliminary TGLF modeling in support of this experiment has already indicated that a two point scan in the primary parameter, Te/Ti, would reveal significant changes in the low-k electron temperature fluctuations. The two point-scan studied used experimental profiles from a past experimental DIII-D L-mode plasma as input (Bt = -2.0 T, Ip = 1 MA, ne ~ 2.3 x10^19 m^-3, NB power = 2.6 MW). Only the electron temperature was scaled in TGLF, LTe was held constant and the other input profiles (ne, Ti, rot, etc.) were held constant. The two points used in the TGLF scan, Te/Ti ~ 0.96 and Te/Ti > 1.8, could be accessed using different heating methods in L-mode (single beam source) plasmas (Electron Cyclotron Heating, NB injection and fast-wave heating). In the experiment, Te will be scanned and other parameters such as Ti, ne, Er, and rotation will be held as constant as possible. However, small changes in parameters other than Te are unavoidable in the experiment and the impact of these anticipated changes can be evaluated ahead of time via the secondary parameter scan using TGLF. For example, using ECH in an experiment often changes the density profile, Er profile and the resulting ExB shear. This means that an increase in Te input to the code must be combined with changes in the density and Er input to the code in order to take account of the effects of ECH when used in the experiment to increase Te. Therefore, at the extrema in the TGLF two point Te/Ti scan, the rotation or density can be the secondary scan parameters, in order to evaluate the sensitivity of the turbulence to small changes in ExB shear and collisionality. The range of these secondary scans will be motivated by past experimental results for the estimated uncertainties in the rotation or density profiles.
Background:	Turbulence characteristics and transport levels depend on various quantities, and choosing which parameters to vary in the simulations should be based on experimental accessibility and theoretical expectations. Theory predicts that transport can depend strongly on dimensionless quantities such as plasma shape, electron to ion temperature ratio, Te/Ti, and collisionality. The ratio Te/Ti is a particularly relevant parameter to investigate because validation studies of nonlinear gyrokinetic turbulence codes are especially critical in the Te ~ Ti regime due to the relevance for application to next-step devices such as ITER. Past experimental non-dimensional scaling studies at DIII-D [1] have found that in H-mode plasmas the thermal energy confinement time decreases with increasing temperature ratio, Te/Ti. One explanation for the strong scaling of ion thermal diffusivity with Te/Ti in high confinement (H-mode) plasmas involved the scaling of the ITG mode with Te/Ti when the ITG mode is near the instability threshold. There is also experimental evidence [2] that the transport increases with increasing Te/Ti in low confinement (L-mode) plasmas, but that the fluctuation level of low-k density fluctuations remains constant. These observations are consistent with recent experiments (these were not a dedicated Te/Ti scan) where ECH was used to predominantly cause increases in Te/Ti and decreases in collisionality, which resulted in little change in low-k density fluctuations, but a large increase in low-k electron temperature fluctuations and heat transport [3]. A key aspect of the proposed new experiment is the use of new diagnostics and recent diagnostic upgrades that were not available during past dedicated Te/Ti scaling experiments [1, 2] and a the extensive use of a priori TGLF scans. Comparing comprehensive measurements of core fluctuations (including low wavenumber (low-k) electron temperature turbulence, low-k, intermediate-k, and high-k density turbulence, ñ/n, zonal flows, and phase angle between and ñ/n) with predictions from theory that have been tailored to specific experimental conditions has the potential to reveal new physics regarding the scaling of transport with Te/Ti. References: [1] C. C. Petty, et al. Phys. Rev. Lett. 83, 3661 (1999). [2] G. R. McKee, et al., APS-DPP (2003) [3] A. E. White et al. APS-DPP (2008)
Resource Requirements:	1 to 1 1/2 experimental days At least 5 gyrotrons, 6 would be better. Neutral beam sources 30L/330L, 210R, 150L, Fast Wave
Diagnostic Requirements:	CECE, BES, multi-channel tunable reflectometer, FIR intermediate- and low-k systems, Thomson scattering, 40-channel ECE radiometer, CER and MSE.
Analysis Requirements:	Standard analysis for GYRO/TGLF input file generation: EFIT, ONETWO, and autoonetwo are essential. TGLF before the experiment, TGLF and GYRO after the experiment.
Other Requirements:	--

Title	247: Impurity and radiation asymmetry during massive gas injection disruption mitigation
Name:	Whyte whyte@psfc.mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):	E. Hollmann, M. Reinke, R. Granetz, R. Pitts	ITPA Joint Experiment :	No
Description:	Massive gas injection is presently being planned as a disruption mitigation method for ITER. The overall goal of disruption mitigation is to minimize thermal transient damage to the internal components, chiefly by promoting radiative dissipation of the plasma thermal energy. A particular challenge with ITER is that given the Q=10 D-T plasma thermal energy of 350 MJ, if dissipated in 0.6 ms with perfect spatial uniformity, would result in surface melting of the Beryllium main-wall, due to the low melt temperature of Be. Since our empirical and modeling experience make us expect ~ms timescales of the radiative dissipation, a realistic concern becomes what degree of spatial asymmetry might lead to localized Be melting in ITER. Recent experiments on C-Mod [M.L. Reinke et al Nucl. Fusion 2008] and DIII-D [Hollmann et al Nucl. Fusion 2008] have identified a particular concern for melting of the Be wall near the gas injection locations. During the initial stages the gas interacts with the local edge plasma and as both gas and ions begins to spread toroidally, poloidally and radially. The edge cooling and MHD evolve to eventually trigger the full thermal quench. Experiments indicate that the final thermal quench is quite uniform toroidally. However, the initial gas interaction necessarily releases some fraction of the edge plasma / pedestal energy local to the gas injection location. By itself this energy is insufficient to cause damage, however it leads to local "pre-heating" of the nearby surfaces, such that when the symmetric final thermal quench, when ~90% of the plasma thermal energy is released, will then go pass the melt temperature. This proposal specifically seeks to address, in collaboration with C-Mod experiments, the question of how much local pre-heating can be expected in ITER and how we might design the MGI system of ITER to avoid such damage, with issues at hand such as number of location injection, poloidal position of the injectors and the gas species. This goal requires that we construct a semi-empirical model that describes the evolution of the impurity transport and radiation through the early stages of the gas injection such that it can be applied for ITER extrapolation (present devices are far from the melt limits). The combination of MGI and radiation diagnostic capabilities of DIII-D and C-Mod provide a unique (worldwide) ability to provide the necessary data on this problem to provide ITER reasonable estimates for the mitigation design on the timescale of 6-9 months. The two devices provide obvious size scalings, yet have nearly identical MGI systems and complimentary/repeated diagnostic abilities to quantify the radiation pattern and evolution, primarily with fast bolometry. This topic is an urgent issue in ITER design since it involves port allocations and tritium plant capabilities, and we have instigated this investigation at the request of the ITER IO physics (Richard Pitts).	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Planned MGI terminations into stable plasmas. "Close to ITER" configuration: q95, H-mode, etc. (details to be sorted out between C-Mod, DIII-D & ITER) Diagnose radiation evolution by using the two toroidally distributed fast bolometry systems (DISRAD) & fast visible cameras viewing the MGI port. These can be nearly directly compared to toroidally distributed fast bolometry in C-Mod. Vary injected gas species to understand the mechanism(s) governing the toroidal and poloidal distribution of the impurities. A paricularly important parameter would seem to be the atomic mass which varies the gas sound speed and also the ion sound speed once present in the plasma. Nominally would study He, Ne, and Ar with also the possibility of adding mixed gases to the study since these are previewed as possibilities in ITER. Repeat with different target plasma. Possibilities to scan - heating level ~ plasma stored energy - rotation - confinement mode - magnetic topology - q95
Background:	see description
Resource Requirements:	~ 1/2 to 1 day of experimental time MGI with changeable gas mixes
Diagnostic Requirements:	Fast bolometry Fast imaging camera for MGI CER spectrometers with reduced gain tuned to low ionization state visible lines Standard disruption diagnostics
Analysis Requirements:	Bolometry inversions Fast-camera imaging analysis Visible spectroscopy analysis from CER
Other Requirements:

Title	248: Study of Sawtooth Physics by ECEI
Name:	Park hyeonpark@postech.ac.kr	Affiliation:	Pohang U of Science and Technology
Research Area:	Stability	Presentation time:	Requested
Co-Author(s):	H. K. Park, B. Tobias, C.W. Domier, T. Munsat, N.C. Luhmann Jr., A.J.H. Donné, M.J. van de Pol	ITPA Joint Experiment :	No
Description:	Recent physics studies of the m/n=1/1 mode (sawtooth oscillation) employing the TEXTOR 2-D Electron Cyclotron Imaging system, which provides real time 2-D images of the electron temperature fluctuations with unprecedented temporal and spatial resolution, have revealed physics insights that contradict all of the major theoretical models developed for the m/n=1/1 mode up to now: 1) Observation of the high field side crash which violates the pure Ballooning mode model and the observed collective process of the heat transport during reconnection process is in contradiction with the global field line stochasticity introduced to explain the experimentally observed negligible change of the core magnetic flux before and after the reconnection process; 2) The images of the island and hot spot resemble those predicted by the full reconnection model but are in disagreement with the observation that the initiation of the reconnection appears to be â??X-pointâ?? rather than â??Y pointâ?? inherent from the Sweet Parker model; 3) Comparative studies have provided a decisive proof that the â??Quasi interchange mode modelâ?? is invalid. Since the studies were performed with the circular shaped TEXTOR tokamak plasma, it is natural to extend the studies to the shaped plasmas on DIII-D in order to understand if the reconnection process of the m/n=1/1 mode has any flux surface shape dependence. In this current experimental proposal, we will employ the new and improved ECEI system which has recently been tested on the TEXTOR device. This new system has a capability to expand the vertical size of the image at the same focal depth point and it is designed to have a capability to measure two images simultaneously from both the high and low field sides of the q~1 surface so that the reconnection process can be decisively interpreted and considerably more comprehensive images can be provided than the prototype system used in the TEXTOR studies. The primary objective of this proposal is to confirm and extend the new physics we have measured in circular plasmas together with the core current density measurement on DIII-D. Note that there is an outstanding issue of the experimental verification of the central q values between circular plasmas (TEXTOR and TFTR) and shaped plasmas (DIII-D).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Plasma operation that is optimum for Sawtooth oscillation. Discharges at B(T) ~ 2T, I(p)~ 1.00 MA and q(a)~4. NBI (Co-injection) power of ~3 MW is preferred to increase the size of the Sawtooth oscillation. 1) Set the focal depth of the ECEI system to image both low and high field sides near the q~1 surface and measure the reference image a) Two images; one with the vertically expanded and the other with reduced images 2) Repeat the measurement with the focal depth moved with +5 cm and -5 cm with respect to the reference position. a) Two images; one with the vertically expanded and the other with reduced images
Background:	Understanding of the physics of sawtooth oscillation is extremely important since this is the most basic core MHD event which has a long history of interest. In particular, understanding of the role of the core current density in the ideal Kink instability is critical. As is well known, the majority of the measured central q values have indicated that the change of the core current density is miniscule (normally q(0) changes from 0.75 to 0.8) while the changes in the core plasma parameters related to the plasma pressure such as the ion and electron temperatures and density have been substantial. The measurement from DIII-D outstands that the central q value was ~0.8 before the crash and returns to q(0) ~1 after the crash as suggested by the full reconnection model. The difference was thought to originate from the flux geometry difference, since the first set of measurements was conducted TEXTOR and TFTR, both of which have a circular cross-section. This may not be strictly true considering that the measurement on JET showed little change of the core current density before and after the crash. Therefore, it is extremely important to verify whether the unresolved issue is simply an instrumental issue or whether it is associated with anything with more fundamental variations of the crash process. Previous Sawtooth studies have primarily employed X-ray tomography for the visualizing the dynamics of this event together with Electron Cyclotron Emission measurements which have been used for the measurement of electron temperature. However, the phenomenology of this event has been found to be extremely complicated and a simple one dimensional diagnostic system such as the conventional 1-D ECE system has a limited capability to understand this highly asymmetric reconnection event during the crash phase. On the other hand, the X-ray imaging system has two major deficiencies in studying this type of highly asymmetric event in the toroidal and poloidal planes: 1) There is no unique solution in the inversion process of the chordal measurement with a limited number of viewing positions; 2) X-ray emission depends on multiple plasma parameters (Zeff, ne, Te) thus leading to possible ambiguities. This need motivated the development of the ECEI approach which not only possesses adequate spatial resolution (~1cm x ~1 cm) for the reconnection process but also sufficiently fast time resolution for this event (a few microseconds) which has thus provided new insights into the reconnection process which were not previously obtainable. As noted above, detailed comparative studies with the theoretical models have revealed that none of the theoretical models up to now consistently describe the measured images. It is thus important to extend the measurements on shaped flux surfaces with the improved ECEI system on DIII-D where extensive supplemental diagnostics are available. In particular, a central q measurement would be of great benefit.
Resource Requirements:	Many resources are needed to have commission of the new ECEI system on DIII-D
Diagnostic Requirements:	MSE, TS, X-ray and any other diagnostics that can address the Sawtooth oscillation,
Analysis Requirements:	General equilibrium analysis. Sawtooth model by M3D
Other Requirements:

Title	249: O-bake of DIII-D + 13CH4-trace experiment re tritium recovery in ITER
Name:	Stangeby peter.stangeby@utoronto.ca	Affiliation:	U of Toronto
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Requested
Co-Author(s):	Steve Allen, Jim Davis, David Elder, Max Fenstermacher, Bernie Fitzpatrick, Mathias Groth, Tony Haasz, Charlie Lasnier, Tony Leonard, Adam McLean, Yarong Mu, Cedric Tsui, Phil West, Dennis Whyte	ITPA Joint Experiment :	Yes
Description:	Tritium-retention by co-deposition with carbon appears likely to prevent the use of graphite in ITER in the DT-phase. The use of graphite at the strike points, however, could considerably increase the likelihood of success of ITER with regard to handling the high power levels which are unprecedented in tokamak operation. Elimination of graphite by ITER may raise questions concerning the relevance of all types of physics results obtained in graphite-protected tokamaks. Thermo-oxidation, �??O-baking�?�, is a potential solution to this problem, capable of recovering tritium from carbon co-deposits at all locations inside the vessel, including places that are out of line-of-sight to the plasma, and in tile gaps, etc. ITER is unlikely to use this method, however, unless it has been fully proven in an existing tokamak. It is necessary to establish, for a given �??severity�?? of exposure (pressure, bake temperature, duration): 1. How much of each type of co-deposit is removed? 2. How much and what type of wall re-conditioning is required to recover tokamak performance? 3. Whether there is long-term collateral damage done to non-carbon materials and components inside the vessel, and if there is damage, how much? 4. The removal efficiency of oxygen cleaning is a strong function of temperature and vessel temperature is restricted in ITER. Is there a window of applicability? Some aspects of O-baking, relevant to application to ITER have already been carried out in existing tokamaks, particularly regarding the risk of collateral damage. O-baking as a means of recovering tritium from carbon co-deposits appears to have been first proposed by G Janeschitz, based on the unplanned (accidental) air-bakes that have occurred in JET each year or so, when the vacuum vessel suffers a leak. Since JET is maintained continuously at 250-350 oC, such accidents become air-bakes. No collateral damage has been identified in JET as a result of these accidental air-bakes and recovery of plasma performance has been easier than after a normal vent to air (where the vessel has been cold and air moisture is a factor). DIII-D has also experienced such accidental air-bakes, with no identifiable collateral damage; however, in order to further reduce risk, over the past two years, tests have been carried out at the University of Toronto on internal DIII-D components, to identify any possible collateral damage effects due to thermo-oxidation. These tests have recently been supplemented by a second O-bake test facility assembled at DIII-D by LLNL. Results of these tests are discussed in the Background section.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	1. In the last days of the 2009 campaign, repeat the 2008 13CH4 injection experiment as closely as possible. In this way the amount of the deposited 13C and its location can be fairly reliably anticipated from the tile analysis presently being carried out at Sandia and MIT. 2. Vent the vessel and remove a representative set of tiles, the reference �??before O-baking�?� tile set. 3. Close the vessel and perform an oxygen bake at 250 oC, counting the 13C's in the exhaust, for a quantitative assessment of the ability of O-baking to remove C deposits (global accounting). 4. Vent and remove a 2nd set of tiles, the �??after 250 oC O-baking�?� tile set. Inspect the inside of DIII-D. If there is no evidence of any adverse effects of the O-baking at 250 oC, then the decision may be taken to close up the vessel and carry out an O-bake at 350 Co. Then open the vessel and remove a 3rd set of tiles, the �??after 350 oC O-baking�?� tile set. 5. It may be decided to then close the vessel and establish the reconditioning process required to recover satisfactory plasma operation. Alternatively, this could be left until the end of the vent period. A potential advantage of attempting to re-start the plasma immediately is that, should it turn out that anything major is required, there would be more time available before the start of the next campaign. 6. The tile analysis will be used to close the particle accounting by directly showing that the 13C deposits are not present after an O-bake and providing a spatially resolved evaluation of this tritium recovery technique. At the time of the first venting (before the 250 oC O-bake) sample coupons will be placed at various locations in the vessel, particularly in cold corners, to confirm that the carbon removed from the tiles doesn�??t just move from one location to another inside the vessel but actually exits the vessel (through the pumps). 7. Following the SNLA/MIT analysis of tiles and coupons for 13C, the D content of the codeposits on these surfaces will also be measured by laser-TDS at the University of Toronto �?? to assess any redistribution of D during the oxidation process
Background:	The testing of DIII-D in-vessel components found that most components were unaffected by heating in oxygen. However, some components have required further study, in particular copper and copper-plated components. Under the oxidation conditions being proposed for DIII-D, Cu surfaces develop a thin (< 1 μm) oxide layer. However, these oxide layers were not observed to flake-off without physical contact. Two further issues related to the oxidation of Cu arose during the course of the lab tests. Firstly, it was found that the copper oxide is readily reduced back to Cu by heating the oxide layers in the presence of hydrogen. The reduced Cu forms a very fine powder on the surface and could potentially result in plasma contamination and also a possible electrical-shorting problem for some diagnostics. The fine size of the powder indicates that the shorting problem appears to be unlikely but plasma contamination is a possibility. The latter could possibly be dealt with by wiping off Cu surfaces that contact the plasma. A second issue appeared, initially, to be more serious: during testing of components with Cu coatings, the test facility became contaminated with an agent which prevented the oxidation of Cu and led to the build-up of black layers on adjacent stainless steel surfaces. In the worst cases, it was found necessary to re-machine all exposed components of the vacuum system in order to recover normal operation. A search of DIII-D records indicated that the specific Cu coatings tested in the Toronto lab were prototype components and that the final installed items had different Cu coatings. Special specimens were prepared with both the final and prototype Cu plating and subsequent tests indicated that the Cu plating actually used in the tokamak did not lead to system contamination. Surface analysis of the deposits on stainless steel, and the �??non-oxidized�?� Cu surfaces suggest that the active agent produced during the attempted oxidation runs is sodium hexanoate. Sodium hexanoate is a known oxidation inhibitor for Cu surfaces and its components are consistent with the surface analysis performed. There appears to be little chance of significant quantities of sodium being present in DIII-D. Prototypes for the DIII-D cryopumps were among the items leading to the greatest system contamination. These prototypes have been stored on the grounds of GA for more than a decade and it is suspected than sodium contamination from salt air may have been the issue. It is also possible that the prototype Cu plating process incorporated small amounts of sodium in the coating. There is no apparent reason to believe that such contamination will occur in DIII-D. The accidental air-bakes in JET have not resulted in any detectable Cu contamination of the plasma, although Cu-coating is used extensively on JET internal components, e.g. bolts. For a complete list of DIII-D components tested and the results of the tests see: https://diii-d.gat.com/diii-d/Oxygen
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	250: VDE and VUD Characterization for ITER
Name:	Wesley wesley@fusion.gat.com	Affiliation:	GA
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):	Humphreys et al	ITPA Joint Experiment :	Yes
Description:	Make 'controlled' VDEs (vertical displacement events) and VUDs (vertically unstable disruptions), respectively by vertical control and weak massive gas injection (to give slower initial Ip decay) starting from an 'ITER-like' LSN beam-heated plasma. Vary plasma current (q), current profile (L-mode, H-mode), thermal energy (OH, NB and aux heating). Monitor magnetics, MHD, halo currents, vessel impulse, divertor and FW energy deposit (especially in wall-contact region(s), fast camera imaging. Intent is to obtain a comprehensive data set suitable for empirical or model based application to predicting 'natural' VDE and VUD dynamics and impacts in ITER, over a range of plasma current and energy states and perhaps a variety of operation modes (ELMy H, hybrid, AT, etc.)	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Make an ITER-like 'standard' NBI-heated 'target plasma' and develop standard VDE and VUD trigger means. Vary target plasma current, energy and or ops mode; collect comprehensive data sets. Modify choice /range of variation parameters based on a) pre-expt plan and b) control room data analysis. May need half-days or collaboration with other parallel studies (ie ITER vertical stability or MGI tests) to pursue emerging findings.
Background:	Unmitigated VDEs and/or natural or partially mitigated VUDs (or failed MGI, etc) in ITER will generate substantial thermal and magnetic loads on the in-vessel systems and major global and local forces/loads on the vacuum vessel and vessel supports. Certain aspects of the vessel and support design are identified as critical issues. Present VDE and VUD data for DIII-D and other ITER-like tokamaks may not be sufficient to resolve open design adequacy issues. A set of comprehensive data for a range of candidate VDE and VUD scenarios would allow more direct empirical extrapolation to ITER and be invaluable for simulation model validation. Experience with high-thermal-energy wall-contact VDEs is limited
Resource Requirements:	Candidate low- to full thermal energy ITER-like LSN plasmas, MGI or pellet systems to provide controlled VUDs
Diagnostic Requirements:	Standard pre-event target plasma characterization; dynamic monitoring of after-onset phase magnetics and thermal and plasma-FW + divertor interaction, optical and spectro data on event phase plasma properties (n, T, Zeff, impurity content, ....). Readiness of dynamic models to assess data and/or interpolate missing information.
Analysis Requirements:	EFIT and/or JFIT data on pre and post-event EQ dynamics
Other Requirements:	diagnostic plans and capabilities and during and post-experiment analysis require further development

Title	251: Snowflake Divertor
Name:	Umansky none	Affiliation:	LLNL
Research Area:	Core-Edge Integration	Presentation time:	Requested
Co-Author(s):	R.H. Bulmer, R.H. Cohen, T.D. Rognlien, D.D. Ryutov (LLNL-Theory), A. Garofalo , M. Groth, A. Hyatt, C. Lasnier, M. Makowski, P. West (LLNL & GA -Experiment)	ITPA Joint Experiment :	No
Description:		ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Proposed experiment: vary PF coil currentsto transition smoothly from standard tosnowflake configuration. As snowflake is approached we should monitor:â?¢Changes in geometry from magnetic reconstructionâ?¢Changes in heat flux, temperature, density on target plate â?¢Changes in divertor radiationâ?¢Far SOL fluctuations (blobs)â?¢Changes in ELM parameters (amplitude, periodicity) â?¢Strike point splitting (â??snowflake-minusâ??)DIII-D will take advantage of its extensive edge diagnostics - fast IRTV, Langmuir probes, D-alpha, fast stroke probe, divertor Thomsonscattering, pedestal Thomson scattering, CER, edge reflectometry. This exciting experiment will exploit the unique DIII-D combination ofmachine flexibility and measurement capabilities
Background:	p For regular x-point 1st derivatives vanish,ΨR= ΨZ=0â?¢For snowflake (2 nd order) null-point also 2ndderivatives vanish: ΨR= ΨZ =ΨRR= ΨRZ= ΨZZ = 0 â?¢Snowflake null-point can be constructed with 3currents (regular x-point just with 2 currents)â?¢Dubbed â??snowflakeâ?? for hexagonal symmetry (D.D.Ryutov, PHYSICS OF PLASMAS 14, 064502) Increased magnetic shear and fluxexpansion may affect ELMs and divertor; â?¢In SOL shear also increases,In pedestal region shear increases by a factor ~2. In SOL shear also increases,may affect turbulence â?¢ELMs are caused by MHDi nstability in the edge;snowflake may lead to new regimesâ?¢UEDGE calculations show significant reduction of target heat flux Snowflake divertor can be tested on DIII-D â?¢Corsica calculations have demonstrated snowflake-like equilibriafor DIII-D, with realistic pressure and current profiles, coil currents 0.338 MA 0.306 -0.2010.237 0.416 MA 0.294-0.2610.26445 8 9 4 5 8 9 The closer to exactsnowflake the largerPF coils currents will be neededâ?¢With existing DIII-D PF coils we shouldbe able to see theeffects of snowflake shearing and fluxexpansion
Resource Requirements:	fast IRTV, Langmuir probes, D-alpha, fast stroke probe, divertor Thomsonscattering, pedestal Thomson scattering, CER, edge reflectometry.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	252: Characterisation of power fluxes to PFCs during VDEs, ITER-relevant disruptions
Name:	Loarte none	Affiliation:	ITER
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):	M. Sugihara, R. Pitts	ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to perform experiments in DIII-D to characterise the balance of in the approach to the thermal quench in disruptions and VDEs as well as determining the characteristics of energy deposition during the thermal quench itself and current quench (including runaway deposition)	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The experiment should be done at a level of current that is considered safe for DIII-D and should consist of a series of discharges in which disruptions, VDEs and runaways are triggered in a reproducible way. Every disruption/VDE should be repeated several times to characterise the variability of the disruptions. The conditions to be studied are those most likely for ITER : a)Density limit disruption from a high density H-mode b)Radiative disruption from a medium density H-mode with impurity puffing c) NTM driven disruption by uncontrolled growth of NTMs in H-modes d) beta limit disruption c) VDE in full performance H-mode As for the runaway experiments this should be done in conditions that generate a reproducible runaway plateau, preferably in diverted configuration. A possibility would be to inject a neon pellet
Background:	The determination of timescales and spatial distribution of power fluxes to PFCs during disruptions in ITER remains uncertain. One of the difficulties is that of the lack of systematic experiments and good available measurements in most devices. This is particularly serious for VDES and runaway electrons which are expected to lead to the largest energy deposition during disruptions on ITERâ??s PFCs
Resource Requirements:	NBI with cryopump, NBI and possibly neon gas massive injection (at moderate level) or neon pellet injection to generate runaways
Diagnostic Requirements:	Core and edge plasma measurements. In particular measurements of the energy lost from the main plasma and power fluxes to PFCs are mandatory. This will drive the details of the experiments carried out. Particular attention should be dedicated to runaway electrons
Analysis Requirements:	Analysis of measurements of plasma parameters, radiation and runaway electrons and their deposition
Other Requirements:

Title	253: Beam-ion transport by plasma turbulence
Name:	Heidbrink heidbrink@fusion.gat.com	Affiliation:	UC, Irvine
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	C. Muscatello, M. Van Zeeland, G. McKee	ITPA Joint Experiment :	No
Description:	Study the spatial transport of beam ions in L-mode plasmas. Avoid MHD and fast-ion driven instabilities to isolate the transport associated with plasma fluctuations. Use FIDA to infer the transport of co-passing, trapped, and counter-passing ions. Vary the injection energy E and the plasma temperature T to test the predicted scaling of the transport with E/T.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Establish a baseline modest current (0.8 MA), full-field, L-mode plasma with monotonic q profile. Run inner wall or upper-single null configurations to stay in L-mode. Start with 1.5 sources (modulated for diagnostics) at normal operating voltages. 2) Drop the injection energy to ~50 keV; increase the number of sources to keep the plasma temperature constant. 3) In either condition #1 or #2, add ECH to raise Te. 4) Attempt to increase the beam power without exciting MHD to raise Ti. 5) Thoroughly document the best condition (fluctuation & fast-ion diagnostics). 6) If time permits, repeat with off-axis injection into a small, upwardly shifted plasma.
Background:	There has been a resurgence of interest in the transport of energetic ions by background plasma turbulence, with four different theoretical groups publishing (sometimes contradictory) papers in the last three years. The issue is important for alpha & beam-ion transport in ITER and for NBCD in DIII-D. The theories predict different scalings for passing & trapped particles vs. E/T, where E is the energetic particle energy and T is the ion or electron temperature. With our excellent fast-ion and fluctuation diagnostics, we are in a position to test the gyrokinetic simulations.
Resource Requirements:	7 beam sources; 5 gyrotrons
Diagnostic Requirements:	The full complement of core profile, fluctuation, and fast-ion diagnostics.
Analysis Requirements:	TRANSP, FIDA simulation code. Gyrokinetic simulations (both UCI and IPP groups have volunteered).
Other Requirements:

Title	254: Pitch resonant vs non-resonant perturbation effects on pedestal and ELM
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	R.A. Moyer	ITPA Joint Experiment :	No
Description:	LFS perturbations from the C and I coils contains mixes of both pitch resonant and non-pitch resonant harmonics. ELM control displays a clear resonance in q95, while pedestal particle transport (density pumpout) displays at best a much weaker/broader q95 dependence and possibly no q95 dependence. Since a primary need for ITER is to obtain ELM suppression while preserving the pedestal pressure (core performance) and density (radiative divertor operation), we need to understand the origin of these effects in order to optimize the coil designs and phasings to maximize ELM suppression while minimizing density pumpout and other deleterious effects.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	- run pumped LSN ISS plasmas with reversed Ip (to preserve ion grad B drift direction and 3.2 < q95 < 6 - use odd parity Icoil with n = 3 and phasing to provide as asymmetric perturbation as possible - match perturbation to pitch resonance and off - measure plasma response (core and edge rotation, Er, profiles, ELM character)
Background:	Objective 1.3 for the ITPA pedestal working group plan of urgent ITER needs highlights the requirement to suppress ELMs in ITER while minimizing the core performance hit and maintaining compatibility with radiative divertor operation (high pedestal density).
Resource Requirements:	I coil cryopumps co and counter NBI
Diagnostic Requirements:	"standard" RMP ELM control core, pedestal and boundary diagnostics
Analysis Requirements:
Other Requirements:

Title	255: Runaway discharge position control and controlled loss of runaway electrons
Name:	Loarte none	Affiliation:	ITER
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	M. Sugihara	ITPA Joint Experiment :	No
Description:	The purpose of this proposal would be to perform experiments in DIII-D to demonstrate the effectiveness/viability of methods of runaway power deposition control for application in ITER, if runaway generation cannot be avoided	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	These experiments should be done in conditions that generate a reproducible runaway plateau whose position can be controlled by the plasma control system (this may require to start from a low elongation plasma and use massive neon injection or a neon pellet). Once runaway discharges are produced various schemes should be applied and their effects on the deposition of runaways characterised, namely : a) a reversed loop voltage of increasing magnitude in subsequent shots would be applied to the runaway discharge and the effects on its termination quantified, b) high Z impurities in increasing quantities would be injected after the formation of the runaway plateau and their effects on runaway discharge and the effects on its termination quantified c) increasing levels of plasma current in the I coils configured for n=1 or n =2 (if possible) operation and/or the C-coils would be applied and their effects on runaway loss characterised
Background:	The installation of in-vessel coils for vertical stability control and ELM suppression in ITER open a new area for the controlled shutdown of discharges that have develop a runaway plateau following a disruption. If the plasma position of the runways beam can be controlled various schemes can be employed to slow down the runaway electrons and decrease their peak energy deposition on the first wall of ITER. The ones presently considered are : use of the transformer in reverse to slow down the electrons, injection of high Z noble gas (Ar, Xe, Kr) to slow down electrons by multiple collisions and use of the ELM control coils in n=1 or n=2 configuration to enhance the loss of runaway electrons. All these techniques need detailed experimental demonstration before they can be considered for ITER
Resource Requirements:	DIII-D with cryopumping and possibly neon massive gas injection of neon pellet to trigger runaways. Injection of high Z gases to stop runaway beam such as Ar, Xe, Kr
Diagnostic Requirements:	Measurements of runaway generation and loss
Analysis Requirements:	Analysis of measurements of runaway loss
Other Requirements:	Fine tuning of plasma position and control schemes to maintain stable runaway plasma position after thermal quench.

Title	256: Revisiting the current limit at q95~2 in RWM perspective
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	R. Stambaugh, A. Kellman, J.S. Kim, M. Okabayashi, H. Reimerdes, E. Strait and RWM Physics group	ITPA Joint Experiment :	No
Description:	The proposal is 1) to identify whether the MHD activity at q95 ~ 2, which imposes the operational current limit, would be RWM and 2) to assess if the existing current limit at q95~2 would be overcome by active feedback stabilization.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Develop a reproducible current-limiting Ohmic discharge with q95~2. When there is a doubt about the macroscopic MHD identity at q95~2, the RWM feedback control will be attempted to see if the mode growth or frequency varies. Specifically, when the application of external fields is found to affect the mode growth or evolution, there will be a systematic gain scan to see if the mode can be fully stabilized by feedback. Depending on the outcome about the mode at q95~2, the existing current limit will be re-examined in RWM physics point of view. It is to be noted that the MHD activity near q95 = 3 is desired to be suppressed/mitigated prior to the mode at q95~2 but that to eliminate the mode at q95~3 is not a pre-requirement, as long as a discharge survives until q95 approaches 2. Just in case a diverted plasma disrupts prior the mode at q95~2, a limiter plasma, whose edge q-values can be accurately identified, will be used as an alternative.
Background:	The operating current limit is often imposed by q95~2, one of which was analyzed in great detail in Doublet III [1]. While such study was being done in 1980s, there was little understanding about resistive wall mode (RWM). However, recent successful current-driven RWM feedback stabilization at q95 ~ 4 in DIII-D [2], as well as the RWM suppression for q< 1 in RFP, has motivated us to revisit the MHD activity responsible for the current limit at q95~2 in the perspective of RWM physics. Based on the old published data in Ref [1], the features (growth time ~ a few millisecond, non-rotating behavior) of the mode growth near q95 ~ 2 are nothing but typical RWM characteristics. Specifically, the non-rotating external kink mode with m/n=2/1 shown in Fig. 4 of the reference [1] could be re-classified as RWM. Considering that DIII-D is fully equipped with proper feedback control tool, as well as diagnostic capabilities, we will be able to not only address the mode identity at q95~2 but also assess if the existing current limit can be lifted due to the enhanced understandings of RWM physics. Reference [1] R. Stambaugh et al, Proc. of 10th Conf. on Plasma Phys. Control. Nucl. Fusion Res, IAEA (1985) [2] Y. In et al, APS-DPP (2008)
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	257: Beam-ion transport by plasma turbulence
Name:	Heidbrink heidbrink@fusion.gat.com	Affiliation:	UC, Irvine
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	C. Muscatello, M. Van Zeeland, G. McKee	ITPA Joint Experiment :	No
Description:	Study the spatial transport of beam ions in L-mode plasmas. Avoid MHD and fast-ion driven instabilities to isolate the transport associated with plasma fluctuations. Use FIDA to infer the transport of co-passing, trapped, and counter-passing ions. Vary the injection energy E and the plasma temperature T to test the predicted scaling of the transport with E/T.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Establish a baseline modest current (0.8 MA), full-field, L-mode plasma with monotonic q profile. Run inner wall or upper-single null configurations to stay in L-mode. Start with 1.5 sources (modulated for diagnostics) at normal operating voltages. 2) Drop the injection energy to ~50 keV; increase the number of sources to keep the plasma temperature constant. 3) In either condition #1 or #2, add ECH to raise Te. 4) Attempt to increase the beam power without exciting MHD to raise Ti. 5) Thoroughly document the best condition (fluctuation & fast-ion diagnostics). 6) If time permits, repeat with off-axis injection into a small, upwardly shifted plasma.
Background:	There has been a resurgence of interest in the transport of energetic ions by background plasma turbulence, with four different theoretical groups publishing (sometimes contradictory) papers in the last three years. The issue is important for alpha & beam-ion transport in ITER and for NBCD in DIII-D. The theories predict different scalings for passing & trapped particles vs. E/T, where E is the energetic particle energy and T is the ion or electron temperature. With our excellent fast-ion and fluctuation diagnostics, we are in a position to test the gyrokinetic simulations.
Resource Requirements:	7 beam sources; 5 gyrotrons
Diagnostic Requirements:	The full complement of core profile, fluctuation, and fast-ion diagnostics.
Analysis Requirements:	TRANSP, FIDA simulation code. Gyrokinetic simulations (both UCI and IPP groups have volunteered).
Other Requirements:

Title	258: Investigate the relative roles of density and colllisionality in RMP ELM control response
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Successful ELM suppression has been obtained at high density and collisionality and low rotation (pedestal vtor = 0), but at low density and colllisionality, the discharges are very sensitive to locked modes at low rotation/torque input. because ITER will operate at high density AND low collisionality, it is important to lift the degeneracy of the density and collisionality and to quantify the plasma rotation response.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use pedestal ECH and variable pumping to vary pedestal density at low collisionality Quantify the plasma rotation response (spin-up, damping etc.) as input torque is varied. - run pumped LSN ISS plasmas with n = 3 0 deg phasing and resonant q95; establish reproducible ELM suppression at low collisionality and density; - increase density by moving strike point away from pump - use ECH in pedestal to maintain pedestal Te and collisionality as density rises - quantify the plasma rotation response as density rises
Background:	In present tokamaks, it is difficult to decouple collisionality from density. However, in DIII-D it's possible to decouple these to some extent with a combination of varying the pumping (varying the strike point location) and locally heating the pedestal electrons via ECH as a way to raise the density and reheat the higher density pedestal to recover the collisionality.
Resource Requirements:	I-coil 3-4 gyrotrons cryopumping co and counter NBI
Diagnostic Requirements:	"standard" core and edge diagnostics high frequency magnetics with long digitizing window
Analysis Requirements:
Other Requirements:

Title	259: Giant sawteeth that never crash
Name:	Heidbrink heidbrink@fusion.gat.com	Affiliation:	UC, Irvine
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	Pinsker, Van Zeeland, Petty	ITPA Joint Experiment :	No
Description:	Create giant sawteeth with 4th and 6th harmonic heating of deuterium beam ions. Use ECCD to arrest current diffusion, allowing operation without sawtooth crashes at q0 < 0.9	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Attempt to reproduce the giant sawteeth observed in 1998, i.e., discharge 96043, using 4th harmonic heating of beam ions in an L-mode plasma. Use ECCD in two ways to increase the sawtooth period. (1) to flatten the magnetic shear near the q=1 surface. (2) As an off-axis current source to arrest current diffusion. In practice, the ECCD deposition layer will be scanned to both destabilize and stabilize the giant sawteeth.
Background:	With the improvements to our FW and ECCD capabilities, as well as the enormous improvements in core Alfven eigenmode and fast-ion diagnostics, it's past time to revisit this condition last seriously studied in 1998. Sawtooth stability is an important issue for ITER.
Resource Requirements:	1.5 MW of 60 MHz power essential; 2 MW of 90 MHZ desirable. At least 2 MW of ECCD. 330LT, 30LT, and 210RT essential.
Diagnostic Requirements:	FIDA, MSE, and ECE essential, as well as core plasma profile diagnostics.
Analysis Requirements:	Fit AE eigenfunctions with NOVA. Model fast-ion distribution with CQL3D and ORBIT-RF. Model sawtooth stability with GATO. FIDA simulation code.
Other Requirements:

Title	260: Study of NTM physics by ECEI system
Name:	Park hyeonpark@postech.ac.kr	Affiliation:	Pohang U of Science and Technology
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	H. K. Park, B. Tobias, C.W. Domier, T. Munsat, N.C. Luhmann Jr., A.J.H. Donné, M.J. van de Pol	ITPA Joint Experiment :	No
Description:	Among harmful MHDs in high beta plasma operation, it is well known that Neoclassical Tearing Modes (NTM) are prime candidate and many steps are taken to suppress the modes (mainly m/n=2/1 or m/n=3/2) with ECCD. Recent study of the m/n=2/1 NTM employing 2-D Electron Cyclotron Imaging system on TEXTOR, which provides real time 2-D images of the electron temperature fluctuation with unprecedented temporal and spatial resolution, revealed physics insight that clarified the process of growth and suppression of this mode. Since this study was performed with the circular shape on TEXTOR tokamak plasma with a limited window size, it is natural to extend the study to the shaped plasmas on DIII-D in order to understand the process in detail. In this experimental proposal, we will employ the new and improved ECEI system which has recently tested on TEXTOR device. This new system has a capability to expand the vertical size of the image at the same focal depth point and it is designed to have a capability to measure two images simultaneously from both the high and low field sides of the q~2 or q~3 surface so that the reconnection process can be decisively interpreted and considerably more comprehensive images can be provided than the prototype system used in the TEXTOR studies.. The primary objective of this proposal is to verify the heat and current distribution during suppression and growth of NTMs on DIII-D. It is known that the NTMs are often triggered by Sawtooth crash, which has been studied extensively and the physical process is relatively well established. Direct measurement of the initiation of the m/n=2/1 mode triggered by asymmetric fast heat transfer during reconnection process of the Sawtooth crash would be the most significant establishment.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This proposal can be combined with other NTM experiment. Plasma operation that is optimum for the NTMs is the requirement. Discharges at B(T) ~ 2T, I(p)~ 1.00 MA. ECCD power of ~ 2.5 MW. 1) Set the focal depth of the ECEI system to image both low and high field sides near the q~2 surface and measure the reference image of NTMs a) Two images; one with the vertically expanded and reduced images 2) Repeat the measurement with the focal depth moved with +5 cm and -5 cm wrt to the reference position. a) Two images; one with the vertically expanded and reduced images
Background:	Full understanding of the physics of NTMs is extremely important since these MHD events could be detrimental for high beta plasmas in steady state operation. In particular, the shrinkage and expansion of the island structure controlled by ECCD should be fully understood. NTMs have been studied mainly by the conventional ECE system which has a limited resolution in addressing the growth mechanism of the NTMs. At the same time, the detail process of suppression of this mode with ECCD which replaces the bootstsrap current within the island would be a great interest. This mode is known to be triggered by the m/n=1/1 mode (Sawtooth) crash. Recent study of sawtooth crash by ECEI system indicated that the mechanism of the Sawtooth crash is proven to be a random 3-D reconnection process. It would be interesting if the NTMs are triggered at the high field side or low field side. Experimental verification of the NTMs triggered by sawtooth crash and the growth mechanism of NTM can be quite interesting. Simultaneous measurement of the images of the reconnection process of the m/n=1/1 mode and growth of the m/n=2/1 NTM will improve the theoretical modeling of NTMs. This is one of the objectives of ITER MHD task.
Resource Requirements:	Many resources are needed to commission the new ECEI system
Diagnostic Requirements:	MSE, TS, X-ray and any other diagnostics that can address the NTMs
Analysis Requirements:	General equilibrium analysis and NTM modeling by M3D.
Other Requirements:

Title	261: RMP field penetration studies
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Scott Kruger, CS Chang	ITPA Joint Experiment :	No
Description:	The role of the plasma response to the externally applied RMP is crucial for understanding the ELM suppression physics and for predicting coil performance in ITER. Given the limited knowledge to date on the internal field in the plasma (the extend of rotational screening and/or resonant field amplification), it would be useful to benchmark models against a case predicted theoretically to have maximum RMP penetration: low rotation L-mode plasmas. This dataset has been motivated specifically by a desire to obtain a well documented case for extended MHD and XGC0 modeling, but will be useful for IPEC and other modeling as well.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	- run L-mode ISS discharges (use USN is needed to raise LH threshold; can reverse Bt to preserve io grad B drift toward xpoint) with high collisionality and low toroidal rotation - apply RMP in situation predicted theoretically to provide maximum penetration - vary input torque, I coil current, q95, and collisionality - data to validate NIMROD and M3D extended MHD models of field penetration - data to validate XGC0 neoclassical transport model of plasma response - likely also useful for IPEC and other codes.
Background:
Resource Requirements:	co and counter NBI cryopumping I coils possibly USN with reversed Bt and Ip to provide equilibrium with high LH lpower threshold and ion gradB drift toward the Xpoint. This point needs some further evaluation.
Diagnostic Requirements:	"standard" core and edge diagnostics used for RMP ELM control experiments
Analysis Requirements:	Significant analysis (profiles, kinetic EFITs, and vacuum fields) required to provide data needed by modelers (NIMROD, M3D, XGC0, IPEC, etc.)
Other Requirements:	Support from the modelers to simulate these discharges; Kruger (NIMROD) and Chang (XGC0) have expressed interest in modeling these discharges. I suspect that others will also be interested (Sugiyama and Straus; Park)

Title	262: Fast-ion transport by sawteeth
Name:	Heidbrink heidbrink@fusion.gat.com	Affiliation:	UC, Irvine
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	Muscatello, Van Zeeland, Lazarus	ITPA Joint Experiment :	No
Description:	Run at low q95 to make a large sawtooth inversion radius. Vary the shape (triangularity) to alter the sawtooth crash time. Determine the transport of both trapped and passing fast-ion populations.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Establish high current, low q95, modest density condition with a large sawtooth inversion radius. Start with a D shape. Modest beam power (to avoid Alfven eigenmode activity). 2) Various beam modulation patterns for optimal FIDA data. 3) Establish low q95 oval. 4) Repeat step #2.
Background:	Large sawteeth are anticipated on ITER with potentially enormous changes in the alpha profile (and possible localized losses). Kolesnichenko has a theory of fast-ion transport at the sawtooth crash that enjoys general acceptance. Because of its excellent MSE, ECE, and fast-ion diagnostics, we can put this theory to a rigorous test.
Resource Requirements:	5 sources
Diagnostic Requirements:	MSE, ECE, FIDA, and core profile diagnostics essential.
Analysis Requirements:
Other Requirements:

Title	263: POP Test of a Low-Z gas IRD
Name:	Wesley wesley@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Not requested
Co-Author(s):	Team MGI/DM	ITPA Joint Experiment :	No
Description:	The inverse rupture disc (IRD) concept first suggested by Paul Parks for massive gas injection disruption mitigation and collisional suppression of RE avalanching has potential application for ITER. There are various pros and cons about how the method can be applied to ITER or a reactor class tokamak and more general physics application questions about how the intense localized gas stream exiting the injector interacts with a tokamak plasma and whether the resulting very massive injection of slow-Z (D2 or He) gas results in higher assimilation or total electron densities relative to past 'diffuse' valve-based MGI. A 'one-off' proof-of-principle test of the concept with order(1000 torr-l) D2 or He in DIII-D would address an number of open application issues and provide a first comparison in DIII-D of close-coupled vs remote jet position.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Install a close-coupled 'single-shot' IRD injector cartridge on DIII-D. Initiate rupture disk opening. Monitor details of jet/plasma interaction and DM and inferred attributes of resulting fast plasma shutdown. Choice of D2, He or weak mixed gas TBD. Promising results could lead to follow-on tests with larger/smaller gas quantities or pulse duration, synchronous trigger means, ....
Background:	ITER needs a fast response (minimum initiation delay), highly reliable (low complexity), inexpensive and radiation-tolerant fast-plasma-shutdown method that is capable of delivering ~100 g of gas in less than 10 ms. Close proximity of the injector exit to the plasma surface is expected to improve gas uptake and increase the resulting in-plasma electron density (see AUG experience). An IRD injector might also comprise a 'back-up' or last-resort FPS or machine asset protection system for ITER that would complement other more-elaborate shutdown or DM systems
Resource Requirements:	Maximum thermal-energy target plasma (see #121 and 123); one-shot test injector; availability of close-to-plasma port location
Diagnostic Requirements:	As for massive pellet and massive gas studies; optical and/or bolometric monitoring of injection site. CO2 interferometer function above 1e16 cm3-m
Analysis Requirements:	Same as massive pellet or gas
Other Requirements:	A prototype test will likely use a one-shot injector cartridge that will require a clean vent to install and/or replace. End of week or end of 2009 run scheduling and/or readiness for an 'opportunity' clean vent required. Test in 2010 might be with a multi-cartridge or air-locked concept to faciliate systematic experiments.

Title	264: Measurement of structure, flows, and turbulence associated with tearing mode islands
Name:	Carter tcarter@physics.ucla.edu	Affiliation:	UC, Los Angeles
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	R. La Haye, C. Craig Petty (GA), W. A. Peebles, L. Schmitz, T. Rhodes, J. Hillesheim (UCLA)	ITPA Joint Experiment :	No
Description:	We propose to measure the localized structure of tearing islands to probe the basic physics of TMs and provide for detailed comparisons to analytical and computational (e.g. NIMROD) predictions. To enable the largest number of different fields to be measured, a slowly rotating island is preferred (~500 Hz). This will be achieved using a mix of co- and counter-NBI. Measurements of density, ion/electron temperature, magnetic field, flow and turbulence profiles will be sought for 2/1 and possibly 3/2 modes. Such an experiment should provide for an excellent data set for validating our understanding of tearing modes. Simultaneous measurements of profiles and turbulence will enable us to gather data on transport around TM islands, important in particular for understanding the degree of profile flattening which is critical for NTM stability.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce long lasting discharges with a large saturated 2/1 tearing mode. Trade off counter-injection beams for co-injection beams to bring the 2/1 mode rotation frequency to ~500 Hz. Use feedback control of the 2/1 mode frequency if the PCS has been upgraded to that feature. Use continuous 30LT and 330 beams to collect MSE and CER data at high time resolution. May need repeat shots to collect all of the CER data, and also may need repeat shots for the BES data. Doppler backscattering (DBS) will be used to measure intermediate-k turbulence and flows around the island with high time resolution. With favorable plasma conditions for a high signal to noise ratio, time resolution of the propagation velocity of turbulence of �10 μs or better is possible. Using current FM-DBS system, spatial coverage is limited and repeat shots may be needed to construct profiles. With the planned Comb-DBS system, single shot spatial coverage might be possible.
Background:	The experimental scenario will be based off of shot 131497, an ITER demo discharge with q95=3.1 which has a q=2 surface fairly far out radially (rho=0.8, and therefore would allow good diagnostic access) This discharge developed a rotating 2/1 mode at around 1 kHz. The proper conditions for this experiment are produced transiently in 131497 (from 4655 to 4675ms) and we will seek to controllably reproduce this state. Tuning this scenario for access for DBS measurements may be required. Initial investigation indicates DBS access to r/a~0.5 in X-mode would be possible in this discharge with perhaps 5% lower density.
Resource Requirements:	Machine time: 1 day, NBI: all 7 sources needed (not simultaneously)
Diagnostic Requirements:	MSE, CER, BES, ECE, FM-DBS and/or Comb-DBS (if available), Profile reflectometer.
Analysis Requirements:
Other Requirements:

Title	265: Image 3D structure in RMP H-modes
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	Jon Watkins	ITPA Joint Experiment :	No
Description:	Understanding the plasma response to externally applied n=3 RMPs as used for ELM suppression is important for assessing the viability of this approach for ITER and next step devices. Direct measurement of the 3D structures for comparison with various vacuum and plasma field models (TRIP3D, Wingen, IPEC, NIMROD) would provide valuation guidance for understanding how RMP ELM suppression works. This proposal involves sweeping regions of the plasma predicted to have significant 3D structures past/through the lines of sight of various diagnostics, including: 1. separatrix through "core" Thomson system to obtain 2D image in the flux expanded region near the crown where remnant islands and/or ExB convective cells should have significant extent 2. outer gap scans with different magnetic axis heights to build up a 2D picture with outer midplane diagnostics, including CER, profile reflectometry, and correlation/Doppler reflectometry; these shots should also allow a poloidally extend view by the BES system. 3. Xpoint/lower divertor sweeps through divertor TS 4. strike point sweeps past fixed floor probes and IRTV views to measure lobe structures in the divertor	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	-establish reproducible ELM suppression using n=3 even parity 0 deg phase I-coil RMP in ISS plasmas; use relatively high I-coil current for robust suppression; fine tune betaN as needed. - sweep separatrix through core Thomson chord to obtain 2D image - sweep outer gap at different Zmag values to obtain 2D image from CER, midplane probe, profile reflectometry, and fluctuation diagnostics - sweep strike point across fixed floor probes, DiMES TV, and fast IRTV; do this early in suppressed phase; we tend to do this last and miss much of the data if the discharge terminates early. - more problematic: scan lower divertor structure through divertor TS view chord; this may lead to loss of ELM suppression - vary I-coil current, parity and phasing: repeat. - time permitting, vary beam torque input
Background:
Resource Requirements:	co and counter NBI I coil cryopumping
Diagnostic Requirements:	CER, Thomson, profile and fluctuation reflectometry, BES, reciprocating probes, fixed floor probes, DiMES TV, fast IRTV
Analysis Requirements:
Other Requirements:

Title	266: Validate Wingen Magnetic field model using pedestal ECH
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	Andreas Wingen, Todd Evans, Jon Watkins, Oliver Schmitz	ITPA Joint Experiment :	No
Description:	Application of the RMP is predicted to bifurcate the magnetic field line footprint patterns in the divertor of RMP ELM-suppressed H-modes. These "lobes" have been measured and shown to correspond to the predictions of vacuum field models (TRIP3D, Wingen). In particular, Wingen's magnetic field modeling shows that while the overall structure and position of the lobes is governed by the non-resonant interaction with the separatrix, the lobes are "filled" with stochastic field lines from deeper in the pedestal. Confirming this internal structure of the lobes could provide experimental verification of the internal magnetic field structure in the plasma. This idea relies on generation of a hot, non-thermal electron population in the pedestal using ECH to "light up" the lobes to facilitate imaging the internal structure and to determine the depth of field lines connected to the target plate through these lobes.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	- establish robust RMP ELM suppression in the "standard" ISS shape with evan parity, n=3 I coil RMP. - use pedestal resonant ECH to generate a hot electron population at varying depths in the pedestal - monitor the divertor strike point conditions with diMES TV, fast IRTV, and probes to measure signatures of increased heat flux from fast electrons; the highest resolution diagnostic is the fixed floor Langmuir probe measurements of the floating potential - this will require sweeping the strike point across the probes for high spatial resolution measurements.
Background:
Resource Requirements:	I coil ECH cryopumping
Diagnostic Requirements:	divertor diagnostics, esp. fixed floor Langmuir probes, DiMES TV, and fast IRTV
Analysis Requirements:	calculations of the magnetic field line footprint patterns and internal structure of the lobes by Dr. Wingen will be required; Dr. Wingen will be at DIII-D as a UCSD Visiting Scholar from Jan. 9 through mid-April to participate in the planning and analysis of this experiment.
Other Requirements:

Title	267: Deep pellet fueling of RMP H-modes
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	Larry Baylor, Tom Jernigan, Todd Evans	ITPA Joint Experiment :	No
Description:	One signnificant drawback of RMP H-modes in ITER-similar shapes has been the significant reduction in the core and pedestal density (density pumpout) which means reduced performace and reduced Qfus in ITER. However, this pumpout effect displays a very different dependence on q95 than the ELM suppression, suggesting that while density pump-out may facilitate ELM suppression it isn't the same thing as ELM suppression. This suggests that it may be possible to recover the core density and pressure while maintaining ELM suppression by using high frequency (4-10 Hz) high field side pellet injection. In conjunction with varying the pumping, this has been shown to restore much of the lost density, but this important result for ITER has not been systematically documented.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	- establish robust ELM suppression in ISS discharges at low collisionalty - use rapid (4-10 Hz) injection of pellets from the HFS to restore the core density - adjust the strike point location relative to the pump aperature to fine tune the fueling while maintaining ELM suppression. - if available, try using shells loaded with deuterium and injected by the impurity pellet injector (see Torkil Jensen idea submitted by Evans) to provide deeper penetration of the fueling.
Background:
Resource Requirements:	Icoil cryopumping HFS 4-10 Hz pellet injection possible D2 loaded shells injected by the Impurity Pellet Injector
Diagnostic Requirements:	"standard" set of core and edge diagnostics for RMP ELM control
Analysis Requirements:
Other Requirements:

Title	268: Deep core fueling in ELM suppressed RMP H-modes using D2 filled hollow shell pellets
Name:	Evans evans@fusion.gat.com	Affiliation:	GA
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Not requested
Co-Author(s):	R. Moyer	ITPA Joint Experiment :	Yes
Description:	Deep core fueling in ITER may provide significant operational advantages compared to the relatively shallow fueling that can be obtained with cryogenic deuterium and tritium pellets. For example, in ELM suppressed RMP H-modes we sometimes see small transient D_alpha bursts in the divertor following the injection of a D2 fueling pellet. These bursts may results form a loss of edge confinement when the plasma spins down during the pellet deposition phase due to momentum conservation. It is assumed that the impact on the edge rotation and confinement can be reduce by depositing all the the D2 fuel inside rho ~0.2 where the cold particles can be quickly thermalized by the beam ions. If this is the case, we may be able to prevent the D_alpha transients following the pellet. In addition, such a method for central fueling may allow us to recover the core density lost when the RMP is applied (or obtain a much higher nGW in the core while keeping the edge collisionality low) without significantly increasing the pedestal pressure gradient. Thus, in theory we would not reach the point where peeling-ballooning modes become unstable and ELM suppression would be maintained. In effect, the goal is to shape the density profile such that the pedestal density remains low while the core density increases (i.e., controlled profile peaking). This could be of significant benefit for ITER especially if the increased edge transport produced by the RMP prevents impurities from penetrating much beyond the unperturbed separatrix. We expect that there will be a variety interesting physics physics effects that could be studied with this control capability that may have important practical applications for the development of higher performance plasmas in ITER (if the approach is successful). The success of the approach depends on our ability to deliver a thin shelled, hollow, pellet filled with D2 gas to the center of the plasma without rupturing it as it travels through the edge. Although this may appear to be somewhat speculative, such pellets, filled with argon, have been successfully injected in DIII-D during disruption mitigation experiments last year.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	In this experiments we would start with a low collisionality, high triangularity, ITER similar shaped plasma that has a highly reproducible RMP H-mode with robust ELM suppression. We would adjust the RMP coil current to maximize the density pump out (with low core and pedestal densities). We would then use the Li pellet injector to inject a single thin shell pellet, filled with ~10 atmosphers of D2), into the plasma. We would use the UCSD CCD camera to assess the penetration of the pellet. We will change the injection velocity of the shell pellet determine when the shell ruptures and how the D2 gas interacts with the plasma as it is deposited at various depths. A key result is to establish a relationship between the pellet penetration depth and the change in the rotation profile along with the changes in confinement and edge stability. The final step of the experiment is to combine an I-coil current scan with the pellet injection parameters that have the best core penetration and fueling performance to see if we can reach a core nGW~1 while maintaining good RMP ELM suppression.
Background:	This idea is based on an approach that was originally proposed for disruption mitigation (ref. T. Evans and T. Taylor "High pressure inert gas filled solid shell pellet designs for terminating high current tokamak discharges, DoE RoI, 12 Dec. 2006). It has been successfully tested using argon gas filled shell pellets during disruption mitigation experiments last year and we expect it can be extrapolated to deep fueling D2 pellets without too much difficulty.
Resource Requirements:
Diagnostic Requirements:	Fast IR camera, fast CCD camera, ECE, DISRAD, CER, fluctuation diagnostics, divertor Lamgmiur probes, full RMP H-mode diagnostics, fast filterscope data, divertor pressure gauges.
Analysis Requirements:	Image analysis of the CCD and IR cameras, TRIP3D, NIMROD modeling, EMC3-EIRENE modeling.
Other Requirements:

Title	269: Low rotation ITER scenarios - impact on confinement and fusion performance
Name:	Doyle doyle@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ITER Demonstration Discharges	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	While much progress was made in successfully generating ITER demonstration discharges in 2008, a critical factor has not yet been addressed, namely the effect of rotation on confinement and overall fusion performance. The 2008 experiments utilized all co-neutral beam injection; here the proposal is to use a co-/counter-beam mix to scan rotation down to ITER relevant values and determine the effect on confinement and fusion performance. Most important scenarios to examine are the baseline and advanced inductive (steady-state would not be addressed)	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The proposed experimental approach is relatively straight forward straightforward: reestablish ITER demonstration scenarios developed in 2008, and change co-/counter-NBI mix to scan rotation downwards. Determine variation of confinement with rotation.
Background:	The ITER demonstration scenarios operated in 2008 focused on matching ITER beta, confinement and normalized current targets. The most glaring physics mismatch with regard to ITER conditions in these discharges was plasma rotation (Mach no), which was much higher than predicted for ITER. Given the effect of rotation on confinement and other properties, it is critical to evaluate the performance of the ITER scenarios at lower rotation. P
Resource Requirements:	All 7 NBI sources.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	270: Solenoidless startup studies in DIII-D
Name:	Cunningham none	Affiliation:	MAST
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	D. Gates, D. Mueller (NSTX), N. Eidietis, D. Humphreys, A. Hyatt, G. Jackson, J. Leuer, P. Politzer, R. Prater, P. West	ITPA Joint Experiment :	No
Description:	Goal of experiment is to achieve diverted H-mode plasma at the highest current possible without utilizing the E-coil or inner F-coils (F1-F4). We will utilize the divertor coils and vertical field coils (F5-F9) and ECH/ECCD to establish sufficient plasma current to allow neutral beam absorption with a goal of achieving N-beam/ECCD current drive to steady state values. If time permits we may add HHFW.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Tasks/testing expected within this project are: 1) magnetic scenario development, 2) fueling/density control 3) ECH/ECCD optimization, 4) position control development, and 5) NB coupling to generate a steady state plasma. Substantial experiments have been performed on other tokamaks and results from these experiments will be utilized to produce guidance for overall scenario development. Primary magnetic scenario under development utilizes current ramps in F6-F9) to produce a first order null in the outboard limiter region. Loop voltages of-order 5V are predicted and should be sufficient for ECH assisted breakdown and current ramp up. Plasma will naturally expand inward and increase in minor radius from the outboard limiter. ECCD and early beams will be used to reduce the flux consumption during this phase of startup. Final configuration is expected to be fully non-inductive H-mode plasma driven by N-beams to the maximum current and beta possible with the current DIII-D system. G Cunningham (MAST) and D. Gates (NSTX) will lead these experiments and provide a wealth of knowledge regarding solenoidless startup.
Background:	Plasma initiation without utilizing central solenoid coils is the â??holy grailâ?? of compact devices like spherical tokamaks. Reactor designs like ARIES-ST assume current can be driven to full parameters without a PF solenoid. However, experimental validation of solenoidless startup is still an ongoing research topic. Many techniques have been tested in different toroidal devices over the years. Studies have been performed on DIII-D using the biased divertor to produce helicity injection (Schaffer â??92). Steady state H-mode has been established with N-beam current drive using a small ohmically generated seed current (Simonen â??88). On other devices studies have been performed using only induction from outside PF coils combined with auxiliary heating current drive to generate startup currents. For PF coils inside the plasma like on MAST/START startup using outside coils is more straight forward with plasma formation around inside coils providing seed current for subsequent plasma formation and compression. For devices with PF coils outside the vacuum region, use of outer PF coils to produce plasma is much more difficult. JT60U has been able to produce 100kA with ECH and outer PF coils alone (Ushigome â??06). Other devices like NSTX have also produce small transient plasmas utilizing only outside coils (Menard â??04). Previous major tokamak experiments have failed to generate a complete solenoidless startup scenario to a steady state configuration. DIII-D is uniquely qualified in this area with high degree of control over our PF coils and substantial auxiliary heating/current drive in our EC and neutral beam systems. Our F-coil topology allows us to study varying degrees of PF coil poloidal utilization, from full poloidal coverage (F1-9) to only outside coils (F6-7). Our ECH and Neutral beams provide substantial heating and current drive at levels not available in other machines. This work is also expected to enhance the knowledge required for ITER startup by generating flux savings techniques for the current ramp up phase.
Resource Requirements:	Tokamak, ECH (5-6 Gyrotrons), diag NB + 8MW Co-NB, (possibly HHFW), Request 2 days separated into: 2 one-half day experiments for testing of the individual elements of the experiment and for synchronization and a one full day experiment to generate the final solenoidless/ECCD/N-beam plasma.
Diagnostic Requirements:	Fast magnetics, MSE, Thompson Scattering, Spred, Visible camera view bumper limiter, Bolometers, IR camera, CO2 Interferometer, ECE, SXR
Analysis Requirements:
Other Requirements:

Title	271: SOL characteristics and PFC energy loads during ramp-up/down
Name:	Pitts richard.pitts@iter.org	Affiliation:	ITER Organization
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Requested
Co-Author(s):	P. C. Stangeby, C. Lasnier, J. A Boedo, D. Rudakov, J. Watkins	ITPA Joint Experiment :	No
Description:	Study thermal load and SOL plasma characterization during limiter ramp-up and down phases in DIII-D to validate ITER first wall load specifications.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	DIII-D has recently studied a LFS ITER ramp-up scenario (G. L. Jackson et al., NF 48 (2008) 125002) and is thus in an excellent position to contribute to this SOL characterization effort. The proposed experiments call for a series of steady state limiter discharges, both LFS and HFS (since an ITER HFS start-up scenario is not excluded) with varying density, plasma current and input power, to provide the best opportunity to obtain SOL profiles with the reciprocating probe diagnostics. Turbulence data may be straightforwardly gathered simultaneously, offering interesting possibilities to study the dynamics of radial transport in LFS and HFS limiter configurations. The idea, if possible would be to attempt at least two plasma currents, providing q values close to those expected in ITER during ramp-up/down and then alternately vary the density at fixed power and the power at fixed density such that two points (corresponding to the probe reciprocation possibilities) are obtained in steady state per pulse. This builds up a database with which to test the ITER assumptions for the ohmic limiter phases. Measurements should then be attempted at selected points during the ramp-up itself to check the validity of a steady state measurement against a time varying profile. Finally, measurements should be made in the L-mode phase immediately after the ohmic ramp up to check the correspondence of the L-mode SOL profile widths (after suitable corrections) to those obtained in the limiter phase. If possible, this should be done at a couple of overlapping powers/densities with the caveat of remaining in L-mode. During the L-mode phase, extra benefit can be drawn from divertor viewing IRTV, allowing upstream probe measurements to be benchmarked against target heat fluxes (provided there is sufficient power and/or low enough density for the divertor plasma to remain attached).
Background:	Design efforts aimed at shaping of the ITER beryllium first wall blanket shield modules are currently underway with the aim of providing a plasma-facing surface in the main chamber capable of protecting leading edges and misalignments. Thermal load specifications, both transient and steady state, are guiding this design process. This shaping exercise is converging on a scenario comprising a set of 18 poloidal bumper limiters on the low field side and poloidally and toroidally shaped elements on the central column which would allow plasma start-up without the requirement for the two specially designed high heat flux start-up limiters currently envisaged on the outboard midplane. The load specifications that are being used to design these first wall limiting surfaces in the limiter phases are based on extrapolations of scalings of heat flux widths from L-mode diverted data (AUG, JT-60U and JET) and published in the 1999 IPB. To adapt this to the case of a limited plasma, corrections have been applied to account for the effect of a variable number of poloidal limiters. A number of different combinations of input power and plasma current have been specified (e.g. 2.5 MA/2.5 MW, 5 MA/5 MW and 7.5 MA/7.5 MW) for the ITER case, with the last two pairs corresponding to the presently foreseen highest plasma current/power at the limiter/X-point or X-point/limiter transition during ramp-up and ramp-down respectively in ITER. New data from present devices is urgently required to verify this ITER approach and confirm the thermal load specifications for the limiter ramp-up and down phases. Together with the shaping choice, these thermal loads determine the necessary heat flux handling capability (water cooling technology) that must be designed into the start-up shield modules. Probing the SOL of both inner and outer wall limited configurations with the array of turbulence diagnostics available on DIII-D also offers some interesting opportunities to study the nature of radial turbulence driven transport (e.g. the ballooning nature) in situations in which plasma flow is prevented from outboard to inboard sides and vice-versa.
Resource Requirements:	Full day
Diagnostic Requirements:	Main SOL reciprocating probes, IRTV, divertor Langmuir probes etc.
Analysis Requirements:
Other Requirements:	OEDGE and UEDGE code analysis

Title	272: Effect of Plasma Rotation upon MGI Impurity Assimilation
Name:	Eidietis eidietis@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Observe the effects of various levels of plasma rotation upon the assimilation of MGI impurities of differing masses.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A �??standard�?? disruption mitigation plasma will be formed. Shot to shot, the plasma rotation will be adjusted between [high, low, none, reversed] prior to MGI. Multiple scans will be undertaken using a different MGI species (H or D and Ne or Ar) in order to assess if a mass dependency. Time allowing, magnetic braking can be added to modify the rotation shear near the edge.
Background:	Massive gas injection (MGI) is considered a prime candidate for rapid shutdown in ITER. Measurements of impurity assimilation during MGI on DIII-D were reported in [Hollman 2008]. However, the use of uniform plasma parameters for disruption studies leaves the effects of numerous salient plasma properties upon MGI assimilation unexplored. This experiment looks to observe the dependence of impurity assimilation upon plasma bulk rotation.
Resource Requirements:	Standard MGI target plasma, co & counter beams, MEDUSA injector, supplies of candidate gas mixtures, between shot gas change capability. Time: 1 day
Diagnostic Requirements:	Standard, fast camera, CO2 interferometer, AXUV, fast bolometers, ECE, SPRED, CHERS, FIDA
Analysis Requirements:	KPRAD, rotation profiles
Other Requirements:	H2 experiments may impact subsequent IC or other species sensitive experiments. Need end of week or H campaign scheduling.

Title	273: Study of runaway beam loss into wall
Name:	Yu yujh@fusion.gat.com	Affiliation:	GA
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):	E.M. Hollmann, A. James	ITPA Joint Experiment :	No
Description:	This experiment will image the structure and dynamics of a runaway electron (RE) beam, and will study the instability that causes the RE beam to strike the wall. The purpose is to understand RE/wall interactions by measuring the interaction spot size, measuring the drift speed of the RE beam, and diagnosing core MHD that may play a role in destabilizing the beam.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A RE beam will be generated at the beginning of the shot by the startup toroidal electric field accelerating a non-thermal electron population created with on-axis ECCD. Assuming the RE beam is fairly well confined, the RE population should be accelerated to sufficiently high energies by the end of the shot so that RE synchrotron radiation could be detectable in IR to visible wavelengths (requiring RE energy > 30 MeV). We plan to use the TEXTOR fast IR camera, but if this camera is not available we will attempt this experiment with the UCSD fast visible/near IR camera.
Background:	Runaway electrons can damage the vacuum vessel of large machines and are a concern for ITER. Understanding the transport and structure of runaways may lead to the development of methods to prevent runaway damage to plasma facing components.
Resource Requirements:
Diagnostic Requirements:	TEXTOR fast IR camera, UCSD fast visible/near IR camera, scintillator array (after upgrade), neutron detectors, fast IR divertor camera, LLNL cameras.
Analysis Requirements:
Other Requirements:

Title	274: Ion gyro-radius scaling of the coupled turbulence/zonal flow system at the LH transition
Name:	yan none	Affiliation:	UCSD
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	G. McKee, G. Tynan, J. Boedo, R. Moyer, S. Muller, D. Schlossberg	ITPA Joint Experiment :	No
Description:	Investigate the ion gyro-radius scaling of the coupled turbulence/zonal flow system before and during the LH transition by varying toroidal magnetic field (while keeping other parameters constant) for both co-current and balanced plasma toroidal rotation regimes.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Neutral beam(s) will be used to generate the co-current and balanced plasma toroidal rotation regimes. In each regime we would vary the toroidal magnetic field and consequent ion gyro-radius, while keep the other parameters nearly constant (e.g. the safety factor q, input beam power and plasma density). The plasma will be operated in an upper single null configuration, with the ion grad-B drift pointing away from the X-point, to allow the plasma to be maintained in L-mode for longer time window at higher input power, as well as lower-single null at lower power. The safety factor will be kept at a high value ~5-6 to maximize the GAM amplitude. In each condition, 2D turbulence characteristics will be measured with the expanded 8x8 BES array near the LCFS and SOL, which will provide the turbulence amplitude and flow shear measurements. In the meanwhile a Reynolds stress probe will be used to measure the plasma density, floating potential and the resulting Reynolds stress at very low injection power (reduced beam voltage as appropriate) in the lower single null condition (lower LH threshold). This measurement will also be compared with the BES measurement to investigate the statistical correlation between these two including 2D velocimetry analysis to infer components of the Reynolds stress.
Background:	The existence of the geodestic acoustic mode (GAM) and the zero-mean-frequency (ZMF) zonal flow have been clearly identified experimentally in tokamak and stellarator plasmas [1,2]. Those flows are predicted to be generated by the plasma turbulence and may relate to the mechanism for L- to H- mode transition [3]. In the CSDX linear plasma device a transition to drift turbulence has been demonstrated with increasing magnetic field [4]. This effect is attributed to the reduction of the ion-ion viscosity via a reduction of the ion gyro-radius. More recent studies have also demonstrated an increase of the zonal flow strength with increasing magnetic fields and the critical gradient behavior for the coupled turbulence/zonal flow system. In addition, the power threshold for the transition to H-mode has been found to depend strongly on the plasma toroidal rotation [5], and the GAM/zonal flow dynamics likewise vary consistently with rotation. Based on these results we propose to experimentally investigate the ion gyro-radius scaling of the coupled turbulence/zonal flow system at both co-current and balanced plasma toroidal rotation regimes. [1] G.R.Mckee, et al., Phys. Plasmas, 10, 1712, (2003) [2] A.Fujisawa, et al., Phys. Rev. Lett., 93, 165002 (2004) [3] K. H. Burrell, Phys. Plasmas 4, 1499 (1997) [4] M.J.Burin, et al., Phys. Plasma, 12, 052302, (2005) [5] G.R.Mckee, IAEA 2008
Resource Requirements:	NBI
Diagnostic Requirements:	BES (8Ã?8 array) and Langmuir probe (Reynolds stress)
Analysis Requirements:
Other Requirements:

Title	275: Effect of NTM's upon Shell Pellet Deposition
Name:	Eidietis eidietis@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Low-z shell pellets were successfully tested in DIII-D in 2008. However, the largest density perturbation of the pellets was observed at rho=0.5, and the peak radiation at rho ~ 0.8. Ideally, the density and radiation peaks would occur much closer to the core. This experiment tests if NTMâ??s of varying size (controlled in realtime) can be utilized to improve the inward transport of the radiating impurities and shift the density and radiation peaks towards the plasma core.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Low Z shell pellets containing argon will be shot into high-power H-mode plasmas. The effects of NTM of width and type (2/1 3/2) upon the impurity deposition profiles will be observed.
Background:	While MGI provides a good candidate for relieving the thermal and EM stresses of a rapid shutdown, it is unlikely to be able to provide sufficient core density to collisionally suppress runaway electron generation. A possible approach for RE suppression is the delivery of large quantities of radiating impurities to the core plasma using low-Z shell pellets containing high-z radiating impurities. These pellets allow for impurity transport to the core before the massive MHD of the thermal quench and subsequent RE production.
Resource Requirements:	Lithium pellet injector, high-power H-mode discharges Time: 0.5 day
Diagnostic Requirements:	Standard, fast camera, CO2 interferometer, AXUV, fast bolometers, SPRED, Thomson, MSE
Analysis Requirements:
Other Requirements:	Custom shell pellets (to be made in cooperation with the GA inertial confinement group).

Title	276: Image NTM structure during ECCD growth and suppression
Name:	Yu yujh@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	R. La Haye, M.A. Van Zeeland, C.C. Petty	ITPA Joint Experiment :	No
Description:	The goal of the experiment is to measure the evolution of the line-integrated 2D structure of islands during ECCD suppression and growth of the islands. The purpose is to investigate how the island structure changes in response to ECCD. Of particular interest would be imaging the onset of island growth (using counter-ECCD) and during the final stages of island suppression (using co-ECCD).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Trigger 2/1 NTM with a beta ramp up in a hybrid scenario plasma, and deposit counter- (co-) ECCD at the q = 2 surface for island growth (suppression). Image the same sized islands with no ECCD to compare island structures. These shots could be done piggy-back during other NTM control days. The requirement for imaging is that the electron density be as high as possible while avoiding cutoff.
Background:	NTMs limit beta and can lead to disruptions. ECCD aimed at the island rational surface will most likely be used in future devices to restore the bootstrap current deficit, which supports the island in an NTM. In addition, understanding changes caused by ECCD to the classical tearing index is an important area of study.
Resource Requirements:	0.5 day, gyrotrons
Diagnostic Requirements:	UCSD fast camera, core plasma diagnostics
Analysis Requirements:	--
Other Requirements:	--

Title	277: Benchmark measurements of NTM island structure using fast camera and ECE
Name:	Yu yujh@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	R. La Haye, M.A. Van Zeeland, M. Austin	ITPA Joint Experiment :	No
Description:	The goal of the experiment is to measure the 2/1 NTM island width using fast imaging and ECE during the same shot.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Trigger 2/1 NTM with a beta ramp up in a hybrid scenario plasma. Keep core density at 5x10^13 cm-3, which is below cutoff for ECE channels near the q = 2 surface but sufficiently high to detect visible bremsstrahlung with the camera.
Background:	The fast camera has been used to image NTMs during the 2008 experimental campaign. In future devices using cameras that have real-time data output, imaging may be used to steer ECCD for NTM control provided that camera data produces reliable measurements of the island location and structure. As a first step toward this goal, imaging of island location and structure needs to be benchmarked with other diagnostics such as ECE.
Resource Requirements:	0.5 day
Diagnostic Requirements:	UCSD fast camera, ECE, core plasma diagnostics, Mirnov coils.
Analysis Requirements:	--
Other Requirements:	--

Title	278: Effect of mode coupling on NTM dynamics
Name:	Yu yujh@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	R. La Haye, C. Petty, R. Prater, E. Strait, M.A. Van Zeeland	ITPA Joint Experiment :	No
Description:	The goal of the experiment is to image island structure during frequency locking of a 3/2 NTM to a 2/1 NTM in order to study NTM mode coupling, and compare with NTM dynamics in the absence of frequency locking.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In a hybrid scenario plasma, ramp up beta to trigger the 2/1 NTM. Hybrid plasmas typically have a naturally occurring 3/2 NTM. The time window of interest will be when the 3/2 NTM slows down and frequency locks to the 2/1 mode. The density pumps out during the 2/1 mode, and thus gas puffing should be done to maintain high density while avoiding disruptions. The gas puffing should be neon or argon to increase Zeff and maximize bremsstrahlung emission. Counter NBI should be used to keep the mode frequencies low to allow long exposure times, while avoiding mode locking to the wall. Perform a scan of plasma rotational shear to find threshold for mode coupling. Compare dynamics of NTMs with and without mode coupling.
Background:	When two NTMs simultaneously exist in the plasma with similar frequencies, the magnetic islands from each NTM can couple to each other and frequency lock. Based on MSE data, the pressure profile is flat when the two modes lock together, which makes it difficult to understand how the islands are maintained. Imaging the island structure during frequency locking could provide physics insight into how the islands support themselves.
Resource Requirements:	0.5 day
Diagnostic Requirements:	UCSD fast camera, core plasma diagnostics.
Analysis Requirements:	--
Other Requirements:	--

Title	279: Real-time Beta Control via NTM Width Modulation
Name:	Eidietis eidietis@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Control/Operations	Presentation time:	Not requested
Co-Author(s):	A. Welander	ITPA Joint Experiment :	No
Description:	We propose to control plasma beta in realtime by modulating the width of 3/2 or 2/1 NTMâ??s using ECCD. The islands will be grown or suppressed in order to obtain and maintain a beta target by a reduction or improvement in confinement.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We will begin with a well-established NTM control scenario in order to establish a nominal beta for the scenario. The target beta will then be lowered and raised in steps from that nominal value in order to exercise the full spectrum of NTM growth and reduction control. That sequence will be followed by a series of over-heated shots in which the beams are turned on at successively higher powers without feedback (a very crude approximation of a nuclear burn) in order to explore the limits of the beta control and to examine the differences (if any) between 3/2 and 2/1 modulation.
Background:	In present-day tokamaks, beta control is primarily accomplished by modulating the auxiliary heating sources, particularly NBI. However, in burning plasmas the effect of auxiliary heating will be merely a perturbation relative to the much greater nuclear heating. Beta control will have to migrate from the present â??turning the hose on and offâ?? 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 â??release valveâ?? on the plasma. The suppression of NTMâ??s 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 unutilized.
Resource Requirements:	Lower pump (upper pumps desirable), He cooled. All NB 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. TIME: 2 hour test + 0.5 day
Diagnostic Requirements:	Magnetic (fast and slow),CER on 30L and 330L,MSE, Thomson,CO2 interferometers, ECE radiometer
Analysis Requirements:
Other Requirements:	Real-time Mirnov acquisition, Control of gyrotron modulation.

Title	280: Sawtooth instability studies with gas puffing for imaging
Name:	Yu yujh@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	M.A. Van Zeeland, E. Lazarus, A. Turnbull	ITPA Joint Experiment :	No
Description:	The goal of the experiment is to image the sawtooth instability while making detailed measurements of the plasma core current, density, and temperature.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce q ~ 1 at the plasma center. Vary the core current profile using ECCD deposition either directly at the magnetic axis or slightly off-axis to study how the core current profile affects sawtooth behavior. The changing sawtooth behavior will be correlated with measurements of fast ion transport using FIDA. To produce sufficient bremsstrahlung signal above the camera detection limit, gas puffing of impurities such Ar or Ne to increase Zeff should be used, with the quantity and rate of gas injection below the fast shutdown limit. Increasing Zeff is needed for imaging because electron density should be kept sufficiently low to allow measurements from MSE, ECE, Thomson, and FIDA. Fine tuning the gas injection rate will take a few shots. Producing enough camera signal while maintaining the capability of core diagnostics will be challenging, and thus this experiment could wait until the camera has an image intensifier. These shots could potentially be done piggy-back during other sawtooth or giant sawtooth run days.
Background:	Complete explanation of the sawtooth instability remains elusive, and understanding how core MHD affects the fast ion population (and vice versa) is important for ITER. Visible imaging used in conjunction with detailed measurements of the q profile, density, and temperature, may provide additional clues to the underlying mechanism driving sawteeth.
Resource Requirements:	0.5 day or piggy back, gyrotrons
Diagnostic Requirements:	UCSD fast camera, core plasma diagnostics.
Analysis Requirements:	--
Other Requirements:	--

Title	281: Measurements of spiral MHD structure induce by pellet injection
Name:	Yu yujh@fusion.gat.com	Affiliation:	GA
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	A. James, E.M. Hollmann, E.J. Strait	ITPA Joint Experiment :	No
Description:	The goal of the experiment is to make additional measurements of the structure and dynamics of an apparently novel MHD mode that has a spiral structure, which was seen in shot 134281 after injecting a polystyrene pellet into the plasma core.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The experiment is exploratory and could consist of three or four piggy-back shots done at the end of a run day. The target shot is 134281 with pellet injection from the impurity pellet injector. After establishing the target shot, perform q(0) scan to observe any effects on mode structure.
Background:	During shot 134281 a polystyrene pellet was injected into the plasma core with the intended purpose of diagnosing runaway electrons. No runaways existed during this shot and the pellet traversed across the plasma core while being ablated. The core density rose to 2x10^14 cm-3. Approximately 5 ms after the pellet ablation was complete, a spatially extended m = 3 perturbation grew in the plasma core and rapidly changed structure within a few rotation periods to m = 2 extending to r/a > 0.7. During this transition, the spiral structure developed in the plasma core (r/a < 0.3) and lasted for a few ms, and had the appearance of a distinct eigenmode. The spiral structure did not resemble previous observation of "snake"-like perturbations following pellet injection in JET [Weller et al., PRL 1987]. Understanding core MHD activity induced by pellet injection may be important for future devices that use pellets for ELM pacing or fast shutdown.
Resource Requirements:	3 or 4 shots, impurity pellet injector
Diagnostic Requirements:	UCSD fast camera, soft x-ray arrays
Analysis Requirements:	--
Other Requirements:	--

Title	282: He Startup in DIII-D with transition to D/H plasma in Ip ramp
Name:	Leuer leuer@fusion.gat.com	Affiliation:	GA
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Requested
Co-Author(s):	N. Eidietis, D. Humphreys, A. Hyatt, G. Jackson, P. West	ITPA Joint Experiment :	No
Description:	Goal of experiment is utilize Helium as the initial gas for plasma breakdown and initial plasma current ramp in the presence of strong ECH. During the current ramp-up we will transition to H/D plasma with gas puff and beams. The object is to compare the final ITER H-mode plasma to see if it is identical to a standard hydrogen discharge. The use of He (with ECH heating) is expected to require lower breakdown voltage and less resistive volt-second losses. In addition, differences in He/D-Beam interaction and potential MHD differences would be explored. Use of a different gas for startup provides us with a different control knob, which may be very beneficial for ITER.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We plan on using He puff to establish the initial density required for breakdown and initial current ramp. Strong ECH should provide additional ionization requirements over that required for hydrogen. H (or D) will be puffed in following the initial He plasma initiation and the change of He => H/D will be documented. The object is to determine is the He reduces the loop voltage requirements and ultimately reduces the resistive flux consumption of the plasma. The switch over point of the He/H transition will be varied during the ramp. We expect to use the nominal ITER scenario to H-mode for comparison. We want define any voltage and/or flux gain savings and insure the final steady state hydrogen H-mode plasma generated using Helium for startup has identical performance as the nominal hydrogen initiated plasma. We will also determine and advantage He may have in ramp-up as a consequence of deuterium N-beams interacting with the He plasma. This could allow beam absorption at lower density or early in the current ramp. This also will provide valuable information on He retention in H plasmas during ramp up and at steady state with application to ash removal in ITER. We will use Argon frost cryo-panels to trap He during the discharge. Control of residual H from wall�??s is expected to present the most challenge to this experiment, as it was in JT60U. If time permits we could explore other gases (like Neon) and will use both D and H for comparison in the standard ITER H-mode plasma.
Background:	The nominal voltage breakdown requirements for ITER is an Electric field E=0.3V/m with ECH for ionization. This is a margin of two over the minimum breakdown electric field achieved in DIII-D of E=0.15V/m with ECH assist (Lloyd �??91). The electric field requirement sets coil voltage requirements and power supply requirements. Reduction of the electric field requirements can have a substantial impact on machine requirements and costs. JT60U using LHRF has shown similar electric field requirements to those established in DIII-D using ECH in hydrogen discharges; however, in helium the discharge electric field requirement is reduced to E=0.08 V/m (Yoshino, �??97). Townsend avalanche theory is reported to partially explain the reduced field requirement for He over H. In addition, energy loss is reported lower in He plasma and could lead to a reduction in resistive flux consumption. Reportedly, use of ECH could reduce the electric field requirements even lower. In addition, NSTX is reporting that small amounts of Lithium in a hydrogen plasma reduces the ohmic flux loss (Menard, �??08). Since confinement is low during the current ramp-up phase the discharge the plasma gas is changing out may time during the ramp up. Use of a different gas for startup relative to that used in the flat top (H/D or D/T in ITER) is of little consequence if the remnants of the initiation gas are purged from the system. We will determine if there is a benefit of He startup in voltage requirements and resistive flux consumption and if the final plasma D/H plasma is identical to the standard ITER H-mode plasma. In addition, we will explore impact of N-beam into a He plasma for flux reduction and information on He retention for ash removal in ITER. The experiment is easy to perform and should have minimal impact on other experiments since He is used routinely for glow discharges between shots.
Resource Requirements:	Tokamak, ECH, diag NB + 8MW Co-NB, (possibly HHFW), Request experiment after boronization Request 1 Day experiment (could be 2 one-half day experiments)
Diagnostic Requirements:	Fast magnetics, MSE, Thompson Scattering, Spread, Visible camera view bumper limiter, Bolometers, IR camera, CO2 Interferometer, ECE, SXR
Analysis Requirements:	Utilize new EFIT based Flux/Ejima calculational tool
Other Requirements:	--

Title	283: Confirm that flow shear acts to reduce NTM island size
Name:	La Haye rbrtlahaye@gmail.com	Affiliation:	Retired from GA
Research Area:	Stability	Presentation time:	Requested
Co-Author(s):	R.J. Buttery	ITPA Joint Experiment :	Yes
Description:	The rotation and thus rotation shear will be varied at near constant bootstrap current to investigate the effect on the saturated m/n=3/2 mode from all co-NBI to as far as one can go in the counter direction.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce an m/n=3/2 mode in a plasma that has been run stably down to balanced injection for 2/1 modes; keep beta and density constant by feedback and run the rotation down from all co-NBI to balanced or slightly counter in steps, and back.
Background:	Discharges in 2007-2008 were run with different torque and beta ramped up to find the m/n=2/1 onset. A preexisting m/n=3/2 mode allowed additional analysis that found: more flow shear that allowed higher 2/1 stable beta, tends to obviate the effect of higher beta on 3/2 island width. The database does not include the "crossing" condition of fixed beta, density and thus local bootstrap current versus rotation. This is input to ITPA-MHD 2009 MDC-14 "Rotation effects on neoclassical tearing modes ".
Resource Requirements:	1/2 DAY, 7 beams, cryopumping, ITER shape at q95=4.4, keep betap~0.9 as in #133512 to stay 2/1 stable with good n=1 EFC down to balanced beams.
Diagnostic Requirements:	Standard suite.
Analysis Requirements:	Standard.
Other Requirements:	None.

Title	284: Structure of the plasma response to kink resonant perturbations in AT steady-state scenario
Name:	Lanctot matthew.lanctot@science.doe.gov	Affiliation:	Department of Energy
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	See #32	ITPA Joint Experiment :	No
Description:	Active MHD spectroscopy is often used in AT steady state discharges to evaluate the n=1 stability limits using magnetic sensor arrays. Typically, the I-coil currents used are insufficient to drive an internal response observable on the SXR diagnostic. The goal of this experiment is to apply a large amplitude perturbation in an AT discharge in order to measure the internal structure of the mode. These measurements can be compared to measurements in LSN shaped plasmas and to MARS predictions of the driven mode structure.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Following the successful application of a small amplitude, slowly-rotating, kink resonant I-coil perturbation in an AT steady state discharge, apply a slowly increasing perturbation and measure the internal response with SXR diagnostic.
Background:	See #32.
Resource Requirements:	1 shot in piggy back
Diagnostic Requirements:	SXR, I-coils configuration that matches kink mode structure (i.e. 240 deg quartets), equilibrium diagnostics
Analysis Requirements:	--
Other Requirements:	--

Title	285: Fast shutdown with shell pellets
Name:	Hollmann ehollmann@ucsd.edu	Affiliation:	UCSD
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	Paul Parks, Todd Evans, Alex James, Phil West, Jonathan Yu	ITPA Joint Experiment :	No
Description:	Investigate the feasibility of using low-Z shell pellets to deliver dispersive high-Z payload to core of ITER for collisional suppression of runaway electron avalanche.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	It is proposed to fire different shell pellets into well-characterized, quiescent DIII-D discharges to study key physics issues for shell pellet shutdown. Key issues to be studied here are: 1) ablation rate of hard, low-Z shells, 2) dispersal of dust payload vs solid payload, 3) characteristics of inverse thermal quench when sufficient high-Z payload is deposited directly into the core, and 4) critical current gradient for onset of standard thermal quench. For issue 1), it is proposed to use D = 2 mm shell pellets with a t = 0.1 mm B4C shell. As a backup plan, small solid D = 1 mm graphite pellets could be used. Ablation rate will be obtained with fast camera imaging of the pellet trajectory. For issue 2), it is proposed to use D = 2 mm, t = 0.4 mm polystyrene shell pellets filled with solid carbon and with powdered carbon for comparison of solid vs dust core deposition rates. For issue 3), it is proposed to use D = 2 mm polystyrene shell pellets filled with XeF2 powder. Finally, for issue 4), it is proposed to use solid D = 2 mm polystyrene shell pellets in increasing quantities (1, 2, or 3 at a time) to find the point at which sufficient current gradient is created to cause the thermal quench MHD.
Background:	In 2008, first experiments were performed on DIII-D with shell pellets for fast shutdown studies. D = 2 mm, t = 0.4 mm polystyrene shell pellets were used to deliver high pressure (10 atm) Ar and boron powder to the core of DIII-D. The ablation rate of the soft, low-Z shell was found to be reasonably well-described by theory; however, the ablation rate of hard, low-Z shell is predicted to be significantly slower and needs to be tested. Rapid delivery of the core payload was observed, although it was not established if a dust core gives faster delivery than a solid payload core. The high-Z (Ar gas) payload was too low density to cause an inverse thermal collapse, so higher density high-Z payload is desirable (we propose to use XeF2 for this, since this can be delivered at solid density but then decomposes to gas and can be pumped away). Also, no normal thermal quench was caused by the low-Z shell in these experiments, so it would be desirable to inject more/larger low-Z shells to see at what point a low-Z shell induces the standard disruptive thermal quench.
Resource Requirements:	One run day with impurity pellet injector.
Diagnostic Requirements:	Fast framing camera.
Analysis Requirements:	Ablation rate modeling.
Other Requirements:

Title	287: Current-driven RWM mode structure time evolution and IPEC analysis
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	Yongkyoon In, A. Boozer, J. Park, H. Reimerdes	ITPA Joint Experiment :	No
Description:	Current-driven RWM in the FY2008 has provided reproducible RWMs and been useful to assess the RWM physics as well as the feedback performance. The RWM was formed as a non-rotating mode and gradually shifted to a slowly-rotating mode. With weak feedback gain, the mode started rotating at the earlier time, which is closer to the mode onset. This behavior suggests that uncorrected error field plays a role for the initial RWM formation and the initial feedback synchronization process. Once the mode started rotating, the mode inversion radius seems to coincide with the q=2 radius. At higher gain, the mode was suppressed within a few gauss before the mode rotating. This leads to an issue of the feedback initial phase synchronization and its relation to the uncorrected error field, and secondly, the mode structure , which may been converted into a tearing mode with a magnetic island. Here, low gain scan is proposed and to compare with I-PEC to investigate the mode structure of current-driven RWM	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) prepare a shot like 133020 and 133021 (2) carry out feedback with low gain, w/wo derivative term (3) several pulsing technique is used to observe the plasma response - shut-off feedback and apply a square pulse. Then we observe the field penetration - turn-on the feedback after the rotaing mode onset and observe the (sheet current) magnetic island buildup on q=2 with ECE. Also we look for the immediate response other resonance - cramp the Ip-ramp to hold q_95 longer after the mode is fully developed? - growth rate scan with varying the current ramp rate? - relation to other error field experiments, like low density? - add balanced NBI for continuous measurement of the rotation - llok for the toroidal phase shift due to some viscosity to be taken into account
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	288: Measurements of hydrogenic retention and isotope exchange using H2 vs D2 fueling.
Name:	Hollmann ehollmann@ucsd.edu	Affiliation:	UCSD
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	Dmitry Rudakov and Neil Brooks	ITPA Joint Experiment :	No
Description:	Investigate spatial distribution of hydrogen deposition across divertor using H/D ratios from divertor spectrometer.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The experiment is designed to take place in two half day periods. Ideal times to do the experiments would be after a bake when the walls are relatively clean, as this would give more easily interpreted data on hydrogen retention (vs only isotope exchange rates). To begin with, a stable (non-disrupting) LSN discharge is developed with H2 as a working gas. Only low-density attached plasmas are used in this experiment (to avoid the complications of detached/recombining divertor legs). Also, L-mode is used to avoid dealing with ELMs for now. Several shots are run with X-point and strike points are fixed relatively far toward large radius (outer wall) to create a fixed H deposition profile in the walls. Heating, if any, is provided by gyrotrons, to avoid NBI fueling with D. The discharge is repeated several times to provide a good wall loading of H. After this, D2 is used as the working gas. In subsequent discharges (about 3), the X-point (and strike points) are moved slowly toward the inner wall (in discrete 1 second, several cm steps). H/D ratios are monitored during this time with the divertor spectrometer (set on Hgamma/Dgamma + CH/CD or Halpha/Dalpha; if possible, a second spectrometer should be hooked up to monitor H2/D2). From the release rate of H at different locations, the initial deposition rate profile can be inferred. Finally, 2 shots are performed with USN to use look for possible H stored in the main wall/upper divertor region. Additionally, DiMES can be used to expose a clean graphite sample during this experiment, for subsequent analysis of net H/D ratio in the lower divertor. Also, the RGA can be used to monitor H/D in the pump-out gas. In the second 1/2 day experiment, the reverse process is run, i.e. we begin with a D plasma and then use an H plasma to diagnose deuterium retention in the walls.
Background:	Understanding the retention of hydrogen and its isotopes in graphite and other wall materials is a high priority for tokamaks which use tritium. Present understanding is that hot strike point regions act as areas of net wall erosion and hydrogen release, while colder nearby graphite wall regions act as net absorbers of sputtered carbon and hot hydrogen neutrals. The goal of this experiment is to measure the net deposition rate of hydrogen across the divertor region for a well-characterized L-mode target discharge.
Resource Requirements:	Two 1/2 day experiments with both H2 and D2 available as fueling gases. No conversion of NBI to H is necessary. Several gyrotrons would be desirable but not required.
Diagnostic Requirements:	MDS, ASDEX gauges, RGA gauge, DiMES, filterscopes.
Analysis Requirements:	Spectroscopy analysis, DiMES TDS analysis.
Other Requirements:

Title	289: Effect of hot ions on RWM stability
Name:	Reimerdes reimerdes@fusion.gat.com	Affiliation:	CRPP-EPFL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	J.W Berkery, S.A. Sabbagh	ITPA Joint Experiment :	No
Description:	This experiment should test whether trapped hot ions in neutral beam injection (NBI) heated DIII-D plasmas contribute significantly to the stabilization of the RWM. We will modify the fast ion content in wall-stabilized plasmas and look for changes in RWM stability either by observing unstable modes or changes in the damping rate using active MHD spectroscopy. We primarily seek to reduce the stabilizing contribution. The ratio of the energies stored in hot ions and the thermal plasma, W_hot/W_th=tau_h/tau_E,th , can be reduced by decreasing the hot ion slowing down time Ï?h. The slowing down time of beam ions is inversely proportional to the density n_e. It depends on the hot ion energy E and the electron temperature T_e, whether the ions slow down on ions or electrons. For E>E_crit with, E_crit=18.6T_e (for fast deuterium ions in a deuterium plasma) the hot ions slow mainly down on electrons. The hot ions maintain their direction and, for tangential injection, remain dominantly untrapped. Below this energy ion collisions are dominant scattering ions into trapped orbits.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We plan to decrease the population of trapped hot ions by increasing the density in wallâ??stabilized plasmas. The density can be increases with gas puffing, pellets and/or increasing the plasma current I_P. Co-and counter-NBI is used to generate moderate co-rotation (about ΩÏ?A~1% at q=2), where the RWM is predicted to be the least stable and the spontaneous onset of the 2/1 NTM is less of a problem. We will start with tangential beams to mainly create hot ions in passing orbits, which do not contribute to the kinetic stabilization. The toroidal field B_T and the plasma current I_P can be reduced to reach higher values of βN (and modify the hot ion profile). Low field and, hence, low electron temperatures also favors collisions on electron, which could lead to an additional reduction of hot trapped particles, since less passing ions are scattered into trapped orbits. Further parameters that can be changed to modify the RWM stability are rotation and βN. In addition to looking for unstable RWMs a change in RWM stabilization can be evaluated by active MHD spectroscopy.
Background:	Kinetic calculations using the MISK code indicate a significant contribution of the hot NBI ions to RWM stabilization in DIII-D [J.W. Berkery, et al., APS 2008]. The contribution of hot ions is expected to increase with aspect ratio and could potentially explain the robust passive RWM stability in DIII-D compared to NSTX. In addition DIII-D and JT-60U observations indicate that q=2 fishbones/energetic particle-driven wall modes can trigger RWMs, which suggests a link between to the hot ion population and the RWM stability.
Resource Requirements:	5 NBI sources including both counter-sources. I-coils with AC power supplies (AA or SPA).
Diagnostic Requirements:	RWM sensors, MSE, CER, beam ion loss detector, FIDA.
Analysis Requirements:
Other Requirements:

Title	290: Comparison of RWM stability in DIII-D and NSTX
Name:	Reimerdes reimerdes@fusion.gat.com	Affiliation:	CRPP-EPFL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	J.W Berkery, S.A. Sabbagh	ITPA Joint Experiment :	Yes
Description:	This experiment seeks to identify the communalities of the RWM instability in DIII-D and NSTX in order to develop a common physics basis. It does so in comparing the RWM stability threshold and its dependence on plasma parameters in both devices with theory. It builds on the success of the experiment on the â??Effect of hot ions on RWM stabilityâ?? in DIII-D as well as a better understanding of the MHD trigger of RWMs in DIII-D and NSTX and is proposed for the 2010 experimental campaign. A corresponding experiment has been proposed on NSTX.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This experiment should identify under what conditions the RWM is intrinsically unstable or can be triggered by other MHD. Once an unstable condition is identified parameters that should influence RWM stability (e.g. beta, rotation, collisionality, fast ion beta) are modified to map out a stability boundary. Parametric dependencies of the stability boundary are then compared between experiments and with theory.
Background:	While NSTX observes spontaneous RWM growth in high beta discharges consistent with a linearly unstable RWM, the RWM in DIII-D seems to be inherently more stable over a wide range of plasma parameters and discharge scenarios. A direct comparison of RWM stability has, therefore, only been possible by comparing the damping rate of the stable mode. While this is possible using active probing techniques [H. Reimerdes, et al., Nucl. Fusion 45 (2005) 368, A.C. Sontag, et al., Nucl. Fusion 47 (2007) 1005], it requires accurate modeling of the geometry and comes along with uncertainties in the plasma response model. The observation of a threshold and its dependence on plasma parameters would offer a more direct comparison and test of theory.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	291: Prep-101: current-driven RWM tool refinements of C- and I-coil dynamic error field correction
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	Yongkyoon In, H. Reimerdes	ITPA Joint Experiment :	No
Description:	Current-driven RWM in FY2008 has provided reproducible RWMs and been useful to assess the RWM physics as well as the feedback performance. However, the feedback system has not been refined yet for detailed feedback analysis. There are several issues exist related to the sensor signal uncertainties and compensation against the unknown error field. Compensation against the other coil system is at present due to direct coupling to poloidal coils. There may exist the stray fields ( we know the existence of unidentified error fields) to the sensor due to the two-way coupling like coil->wall-> sensor, unidentified current path through metallic materials. It may be true that this field does not contribute to the RWM mode structure. This field could be non-resonant component. The component of coil current sent out by this component may not matter to the plasma and be neglected by the mode. However, if these are non-resonant, the request may end up to large current, and the current could be interpreted as â??error field identified by DEFCâ??. Separation of the sensor pickup offset from the actual unknown error field can be made after several trial/errors.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	(1) Setup a shot like 133011: fix the feedback parameters like Gp=80 with Gd of FY2008 (2) It is wise to add an option in PCS to offset-wave forms against the raw sensor Bp signals. With present arrangement through coil current-offset, there will be uncertainties whether the needed change is due to the sensor offset or actual compensation fields. When we switch the I-coil to C-coil DEFC, this will add uncertainties - if not modified, adjust the offset parameter to I-coil (N1TIU30OF etc), try to reduce both of the total Bp and I-coil total current. - if not modified, adjust the offset parameter to I-coil (N1TC79OFF etc), try to reduce both of the total Bp and total I-coil. (3) Dynamic error field correction with C-coil at least up to 650ms.
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	292: Real-time stability measurement using active MHD spectroscopy
Name:	Reimerdes reimerdes@fusion.gat.com	Affiliation:	CRPP-EPFL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Use spatial and temporal Fourier analysis of (integrated) magnetic measurements to extract the plasma response to externally applied rotating n=1 fields.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	It is proposed to implement the magnetic signal analysis for active MHD spectroscopy in a real-time PCS algorithm. The required calculations have been outlined in [H. Reimerdes, et al., â??Study of RWM stabilization by plasma rotation using active MHD spectroscopyâ?? 49th Annual Meeting DPP/APS 2007 https://diii-d.gat.com/DIII-D/physics/steadystate/rwmpres/aps07_reimerdes_poster.pdf].
Background:	Early detection of imminent stability limits can prompt action to avoid the limit and return the discharge to a stable operating point. Low-frequency active RWM spectroscopy can be used to detect the ideal MHD no-wall stability limit as well as the damping rate of the stable RWM in the wall-stabilized regime.
Resource Requirements:	This proposal will require some PCS development and piggy-back time on high beta experiments. While the sensors signals are already available in the PCS, several channels will be required for the results of the real-time calculations.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	293: DIII-D and NSTX m/n=2/1 NTM Comparison Allows Testing Aspect Ratio Physics (Including Rotation)
Name:	La Haye rbrtlahaye@gmail.com	Affiliation:	Retired from GA
Research Area:	Stability	Presentation time:	Requested
Co-Author(s):	R. J. Buttery, S. Gerhardt	ITPA Joint Experiment :	Yes
Description:	Excite m/n=2/1 modes, allow them to "saturate", then ramp down beta to self-stabilization, the marginal point while staying in ELMing H-mode. Data at onset, saturation, and at marginality allows comparison to experiments to be done in NSTX at lower aspect ratio. Rotation at onset and saturation will be varied.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce #133574 and do a two point scan of all co-NBI and balanced NBI to excite and saturate m/n=2/1 modes. Use the dud detector to start a controlled NBI feedback ramp down of beta to reach the marginal point at which self-stabilization occurs. Key is to stay in ELMing H-mode which is easier at this high q95. Might have to switch to all co-NBI in ramp down to avoid mode locking: best n=1 EFC throughout.
Background:	The working model for the marginal island width of NTMs is twice the ion banana width. Preliminary results (one good shot each) on DIII-D and NSTX confirm this for m/n=2/1 modes at q95~7 matched and show the expected sqrt of epsilon dependence. Need more data for reproducibility and interested in the effect of rotation on the tearing stability. This is included in the ITPA MHD proposals for 2009 under both MDC-4, "Neoclassical tearing mode physics - aspect ratio comparison", and MDC-14, "Rotation effects on neoclassical tearing modes".
Resource Requirements:	1/2 DAY in q95~7 LSND as in #133574, 7 beams. Follow up to D3DMP 2008-54-01 in which we also got data (3 good shots) at q95~4 to match JET marginal conditions.
Diagnostic Requirements:	Standard
Analysis Requirements:	Standard
Other Requirements:	None

Title	294: Identification, Extraction, and Reallocation of DEFC in RWM stabilization at q95~4
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	L. Marrelli, M. Okabayashi, H. Reimerdes, E. Strait	ITPA Joint Experiment :	No
Description:	The proposal is 1) to identify and extract the DEFC portion of I-coils and then 2) to reallocate it to C-coils based on reproducible current-driven RWM at q95~4	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish a reproducible current-driven RWM with q95~4 without feedback (e.g. 133021). StepA. To identify the DEFC portion of I-coils, the following methodology will be used 1) taup scans > tauW (conventional) 2) taup scans < tauW: high Gp scans w/ Gd (newly recognized) Step B. Once the DEFC portion of I-coil currents from 1) and 2) are similar, it will be reallocated to C-coils. Just in case 1) and 2) show a big discrepancy, it implies that the role of Gd is not limited to remove the pole related to the L/R time of the I-coil currents. Thus, the Gd scans with fixed high Gp will be done, which may converge to the results of 1) in Step.A
Background:	Recent analysis of the current-driven RWM feedback stabilization showed that a significant portion of I-coils was used to suppress the mode at q95~4. Considering that such near-static portion of the coil currents can be supplied by the C-coils and that the I-coil currents could be more effectively used for direct RWM feedback control, it is desirable to identify, extract and reallocate the portion of DEFC of I-coil currents not only for the feedback stabilization of RWM at q95~4 but also for more challenging RWMs at lower integer q95. Once the methodology is confirmed to be effective, it can be also used for RWM-prone high beta operations.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	295: ITER startup studies in DIII-D
Name:	Jackson jackson@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Requested
Co-Author(s):	T.C. Casper, T. Luce, D. Humphreys, T. Petrie, A. Hyatt	ITPA Joint Experiment :	No
Description:	Continue ITER large-bore startup work, particularly the high priority ITER items such as EC assist with 20 deg launch and 3 V Ohmic startup	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Further define and optimize ITER startup scenarios, building upon the results from 2007 and 2008. o Demonstrate a VLoop=3V ohmic startup o Investigate EC assist at 20 deg. off radial (ITER high priority) o Quantify V-sec reduction with auxiliary heated rampup o Better comparison of HFS vs LFS startup (requires discharge development) o Better simulate the very early phase with better shape control o further define the n=0 stability "window" for ITER startup o li control with NB feedback
Background:	ITER startup experiments in 2007 and 2008 helped define and improve ITER startup scenarios. This work led to a Nucl. Fusion paper, and presentations at APS (2007,2008) and IAEA. However additional work is required to further define and understand this critical phase. The two highest priority tasks are to demonstrate ohmic startup at 3V (2008 work was not successful below 4.5 V) and demonstrate EC assist with a launch angle of 20 deg. off perpendicular (the minimum allowed in the ITER design). Other tasks, listed in the experimental approach will better define the ITER operating space
Resource Requirements:	ECH (for ITER large-bore startup) and NB rampup V-sec characterization A real time indication of the vertical stability is highly desireable
Diagnostic Requirements:	Magnetics, ECE, TS, mse, filterscopes, interferometers, visible fast camera viewing the bumper limiter, bolometer, CER
Analysis Requirements:	Corsica, EFIT, vertical stability analysis, TOKSYS simulation to develop an acceptable power system configuration. JFIT (for n=0 stability)
Other Requirements:	--

Title	296: Feedback stabilization and mode helicity of current-driven RWM at q95~3
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	M. Okabayashi, H. Reimerdes, E. Strait	ITPA Joint Experiment :	No
Description:	The proposal is 1) to feedback stabilize the current-driven RWM at q95 ~ 3,and 2) to elucidate whether the mode helicity optimized for the effective feedback stabilization at q95~4 needs to be changed for q95~3.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Develop a reproducible current-driven RWM with q95~3. Starting from the coil configurations optimized for RWM feedback stabilization at q95~4 ('240 quartet'), the phasing between upper and lower I-coils will be changed to other options (180 and 360 quartets) to see if the helicity change is needed for q95~3. Once an optimized phasing is identified for the mode helicity at q95~3 with minimal plasma disturbances, a systematic gain scans and complex gains scans will be performed. NOTES: 1. AAs will be used, as long as no coil current saturation occurs. Otherwise, the SPAs will be attempted. 2. Good error filed correction (EFC) prior to q95~3 needs to be established primarily by C-coils, while the I-coil currents should be mostly used for direct RWM feedback.
Background:	Although successful feedback stabilization was achieved for current-driven RWM at q95 ~ 4, frequent I-coil saturations hampered us from assessing the RWM feedback stabilization at q95 ~ 3. Also, considering that a significant portion of DC-like I-coil currents can be supplied by C-coils, the portion of I-coil currents attributable to direct feedback of RWM at q95~3 is expected to be smaller without exceeding the affordable coil current limit (~1.4 kA). Meanwhile, recent study showed that a kink mode helicity is not aligned by the pitch of edge equilibrium field. Thus, from the systematic external field pitch changes, we may assess whether the kink mode helicity at q95~3 is the same as one at q95~4, which will clarify which coil configurations would be ideal to stabilize the kink mode and what conditions are necessary to stabilize the RWM at q95~3.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	297: Development of ITER rampdown Scenarios without VDEs or disruptions
Name:	Jackson jackson@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Requested
Co-Author(s):	T.Casper, D. Humphreys, T. Luce, A. Hyatt, T. Petrie	ITPA Joint Experiment :	No
Description:	Developing a "soft landing" scenario for ITER is a high priority and can be investigated experimentally on DIII-D	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This will be a multi-day experiment. The goal will be to explore the operating space where ITER can successfully execute a 'soft landing'. Both ohmic and auxiliary heated rampdowns will be investigated. A successful H-L transition (no locked modes or VDEs) will also be investigated Key Scans are: o kappa scan with H-mode a. Constant betaN b. li control o dIp/dt scan " " o ohmic rampdown at the end of an H-mode flattop phase o density scan and control (e.g. use n=3 coils)
Background:	The rampdown phase in ITER is of critical important since an uncontrolled VDE has the potential of compromising machine integrity and limiting ITER's lifetime. At the end of the burn phase, stored energy must be reduced, plasma current ramped down, and a transition to either L-mode or ohmic heating power must be accomplished while maintaining control of plasma shape for heat exhaust. Later in the rampdown the discharge will probably be limited, either HFS or LFS. Scenario development for the rampdown phase is currently being ramped up (pun) and DIII-D has the capability of simulating the high power phase and then investigating conditions for a successful rampdown. This work was begun in 2008 and a successful discharge with a soft landing was obtained after simulating the ITER flattop H-mode scenario 2. However only a single discharge was successful and further work is obviously required.
Resource Requirements:	Rampdown modeling with Corsica before the experiment. ECH (for ITER large-bore startup) and NB for ITER flattop scenario demonstration. 6 sec. discharges are required and possibly longer if simulations show a long rampdown time is optimal. A real time indication of the vertical stability is highly desireable
Diagnostic Requirements:	Since this is rampdown, extended time base (or variable acquisition rate is required to at least 6 s) Magnetics, ECE, TS, mse, filterscopes, interferometers, visible fast camera viewing the bumper limiter, bolometer, CER
Analysis Requirements:	Corsica, EFIT, vertical stability analysis, TOKSYS simulation to develop an acceptable power system configuration. JFIT (for n=0 stability)
Other Requirements:	--

Title	298: Particle Pinch In H-Mode Pedestal?
Name:	Callen jdcallen@wisc.edu	Affiliation:	U of Wisconsin
Research Area:	Pedestal Structure	Presentation time:	Requested
Co-Author(s):	R. Groebner, M. Austin, A. Leonard, G. McKee, T. Rhodes	ITPA Joint Experiment :	No
Description:	To explore the transport properties in H-mode pedestals it is proposed to follow the space-time evolution of the transient density and temperature profiles in the pedestal region in the first few ms after an L-H transition. The primary objective will be to determine the transient-transport-inferred ("true") particle diffusivity in an H-mode pedestal. Determining this diffusivity and comparing it to the quasi-equilibrium transport-analysis-inferred "effective" particle diffusivity in H-mode pedestals has emerged as a major issue in the H-mode edge Pedestal (HEP) Benchmarking Exercise (BE) (see Background below). In particular, there is considerable circumstantial evidence that a large particle pinch in the pedestal nearly cancels the outward particle diffusion to yield a very small effective diffusivity (~ 0.01 - 0.1 m^2/s) in H-mode pedestals -- e.g., how else does the density at the top of the pedestal get fueled? In addition, the transient-transport-inferred electron and ion heat effective diffusivities need to be compared with the rather small values inferred in the HEP BE in the middle of the pedestal transport barrier -- minimum ~ 0.3 m^2/s. In this experiment it will be most critical to obtain the space-time evolution profiles of the electron density (via reflectometry) and temperature (via ECE). It will also be very useful to simultaneously use the CER and BES systems to obtain the space-time evolution of the ion temperature, poloidal, toroidal flows, and fluctuation properties. The sum total of such comprehensive data should also provide important transport property information for the HEP BE and for the CPES and ESL edge plasma computational projects. While the transport re-equilibration just after a Type I ELM crash could also be considered, this has additional complications associated with the dramatic, extensive and largely un-quantifiable effects of the ELM on the edge pedestal region. Nonetheless, previous analyses [1-4] of the aftermath of an ELM crash have provided encouragement that the individual plasma parameters can be followed reasonably well with the desired spatial and temporal resolution. The broader objective of this proposed experiment is to collect all such data simultaneously for a single L-H transition (or perhaps after an ELM crash) to facilitate a complete transient transport analysis of pedestal transport properties and compare them to those in the quasi-transport equilibrium states being analyzed in the HEP BE. [1] Reflectometry: L. Zeng et al., J. Nucl. Mat. 337-339, 742-746 (2005); L. Zeng et al., Nucl. Fusion 46, S677 (2006). [2] ECE in edge: T.H. Osborne, "Fluid Modeling of ELMS," talk at Workshop in Boulder, CO, January 25, 2006. [3] Carbon impurities, flows via CER: M.R. Wade et al., Phys. Rev. Lett. 94, 225001 (2005); M.R. Wade et al., Phys. Plasmas 12, 056120 (2005). [4] BES: R.J. Colchin et al., Phys. Rev. Lett. 88, 255002 (2002).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The key will be to obtain a low power L-H transition into a low density H-mode, i.e., only slightly above the L-H power threshold so the transport evolution after the transition won't be too rapid. Also, this should allow the reflectometry to be able to easily follow many density levels and the electron temperature to be well above 200 eV throughout the pedestal so the ECE will be optically thick through most of the pedestal. Also, it will be helpful if the L-H transition were triggered in controlled way -- so the fast diagnostic windows could be assured to capture the approximately 10 ms immediately after the L-H transition.
Background:	Over the past year and a half Jim Callen and Rich Groebner have led an H-mode Edge Pedestal (HEP) Benchmarking Exercise (BE). The objective of this activity is "To identify, clarify and quantify the key physical processes, transport and structure involved in DIII-D H-mode pedestals through benchmarking comparisons between various types of transport modeling codes." To date it has focused on analyzing the pedestal transport properties (in particular effective diffusivities) of two DIII-D H-mode pedestal time slices in transport quasi-equilibrium just before an ELM -- 98889@4530 and 118897@2140. Consistency is being sought between transport modeling of these two pedestals with 6 different modeling codes: the 1D interpretive codes ONETWO (Grobner, Osborne), Stacey and TRANSP (Budny); 2D codes in interpretive mode SOLPS (Owen, Canik) and UEDGE (Rognlien); and 1D predictive ASTRA (Rafiq, Pankin). To date agreement has been achieved in the pedestal transport barrier region to within 10s of percent for the electron heat diffusivity (minimum ~ 0.3 m^2/s) and about 50% for the ion heat diffusivity. The next step for the HEP BE is to begin exploring the density transport and neutral fueling with these codes. However, a major complication there is that there is considerable evidence that there is a strong particle pinch (e.g., how else does the density at the top of the pedestal build up?), particularly in the lower half of the pedestal; this can cause the effective particle diffusivity (minimum ~ 0.01 to 0.1 m^2/s?) to be much less than the true particle diffusivity. The only known way to identify the true particle diffusivity (and hence the possible pinch magnitude) is through a transient transport analysis. Thus, we propose to explore the transient transport evolution of an H-mode pedestal just after an L-H transition (or perhaps, with more complications, just after a Type I ELM crash), mainly to determine the true particle diffusivity in an H-mode pedestal. Also, the electron and ion heat transient transport diffusivities will be analyzed to see if they are consistent with their quasi-equilibrium values in H-mode pedestals.
Resource Requirements:	Fast analysis (< 1 ms) of transient pedestal plasma properties from just before an L-H transition to about 10 ms after it: reflectometry for n_e and poloidal flow; ECE for T_e (optically thick only for top half of pedestal and inward), CER for n_i and T_i of Carbon 6+ and toroidal flow, and BES for fluctuations and poloidal ion flow.
Diagnostic Requirements:	See Resource Requirements
Analysis Requirements:	Since most modeling codes are not presently able to model the fast, localized transport pedestal responses after an L-H transient, approximate analytical methods will be mostly used. For example, a relatively simple direct way of estimating say a density diffusivity is through the local relation obtained from a density diffusion equation: D ~ dn/dt / (d^2 n / dx^2) -- see Eq. (23) and Figs. 6 and 7 of M. Soler, J.D. Callen, Nuclear Fusion 19, 703 (1979). Other analytic-based transient transport analysis techniques will also probably need to be developed to determine pedestal transport properties.
Other Requirements:	--

Title	299: Fast-ion effects on intrinsic rotation (Combined RF Exp. on Rotation, Core TAEs & sawtooth stabiliz
Name:	Nave mfn@ipfn.ist.utl.pt	Affiliation:	Instituto Superior Tecnico, Lisboa, Portugal
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	M.F.F. Nave (IST, Portugal), G. Kramer (PPPL, USA), A. Turnbull (GA, USA), T. Johnson (VR, Sweeden) , I. Chapman (UKAEA, UK), J. Graves (CRPP, Stwizerland)	ITPA Joint Experiment :	No
Description:	Aims: (a) Measurement of intrinsic rotation in ICRF heated plasmas during a sawtooth cycle (scan from usual sawtooth to monster sawtooth). (b) Demonstrate fast-ion effects on rotation - by comparing rotation profiles in pulses just below and just above core TAE threshold - by comparing rotation profiles during monster sawtooth stabilization phases in pulses where core-TAEs are present with pulses where core TAEs are suppressed. (c) And in a related experiment proposed by G. Kramer (n. 118) Control Core TAE and sawtooth stability, in particular - Study relationship of sawtooth stability and core tae modes. - Observe sawtooth behavior below and above core tae threshold. - Develop a method for controlling core TAE modes - Optimize sawtooth period by controlling core TAE modes	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Configuration as in pulse 96043. Starting with NBI only, scan co/counter NBI to obtain a non-rotating plasma. Then apply various levels of ICRF to document the sawtooth period, Core TAE activity and rotation profiles as a function of ICRF power. When the giant sawtooth regime is established with core TAEs we want to apply ECCD at an off axis location to stop the current diffusion to the core and keep q on axis to a value between 0.9 and 1.0 to avoid the onset of low-n core TAEs.
Background:	In JET ICRF heated discharges both co-current and counter-current rotation is observed. The question of whether counter-rotation observed in the plasma core could be attributed to differences in sawtooth activity and other core MHD modes remain open. In JET the ICRF induced rotation is difficult to measure since the charge exchange diagnostic requires co-NBI which adds torque to the plasma. DIII-D has the unique capability to use a combination of co and counter injected beams to deposit zero torque into the plasma while rotation measurements are possible. We would like to measure rotation during sawtooth cycles, using a torque-free target plasma, where ICRF power is added to investigate changes in the plasma rotation that can be correlated with ICRF-generated fast ions. We like to focus further on two possible events that can affect the transport of fast ions significantly: the onset of core-localized TAEs and the suppression of TAEs with ECCD. At the ICRF power threshold when the TAEs are excited we expect to observe a change in the core rotation due to TAE-induced enhanced transport of fast ions.
Resource Requirements:	co/counter NBI, ICRH, ECRH
Diagnostic Requirements:	Charge exchange recombination spectroscopy, FIR scattering, Beam Ion Loss Detectors, Fast Ion D-alpha (spectroscopy and imaging), SXR, ECE, magnetics, Fast CO2 interferometry, MSE
Analysis Requirements:	--
Other Requirements:	--

Title	300: Investigate scaling of residual stress / effective intrinsic torque with edge pressure gradient
Name:	Solomon solomon@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	P.H. Diamond, T.S Hahm, K.H. Burrell, J.S. deGrassie, S. Mueller	ITPA Joint Experiment :	No
Description:	The goal of the experiment is to validate whether the effective torque associated with the intrinsic rotation follows the basic theoretical scalings of the residual stress. This would be a key step in positively identifying the residual stress as the generation mechanism for intrinsic rotation.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We will measure the effective intrinsic torque as demonstrated in 2006 by finding the amount of counter NBI required to zero the rotation profile. To augment the data set, we will also try scanning the rotation around zero using the beams to correct for imperfect cancellation of the torque in steady state. This will allow us to look for how the edge velocity gradient scales with torque, which should show a linear dependence with the offset being related to the contributions from the residual stress and pinch, and the gradient related to the effective momentum diffusivity. In addition, it will also provide additional momentum transport information from the transient response of the rotation. The Rice scaling goes as W/Ip, so we effectively want to do either a power or betaN scan with these rotation sweeps. ECH can be used as an additional way of changing W, which will likely modify local gradients and the residual stress differently. If the residual stress is indeed the source of intrinsic rotation, then we should see a correlation between the effective intrinsic torque and edge pressure gradient, which we can compare with theoretical expectations. If possible, the turbulent driven Reynolds stress at the edge should be monitored with probes as a secondary means of confirming the residual stress as the key player in intrinsic rotation generation. The Doppler Backscattering system would be useful for an alternate characterization of the ExB shear.
Background:	Theory suggests that the edge pressure gradient is a driver for residual stress that is a likely candidate to explain the intrinsic rotation observed in tokamaks. Experimental evidence for a residual stress like effective torque associated with the intrinsic rotation has been directly observed on DIII-D under a single condition, but we need a bigger dataset to check the expected trends.
Resource Requirements:	1 day experiment, co/ctr beams, ECH
Diagnostic Requirements:	CER, Thomson, edge turbulence measurements, RS probe, DBS...
Analysis Requirements:	TRANSP, probe analysis
Other Requirements:	--

Title	301: Fully integrated ITER operational scenarios
Name:	Doyle doyle@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	A clear "next step" in the development of ITER operational scenarios is to move towards demonstration of "fully integrated" discharges, simulating ITER startup, flattop and ramp-down phases. This can be thought of as a combination and integration of the 2008 work by the ITER startup and ITER Demonstration groups.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We will need detailed discussions as to what level of detail we should pursue in fully integrated demonstrations, e.g. exact ITER shape versus ITER-like, pursuing TAE dominated versus fishbone dominated hybrids, etc.
Background:	In 2008 the ITER startup and ITER demonstration discharges separately dealt with the ITER startup and flattop phases. For the 2010 IAEA, it seems clear that DIII-D should target having more more fully integrated demonstration discharges covering all ITER plasma phases from initiation to ramp-down
Resource Requirements:	7 NBI
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	302: Relationship between momentum and particle pinch and dependence on collisionality
Name:	Solomon solomon@fusion.gat.com	Affiliation:	GA
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	S.M Kaye, T. Tala, G. Tardini, K.H. Burrell, J.S. deGrassie, C.C. Petty	ITPA Joint Experiment :	No
Description:	The goal of the experiment is look at the dependence of the momentum pinch velocity as a function of collisionality, and compare with analytic theory and/or gyrokinetic predictions. In addition, we will attempt to compare the momentum pinch velocity with the particle pinch velocity (which should be related), using separate density perturbation techniques.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We will continue the successful use of rotation perturbations using beam modulation and/or blips to decouple the diffusive and pinch contributions to the momentum flux. The modulation should be fast so as to get a dominant jxB torque which has better spatial localization and has previously been characterized (deGrassie et al). This also minimizes Ti modulations and plasma position oscillations. In addition, we may also use non-resonant magnetic field perturbations to brake the plasma as on NSTX. Collisionality will be varied while trying to keep rho*, Ti/Te, Mach number and q fixed. The pinch velocity may also have some dependence on R/Lti, Ti/Te etc, so ultimately any changes there will have to be factored in and correctly reproduced in theory or gyrokinetic simulations. In otherwise matching discharges, we will use pellets to perturb the edge density and measure the corresponding inward particle pinch from the transient response of the density profile. The relationship between the momentum and particle pinches will be compared with theoretical expectations.
Background:	Experiments have been conducted in FY08 to separate the role of diffusion and convection in momentum transport, using beam blips and modulation. These studies have identified the existence of an inward momentum pinch on DIII-D. However, little information is available regarding the scaling of the pinch compared with theory, including the role of collisionality and density scale length. In addition, there has not been any direct investigation experimentally between the momentum and particle pinch. This ITPA experiment will receive run time on NSTX, and likely also on JET.
Resource Requirements:	1 day experiment, pellet injection, ECH, co/ctr beams
Diagnostic Requirements:	CER, fluctuation diagnostics useful...
Analysis Requirements:	TRANSP + modulation analysis
Other Requirements:	--

Title	303: Measurement of enhanced jxb torque driven by enhanced fast ion transport
Name:	Solomon solomon@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	M.A. VanZeeland, W.W. Heidbrink, R. Nazikian, K.H Burrell, J.S deGrassie	ITPA Joint Experiment :	No
Description:	The objective of this experiment is to answer whether additional torque is exerted on the plasma due to non-ambipolar fast ion losses.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The plan is to measure the rotation profile evolution using fast CER during a bursting mode such as the n=0 E-GAM. We will use constraints on the fast ion transport provided from fast ion diagnostics such as FIDA etc to see whether the observed rotation is consistent with such transport, or whether an additional jxB torque associated with an inward return current of thermal ions. If the latter, then we can check if this matches modifications in the (ne, nc, n_FI) density profiles. If no additional torque is needed, then there must be additional electron transport, which can be assessed through transport modeling and checked empirically using density fluctuation diagnostics. In that case, we should inject additional NBI torque transiently to look at the jxB torque response on the rotation; this may be reduced compared to cases without the MHD. Scans of Ip/Bt at fixed q would be useful for characterizing the scaling of the torque with B. Measurements of the fast time scale evolution of the radial electric field from the new DBS diagnostic would be an excellent supplement to this measurement.
Background:	Various Alfven Eigenmodes and other MHD have been shown to redistribute fast ions. This redistribution can affect the torque delivered by the neutral beams in two ways. Firstly, when the fast ions are transported, the location where they collisionally slow down on the thermals and deliver torque is obviously changed. However, a more subtle effect is realized in the bookkeeping of charge neutrality, where the outward transport of the fast ions may be balanced by an inward return current of thermal ions. These inward moving thermal ions experience a jxB torque in the counter-Ip direction. However, if charge neutrality is instead maintained through increased electron transport, then this extra counter-torque will not be realized.
Resource Requirements:	1/2-1 day experiment
Diagnostic Requirements:	fast CER, FIDA and fast ion diagnostics, fluctuation measurements (BES, reflectometer, DBS...)
Analysis Requirements:	TRANSP, FIDA
Other Requirements:	--

Title	304: Hybrid operation with electron heating and Te~Ti
Name:	Doyle doyle@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	ITER will operate with dominant electron heating and Te~Ti. In validating hybrid scenarios for use on ITER it is essential that we understand how hybrid confinement and fusion performance are affected by strong electron heating and operation with Te~Ti. This proposal is to extend and improve on work performed in 2007 which increased the Te/Ti ratio in hybrid discharges to 0.8, using ECH. The availability of increased gyrotron power offers the prospect of operation with Te/Ti~1, with strong electron heating.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Utilize increased gyrotron power to take DIII-D hybrid plasmas to Te/Ti ~1 with strong electron heating. A key and unique feature of the 2007 experiment which will be repeated here is to utilize the PCS feedback control capabilities to maintain constant rotation and density with/without ECH (using co-/counter-NBI capability to match rotation).
Background:	An experiment in 2007 successfully raised Te/Ti to 0.8, with moderate impact on confinement. However, it is important to raise Te/Ti beyond this level, to ~1. Increaded gyrotron power makes this feasible. In addition, these experiments were significant and unique in that they were the first to isolate the Te/Ti ratio effect, by matching density and rotation with/without ECH.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	305: Active MHD Spectroscopy for Unstable RWM Using Feedback with Complex Gains
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	Yueqiang Liu, M.S. Chu, M. Okabayashi and the DIII-D Team	ITPA Joint Experiment :	No
Description:	This proposal is to benchmark plasma response model (PRM) theory using active MHD spectroscopy for unstable RWM. This will also address the issue whether RWM rotation can be induced by active coils in tokamak.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish a reproducible current-driven RWM with q95~4 without feedback (e.g. 133021). Determine a real (Gp) without phase shifts, which affects the mode growth without stabilizing the unstable mode completely. Once the unstable mode is identified, the feedback will be turned on right after the mode growth is detected. Then, the rest of the complex gain scans will be performed. Specifically, the phase shifts will be 0, +/- 45 (+/- 30 or 60) to monitor the mode rotation, as well as the mode growth.
Background:	Recent RWM feedback results with complex gains in RFX [1] confirm the theory of plasma response model (PRM). Also, Recent current driven RWM experiments in DIII-D [2] show re-producible RWM and feedback stabilisation. By comparing the current driven RWM experimental feedback results with complex gains, with simple plasma response models, many important RWM physics questions can be addressed. Here, we show the idea on a simplest example of single mode control with complex proportional feedback gain only. This model explains very well the RFX results [1]. Assuming a single unstable RWM with the open loop growth rate gamma0, the PRM P(s) is written as P(s) =c/(s-gamma0)/tauW (1) where s is the Laplace variable (can be regarded as the closed loop mode eigenvalue), c measures the coupling strength between the active coil and the mode, tauW is the wall time. ActiveMHD spectroscopy here basically means that we determine the open loop RWM growth rate gamma0 using feedback measurements as explained below, assuming that a direct measurement of the mode growth rate is difficult. In addition, we deduce the coupling parameter (residue) c from the feedback experiments, hence fully obtain the PRM. Knowing PRM helps us to optimise the feedback controller analytically. The mode eigenvalue s= gamma + j omega of the closed loop, with a complex feedback gain G = G0 exp( j D_Phi), is determined by the characteristic equation 1+GP(s) = 0 (2) This gives gamma tauW = gamma0 tauW - cG0 cos (D_Phi), (3) omega tauW = - cG0 sin (D_Phi) (4) In the feedback experiments, we vary D_Phi while keeping G0 unchanged, and try to measure gamma and omega. In RFX, the mode frequency, omega is determined by measuring the phase shift of the plasma response w.r.t. the coil current, as a function of time. The slope of the curve gives the mode frequency. This technique works when the mode is observed during a short time period compared with the wall time. If the mode is observed during a few wall times period, it is even simpler to measure the mode frequency. I assume that the mode frequency can be measured in DIII-D with a relatively good accuracy. Knowing the mode frequency omega as a function of D_Phi, we can deduce the coupling coefficient c using Eq. (4). The mode growth rate gamma is easy to measure in RFX, but probably difficult to measure in a tokamak device even for the current driven RWM. However, in order to obtain the value for gamma0, we do not need very accurate measurement of gamma, since gamma0 can be deduced by averaging gamma over a range of D_Phi, according to Eq. (3). Preliminary study based on the 2008 expeirmental data and one-pole model with real residue described by Eq (1) appeared inadequate but either complex residue or 2-pole model seems to be close to what was observed in experiments. References [1] T. Bolzonella, et al., Phys. Rev. Lett. 101, 165003 (2008). [2] Y. In, et al., APS-DPP (2008)
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	306: Sawtooth Crash Flow Generation, Damping, and Multiscale, Multifield Turbulence Response
Name:	Hillesheim jon.hillesheim@ukaea.uk	Affiliation:	CCFE
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	W.A. Peebles, T.L. Rhodes, L. Schmitz, T.A. Carter	ITPA Joint Experiment :	No
Description:	This experiment will seek to fully characterize modifications in the spatial structure of the radial electric field generated by the sawtooth crash with Doppler Backscattering (DBS) measurements and also to acquire multiscale, multifield measurements of turbulence associated with the crash via DBS, high-k backscattering, FIR scattering, CECE, etc. With these flow and turbulence measurements we aim to attain insight of fundamental importance to understanding the dynamics of the crash phase of the sawtooth cycle. The working hypothesis for the flow generation mechanism is that, in addition to the primary m=1, n=1 mode at the crash, either harmonics of the primary mode at 2/2 and 3/3 or other secondary instabilities interact to create a region of stochastic magnetic field lines which decouples the electrons and ions, resulting in local non-ambipolar transport which sets up a radial electric field which is then measured using Doppler Backscattering. The modification to local transport by the stochastic fields is a possible explanation for the rapid time scale of the crash reconnection, which is much faster than theoretical predictions from the conventional Kadomtsev model. The generation of large electric fields can also potentially explain the observed impurity transport at the crash. Successful confirmation and characterization of the stochastic crash hypothesis would represent a transformational change in the understanding of the physics underlying the sawtooth crash and provide important results for next-step devices such as ITER where sawtooth-induced impurity transport will affect Helium ash accumulation. Even if the experiment returns a null result on the stochastic field hypothesis, characterization of the crash driven flows and turbulence will provide insight into the physical mechanisms responsible for increased transport during the crash, the enhanced reconnection rate, and sawtooth driven impurity transport.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	It is planned to run a high Ip, diverted L-mode plasma so that the inversion radius for the sawteeth will be accessible with DBS measurements, but to avoid H-mode where the flat density profile makes DBS measurements problematic. Some promising initial measurements came from shot 133505, which could serve as a starting point for a new experiment. It is assumed that both the current multichannel frequency-modulated DBS (FM-DBS) system with narrow channel separation (350 MHz, â?²1 cm) and a new multichannel comb generator DBS (Comb-DBS) system with wider channel separation (2.5 GHz, â?²5 cm) will be available. The FM-DBS array will be adjusted to probe in the proximity of the inversion radius to measure the flow structure around the rational surface. The Comb-DBS system will monitor the flow response across a large portion of the plasma radius. The density profile will be measured during the course of the crash at high time resolution with the profile reflectometer, which is necessary for accurate interpretation of DBS during the crash phase. The toroidal magnetic field will be slowly ramped such that the q=1 surface passes through the FM-DBS array for different sawteeth over the course of the shot. If this approach proves unworkable, the FM-DBS frequencies will be tuned to hop to different positions around the inversion radius (this is less desirable since the fluctuation levels between hops would not be directly comparable). Other diagnostics will be adjusted to likewise focus on this region. Beam blips will be employed early in the shot after reaching a steady-state and at the end for CER and BES measurements to confirm that the plasma was steady-state during the main time period of data acquisition and that the toroidal field ramp does not have undesirable effects. One beam will be used throughout the shot for MSE measurements of the q profile. Beam blips or beam modulation are to be avoided during most of the shot to avoid contaminating DBS data with non-sawtooth transients. The main goal of the experiment will be to acquire sufficient data from a stationary (other than sawteeth) plasma for conditional averaging to compare to flux surface average theoretical predictions and test the stochastic crash hypothesis. This could require several repeated shots to build up sufficient data for statistical analysis. If that is accomplished, the same approach will be applied to different plasma conditions, such as by applying central ECH to produce larger temperature collapses from the crash.
Background:	During the 2008 run period the generation and damping of large (~10 km/s) v_(ExB) flows at sawtooth crashes were observed on the timescale of ~100 us with DBS measurements in Ohmic and L-mode plasmas. Increases in intermediate wavenumber density fluctuations were also observed. Theoretical models of the occurrence of stochastic magnetic fields at the sawtooth crash have been developed [A. Samain, Plasma Physics 18, 551 (1976); A.J. Lichtenberg, Nucl. Fusion 24, 1277 (1984); A.J. Lichtenberg, K. Itoh, S.-I. Itoh, and A. Fukuyama, Nucl. Fusion 32, 495 (1992); T.E. Stringer, Nucl. Fusion 32 (1992)] and several past and recent experiments are consistent with this hypothesis. Stringer lists a number of references to experiments showing impurity transport during sawteeth that could be explained by the generation of large electric fields. More recently non-ambipolar charge transport due to magnetic fluctuations has been measured in MST [W.X. Ding et al., Phys. Rev. Lett 99, 055004 (2007); W.X. Ding et al., Phys. Plasmas 15, 055901 (2008)], but the electric field and its damping were not able to be directly measured. Work on ASDEX [V. Igochine et al., Nucl. Fusion 47, 23 (2007); V. Igochine et al., Nucl. Fusion 48, 062001 (2008)] has investigated the stochastic crash hypothesis through modeling and through identification of a transition to chaos in SXR and ECE measurements. The flow generation from sawteeth we have seen in experiments on DIII-D in 2008 has not been previously directly observed. A number of experiments have looked at fluctuations during the crash phase [F. Gervais et al., Plasma Phys. Control. Fusion 39 43 (1997) and references therein], but not with the detail currently possible with the multiscale, multifield fluctuation diagnostics on DIII-D.
Resource Requirements:	Machine time: 1 day Neutral Beams sources: MSE beam. Beam blips for CER, BES.
Diagnostic Requirements:	Both the current FM-DBS and system and the under development Comb-DBS system. Profile reflectometer. All other core turbulence systems (CECE, FIR scattering, microwave backscattering, BES, PCI). ECE. SXR. MSE. CER.
Analysis Requirements:
Other Requirements:

Title	307: maintenance of high beta reverse shear discharges with ITBs
Name:	Doyle doyle@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Assess Steady-State Current Profiles for Optimum Performance	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	In 2005, reverse shear discharegs with beta-n of 4 and ITBs were maintained for 2 s, but with non-stationary conditions. This experiment aims to address the stationary of these plasmas by utilizing both the higher power ECCD available since 2005 and also the improved plasma shapes developed for steady-state high beta operation.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Utilize off-axis ECCD to maintain reversed shear current profile and large radius ITBs generated by Bt ramp. Perfrom experiment in optimized shape identified in 2008 steady-state experiments.
Background:	See results reported in A. Garofalo, et al, Phys. Plasmas 13, 056110 (2006).
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	308: Pedestal scaling in hydrogen plasmas
Name:	Snyder snyderpb@ornl.gov	Affiliation:	ORNL
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Requested
Co-Author(s):	R. Groebner, A. Leonard, T. Osborne	ITPA Joint Experiment :	No
Description:	Compare H-mode pedestal width and height between matched hydrogen and deuterium plasmas. Vary both the species mass and the density to create a very large difference in predictions between leading models of the width.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Following a pedestal density (ie collisionality) scan in deuterium plasmas, create matching discharges in hydrogen. Match global beta, and vary density over a wide range.
Background:	Leading models of the pedestal width suggest scaling of the width with sqrt(beta_pol_ped) (KBM based), or with poloidal gyroradius (ExB suppression of drift turbulence). While these parameters scale similarly in many experiments, they can be separated by varying the density and the species mass. The ratio of sqrt(beta_pol_ped)/rho_pol_ped scales like sqrt(neped/mass). Hence by varying the mass a factor of 2, and the density by ~8, we can achieve roughly a factor of 4 separation between these scalings, as well as provide a vigorous test of models of the width, as well as models of the height such as EPED1.
Resource Requirements:	reasonably pure hydrogen plasmas
Diagnostic Requirements:	full pedestal diagnostic set including reflectometry
Analysis Requirements:	EPED1 runs before expt if possible
Other Requirements:

Title	309: Physics of the Super X Divertor
Name:	Garofalo garofalo@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Requested
Co-Author(s):	P. Valanju, M. Kotschenreuther (Univ. of Texas), J. Canik (ORNL), C. Lasnier (LLNL), A. Leonard, P. West (GA)	ITPA Joint Experiment :	No
Description:	Use flexibility and diagnostics capability of DIII-D to test the physics of the Super X Divertor concept. This is an advanced divertor concept that could address many of the divertor issues identified for high performance burning plasma experiments.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Produce a lower single-null divertor (LSND) version of DIII-D discharge 134418. This FY08 discharge was a 0.9 MA, upper SND plasma, with small minor radius (55 cm), and magnetic axis located 23 cm above the midplane. The LSND version of 134418 would be a small plasma with a very long outer divertor leg (>60 cm). This would be our reference plasma to start a series of scans: - Scan the major radius of the outer divertor strike point, R_D: we should be able to vary R_D from 105 cm to 175 cm, remaining on the divertor floor. Look for changes in the divertor heat flux with varying connection length. Test predicted dependence of divertor peak heat flux on R_D. - Scan the vertical coordinate of the magnetic axis: from +23 cm to -23 cm (shot 132601). Look for changes in the divertor heat flux with varying connection length. - Repeat scan of magnetic axis height in strongly radiative regime. For each point of the scan, establish detached divertor operation and look for development of MARFE or deterioration of H-mode confinement. Can MARFE be avoided with sufficiently long connection length? Test effect of pumping from the private flux region.
Background:	Key idea of the Super X Divertor (SXD) concept is to spread the divertor heat flux over a larger area (than in a standard divertor) further from the plasma, by increasing the major radius of the divertor strike point. The larger major radius also strongly reduces the parallel heat flux and hence the plasma temperature at the plate. Other advantages of a very long connection length may include increased robustness of detached divertor operation against development of MARFE.
Resource Requirements:	Same resources as used for 134418 (off-axis NB-driven current experiment).
Diagnostic Requirements:	All pedestal and lower divertor diagnostics, including fast divertor IRTV, divertor floor Langmuir probes, thermocouples, and divertor Thomson. The capability to switch connections for the 4 power supplies to monitor different Langmuir probes during the run day is highly desired.
Analysis Requirements:	Some shape development will be required to produce a LSND version of DIII-D discharge 134418. Preparatory EFIT modeling work will reduce the burden on experimental run time. CORSICA analysis will be performed to determine the F-coil currents required to attain the various configurations in the R_D scan.
Other Requirements:	This is a 1 day experiment

Title	310: Fast Ion Instability Validation Plasmas
Name:	Van Zeeland vanzeeland@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	M.S. Chu, G.Y. Fu, N. Gorelenkov, W.W. Heidbrink, G. Kramer, Z. Lin, R. Nazikian, D. Spong, A. Turnbull, J.H. Yu	ITPA Joint Experiment :	No
Description:	The primary goal of this experiment is to obtain a set of well diagnosed plasmas with Alfvenic instabilities in a range of parameters, the data from which will be made publicly available to help fast ion driven instability modeling validation and development efforts.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin with discharge 132707 (circular plasma dedicated to SCIDAC initiative), carry out an elongation scan to obtain AE sensitivity to elongation. Vary injected neutral beam power, co/ctr direction and density to obtain AE dependence on fast ion population. Vary toroidal field to obtain sensitivity to VB/VA as well as eigenmode dependence on B. Vary plasma current to obtain sensitivity to q95. Document eigenmode structure and impact on fast ion population for comparison to modeling. This experiment may require reverse Ip operation to obtain 2D FIDA imaging with co-injected beam ions. This experiment, if accepted, will involve D3D experimentalists working closely with the developers of the various fast ion physics codes such as M3D, GYRO, HMGC, NOVA-K, GATO, TAEFL, AE3D, GTC, and LIGKA to design the most relevant possible discharge conditions so it is anticipated that some of the basic experimental approach may change.
Background:	The role fast ion driven instabilities will play in future devices such as ITER is still unknown. Present day experiments actually show that these modes can have a severe impact on the fast ion population. Motivated by these data, there is a large worldwide effort aimed at developing validated codes capable of making believable predictions for ITER and future devices. In support of this effort, the US has sponsored two SCIDAC computational initiatives in the area. The degree with which a code can be validated, however, depends strongly on the quality and variety of data available as well as the applicability of the underlying physics models used. It is the goal of this experiment to take advantage of DIII-D's shaping flexibility and advanced suite of diagnostics to obtain a set of discharges in which measurements will made of unstable eigenmode structure, fast ion profile, impact of eigenmodes on fast ion profile, as well as the basic profiles necessary for input to the models. The new diagnostics relevant to this experiment that greatly enhance the quality of data obtainable in 2009-2010 include: FIDA Imaging, Fast FIDA, ECE imaging, FILD detectors, NPAs, Linear BES array, Visible Bremsstrahlung Imaging (if intensifier is available). Experiments will be carried out in a variety of regimes and shapes including L-mode, H-mode, and circular as well as elongated plasmas. These experiments will build on results obtained in 2008 experiments in which one successful circular L-mode discharge (132707) was obtained with well documented Alfven Eigenmode behavior. Circular plasmas have the advantage that they are easily modeled both analytically and numerically. This particular plasma is currently serving as a SCIDAC fast ion physics test case. Additionally, typical of RSAE and TAE experiments such as this is a large reduction in central fast ion pressure relative to classical TRANSP predictions. While the AE activity is thought to be responsible, the underlying fast ion transport mechanism remains to be conclusively identified. The new fast ion loss detector (FILD) system has the capability to conclusively link RSAEs and TAEs to fast ion loss during these current ramp experiments and the FIDA imaging can document this in 2D.
Resource Requirements:	3 co, 2 ctr sources
Diagnostic Requirements:	FIDA Imaging, Fast FIDA, FILD detectors, NPAs, Linear BES array, Visible Bremsstrahlung Imaging (if intensifier is available).
Analysis Requirements:	--
Other Requirements:	--

Title	311: Pedestal studies in helium plasmas
Name:	Snyder snyderpb@ornl.gov	Affiliation:	ORNL
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Requested
Co-Author(s):	R. Groebner, A. Leonard, T. Osborne	ITPA Joint Experiment :	No
Description:	Explore scaling of pedestal width and height, as well as main ion rotation and temperature in helium plasmas	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Vary pedestal density in a series of pedestal expts matching the shape, global beta, Ip, Bt from an earlier set of density scans in deuterium plasmas.
Background:	Leading models of the pedestal width suggest scaling of the width with sqrt(beta_pol_ped) (KBM based), or with poloidal gyroradius (ExB suppression of drift turbulence). While these parameters scale similarly in many experiments, they can be separated by varying the density and the species mass. The ratio of sqrt(beta_pol_ped)/rho_pol_ped scales like sqrt(Z^2*neped/mass). Hence by varying the mass a factor of 2, Z a factor of 2, and the density by ~8, we can achieve roughly a factor of 4 separation between these scalings, as well as provide a vigorous test of models of the width, as well as models of the height such as EPED1. Additionally, in helium plasmas, the CER system can be used to measure main ion velocity and temperature. Detailed studies of main ion rotation will be conducted, and coupled to tests of neoclassical models, as well as models of the rotation impact on stability.
Resource Requirements:	reasonably pure helium plasmas
Diagnostic Requirements:	CER tuned for helium, full pedestal diagnostic set including reflectometry
Analysis Requirements:	EPED1 runs before expt if feasible
Other Requirements:

Title	312: Energetic-particle driven RWM: comparison DIII-D / JT60U with the emphasis of residual error field
Name:	Matsunaga none	Affiliation:	Japan Atomic Energy Agency
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	M. Okabayashi, H. Reimerdes, M. Takechi(JAEA)	ITPA Joint Experiment :	Yes
Description:	Investigate RWM stability with energetic particle driven instability that can induce RWM in the high-beta plasmas above no-wall beta limit. And clarify how the energetic particle mode can couple and trigger the RWM with different plasma rotations. Measure detailed mode structures of the energetic particle mode, in particular, in the radial direction by ECE and soft-X-ray. Finally, compare obtained data with those of JT-60U. JT60U high betan exploration has revealed that the RWM is excited as an energetic-particle mode â??energetic particle wall modeâ?? which provided various types of transient MHD modes immediately coupling to a RWM or forming a RWM precursor. Thus, the RTWM in JT60U seems to differ from fishbone energetic particle mode-driven RWM observed in the DIIID. The purpose of this experiment is to identify possible issues which can have contributed to the characteristics of RWMs. The nature of RWM with near-zero rotation is highly vulnerable to the uncorrected error field. We emphasize the energetic particle mode sensitivity to various types of error field simulated by I-coil or C-coil.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	To identify key parameters to destabilize the energetic particle mode in the high-beta region, parameter scans e.g., betaN, rotation, q-profile and so on will be planned. If possible, energy and direction of NB also will be scanned. For detailed measurement of mode structures by ECE, soft-X-ray and available diagnostics. Obtained data will be compared with those of JT-60U. The reference shot is 131129, where the fishbone was observed from 1000- 2000ms period. The precession frequency will be scanned by reducing the NBI energy and plasma current. The q_min difference between the DIIID and JT60U will be explored by adjusting the near-zero rotation timing. The needed run days are - ½ day target development - 1 day documentation
Background:	In the high-beta plasmas above the no-wall beta limit, n=1 fishbone-like instability were observed in JT-60U (Energetic particle driven wall mode : G. Matsunaga et al., IAEA08 EX5-2) and DIII-D (m/n=2/1 fishbone : M. Okabayashi et al., IAEA08 EX/P9-5). This mode has several kHz mode frequency and â?¤1ms growth time. This mode can directory induce the RWM despite enough plasma rotation for RWM stabilization. In JT-60U, this mode frequency corresponds to the precession frequency of the energetic particles from the perpendicular NBs. Moreover, it can be stabilized by minimizing the perp-NBs while keeping high-beta. However, mechanism and key parameter for interaction between it and RWM are still unclear. Therefore, in DIII-D similar discharges to JT-60U will be planed to compare between different machines.
Resource Requirements:	NB to exceed no-wall beta limit and to control plasma rotation .ECCD for suppresing NTM
Diagnostic Requirements:	ECE, CER, Mino/saddle coils, soft-X ray
Analysis Requirements:	REVIEW, NEWSPEC, etc
Other Requirements:

Title	313: Test ELM suppression by RMPs in counter-rotating plasmas
Name:	Nardon none	Affiliation:	UKAEA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	A. Kirk, F.L. Waelbroeck	ITPA Joint Experiment :	No
Description:	Try and obtain ELM suppression with the I-coils in plasmas rotating in the counter-Ip direction in order to test a theoretical idea concerning the mechanisms leading to ELM suppression.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Prepare a target plasma with as large as possible counter-Ip rotation, Type-I ELMs (avoid QH-mode), and q95 in the window for ELM suppression. Turn on the I-coils at maximal current. If ELM suppression cannot be achieved with counter-Ip rotation, do a scan in rotation, from counter to co, and identify when ELM suppression is retrieved.
Background:	The theory of screening of RMPs by plasma rotation predicts that RMPs penetration is possible only at locations where the electron perpendicular velocity (which includes an ExB component and a diamagnetic component) is small. In plasmas rotating in the co-Ip direction, this is the case in a small region near the top of the pedestal (everywhere else except at the very edge, the RMPs are expected to be strongly screened). This was presented by E. Nardon at the Workshop on the effect of plasma response on RMPs at GA in August 2008 and at the MHD control workshop in Austin in November 2008. Interestingly, this region could coincide with the region where the RMPs are observed to affect the pressure gradient, leading to ELM suppression. Plasmas rotating in the counter-Ip direction do not have that particular region. Observing ELM suppression in such plasmas would invalidate the idea that ELM suppression is related to this region. A negative attempt at ELM suppression would reinforce this idea, even though it would not prove it.
Resource Requirements:	I-coils at maximal current. Counter-Ip rotation (counter NBI required).
Diagnostic Requirements:	All pedestal and lower divertor diagnostics.
Analysis Requirements:	All pedestal and lower divertor diagnostics.
Other Requirements:

Title	314: Investigate the effect of a slow I-coils ramp down after a fast initial ramp up
Name:	Nardon none	Affiliation:	UKAEA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	Andrew Kirk	ITPA Joint Experiment :	No
Description:	In a typical plasma with the ELMs suppressed by the I-coils, apply a slow I-coils ramp down after the ELMs are suppressed. The purpose is, on the one hand, to see whether the density pump-out can be limited (and whether the ELMs return at the density at which they had disappeared), and on the other hand, to look for hysteresis effects.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Reload a discharge with the ELMs suppressed by the I-coils. Replace the end of I-coils flat top (after the ELMs are suppressed) by a slow I-coils ramp down. Slow I-coils periodic oscillations could also be considered.
Background:	In many DIII-D discharges where the ELMs are suppressed by the I-coils, the density continues to go down after the ELMs have disappeared. It is important for ITER to demonstrate the possibility to limit the density pump-out. We propose here to decrease the I-coils current, hoping to limit the pump-out. Besides, it is of great importance for the understanding of ELM suppression by RMPs to have a proof of the RMPs penetration (if it does happen). One indirect way is to look for a hysteresis, which is expected by the classical theory of RMPs penetration: the current required for penetration is smaller than the current at which the RMPs are expelled after having penetrated.
Resource Requirements:	I-coils at maximal current.
Diagnostic Requirements:	All pedestal and lower divertor diagnostics.
Analysis Requirements:	All pedestal and lower divertor diagnostics.
Other Requirements:	This experiment probably does not require more than a few shots.

Title	315: MHD induced fast ion transport
Name:	Van Zeeland vanzeeland@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	W.W. Heidbrink, J.H. Yu	ITPA Joint Experiment :	No
Description:	The primary goal of this experiment is to document the impact of core MHD on fast ion transport using new fluctuation and fast ion diagnostic capabilities. Specifically visible bremsstrahlung imaging and FIDA imaging will be used to document the 2D mode structure and impact on the fast ion profile respectively.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A sawtoothing discharge will be established with moderate density of approximately 4x10^13 cm-3. Density will be scanned to optimize FIDA imaging and Visible Bremsstrahlung mode structure measurements which have conflicting requirements (low and high density respectively is preferred for better signal to noise). Neutral beam power will also be varied to obtain sufficient FIDA signal. Once an optimum case is obtained a q95 scan will be carried out to vary the inversion radius and degree of impact on the fast ion profile. The same q95 scan will be repeated with the fast framing camera set to record broadband visible bremsstrahlung emission in one case and FIDA in the second. If time permits, neutral beam power will be varied. Also, if time permits, a similar scan will be carried out for a discharge with bursting fishbone activity.
Background:	Tearing modes, sawteeth, and fishbones are capable of causing large fast ion transport, the impact of which ranges from deleterious to helpful in the case of the 3/2 NTM which sustains q0 above 1 in hybrid plasmas. Our ability to measure this phenomena in 2D has just become available through fast visible imaging. Additionally, the DIII-D diagnostic suite will now include measurements of lost fast ions. Measurements will be made documenting sawtooth (and/or fishbone) structure and impact on the fast ion profile with a range of q95 for comparison to modeling.
Resource Requirements:	--
Diagnostic Requirements:	FIDA imaging, VBE imaging, FIDA, FILD, BES linear array, ECE
Analysis Requirements:	--
Other Requirements:	--

Title	317: Beta Induced Alfven Acoustic Eigenmode (BAAE) Studies
Name:	Van Zeeland vanzeeland@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Requested
Co-Author(s):	N.N. Gorelenkov, W. W. Heidbrink, R. Nazikian	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to document the Beta Induced Alfven Acoustic Eigenmode (BAAE) and its impact on the fast ion profile in DIII-D reversed shear plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We will begin with discharge 132710 in which a spectrum of BAAEs up to n=20 have been observed. The discharge will be repeated with varying q95 to obtain the dependence of eigenmode existence, structure, and fast ion loss on magnetic shear near the mode location. Neutral beam power and direction will also be varied to change the eigenmode drive from both fast and thermal ions.
Background:	The BAAE is a relatively new type of Alfven eigenmode found in the lowest Alfvenic gap formed by the coupling of the shear and acoustic continuum. It was recently observed in NSTX, JET, ASDEX, and DIII-D. In all devices except JET, the BAAE has been associated with enhanced fast ion transport. Due to its low frequency and compressional component, it can also interact strongly with the thermal ion population and possibly turbulence. Several questions remain such as the lack of up-chirping BAAEs in DIII-D (they only appear as short lived relatively constant frequency modes), the degree to which they cause fast ion transport, and their drive dependence on the thermal ions and neutral beam power. It is expected that the first two can be varied significantly by changing the flatness of the q-profile around qmin (and consequently mode width). Additionally, the roles of thermal and fast ion drive can be investigated by changing the relative mix of co/ctr beam power as well as overall beam power. The new diagnostics relevant to this experiment that greatly enhance the quality of data obtainable in 2009-2010 include: FIDA Imaging, Fast FIDA, ECE imaging, FILD detectors, NPAs, Linear BES array, Visible Bremsstrahlung Imaging (if intensifier is available). Combined, these will measure the eigenmode structure and impact on the fast ion population in detail for comparison to theory.
Resource Requirements:	2 co, 2 ctr source, 4 gyrotrons
Diagnostic Requirements:	ECE, BES, FIDA, FILD
Analysis Requirements:
Other Requirements:

Title	318: Rho* Scaling of Alfvenic Activity Using Hydrogen Discharge
Name:	Van Zeeland vanzeeland@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	W.W. Heidbrink, R. Nazikian	ITPA Joint Experiment :	No
Description:	The primary goal of this experiment is to document the difference in Alfvenic activity and resultant fast ion transport by repeating the well-documented deuterium AE discharge 122117 in hydrogen.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin with discharge 122117 as a reference and repeat in hydrogen. Carry out beam power scan as well as density scan. Document the variation in AE activity as well as impact on fast ion transport for comparison to equivalent discharges already carried out in deuterium.
Background:	The fast ion gyroradius/plasma radius is thought to be fundamental in determining the degree to which Alfvenic activity or other MHD induced fast ion transport impacts plasma performance. Additionally, the most unstable mode is expected to occur when the poloidal wavelength is approximately the fast ion gyroradius. Both of these ideas are fundamental to making reliable projections to ITER. Operating the well-documented AE discharge 122117 in hydrogen will provide a sensitive test of both of these predictions since essentially all relevant parameters including V_beam/V_Alfven will remain constant except rho*.
Resource Requirements:	Hydrogen plasma
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	319: L-H transition mechanism study through comparison of different triggers: NBI, ECH, and sawtooth
Name:	Wang wangg@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	UCLA, Burrell, McKee, Austin	ITPA Joint Experiment :	No
Description:	The purpose is to study L-H transition physics through comparison of edge plasma characteristics utilizing DIII-Dâ??s set of profile and turbulence diagnostics across transitions with different triggers: NBI, ECH, and sawtooth.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Trigger L-H transition with 3 different methods: NBI (balanced beam), ECH, and sawtooth in similar L-mode target and characterize edge plasma characteristics utilizing DIII-Dâ??s set of profile and turbulence diagnostics across transitions.
Background:	The trigger mechanism of edge transport barrier formation is still not clear. Take advantage of DIII-Dâ??s full set of high-resolution edge profile and turbulence diagnostics and diverse ways of generating the L-H transition could help better understand this issue.
Resource Requirements:	NBI, ECH
Diagnostic Requirements:	all profile and fluctuation diagnostics
Analysis Requirements:
Other Requirements:

Title	320: Edge Low-Density Locked Mode (LDLM) I. Mode Penetration at electron velocity-reversal surface
Name:	Waelbroeck none	Affiliation:	IFS, U. Texas
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	I. Joseph, E. Nardon	ITPA Joint Experiment :	No
Description:	Use I-coils and C-coils to try to create an n>=3 locked mode in the edge while avoiding a low-n locked mode. If successful, turn off the I-coils to unlock mode and observe spin-up and healing of the edge (and return of ELMs). Compare profiles in locked and unlocked states.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Create an edge LDLM by repeating current ramp-up ELM-suppression experiment at successively lower values of rotation. For each set of conditions, do one shot at maximum I-coil current then three shots where the I-coil is turned off at the beginning, middle, and end of the suppression window. Look for mode spin-up on Mirnov.
Background:	The design of ELM suppression coils for ITER is guided by vacuum island calculations. Given the question posed by the steepening of the electron-temperature pedestal, it is important to demonstrate that islands are indeed present in order to validate the design. Co-NBI injected H-modes are characterized by the reversal of the perpendicular velocity of the electrons on a surface lying near the top of the pedestal. Current ramp-ups make rational surfaces sweep through this reversal surface. This should result in rapid, unscreened island growth. The aim of the experiment is to try to observe the resulting island(s) in order to confirm their role in ELM suppression.
Resource Requirements:	Same resources as used for 2007 ISS ELM control experiments, see for example shot 128374 etc. I-coil in n=3, maximum current with C_supplies operating simultaneously at full current capability (6.4 kA), C-coil for optimum error field correction, 5 co-beams. All cryopumps LHe cold.
Diagnostic Requirements:	All pedestal (esp. Thomson and CER) and lower divertor diagnostics, divertor probes, midplane recip probe. Edge current measurements (especially simultaneously) with the Li-beam and co- plus counter-beam MSE would be highly desirable.
Analysis Requirements:	Control room SURFMN of applied mode spectrum on EFIT01. Post experiment profile analysis, kinetic EFITs, CER (Ti, rotation and fZ), ELITE stability analysis, TRIP3D field line loss fraction analysis, divertor strikepoint pattern analysis from cameras and particle balance analysis.
Other Requirements:

Title	321: RWM stability with zero neutral beam torque
Name:	Strait strait@fusion.gat.com	Affiliation:	GA
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The goal of this experiment is to study RWM stability in ITER-like conditions of zero neutral beam torque, using balanced NBI. Non-zero rotation will be generated by intrinsic rotation and (optionally) n=3 magnetic perturbations. Is this sufficient for robust RWM stability?	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use a low-triangularity LSN plasma with low no-wall limit. Go to exactly balanced NBI early in the flattop. If necessary, use ECCD to suppress NTMs. (see shot 132253, for example). Apply a non-resonant n=3 perturbation with the I-coil, and vary the amplitude in order to vary the NTV torque. Vary the intrinsic rotation by varying density, and thus Ti. If a non-rotating n=1 mode appears, apply RWM feedback with the I-coils. Vary the gain and time constant of the feedback control.
Background:	Experiments in 2008 indicated that the RWM can be stable at very low rotation. Intrinsic rotation and n=3 NTV torque are therefore often cited as possible solutions to RWM stability in ITER. This experiment will test the stability of RWMs in these conditions.
Resource Requirements:	Both counter-injection sources. I-coils with SPAs.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	322: Study of High Temperature Pedestals in JET
Name:	Solano none	Affiliation:	JET &Ciemat, Spain
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	P.J.Lomas, B. Aper, V. Parail, A. Turnbull, T. Osborne, G.L. Jackson, T. Leonard, T.E. Evans, M. Fenstermacher	ITPA Joint Experiment :	No
Description:	Use the VH-mode as a vehicle for studying a pedestal as close to ITERâ??s as it is possible in DIII-D. A successful experimental might lead to a steady state VH-mode with no ELMs	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Revisit the DIII-D VH regime to study a high temperature low density pedestal.
Background:	In terms of both resistivity and collisionalities VH-mode pedestals can be very ITER relevant. Note that the present plan is for ITER to assess the H-mode with full heating at low density, and then increasing the density until a steady ELMy H-mode regime (without ELMs) is attained. According to TRANSP simulations by Bob Budny, initially the plasma could be in a VH or Hot Ion regime. The low density access is necessary because of the high L to H power threshold expected. Note that the natural route to obtain the VH mode automatically produces such a pedestal evolution. We believe that the characterization and understanding of pedestal evolution and MHD instabilities present in a VH plasma before the first ELM provide a good benchmark of ELM models to be applied to ITER prediction, as well as pedestal scaling (heights, widths). Further, such a plasma is a good target to test ELM control or suppression techniques. In JET the Hot Ion H-mode regime is very similar to the DIII-D VH-mode. Low recycling and sawteeth control are essential ingredients to obtain high Te pedestals. In the past pedestal temperatures as high as 3.5 keV were obtained in JET in Hot-Ion H-modes. Recently the regime was recovered, and we obtained up to Te_ped= 3 keV during the ELM-free phase, but edge MHD (â??outer modesâ??) reduced the Te to 2.7 keV just before the ELM. Outer modes are described in (Nave, Nucl. Fusion 37 (1997) 809-24): they are low frequency low n modes, localized to the pedestal region, and were identified (Huysmans et al, Nucl. Fusion 38 (1998) 179-187) as external (current-driven) kink-modes. We have various recent examples with similar temperatures, and have not had opportunity to fully analyze the data yet (experiments were carried out on 29/10/2008 and 25/11/2008). We clearly see outer modes, very distinct from the high n ELM precursors observed more typically in the ELMy H-mode regime. In some pulses we observe both types of modes, either simultaneously (bursts of high n modes amongst low n kinks), or at different times. Washboard modes [C P Perez et al 2004 Plasma Phys. Control. Fusion 46 61-87] are seen in many of these plasmas, but appear to be weaker when outer modes are present.
Resource Requirements:	7 NB sources, ECH (for possible current profile control), Icoils, 3 cryopumps, prefer fresh boronization for low recycling (not required)
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	323: Edge Low-Density Locked Mode (LDLM) II: Locked EHO
Name:	Waelbroeck none	Affiliation:	IFS, U. Texas
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	I. Joseph, E. Nardon	ITPA Joint Experiment :	No
Description:	Use I-coils and C-coils to try to create an n>=3 locked mode in the edge while avoiding a low-n locked mode. In this second part of a two-part experiment, try to achieve the locking of a pre-existing edge mode, the EHO. If successful, turn off coil currents to unlock the mode and observe its spin-up.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Create a high-rotation counter-NBI QH mode. Ramp-up the I-coil current to its maximum value to see if the EHO can be locked. Repeat at successively lower values of rotation. Use even-parity I-coil phased so as to maximize coupling to edge while minimizing likelihood of 2/1 LDLM.
Background:	The EHO provides an attractive alternative to RMP ELM-suppression with which it shares several characteristics. As a saturated MHD instability, it is likely to result in some degree of magnetic reconnection. This suggests the possibility to lock the mode with the I-coils. Locking the EHO, with its rich harmonic content, would strongly enhance edge stochasticity and yield abundant information on both the QH and RMP mode suppression mechanism and on their relationship.
Resource Requirements:	Reverse Ip for QH-mode. 7 NBI sources
Diagnostic Requirements:	All pedestal (esp. Thomson and CER) and lower divertor diagnostics, divertor probes, midplane recip probe. Edge current measurements (especially simultaneously) with the Li-beam and co- plus counter-beam MSE would be highly desirable.
Analysis Requirements:	Post experiment profile analysis, kinetic EFITs, CER (Ti, rotation and fZ), ELITE stability analysis
Other Requirements:

Title	324: Modification of plasma rotation using NTV torque
Name:	Sabbagh none	Affiliation:	Columbia & NSTX
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	G.L. Jackson	ITPA Joint Experiment :	No
Description:	Use n=2 rotating field with Icoils to induce a bulk plasma rotation via NTV torque	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Later
Background:	Later
Resource Requirements:	Later
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	325: Super H-Mode
Name:	Snyder snyderpb@ornl.gov	Affiliation:	ORNL
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	K. Burrell, R. Groebner, T. Osborne, A. Leonard (?)	ITPA Joint Experiment :	No
Description:	Probe the extreme limits of H-mode performance by an overall optimization of the pedestal height going beyond what is possible in usual experiments. Attempt to achieve record pedestal height and global confinement.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Make detailed use of control systems and improved understanding of pedestal physics to achieve very high pedestals while avoiding global beta limits and the H-L back-transition. Multiple approaches are possible. A leading candidate is to begin with a QH mode discharge in the most optimal shape possible. First achieve a steady QH mode at low density. Then slowly ramp up the density while remaining in QH mode (via rotation modification, and variation of fuelling and pumping), accessing the far upper corner of the pedestal stability diagram, which is inaccessible at fixed density. As pedestal height becomes very high, global confinement will become "too good" and power will have to ramped down to avoid global (hard or soft) beta limits. Shape and q profiles in the core will be modified (similar to AT) maximum global betaN limits (may need NTM stabilization via ECH). As power is ramped down, must avoid H-L back transition or appearance of TIII ELMs - accomplish this by selecting plasma with very low LH threshold power.
Background:	Understanding of the physics of the pedestal width and height is improving (though still incomplete). In particular, peeling-ballooning stability (and the EPED1 model) predicts constraints on the pedestal that have been tested in some detail. That understanding allows us to calculate stability diagrams that are usually presented in a j-alpha space. The stability boundary in that space increases rather dramatically shape. In fact, for very optimized shapes (high triangularity, low squareness...), a "nose" on the upper-right is pulled out and extremely high pedestals are predicted to be stable. However, that region is not accessible for any fixed pedestal density. At low density one hits the low-n peeling bound, and at high density the high-n ballooning bound. However, by starting at low density, moving up to the peeling bound, and only then increasing density (avoiding large ELMs), it is possible to access this space. This has been partially achieved in QH mode discharges in 2008, but in principle can be carried to much higher levels if QH mode can be maintained.
Resource Requirements:	Significant study of control techniques needed, as well as shape optimization before the expt, ECH
Diagnostic Requirements:	All pedestal diagnostics
Analysis Requirements:	EPED1 and P-B stability studies before expt
Other Requirements:

Title	326: ITER baseline H-mode access
Name:	Wang wangg@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):	Petty, Doyle, DeBoo	ITPA Joint Experiment :	Yes
Description:	Better understanding of power requirement for a good H-mode confinement has been identified as an urgent task for ITER and listed as an ITPA joint experiment. The purpose of this experiment is to characterize the ITER baseline H-mode access with heating power near the L-H transition threshold power, i.e. characterize the H-mode confinement, ELM properties, etc as the heating power is slightly higher than the L-H transition threshold power.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	After establishing ITER-like L-mode Deuterium target plasma at ne~0.4*n_GW and H-mode threshold power being found with power ramp, vary input power shot-to-shot but at a fixed level during each shot (PLTH~1.1, 1.2, 1.3, 1.5 x P_th). Establish new equilibrium condition. If new ne level is below 0.8*n_GW, then gas puff to achieve 0.8*n_GW to see if H-L transition occurs. Characterize ELMs and confinement properties after the L-H transition. If good confinement in the H-mode phase is not achieved, raise input power further. Repeat above process at low and high Bt levels.
Background:	ITER will operate in H, (He), D, before DT operation. Due to power limitations, ITER will likely need to access H-mode at a low density (0.5*10^20 m^-3, ~0.4*n_GW) with input power close to H-mode power threshold then achieve their target flattop density of 1*10^20 m^-3 without more input power. So it is important to have dedicated experiments to demonstrate/document this scenario.
Resource Requirements:	Beams
Diagnostic Requirements:	standard
Analysis Requirements:
Other Requirements:

Title	327: Hyper-Velocity High-Density C60-Fullerene Plasma Jet for Disruption Mitigation
Name:	Bogatu nbogatu@far-tech.com	Affiliation:	FAR-TECH, Inc.
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	S.A. Galkin (FAR-TECH, Inc.), J.S. Kim (FAR-TECH, Inc.), E. Hollmann (UCSD), P.B. Parks (GA), E.J. Strait (GA)	ITPA Joint Experiment :	No
Description:	A successful disruption mitigation must simultaneously convert ITER plasma energy density (~1 GJ in ~840 m^3) into radiation within 1 ms and to get the Rosenbluth density of electrons [1] (free and bound) by increasing it at least 100 times all over the plasma cross section to suppress the REs avalanche. The principle solution is considered to be the impurity injection. First experiments with massive (neutral) gas injection (MGI) in the scrape-off layer region seemed [2] to produce high density streams penetrating to the plasma center. But following recent attempts [3] have shown that the injected gas apparently does not penetrate far into the plasma and only relatively fast MHD mixing of the partially ionized impurity seems to occur. All these phenomena happen because once the impurity atoms are ionized in a thin outer layer of the hot plasma they can no longer penetrate the confining magnetic field unless they posses a very high velocity. For relatively low gas injection velocity of MGI the mitigation process has to rely on the inward propagation of a cooling front, enhancement of MHD activity, and mixing of impurity with the plasma. It might take too long time and be very difficult to properly control the sequence of these processes, so that to provide reliable and prompt disruption mitigation. As stated in [4] �??unfortunately, it will be difficult to project such an MHD mixing process to ITER.�?� Hence there is a need for alternative of impurity injection methods providing the proper balance of necessary mass, acceptable atomic number Z, in conjunction with a suitable specific momentum (or density-velocity) to penetrate/deliver it in the core plasma on the fast disruption time scale for real-time mitigation. We propose to fulfill the above requirements by using hyper-velocity high-density C60-fullerene plasma jets from a coaxial accelerator [5,6]. For disruption mitigation we need to get in the end not only a hyper-velocity, but also a large mass of impurity, produced in a pulsed source in a suitable form for acceleration as a plasma slug. The plasma gun technique must be modified, adapted, and re-designed for disruption mitigation. We propose to use the C60-fullerene (buckyball) heavy molecule to form a high-density compact plasma slug of large mass made of sub-nanometer C60 ionized molecules. Using C60 in conjunction with the pulsed source of TiH2 grains [7] requires significant modifications of the design. [1] M.N. Rosenbluth et al., Fusion Energy (Proc. 16th Intl. Conf.,1996) IAEA, Vienna, vol 2, 979 (1997) [2] D.G. Whyte et al., Phys. Rev. Lett. 89 (2002) 055001-1 [3] E. Hollmann et al, Fusion Energy (Proc. 20th Conf.,2004) IAEA, Vienna, EX/10-6R; R. Granetz et al., 21st IAEA FEC, Chengdu, China, 2006, EX/4-3 [4] R. Stambaugh, 21st IAEA FEC, Chengdu, 2006 [5] I.N. Bogatu et al., J. Fusion Energy vol.27, 6 (2008) [6] I.N. Bogatu et al., 49th APS DPP: Bull. Am. Phys. Soc. 52, 358 (2007) [7] I.N. Bogatu, US Patent application (2008)	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	First experimental test on DIII-D could take place earliest in the end of 2010. The coaxial plasma gun with pulsed power solid-state TiH2/C60 needs approximately two years to be modeled, designed, built, tested and with its plasma jet characterized on a test-bed. Modeling phase (on-going at present): 1. Perform analytical estimations and numerical simulation for the molecular gas mixture production in pulsed power TiH2/C60 source to achieve the required mass (0.3 to ~2 g) 2. Characterize the Laval micro-nozzles, which form the special grid filter, to maximize the injection velocity of molecular gas into the coaxial accelerator 3. Determine the optimum parameters of the plasma gun to accelerate the plasma slug to hyper-velocity (~10 to 50 km/s) 4. Identify the key atomic processes during the plasma jet transport 5. Evaluate the jet penetration capability into the hot tokamak plasma and confining magnetic field in the perspective of extrapolation to ITER 6. Determine the feasibility of a proof-of-principle disruption mitigation experiment on DIII-D with high-density hyper-velocity C60 plasma jet. We know from our on-going collaboration that a pulsed power coaxial plasma gun with TiH2/C60 and required parameters can be built and tested by HyperV Technologies.
Background:	Recent successful fueling experiments on Globus-M tokamak [1,2,3] demonstrated the deep (to half minor radius) and fast penetration into tokamak plasma and its confining magnetic field of a high-velocity (~ 140 km/sec) dense (~ 2x10^16 cm-3) plasma jet with a mass of 17 μg, accelerated using a coaxial plasma gun. Electron density has risen to a double value much faster (<0.5 ms) than for fast gas jet injection (~2.5 ms) and over the whole plasma cross section. TiH2 has an outstanding capacity of storing H2 and can release 448 cm^3 H2/g or 1.2x10^22 molecules H2/g. The H2 release begins at 573 K and is complete at 873 K (usually TiH2 is heated above 738 K). Transient heating and required high temperature is achieved by discharging a capacitor bank and driving a high current pulse through TiH2 grains. For example, in the fueling plasma gun of Globus-M tokamak, a pulsed current of 14 kA for 10 μs releases ~ 4x10^19 H2 molecules from a source volume of 3 cm^3 and a mass m_TiH2 ~12 g (Ï?TiH2 = 3.9 g/cm3). Sublimation temperature of C60-fullerenes (800 K) is located within the range where H2 is released from TiH2. Thus, by mixing or coating TiH2 grains with C60 fine powder (C60 is soluble in organic solvents) we can produce both sublimation temperature for C60 and H2 release at a high pressure. Driving a current pulse through the TiH2 grains �??oven�?� transiently heats and explosively sublimates the very fine C60 powder transforming it into a heavy molecular gas. The heavy C60 molecule can trap metal atoms or any of the noble gases [4,5] (Ne, Ar, Kr, or Xe) forming a stable �??micro-shell�?� with even larger mass and containing atoms of high atomic number Z which can be released [6], a property which can be used to enhance the potential for disruption mitigation, if necessary. Moreover, C60 basic properties are now precisely determined and pure C60 is commercially available [7]. [1] A.V. Voronin, et al., Nucl. Fusion, 45, 1039 (2005); A.V. Voronin, et al., IAEA 2006 Topic: EX/P3-18 [2] V. K. Gusev et al., 11th IST Workshop, Chengdu, China, October 11-23, 2006 [3] V. K. Gusev, et al., Nucl. Fusion 46, S584 (2006) [4] M. Saunders, et al., Science 271, 1693 (1996) [5] H. A. Jiménez-Vázquez and R. J. Cross, J. Chem. Phys. 104, 5589 (1996) [6] R. Shimshi, et al., Tetrahedron, 52, 5143 (1996) [7] MATERIALS TECHNOLOGIES RESEARCH Ltd., C60 (99.5+%) at $26.50 per gram.
Resource Requirements:	1. Access port on DIII-D to install the coaxial plasma gun 2. Restricted area/space to install the high-voltage capacitor bank (few cubic meters) and transmission line
Diagnostic Requirements:	1. SXR toroidal arrays (TAs) 2. DISRAD camera 3. SXR poloidal array (PA) 4. Hard X-ray detectors 5. Bolomerters 6. Tangential camera 7. ECE 8. Thomson scattering 9. Magnetic sensors
Analysis Requirements:	--
Other Requirements:	--

Title	328: Power hysteresis in ITER baseline scenario
Name:	Wang wangg@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	Doyle, Petty, DeBoo	ITPA Joint Experiment :	No
Description:	Better understanding of power requirement for a good H-mode confinement has been identified as an urgent task for ITER. The purpose of this experiment is to determine the power hysteresis effect in the ITER baseline H-mode discharges via ramping down NBI power at various density levels at low and high magnetic field.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Find L-H power threshold with power ramp. At stationary H-mode flattop ramp down power (but keep torque unchanged) until the H-L back transition occurs. Repeat above process at low and high Bt levels, and at different H-mode density levels (use density control tools such as feedback control or I-coil application as necessary).
Background:	ITER will operate in H, (He), D, before DT operation, and available auxiliary power is predicted to be limited. It is desirable to find out in present ITER-like discharges what the plasma properties are as a function of input power.
Resource Requirements:	beams
Diagnostic Requirements:	standard
Analysis Requirements:	--
Other Requirements:	--

Title	329: ELM-driven RWM: comparison between DIII-D and JT60U
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	G. Matsunaga, H. Reimedes, M. Takechi	ITPA Joint Experiment :	Yes
Description:	The high betan plasma was found to be stable significantly lower than reported. However, Even in this stabilized regime, stable steady-state operation is not unconditionally guaranteed. Around the marginal condition, several MHD-driven RWMs were excited. With high plasma rotation, the ELM-driven RWM caused the βN collapse after a series of ELM events (ELM-driven RWM). Typically, ELM-driven RWM is relatively mild mode and can be suppressed. However, too-frequent ELM events can cause difficulty for the feedback. It seems the RWM pattern before / after EWM evolves in time. In JT60U high betan exploration, RWM associated ELMs have been observed. At present the details of ELM characteristics and relation to the ELM onset are investigated. The ELM and RWM onset in JT60U are often coincided together as was observed in DIIID. In the DIIID, effective avoidance of irreversible process of ELM-driven beta collapse was the usage of C-coil as a dynamic error field correction in addition to the I-coil RLM control. A possible hypothesis is as follows. With too-frequent ELM events, the amplitude of an ELM-driven RWM remains finite even at the next ELM events. Pre-existing-residual RWM influences the next RWM formation during ELM event. by still-uncorrected error field or due to the RWM left over from the previous RWM. The formation of the post-ELM RWM (ELM-driven RWM) is determined during the ELM period along with pre-existing-residual RWM.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	One possible hidden parameter is the uncorrected error field, which determines the ELM-driven RWM amplitude. The best approach is to optimize the C-coil dynamic error field correction along with fast I-coil RWM control. This has been proposed several occasions, but only a few cases were pursues to satisfactory level. -Observing the ELM events at high bn collapse -ECCD is highly desirable to avoid NTM -Bt ramp down may be needed, - Documenting profile / mode pattern
Background:
Resource Requirements:	NB to exceed no-wall beta limit and to control plasma rotation ECCD for NTM suppression
Diagnostic Requirements:	ECE, CER, Mino/saddle coils, soft-X ray
Analysis Requirements:
Other Requirements:

Title	330: Scaling of conventional H-mode power hysteresis in DIII-D
Name:	Wang wangg@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The purpose is to systematically study the power hysteresis effect in conventional H-mode discharges in DIII-D, i.e. obtain scaling of this effect vs Bt and density.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Achieve conventional H-mode with balanced beam. Find L-H power threshold with power ramp. At stationary H-mode flattop ramp down power (but keep torque unchanged) until the H-L back transition occurs. Repeat above process at different Bt levels, and at different H-mode density levels (use density control tools as feedback control or I-coil applied if necessary).
Background:	ITER will operate in H, (He), D, before DT operation, and available auxiliary power is predicted to be limited. H-mode will be achieved at a low density with low input power then increase density utilizing hysteresis effect. This hysteresis effect has been observed in some tokamaks including DIII-D, but not in JET recently. A systematic study of the power hysteresis effect dependence on Bt and density is necessary.
Resource Requirements:	beams
Diagnostic Requirements:	standard
Analysis Requirements:
Other Requirements:

Title	331: Multiple low-n RWM identification and feedback control
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	A. Garofalo, J.S. Kim, M. Okabayashi, H. Reimerdes, E. Strait	ITPA Joint Experiment :	No
Description:	The proposal is first to identify the presence of multiple low-n (up to n=3) RWMs, and then to provide the corresponding feedback control. Both rotational stabilization and active multiple mode feedback controls will be applied.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The first step will be to identify the presence of the multiple low-n RWM modes. Quite often, ELM-induced RWMs in high beta plasma are accompanied by significant multiple low-n modes (at least for both n=1 and n=3 modes). Once these modes are identified, the feedback control will be provided to suppress each mode simultaneously 1) using rotation control and/or 2) using independently controlled coils. If the non-RWM noise affects the feedback performance, a new Kalman filter compatible for n = 1 and 3 RWMs would be applied.
Background:	A paper recently published by Y. In et al, Phys. Plasmas 15, 102506 (2008) identified the multiple low-n RWMs above each low-n no-wall limit (betaN ~ 3.76) in DIII-D discharge. Specifically, while the n=1 mode was suppressed likely due to both rotational stabilization and n=1 RWM feedback, the n=3 mode appeared dominant, leading to beta collapse. Also the presence of the n=2 mode was also identified by post-processing Mirnov probes. However, since no consideration was made for the multiple low-n modes at the time of the experiments, the n=1 mode feedback was greatly influenced by the aliasing component from n=3. Since DIII-D is equipped with multiple low-n mode identification and feedback control capabilities (at least for n=1 and n=3), it would be ideal to demonstrate the multiple low-n RWM feedback control scheme, once high beta plasma exceeds the corresponding ideal MHD no-wall limits. Along with the PCS rotation control capability, the fully independent I-coil control needs to be confirmed, prior to the experiments. When the non-RWM noise becomes prevalent, a multiple-mode compatible Kalman filter which FAR-TECH has been developing will be applied for the first time.
Resource Requirements:	7 NBI sources, 4 gyrotrons for ECCD, and ICRF (if available for high-beta operation)
Diagnostic Requirements:
Analysis Requirements:	n=2 component extraction from Mirnov signals, fast particle populations
Other Requirements:	Rotation control would be used first to see if the n > 1 modes are rotationally stabilized. Then, the active multiple low-n mode control will be provided.

Title	332: Pedestal variation with collisionality
Name:	Snyder snyderpb@ornl.gov	Affiliation:	ORNL
Research Area:	Pedestal Structure	Presentation time:	Requested
Co-Author(s):	R. Groebner, A. Leonard, T. Osborne	ITPA Joint Experiment :	No
Description:	Study the variation of the pedestal width and height as collisionality (density and Greenwald fraction) is varied over a wide range.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In discharges optimized for pedestal measurements (similar to 2008 pedscale extp), vary the density by roughly a factor of 8 via modification of particle source and pumping. Conduct scans at multiple values of Ip in order to significantly vary the width.
Background:	eading models of the pedestal width suggest scaling of the width with sqrt(beta_pol_ped) (KBM based), or with poloidal gyroradius (ExB suppression of drift turbulence). While these parameters scale similarly in many experiments, they can be separated by varying the density. The ratio of sqrt(beta_pol_ped)/rho_pol_ped scales like sqrt(neped). Hence by varying the density by ~8, we can achieve roughly a factor of 2.8 separation between these scalings, as well as provide a vigorous test of models of the width, as well as models of the height such as EPED1. Also, the width model (~sqrt(beta_p_ped) in EPED1 corresponds to a kinetic ballooning mode constrained pedestal model. The dominant dependence of the KBM is captured by the beta_p_ped dependence. However, there should be a weak, but significant at high collisionaly, dependence on the collisionality, which at high values suppresses the bootstrap current, altering the magnetic shear. Data at high collisionality will help both test the existing EPED1 model and aid the testing/development of planned EPED2 model.
Resource Requirements:
Diagnostic Requirements:	full pedestal diagnostic set, including reflectometry
Analysis Requirements:	EPED1 predictions before expt
Other Requirements:

Title	333: Low Radial Electric Field For Maximum NTV
Name:	Cole andrew@woodruffscientific.com	Affiliation:	Woodruff Scientific Incorporated
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Callen, Hegna, Garofalo, Reimerdes, Solomon	ITPA Joint Experiment :	No
Description:	We propose to explore a collisionality regime where NTV is maximal to facilitate better comparison between experiment and theory models. The goal would be to obtain E_r ~ 0 before I-coil activation. Optimally we would like to scan a set of shots with E_r around E_r ~ 0---between 'slightly' co-rotating and nearly balanced (as in shot #131861 from the 2007 campaign). Here "slightly co-rotating" would be determined from radial force balance, where to minimize E_r would require that the toroidal rotation Omega_t = V_t/R, nearly balance the diamagnetic type flow T_{i0} d n_{i0}/dr /(n_{i0} q_i R_0 B_theta), where T_{i0} is the equilibrium ion temperature, n_{i0} the equilibrium ion density, d n_{i0}/dr the equilibrium radial ion density gradient, q_i the main ion species' charge, R_0 is radius of the magnetic axis, and B_theta is the poloidal magnetic field.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This is a continuation and extension of experiments started in FY08 (MPs 2008-02-01 and 2008-02-02) operating in reversed-Ip. We will use the C-coil for optimal correction of the n=1 error field, determined via DEFC. We will use odd-parity on the I-coil for an almost purely non-resonant n=3 field to apply braking. The braking is applied after all profiles have reached nearly stationary conditions. We will apply the braking field with a fast step and maintain it for at least a couple of momentum confinement times, before ramping down the field slowly, through a succession of torque equilibrium states, in order to carry out a full scan of the braking amplitude, and test torque dependence on the plasma rotation. Feedback control of the NBI power is required in order to maintain both beta and the torque constant during the application of the braking. We will carry out shot-to-shot scans of the NBI torque such that the initial toroidal plasma rotation prior to activation of the I-coil varies from 'slightly' co-rating to nearly balanced, and repeat the braking measurement mentioned above for each case.
Background:	Experiments on DIII-D in the 2008 campaign demonstrated for the first time, the existence of a predicted neoclassical â??offsetâ?? rotation associated with a toroidal neoclassical viscous torque [NTV] driven by a non-resonant magnetic field (NRMF) applied to a tokamak [Garofalo et al., PRL (2008)]. Most of the discharges were in a regime far from optimal with regards to the predicted peak NTV torque. New theoretical work [A. Cole, C.C. Hegna and J.D. Callen, "Low Collisionality Neoclassical Toroidal Viscosity in Tokamaks and Quasi-symmetric Stellarators", UW-CPTC 08-8, to be submitted to PoP] predicts NTV is maximal when nu_i/epsilon ~ 4-10 *omega_E, where nu_i is the ion-ion collisionality, epsilon = r / R_0, and omega_E = E_r/(R_0 B_theta) is the toroidal precessional E x B drift frequency. Shot 131408 in the 2007 campaign analyzed for omega_E and nu_i/epsilon was in a regime where apparently NTV torque was ~ 10% of the predicted peak.
Resource Requirements:	Same resources as used for MPs 2008-02-01 and 2008-02-02, except use a 7 kA capable I-coil hook-up and operate a reversed-Ip discharge only.
Diagnostic Requirements:	All standard magnetics and internal profile diagnostics, including reflectometer for density profile measurements. FIR and BES fluctuation measurements should also be acquired.
Analysis Requirements:	Analysis of the magnetics for extraction of the n=3 plasma response. Kinetic equilibrium reconstruction for accurate n=3 stability modeling. Time-dependent profile fitting for TRANSP modeling of the discharge evolution.
Other Requirements:

Title	334: Taking ITER baseline demonstration discharges to more reactor relevant conditions
Name:	Doyle doyle@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ITER Demonstration Discharges	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	This proposal aims at improving the match between the DIII-D baseline scenario demonstration discharges and anticipated ITER conditions in the following ways: (1) Increase density, so as to match ITER operating point of 0.85 n_GW. (2) Decrease density, to better match ITER collisionality, and also allow gyrotron heating. (3) Increase Te/Ti (at lower densities), using ECH, (4) Use ECH to increase electron heating fraction. Effect on confinement and fusion performance will be evaluated for each area.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	What is proposed is relatively straightforward: Density can be increased using pellets and/or gas puff. Lowering the density requires higher ELM frequency (so as to obtain density control), which can be obtained by using incresed fraction of counter-NBI. Lower density will allow use of ECH to increase electron heating and Te/Ti (this ratio is unity for current high densities, but would not be at lower density).
Background:	The ITER demonstration discharges operated in 2008 concentrated on matching ITER target beta_n, confinement and normalized current. However, the match to anticipated ITER conditions can be substantially improved in other ways, and this forms the basis for this proposal. Better matching ITER rotation is dealt with in a separate proposal.
Resource Requirements:	NBI, gyrotrons, pellet injector
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	335: Eigenmode model-based n=1 RWM feedback control with Kalman filter
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	Integrated and Model-Based Control	Presentation time:	Not requested
Co-Author(s):	J.S. Kim, D. Humphreys, M. Walker, E. Schuster, RWM physics working group	ITPA Joint Experiment :	No
Description:	The goals of this proposal are to (1) demonstrate the superior performance of the dynamic Kalman filter based on wall surface current eigenmodes, compared to the counterpart based on picture-frame wall model (2) evaluate the performance of the eigenmode model-based PD controllers, compared with the off-line predictions (3) assess the robustness of model-based controllers	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish high beta, high torque RWM plasmas, where ELM-noises are dominant. It is expected that ELM-induced RWMs would readily occur. When RWM cannot be obtained in high torque plasmas, lower the plasma rotation by injecting counter beams with no magnetic braking. Once RWM is obtained in high torque plasmas, the dominant non-RWM noise will be likely to be ELMs. Thus, the performance of the eigenmode model-based Kalman filter will be tested in comparison with picture frame model-based counterpart with respect to ELM-noise discrimination quality (related to Goal 1). Simultaneously, the closed-loop performance will be evaluated with respect to RWM suppression (related to Goal 2), which may necessitate gain scans. As the pressure-driven RWM experiments have been plagued due to the lack of reproducibility, careful comparison needs to be made even when RWM is believed to be actively stabilized. Thus, a successful RWM-free shot will be followed by an open-loop testing, which will also provide the measured open-loop RWM growth rate. Since RWM open-loop growth rate is the key parameter to relate experimental observation to eigenmode model-based modeling, it will be a good indicator how appropriately the model has been established for a given growth rate. If a reproducible pressure-driven RWM is obtained in low torque plasmas, the non-RWM noise might not be ELMs, but could be fishbones or tearing modes. This will require the Kalman gain changes, so a set of the Kalman filters (e.g. based on a variety of process noise covariance magnitudes) will be prepared. The rest of the procedures will be exactly the same as mentioned for high torque case. To assess the robustness of the model-based controllers (related to Goal 3), we will do beta scans. As for high torque plasmas, the beta scan will be equivalent to growth rate scan. Thus, the beta scan can directly address the robustness of the model-based controllers, which also allows us to assess the need of gain scheduling. As a result, the robustness of the eigenmode model-based controllers will be evaluated. As a note, although the eigenmode-based DIII-D/RWM model has been developed without taking into account the plasma rotation so far, the performance of eigenmode-based Kalman filter appears to fit in high torque plasmas, where an ideal MHD assumption is deemed appropriate.
Background:	In high beta, high torque RWM experiments, a picture-frame model (modeled open-loop RWM growth rate, gamma =120 rad/s) was experimentally confirmed to be reasonable to describe the DIII-D/RWM system, in that the dynamic Kalman filter (based on picture frame wall model) was effective in discriminating the ELM-noise from RWM [Y. In et al., Phys. Plasmas 13, 062502 (2006)]. As a more advanced model, FAR-TECH's eigenmode-based DIII-D/RWM model was predicted to be superior to picture frame wall model in terms of its effectiveness and computation times. For example, the performance of 3 eigenmode-based Kalman filter (i.e. requiring 6 wall states) was almost the same or slightly better than that of the successful 72 picture frame wall model without compromising the capability to effectively discriminate non-RWM noise from RWM. Also, a recent simulation based on the eigenmode-based DIII-D/RWM model showed an excellent agreement with vacuum discharges, which is now checking with the experimental results regarding current-driven RWM feedback stabilization. Based on the latest DIII-D/RWM model, several types of model-based controllers including linear-quadratic-Gaussian (LQG) controller, are being designed in the same manner studied for preliminary model-based controllers [S. Yang, E. Schuster et al, APS-DPP (2008), D. Humphreys et al., Nucl Fusion 47, 943 (2007)]. While all the ITER controllers are expected to be model-based, no validated RWM model exists, applicable to controller design. The eigenmode-based DIII-D/RWM model is one of the candidates, which needs to be validated.
Resource Requirements:	5 co-beams and 2-counter beams, 2 gyrotrons for ECCD
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:	Desirable to be considered for both RWM physics and model-based control groups

Title	336: ELM pacing driven by non-resonant magnetic fields
Name:	Solomon solomon@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	A.M. Garofalo, H. Reimerdes	ITPA Joint Experiment :	No
Description:	he basic goal is to try to advance our understanding of the mechanism resulting in the pacing of ELMs through non-axisymmetric fields.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin with a plasma with relatively large and infrequent (~10-20 Hz) type-I ELMs, probably utilizing a higher triangularity shape with low collisionality. Add modulated I-coil, and scan the frequency up to the maximum tolerable (<~ 100 Hz). This should also include frequencies below the natural ELM frequency to see if ELMs can be slowed down, which might be useful for studying other ELM phenomenon in the future. Unlike in the previous data, we probably should avoid density and betan feedback, to assess the impact of the I-coil modulation on the density and stored energy. At each frequency level, determine the minimum amplitude of I-coil current required to pace ELMs. Compare pedestal parameters with no I-coil case. Since previous data shows that ELMs occur at the maximum and minimum of the I-coil current independent of the amplitude of the modulation (ie when dI/dt=0), it would be desirable to try other modulation schemes (eg triangular) where dI/dt is effectively always finite to see if ELMs might be completely suppressed.
Background:	Experiments in 2008 aimed at investigating the role of non-resonant magnetic fields on rotation found that modulation of the applied n=3 field resulted in the ELMs being paced at twice the modulation frequency. This pacing was observed under a wide range of conditions (approx factors of two in density, temperature and betaN, and collisionality as low as typical RMP ELM suppressed H-mode), and in both high and low triangularity shapes. The energy per ELM appears to scale with 1/f, but to date, only a two pacing frequencies have been measured and the reference discharge with no pacing is not well documented. No clear mechanism has been identified for this effect, although there is possibly a relation to the TCV observations of ELM pacing by jogging the plasma vertically in the vacuum vessel.
Resource Requirements:	1 day experiment, I-coils
Diagnostic Requirements:	fast magnetics, standard profile diagnostics...
Analysis Requirements:	fast WMHD signals etc, edge profiles, ELITE stability analysis
Other Requirements:

Title	337: ne and Te fluctuation level and radial correlation length in both low and high field side of L-mode
Name:	Wang wangg@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	UCLA, etc	ITPA Joint Experiment :	No
Description:	The purpose is to make simultaneous measurement of density and temperature fluctuation level and radial correlation length in both low and high field side of an L-mode plasma, and compare with TGLF, GYRO predictions.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Develop an L-mode plasma with appropriate Bt and density so that correlation reflectometer/Doppler backscattering and CECE system can take measurements in both low and high field side. Take data together with BES in the low field side in repeat shots. Fluctuation data will also be collected with Doppler reflectometer, FIR scattering, PCI, and Langmuir probes, etc.
Background:	In the last year for the first time simultaneous density and temperature fluctuation measurement using CECE and BES in the low field side was very successful. Expand this effort to high field side for the first time, which is achievable with CECE and correlation reflectometer/Doppler backscattering will provide further test of predictions of modeling and simulations.
Resource Requirements:	--
Diagnostic Requirements:	all profile and fluctuation diagnostics
Analysis Requirements:	--
Other Requirements:	--

Title	338: Test of kinetic ballooning mode constrained pedestal
Name:	Snyder snyderpb@ornl.gov	Affiliation:	ORNL
Research Area:	Pedestal Structure	Presentation time:	Requested
Co-Author(s):	R. Groebner, A. Leonard, T. Osborne	ITPA Joint Experiment :	No
Description:	Test in as much detail as possible, the theory that the gradients in the edge pedestal are constrained by the onset of the kinetic ballooning mode (data will of course be useful for testing other pedestal width models).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Create discharges with very wide pedestal (high beta_p_ped) by operation at high Bt, low Ip, strong shaping. Optimize for diagnostic coverage including reflectometry. Create long steady states and use breathing to diagnose in as much detail as possible the actual gradients within the pedestal. Vary Ip and collisionality to vary the expected KBM limit. Look for KBM signature with turbulence diagnostics.
Background:	It has been proposed that kinetic ballooning mode turbulence, onsetting at a critical value of alpha (which varies with shear) yields a model of the pedestal width consistent with many observations. However, it is normally not possible to measure gradients inside the pedestal with very high accuracy in order to fully test this idea. By creating very wide pedestals (with many Thomson channels inside the barrier) it is possible to more accurately measure these gradients and compare to theory. The onset of the KBM should correspond to high frequency turbulence which should be measurable and distinguishable from lower frequency drift modes.
Resource Requirements:
Diagnostic Requirements:	full pedestal profile and turbulence diagnostic set. BES across pedestal region and reflectometry
Analysis Requirements:	EPED1 runs before expt if possible. KBM stability studies with GK code or TGLF
Other Requirements:

Title	339: Can we prevent locked mode from being locked ?
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	J.S. Kim, M. Okabayashi, H. Reimerdes	ITPA Joint Experiment :	No
Description:	The proposal is to assess if a mode-locking is avoidable by providing electromagnetic torque in pre-locked stage.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In high beta, low torque plasmas, locked modes were frequently observed. However, to assess if locked modes are avoidable in pre-locked stage, low density plasmas below the ideal MHD no-wall stability limit would be primarily used, unless other plasma conditions (e.g. high beta, low torque plasmas) are identified to be more favorable to locked modes. To avoid locked modes from being locked, the following attempts will be made in pre-locked stage using; 1) FAR-TECH's model-based dynamic Kalman filtered algorithm 2) Complex gains to set the mode phase shifts 3) Rotating field that would interact with a pre-locked mode. Assuming that the attempts in Step 1) show positive progress in terms of plasma responses, a series of gain scans will be used to optimize the effectiveness of the applied field. If no satisfactory responses are obtained in Step 1), we will attempt Step 2) or 3) approaches to see if the locked mode can be diverted using complex gains or rotating fields.
Background:	During high beta, low torque experiments, locked modes frequently occurred, even after reasonably good error field correction was provided. Interestingly, the RWMID algorithm, which combines the matched and Kalman filters based on FAR-TECH's DIII-D/RWM model, showed that the growth of locked mode has remarkable similarity to that of typical n=1 RWM in pre-locked stage. Considering that locked modes cannot be easily unlocked and no control is taken to avoid locking, we will take advantage of the model-based RWM algorithm to suppress or divert the locked modes electromagnetically. Regarding complex gains, recent RWM experiments showed that the coil current demands were significantly reduced at a toroidally shifted angle from the measured mode phase which is typically attributable to unknown error field. Thus, using complex gains will rotate the pre-locked mode and then tackle the unknown error field that might be linked with the near-locked modes more directly. As for rotating fields, the pre-locked mode, which is not locked yet, would interact with the applied rotating field where the electromagnetically moving field would divert the locked stage. Compared with MHD spectroscopy, this approach of rotating field will be almost the same except the necessity that the applied magnetic field perturbations should affect the plasma motion.
Resource Requirements:	4 co-beams and 2-counter beams, 4 gyrotrons for ECCD
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	FAR-TECH's model-based RWMID algorithm Could be also allocated to error field group.

Title	340: Catastrophic MHD-affiliated non-axisymmetric fields
Name:	In inyongkyoon@unist.ac.kr	Affiliation:	Ulsan National Institute of Science and Technolog
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	M. Okabayashi, M. Schaffer, H. Reimerdes, E. Strait	ITPA Joint Experiment :	No
Description:	The proposal is to investigate the unknown inherent non-axisymmetric fields in DIII-D plasmas that would readily interact with other catastrophic MHD activity (e.g. RWM, ELMs, NTM or locked modes). If successful, we may be able to not only characterize these error fields but also help to establish the control algorithm to either mitigate or divert the catastrophic influences, though it would be ideal to eradicate catastrophic MHDs.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Identify a set of plasma conditions that would be influenced by the non-axisymmetric error fields (n = 1, 2 and 3). For example, three plasma sets would be listed; 1) high beta, high torque plasmas (for ELM-induced RWM), 2) high beta, low torque plasmas (for NTM or RWM-prone plasmas), 3) low density plasmas near locked mode condition Once the plasma conditions are repeatable, sweep the externally stimulated non-axisymmetric field (e.g. n= 1, 2, 3) toroidally to observe the plasma responses. Since the relevant waveforms might not be the same for different types of plasmas, it would be better to investigate each case (which is expected to require half-day) in separate run days. Regardless of the types of the runs, the unknown field that is inherently combining the vacuum and plasma non-axisymmetric field would be measurable. If a certain toroidal angle has stronger plasma responses for all the n spectra, the error field source might be from the interaction with external components (e.g. beam ports). If no preferable toroidal angle is found, it would suggest that no dominant non-axisymmetric fields are present other than the plasma responses.
Background:	Recent experiments showed that RWM could be triggered by ELMs in high torque plasmas, where the plasma rotation supposedly exceeded the RWM rotational thresholds. Although the mechanism needs to be thoroughly investigated, one could imagine that a wall stabilized RWM would interact with ELM-induced non-axisymmetric components and then the resonant field would be amplified enough to drag the plasma rotation below the effective rotation threshold, resulting in unstable RWM. Similar observations were made even with externally stimulated n=1 pulses, which are supposed to mimic the ELM-induced n=1 error fields. Interestingly, during a toroidal sweep of such external n=1 pulses, there was stronger plasma response in a certain toroidal angle than in the other toroidal angles. This indicates that even the unknown error field, which is likely to be a sum of all the residual error fields from machine and plasmas, would have a certain preferred toroidal angle. Considering that the goal of the error field study is ultimately relevant to understand the plasma responses in various plasma conditions, it would be more relevant and practical to apply various waveforms of externally controlled non-axisymmetri fields (n=1, 2, and 3) and then investigate the unknown fields.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	fully independent I-coil control is desirable

Title	341: n=1 braking in counter rotation
Name:	Reimerdes reimerdes@fusion.gat.com	Affiliation:	CRPP-EPFL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	A.M. Garofalo, E.J. Strait	ITPA Joint Experiment :	No
Description:	The experiment should verify the δBcrit â?? sqrt(TNBI+Tintrinsic) dependence found in co-rotating plasmas by extending the range of TNBI to negative (counter-IP) values, where the changes of the error field tolerance should be more dramatic. Confirming such a scaling would verify (1) the strong non-resonant contribution to the resonant braking and (2) link the error field threshold without external torque input to the intrinsic plasma rotation, which could have important consequences for ITER and the understanding of low density locked mode experiments. At counter rotation comparable to the neoclassical offset rotation for NTV torque (which has been neglected in the simple scaling of the error field tolerance [H. Reimerdes, et al., IAEA FEC 2008]) the non-resonant braking component is actually expected to increase the tolerance for resonant error fields.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We plan to apply a slowly rotating n=1 field (I-coils with 240Deg phasing) in SND counter-rotating plasmas, similar to the targets used in the 2007 (MP 2007-04-02/04) and 2008 (MP 2008-02-05) n=1 magnetic braking experiments. The value of βN has to be chosen to avoid the onset of a 2/1 NTM but still allow sufficient amplification to cause the rotation collapse. Since the beta dependence of the error field tolerance has been explained by amplification it is not necessary to carry out the TNBI scan at constant betaN.
Background:	n=1 braking experiment in 2007 and 2008 showed that the tolerable plasma response to externally applied n=1 fields is consistent with a δBcritâ??sqrt(TNBI+T0) scaling, where TNBI is the NBI torque and T0 an estimate of the intrinsic torque. Such a scaling is consistent with an error field threshold caused by joint resonant and non-resonant braking.
Resource Requirements:	Counter-IP and 5 NBI source including both 21 sources.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	342: Pedestal optimization with ELM suppression
Name:	Snyder snyderpb@ornl.gov	Affiliation:	ORNL
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	Burrell, Groebner, Leonard, Osborne	ITPA Joint Experiment :	No
Description:	Use improved understanding of the pedestal to optimize the pedestal height in ELM suppressed discharges. In a very strongly shaped plasma, follow a tailored trajectory through the pedestal j-alpha space to reach very high pedestals in the upper right corner.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In a strongly shaped plasma with ELMs suppressed (QH or possibly RMP), begin operation at low pedestal density until kink/peeling limit is contacted in the pedestal. Then slowly ramp density (vary rotation, pumping and/or particle source) up to larger values, attempting to achieve very high pedestals.
Background:	At strong shaping, a "nose" opens up in the upper right of the j-alpha peeling-ballooning stability diagram. At very strong shaping, this becomes a "beak" extending far out, and allowing very high pedestals to be stable. However, this region is not accessible at any particular fixed pedestal density. It can however be reached following a trajectory first up (low density) and then across (ramp density), provided large ELMs can be avoided. This could be combined with efforts to optimize performance in QH or RMP discharges.
Resource Requirements:
Diagnostic Requirements:	all pedestal diagnostics, both profile and turbulence
Analysis Requirements:	EPED1 and varyped studies before expt
Other Requirements:

Title	343: Validation of FIDA upgrade in quiet plasma
Name:	Muscatello muscatello@fusion.gat.com	Affiliation:	GA
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	C.M. Muscatello, W.W. Heidbrink	ITPA Joint Experiment :	No
Description:	A complementary upgrade to the existing FIDA diagnostic was installed during the 2008 vent. This 2nd generation system extends the diagnosticâ??s viewing capacity in both coordinate and velocity spaces. The new installment will also provide an additional level of background subtraction along with an extended instrument suite. Additionally, a simulation code that predicts classically the FIDA and NPA signals has recently been revamped with greater flexibility. It can now accept arbitrary viewing angles and injection beams. With these two major upgrades to the FIDA platform, a validation and cross-check study is desirable. Since the simulation code requires inputs from TRANSP and/or EFIT, classically behaved plasmas are most desirable. Ideally, data as obtained from FIDA in MHD quiescent plasmas would be compared to simulation results using parameters from the same plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Steady and moderate neutral beam injection along with toroidal fields of ~2T and plasma currents of ~1MA should be maintained in L-mode plasmas throughout the portion of the discharge where classical behavior is required. These need to be moderately low-density (<5*10^19 m^-3) plasmas with little to no Alfvenic activity. At first, constant beam power should be maintained for at least 50ms in order to populate the plasma with fast ions. After about 50ms, the fast-ion distribution will come to its asymptotic value. Following the fast-ion pumping, modulation of the 210RT source with a duty cycle somewhere within 50 â?? 75% for 0.5 â?? 1.0s will be needed in order to perform FIDA collection and background subtraction. During 210RT modulation, another source should modulate with the opposite duty cycle at the same power to maintain constant total beam power; the idea is to hold the fast-ion distribution function constant. At this point, there are several goals to achieve: Mostly importantly 1.) test 2G FIDA signal dependence on co/counter injection. Various combinations of co- and counter-injection will be used to test the sensitivity of the new FIDA views on different populations of fast ions. Three combinations should be employed: strongly counter-injection, strongly co-injection, and balanced injection. 2.) test FIDA dependence on injection energy. At least 2 portions of a discharge should be dedicated to 2 different injection energies while keeping the dominant injection angle constant (or better yet, 2 separate discharges maintaining 2 different but constant injection energies). Suggested energies would be 80keV and 60keV. 3.) test FIDA dependence on injection angle. At least 2 portions of a discharge should be dedicated to a dominant injection angle (strongly tangential and strongly perpendicular) while keeping the total beam power constant (more ideally, 2 separate discharges maintaining a dominant injection angle throughout). 4.) test FIDA dependence on electron temperature. Two scenarios: 4a.) Modulate ECH with 1 or 2 long enough pulses so that Te and, more importantly, FIDA reach and maintain their asymptotic values for about 100ms; the idea is to compare FIDA spectra at 2 different values of Te. 4b.) Modulate ECH with several moderate pulses to observe and compare FIDA and Te time evolutions. 5.) test FIDA dependence on electron density.
Background:	FIDA essentially measures the velocity component of a gyrating fast ion along the line of sight of the diagnostic. The new FIDA views are most sensitive to co-circulating particles. Goal #1 sets out to study just how sensitive the new system is to different populations of fast ions, which can be controlled with different directions of injection. Stronger perpendicular beam injection introduces fast ions with a higher perpendicular energy and stronger tangential injection introduces higher parallel energy. Goal #3 sets out to study the dependence of the FIDA signal on the injection angle. The dependencies of goals 2,4, and 5 can be understood based on a simple scaling expression; nf = fast-ion density, D(ne) = beam deposition rate, Pinj = total beam power, f(Te) = increasing function of Te, ne = electron density; the fast-ion density scales as nf ~ D(ne)*Pinj*f(Te)/ne. Goal 2.) higher beam power yields higher injection energy and therefore increased fast-ion signal. Goal 4.) higher electron temperature yields higher f(Te) and therefore increased fast-ion signal. Goal 5.) higher electron density decreases the beam deposition rate and increases the stopping power of injected neutral density and therefore decreases the fast-ion signal.
Resource Requirements:	At the very least, the counter-source 210RT beam and a co-source will be required. Ideally, 4 sources, including the 210RT source, would be desired for ease of flexibility of injection variation. Two gyrotrons for electron cyclotron heating would be desired.
Diagnostic Requirements:	The main diagnostic for our purpose is the FIDA upgrade (2G FIDA) with both s-FIDA and f-FIDA instruments employed to gather both spectral and accurate temporal data. Other necessary diagnostics would include neutral particle analyzers and neutron detectors/scintillators to complement and to provide a crosscheck to the time evolution of the FIDA signals. Inputs to the simulation code require electron density profiles, electron temperature profiles, ion temperature profiles, and carbon density profiles. For these measurements, the Thomson scattering, CO2 interferometry, ECE, and CER diagnostics will be employed. Additionally, magnetic fluctuation diagnostics will be used to monitor and verify quiet magnetic behavior, and MSE will be needed for accurate EFIT profiles.
Analysis Requirements:	Analysis will require EFIT profiles for the simulation code. Our FIDA simulation code will be used to verify our experimental results. Some GA computer time will be needed in order to run the FIDA simulation. Experimental FIDA profiles and spectra will be obtained using personal IDL codes residing on Hydra.
Other Requirements:

Title	344: Validation of Transport Elongation Scalings
Name:	Holland chholland@ucsd.edu	Affiliation:	UCSD
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	G. R. Tynan, G. R. McKee, M. W. Shafer, A. E. White, T. Rhodes, J. DeBoo, R. Prater, G. Staebler, J. Kinsey, R. E. Waltz, J. Candy	ITPA Joint Experiment :	No
Description:	Obtain comprehensive fluctuations measurements in steady, MHD-free L-mode plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Repeat 2008 elongation scan using what we learned about nessecary patch panel setup, but tailor heating at start-up to eliminate/delay sawteeth. Consider increase heating in second phase to investigate scalings in H-mode as well.
Background:	TGLF and GYRO predict a strong dependance on turbulent transport on elongation and elongation shear. A 2008 DIII-D experiment showed strong preliminary evidence supporting this prediction. However, these shots were also exhibited clear sawteeth, which makes them non-ideal for validating microturblence transport models. Goal is to repeat these experiments, with changes made to start-up to prevent or at least delay the onset of the sawteeth.
Resource Requirements:	Neutral beams, ECH
Diagnostic Requirements:	full profile and fluctuation diagnostic suites
Analysis Requirements:
Other Requirements:

Title	345: RWM stability above no-wall limit in AT plasmas with q~2 and near zero rotation
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	G. Matsunaga(JAEA), H. Reimerdes, M. Takechi(JAEA)	ITPA Joint Experiment :	Yes
Description:	ECCD-NTM suppression has been explored in FY2008. The extremely low rotation plasma stable for NTM and RWM was achieved. However, the safety factor was with relatively low q_min barely above one. This is mainly due to the fact that it was rather difficult to lower the rotation down during the ECCD. The confinement time drop during the ECCD is the main cause of the problem. The near-zero rotation condition was achieved well after ECCD off, at same time, q_min decreased. In the steady state AT plasmas, it is highly desirable to have the condition q_min â?¥ 2. Here, it is proposed to optimize the ECCD waveform in time or even turn off the ECCD for a short duration if the NTM suppression remains effective.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The reference shot is 132270 The ECCD power will be reduced after NTM suppression condition is achieved. The near-zero rotation target will be achieved before 2000 ms, where q_min can remain near two.
Background:
Resource Requirements:	NBI for achieving above no-wall limit ECCD-NTM suppression
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	346: Tungsten surface treatments for thermography
Name:	Lasnier Lasnier@fusion.gat.com	Affiliation:	LLNL
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	West, Rudakov, Wong	ITPA Joint Experiment :	No
Description:	Apply ELMing H-mode plasma to a DIMES sample containing tungsten buttons with different surface treatments. Observe the effect of plasma exposure on the emissivity.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Insert the DIMES sample and sweep the strike point across DIMES while in lower single-null ELming H-mode. Use tungsten buttons that are variously polished, sandblasted, knurled, and castellated.
Background:	The infrared emissivity of tungsten varies greatly depending on surface conditions. This is of concern for a high-power tokamak using tungsten in the divertor, because thermography is needed to monitor the surface temperature.
Resource Requirements:	ELMing H-mode, beams
Diagnostic Requirements:	DIMES, IRTV viewing DIMES at high resolution
Analysis Requirements:	Process IRTV measurements and compare with DIMES thermocouple measurements, and between different tungsten buttons in the sample which started with different surface treatments. Perform surface analysis on the exposed tungsten to quantify the effect on the differently prepared surfaces.
Other Requirements:

Title	347: Rotate RMP perturbation to spread the divertor heat flux
Name:	Lasnier Lasnier@fusion.gat.com	Affiliation:	LLNL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Boedo	ITPA Joint Experiment :	No
Description:	Apply a rotating RMP perturbation, see if the divertor heat flux footprint is broadened in as many toroidal locations as we can measure.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply a rotating RMP perturbation, in a plasma case where we normally see strike point splitting. Look for time-averaged broadening of the divertor heat flux profile.
Background:	Possibly an additional way of distributing heat flux.
Resource Requirements:	ELMing H-mode, strike points visible
Diagnostic Requirements:	SOL and divertor diagnostics
Analysis Requirements:
Other Requirements:

Title	348: Dependence of pedestal on heating method
Name:	Groebner groebner@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	A. Leonard, T. Osborne	ITPA Joint Experiment :	No
Description:	Determine if pedestal characteristics depend on the heating method. More specifically, determine if pedestal is different when a plasma with given parameters is produced with ECH as opposed to NBI.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Develop an H-mode plasma which reaches a steady-state ELMing operation with 2.5-3 MW of ECH heating power. Characterize pedestal parameters, including width, height, ELM size and ELM frequency. Then, use the same amount of NBI power only and inject into same plasma as used for ECH-heated discharges. Again, measure pedestal parameters. Determine if pedestal parameters vary with the heating method. The discharge conditions used in 2008 for the ITER baseline demo work might be good candidates for preforming this work. The current and field might need to be reduced to obtain betan=1.8. Ideally, the NBI part of these studies should be performed with both standard co-injection and with balanced beams to put in net zero torque.
Background:	The ITER team is exploring the options of running the ITER with different options for heating systems and to examine the possibility of operating with RF heating only. Given that most of the database used to design ITER has been based on NBI-heating, the team wants to know if the required performance in ITER can be obtained with RF-only heating schemes. Pedestal performance is one of the concerns along with concerns about core performance.
Resource Requirements:	DIII-D tokamak. LSN discharges in ITER shape. 5 ECH gyrotrons. 4-5 NBI sources including at least one counter source. Cryopumps.
Diagnostic Requirements:	TS, CER, CO2, bolometers, profile reflectometer, ECE, MSE, fast magnetics, divertor IR camera, photodiodes
Analysis Requirements:	Perform standard pedestal profile analysis to determine pedestal heights and widths. Determine ELM frequency and energy loss per ELM. Compare and contrast results for pedestals in ECH-heated and NBI-heated discharges.
Other Requirements:

Title	349: feedback from t=0: developing DEFC from t=0
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	Yongkyoon In, H. Reimerdes	ITPA Joint Experiment :	No
Description:	Current-driven RWM in FY2008 has provided reproducible RWMs and been useful to assess the RWM physics as well as the feedback performance. We also observed feedback response as soon as feedback was applied, the coil current conversing to finite level. This does occur during certain time period rather than continuously in time. This may indicate some type of DEFC takes place responding to some to-be-identified mode.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	350: Validation of R/L_Te Transport Scalings
Name:	Holland chholland@ucsd.edu	Affiliation:	UCSD
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	G. R. Tynan, J. Kinsey, J. DeBoo, R. Prater, T. Rhodes and TMV task force	ITPA Joint Experiment :	No
Description:	Test scaling of transport in ECH and ECH+NBI heated plasmas. Holding total injected ECH power constant, scan ECH deposition location, similar to recent JET and previous ASDEX/DIII-D experiments. Obtain multi-scale fluctuation data for validation tests.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use low and high elongation shapes from 2008 kappa-scan. Use a series of repeat discharges to scan ECH depositon location while maintaining total ECH power fixed. Use density control to prevent ECH pumpout. Will lead to variation in core R/L_Te, which can then be used to validate transport model predictions of that scaling. Repeat process in with 1 co-injected NBI beam, to validate scaling in presence of different levels of ITG transport.
Background:	R/L_Te is known to be a primary driver of TEM and ETG turbulence, which can play key roles in setting electron transport. Therefore validating model predictions for transport scalings with R/L_Te is critical. Propose to test this scaling in different shapes, and with different levels of R/L_Ti which will vary the amount of ITG modes present
Resource Requirements:	NBI, ECH
Diagnostic Requirements:	full profile and fluctuation measurements
Analysis Requirements:	Would be interesting to do a set predicitve thought experiments where we predicted profiles that would be obtained by varying ECH deposition location before experiment, and then testing that result.
Other Requirements:	--

Title	351: Rotation dependence of the LH power threshold in ITER & C-MOD like configurations
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	J. Boedo, P. Gohil, R. Moyer, S. Mueller, D. Rudakov, D. Schlossberg, G. Wang, Z. Yan, J. Rice, B. LaBombard	ITPA Joint Experiment :	No
Description:	Determine whether the toroidal-rotation-dependence of the LH power-threshold is the same in plasmas with an ITER-like plasma configuration (plasma current and field parallel), as in the DIII-D configuration (plasma current and field anti-parallel). This will be important to understanding how this behavior is predicted to extrapolate to ITER and other experiments.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Operate with counter-Ip and "normal" BT, and perform an NBI torque scan. Determine the LH power threshold at each condition, as was done for previous experiments by Schlossberg and Gohil. This should be performed in LSN since this is the ITER-like configuration, but for completeness, should also be done in USN, since the rotation-threshold dependence is stronger. Fluctuation diagnostics should be used to examine edge turbulence flows and dynamics.
Background:	Experiments on DIII-D have demonstrated that the LH power threshold depends strongly on the toroidal rotation in plasmas with the ion grad-B drift directed towards and away from the dominant X-point (McKee-APS-2007, IAEA-2008). In particular, plasmas with toroidal rotation in the counter-current direction have been shown to have a lower LH power threshold than those rotating in the co-current direction. On DIII-D, the power threshold varies by a factor of 3-4 in USN plasmas (grad-B away from X-point), and about a factor of 2 in LSN plasma as the torque is varied from -1 to +4 N-m. This effect has been connected to the edge turbulence dynamics which show a consistent effect. C-MOD has shown a related dependence of the LH power threshold on toroidal rotation, though its dependence appears to be different than that observed in DIII-D (LaBombard, Phys. Plasmas, (2005)). C-MOD shows a similar dependence of the LH threshold on USN vs. LSN plasmas, and has also shown substantial variation in the scrape-off-layer flows with these different configurations. In contrast to DIII-D plasmas, the C-MOD experiments show that plasmas are more likely to transition into H-mode when the core toroidal rotation is in the co-current direction. A major difference in these experiments is that rotation is externally driven via NBI torque on DIII-D, while no external torque is applied to the C-MOD plasmas, so the toroidal rotation is strictly intrinsic. Another important difference between C-MOD and DIII-D is that the ohmic plasma current and toroidal field are parallel in C-MOD while they are anti-parallel for typical DIII-D plasmas (including the experiments demonstrating this rotation dependence). ITER will have a parallel Ip-Bt configuration like C-MOD, and the LH power threshold in ITER is highly uncertain. Because of the important of the ion grad-B drift direction in LH transition physics, it is important to determine whether the rotation-threshold relation depends on the parallel/anti-parallel, i.e., whether it is the same as observed in DIII-D (counter rotation = lower threshold) or perhaps the reverse (counter rotation = higher threshold). Given the importance and the uncertainty of the ITER LH power threshold, and the varied sources of toroidal rotation (intrinsic, NRMF, NBI, LHCD), it is necessary to determine this relation in the ITER (C-MOD like) configuration.
Resource Requirements:	Reversed Ip, standard BT
Diagnostic Requirements:	Fluctuation diagnostics (BES, reciprocating probe in particular)
Analysis Requirements:	--
Other Requirements:	--

Title	352: Particle Exhaust from RMP ELM suppressed LSN discharges vs X-point height
Name:	Unterberg unterbergea@ornl.gov	Affiliation:	ORNL
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Recent analysis has shown a clear difference in the neutral particle exhaust to the lower cryopump and D_alpha recycling light characteristics during the application of the RMP (e.g. shot 122489 vs 134162). This experiment will test in a high-trig. RMP ELM suppressed discharge (like 134162) the dependence on particle exhaust to the lower baffle to the height of the x-point. This is will be done by going from the typical ISS (closed) divertor geometry to a more open configuration by lowering the x-point.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Start with a shot like 134162 and vary the x-point height in steps of 1-2 cm until the change in height ~ - 11cm (almost to the divertor floor). Since ELM suppression depends sensitively on plasma shape, the could be a need for fine-tuning the shape as the x-point height is varied. The x-point scan might be most effective if the changes are done in time-delayed discrete steps to allow transient effects to equilibrate (e.g. changes in PFR drifts, changes in power and particle flux to the target due to the height change, etc.). Also, probably would want to do both a Zxpt only and Rxpt & Zxpt scan. Finally, this experiment is proposed at low density to keep the divertor radiation low.
Background:	Even if the density pumpout is reduced during RMP discharges, there is still a need to direct any particle exhaust to the divertor and cyro-pumps as opposed to the main chamber walls. In experiments on DIII-DB, it was shown that the cyro exhaust increase by 75% and the poloidally averaged D_alpha light increases by ~ 75-100%. There is an open question of whether these changes are due to 1) the change in particle transport due to the change in plasma shape or 2) a change in the lower divertor configuration (i.e. A more open case in 122489 vs more closed case in 134162). This should decrease the flux expansion in the throat of the divertor (the distance form the SOL to the edge of the graphite shelf) and allow more or less neutral leakage into and out of the divertor. This idea is similar to the optimization criteria of 2-3 lambda_n (midplane SOL density width) for divertor baffles (see Loarte, PPCF 2001). This experiment would/could compliment Petrie's ROF09 Prop. 146 & 147 on radiative divertor RMP discharges.
Resource Requirements:	1/2 run day needed
Diagnostic Requirements:	Usual RMP edge diagnostics.
Analysis Requirements:	particle balance calculations. EIRENE or DEGAS2 modeling. pre-run analysis: mag. topology and lamda_n determination for zxpt1 height planning.
Other Requirements:	--

Title	353: Measurement of error field penetration/shielding with slowly rotating n=2 fields
Name:	Reimerdes reimerdes@fusion.gat.com	Affiliation:	CRPP-EPFL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	This experiment aims at a direct measurement of the shielding or penetration of external resonant fields by observing the characteristic flattening of flux surface quantities across magnetic islands using various profile diagnostics.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply slowly rotating (~10Hz) n=2 fields to measure signatures of magnetic islands using profile diagnostics (e.g. ECE radiometer, reflectometer, BES). The rotation and value of q95 is varied in order to evaluate the dependence of the shielding on plasma rotation and magnetic shear.
Background:	RWM and ELM suppression experiments involving locked modes or using static external non-axisymmetric fields have had the difficulty to investigate the role of magnetic islands because profile diagnostic can only see the characteristic flattening of flux surface quantities across magnetic island when the o-point of the island is close to toroidal location of the diagnostics. This restriction can be avoided by rotating the islands past the diagnostics.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	354: Kink mode-resonant windows in q95 for n=3 fields with even and odd parity
Name:	Reimerdes reimerdes@fusion.gat.com	Affiliation:	CRPP-EPFL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Measure the relation between external resonant fields and plasma response for even and odd n=3 I-coil parity.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply slowly oscillating (10Hz) n=3 fields with even and odd I-coil parity in rotating plasmas with various values of q95.
Background:	Experiments applying n=1 fields with various poloidal spectra in high beta plasmas showed that the amplitude of the plasma response does not correlate with the resonant component of the external field. The IPEC model predict that the total resonant field (including fields from perturbed plasma currents and in external non-axisymmetric coils) correlates with the plasma response.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	355: Target plate profiles during ELM suppression
Name:	Watkins watkins@fusion.gat.com	Affiliation:	Sandia National Lab
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	Evans, RMP ELM suppression group	ITPA Joint Experiment :	No
Description:	Measure target plate profiles during RMP induced ELM suppression	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Develop an ELM suppressed discharge with the outer strike point on the shelf such that we can more fully utilize our target plate diagnostics such as the IR camera, divertor Thomson, Langmuir probes, etc. The configuration would require inner strike point pumping from the private region and/or SOL pumping from the upper pump.
Background:	ITER needs to know what the target plate plasma looks like in ELM suppressed discharges. We have a few shots with sweeps inward to allow views of the outer strike point but we need much better measurements without blocked views. The shelf itself is a large perturbation in the SOL behavior. ITER will be pumped from the private region and this experiment would be relevant to that configuration as well. ITER boundary personnel have repeatedly requested more detailed measurements at the target plate in ELM suppressed conditions (IAEA, High heat flux component workshop).
Resource Requirements:	icoils, 4 beams, cryopumps
Diagnostic Requirements:	IR camera, DIMES camera, Langmuir probes, thermocouples
Analysis Requirements:	--
Other Requirements:	--

Title	356: Type III ELM heat flux profile and scaling
Name:	Lasnier Lasnier@fusion.gat.com	Affiliation:	LLNL
Research Area:	Thermal Transport in the Boundary (2010)	Presentation time:	Not requested
Co-Author(s):	Leonard	ITPA Joint Experiment :	No
Description:	Characterize TYPE III ELM heat flux profiles, find how the peak and profile shape scale with plasma parameters	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Make Type III ELMs, vary the plasma parameters as widely as possible, characterize with SOL and divertor diagnostics including new IR camera and divertor Thomson.
Background:
Resource Requirements:
Diagnostic Requirements:	SOL and divertor diagnostics
Analysis Requirements:	Compare with measurements of Type I ELMs.
Other Requirements:

Title	357: Effect if RMP strength on pedestal height
Name:	Groebner groebner@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	A. Leonard, T. Osborne, P. Snyder	ITPA Joint Experiment :	No
Description:	Determine effect of RMP on pedestal height. In particular, characterize reduction of pedestal height as function of applied RMP current.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Develop standard ELM-suppressed discharge with application of RMP. Characterize pedestal height and width under this condition. Then, perform scan of RMP current from low to high and determine pedestal height in all ELM-suppressed regimes. Compare pedestal height to ELMing plasma in which RMP is not applied.
Background:	ITER team has many questions about the use of RMP fields to eliminate ELMs in ITER. One issue is how much the pedestal height (and confinement) reduced by application of RMP fields. We propose to clarify this physics through a dedicated scan of RMP current. We anticipate that the decrease of pedestal height increases as RMP current is increased, but this needs to be tested and documented.
Resource Requirements:	DIII-D tokamak. 5-6 NBI sources. I-coil. cryopumps
Diagnostic Requirements:	TS, CER, CO2, bolometers, profile reflectometer, ECE, MSE, fast magnetics, divertor IR camera, photodiodes
Analysis Requirements:	Perform standard analysis of pedestal profiles to determine heights and widths. Determine the change in these parameters as a function of I-coil current.
Other Requirements:

Title	358: burning plasma simulator with control via pedestal
Name:	Snyder snyderpb@ornl.gov	Affiliation:	ORNL
Research Area:	Integrated and Model-Based Control	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Use the control system to run DIII-D as a model burning plasma, where input power is proportional to a calculation of the scaled fusion power for given conditions. Model high Q operation where it becomes challenging to avoid global beta limits via control of limited external heating. Attempt to control such plasmas via small shape changes (eg squareness) which alter the pedestal and therefore core confinement	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Set up control system to imitate a burning plasma, with input power proportional to of the plasma (plus small "external source"). Operate in regime where a hard beta limit will be encountered in the absence of control, and where changes in the "external" heating power are insufficient to control the "high Q" plasma. Attempt to use shape control as an independent lever to control global beta (via the pedestal) in such plasmas. Produce high performance, strongly shaped H-mode plasmas, in which th
Background:	The pedestal height strongly impacts global confinement. Because the pedestal is constrained by sensitive peeling-ballooning instabilities, it can be strongly impacted by, for example, small changes in plasma shape. A burning plasma at high Q is difficult to control via modification of the relatively small external power source. In many cases such plasmas are predicted to run up to global beta limits in the absence of strong control. We propose to use shape changes, for example squareness increases, that may be feasible in ITER, to lower the pedestal, reduce core confinement and beta even though the plasma is alpha-heating dominated.
Resource Requirements:	full set of neutral beams, modified control system
Diagnostic Requirements:	full pedestal diagnostic set
Analysis Requirements:	analysis of global and pedestal stability prior to expt
Other Requirements:

Title	359: Critical nu* for Density Pumpout in ITER
Name:	Joseph joseph5@llnl.gov	Affiliation:	LLNL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Andrea Garofalo, Jong-Kyu Park, and Holger Reimerdes	ITPA Joint Experiment :	No
Description:	Determine the phenomenological scaling of the nonambipolar particle flux caused by resonant magnetic perturbations with beta, nu*, and omega_ExB. Determine the critical nu* and bfield required to achieve density pumpout.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Power scan at constant rotation to vary beta and nu* and rotation scan at constant power to vary omega_ExB. Reference discharges chosen from neoclassical torque studies: â??Resonant vs. nonresonant brakingâ?? (D3DMP #2007-04-02 and #2007-04-04 by Garofalo et al.).
Background:	The neoclassical theory of nonambipolar transport due to nonaxisymmetric magnetic perturbations leads to definite scalings regarding the magnitude of the particle flux and associated torque on the plasma. However, the direct application of this theory to experiment has lead to a number of suprises regarding the magnitude of the predicted flux. Kinetic modeling by J.-K. Park, et al (submitted to PRL) including both resonant field amplification and all drift and bounce orbit resonances indicated roughly 1/nu_* scaling, but with a much smaller pre-factor than previously assumed and a complex spatial structure that peaks near the pedestal. The theory leads to definite, testable scalings of the particle flux with beta, nu* and omega_ExB. It predicts that a critical RFA-amplified magnetic perturbation strength is needed to achieve an observable density pumpout, and that the pump-out will only occur with co-current plasma rotation. This theory needs to be tested in order to provide input for the design of magnetic perturbation coils on ITER.
Resource Requirements:	I-coil, C-coil
Diagnostic Requirements:	CER
Analysis Requirements:	Kinetic equilibrium reconstruction for accurate rotation profile reconstruction and stability modeling.
Other Requirements:

Title	360: The RMP plasma edge and impurity control
Name:	Watkins watkins@fusion.gat.com	Affiliation:	Sandia National Lab
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	Evans, Groebner, Yu	ITPA Joint Experiment :	No
Description:	Using the RMP coils, demonstrate that impurities can be flushed out of the core plasma similarly to what happens with ELMs.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Using diagnostic Neon and/or Argon puffs before and after energizing the coil, show that the impurity confinement time is at the ELMing level or below in the ELM suppressed plasma. Measure the impurity profiles with CER. Utilize the fast UCSD camera (with filters if possible) to try and image gas puffs to see magnetic field structures associated with the RMP.
Background:	ELMs play an important role in preventing the buildup of impurities in a high confinement core plasma. Using the RMP, we can suppress the ELMs but we need to demonstrate that we can also provide the impurity removal role that the ELMs played. Evidently ASDEX upgrade is planning this experiment when their coils are ready in the next year or two.
Resource Requirements:	4 beams, lower cryopumps, I coils, neon gas puff
Diagnostic Requirements:	CER with neon filters, fast camera, spred
Analysis Requirements:	CER profiles, produce fast camera images correlated with puff times
Other Requirements:

Title	361: Compatibility of RMP ELM suppression with large NRMF torque
Name:	Garofalo garofalo@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	Solomon, Schmitz	ITPA Joint Experiment :	No
Description:	This experiment would investigate RMP ELM suppression in conditions of NBI injected torque much smaller than the torque from nonresonant magnetic fields (NRMF) associated with the RMPs. In such conditions, the NRMF torque may drive the plasma rotation from an initial co-Ip direction down to zero and then back up in magnitude, but in the counter-Ip direction. This is likely the case we will face with RMP ELM suppression in ITER. The purpose is to study what happens to the ELMs and to the plasma as the rotation changes in this way.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The starting point for this experiment are the ITER-similar shape, ITER-similar collisionality plasmas from D3DMP #2008-03-13 "RMP ELM suppression at low torque", by Evans, et al. We will use additional counter NBI to reduce the net injected torque at high beta. We will develop optimal error field correction using the C-coil controlled by the RWM feedback system. This error correction scheme has allowed us to reach very low rotation at high beta in RWM experiments. We will use the I-coil to apply even parity and odd parity n=3 fields, using up to 7 kA of current.
Background:	RMPs for ELM suppression in ITER have an associated NRMF torque that is predicted to be 40 to 400 times larger than the expected NBI torque. In such conditions, the NRMF torque may drive the plasma rotation from an initial co-Ip direction down to zero and then back up in magnitude, but in the counter-Ip direction. What will happen to the ELMs as the rotation changes in this way? Can the discharge survive with large applied RMPs as the rotation becomes zero? DIII-D can reproduce the ITER condition of very large NRMF torque compared to the NBI torque, in an ITER-similar shape and at ITER-similar collisionality.
Resource Requirements:	Same as for discharge 133915: Iter-similar shape, Normal Ip= 1.3 MA, Normal Bt=-1.7 T. SPAs for dynamic error field correction using the C-coil. C-supplies for 7 kA n=3 I-coil currents
Diagnostic Requirements:	Magnetics, Thomson, MSE, CER, bolometry, photodiodes, ECE, etc.
Analysis Requirements:	--
Other Requirements:	--

Title	362: Target Plate Heat flux, sheath power transmission factor, and Power Accounting
Name:	Watkins watkins@fusion.gat.com	Affiliation:	Sandia National Lab
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Lasnier,Schaffer, Murphy, Leonard, Evans, West, Stangeby	ITPA Joint Experiment :	No
Description:	To validate our measurements of heat flux, measure the target plate heat flux profile using several different methods and do a detailed power accounting of DIII-D plasmas at high and low density and for different power levels in both L and H mode and ELM - suppressed H mode.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	To better measure the target plate heat flux, the outer strike point must be run on the outer shelf in the lower divertor to get good views with the IRTV. Sweeps should be used to get better profiles with the Langmuir probes, and the calorimeter probe. Fixed strike point shots will be needed to simplify analysis of the thermocouple array and calorimeter probe data. The bolometer array will be used to measure the radiated power and generate a total power accounting. The inner strike point should be run on the lower divertor floor in view of the IRTV, Langmuir probes, and the floor TC array.
Background:	Experiments in heat flux reduction and Elm suppression require good profiles of heat flux to document and scale the expected heat flux to ITER conditions. During DIII-D recent campaigns, power accounting has not been as good as in the past. In high density shots, where most of the power goes into radiaiton, the power accounting is near 100%. In low density shots, about 50% of the power is missing. This is consistent with lower than expected heat flux levels at the outer strike point. Heat flux calculated from the thermocouples embedded in the tiles for these shots is ~ 5X higher. Heat flux measured with the calorimeter probe was about 2x larger than the IRTV for low powered shots. By using the new thermocouple array as an independent measurement of the heat flux, we hope to verify that the IRTV is working properly and extend the heat flux measurements into areas that the IRTV cannot see (baffle entrance).Using the Langmuir probes will allow us to directly measure the sheath power transmission factor with the new Langmuir probe array and calculate heat flux from previous profiles and in any location with a Langmuir probe.
Resource Requirements:	4 beams
Diagnostic Requirements:	IRTV, fast thermocouples, lower Langmuir probes, Thomson, x-point and midplane reciprocating probes, bolometers
Analysis Requirements:
Other Requirements:

Title	363: Sheath Factor measurement in unbalanced Double Null
Name:	Watkins watkins@fusion.gat.com	Affiliation:	Sandia National Lab
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Lasnier	ITPA Joint Experiment :	No
Description:	Test the sheath power transmission theory using a double null plasma. Vary the magnetic balance slightly around the balance point to shift the fast electron heat flux at the strike point from the lower divertor to the upper divertor while measuring the heat and particle flux to the strike point.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Control fast electron content of lower divertor plasma with DN balance. Control collisionality with density. Measure heat flux with IR camera. Measure particle flux with Langmuir probes near strike point. Measure Te with Langmuir probes and Divertor Thomson near strike point. Plot heat vs particle flux*Te for different drsep and density values.
Background:	The heat flux balance varies very strongly with drsep near the balance point of a DN plasma. The particle flux does not vary as quickly. This difference in drsep dependence allows us to study the two quantities independently.
Resource Requirements:	cryo pumps, 4 beams
Diagnostic Requirements:	IRTV, Floor probes, midplane probe, xpt probe
Analysis Requirements:
Other Requirements:

Title	364: Secondary divertor and SOL
Name:	Watkins watkins@fusion.gat.com	Affiliation:	Sandia National Lab
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Lasnier, Stangeby, Rudakov	ITPA Joint Experiment :	No
Description:	Study of the secondary divertor and SOL is relevant to elongated plasmas like ITER. If there is not enough armor at the secondary divertor, the wall fluxes of particles and power may lead to problems at the first wall. Outer wall fluxes are coming from the secondary SOL and may depend on magnetic balance through instabilities at the primary/secondary SOL interface and secondary divertor recycling.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Setup an unbalanced DN such that the x-point probe has access to the primary and secondary SOL. Both USN and LSN should be studied with the same Bt direction.
Background:	An unbalanced DN can have a secondary x-point inside the vacuum vessel and a secondary divertor with significant fluxes of particles and energy. This can have an impact on high power machines like ITER if the secondary divertor is not armored sufficiently. Also, near the magnetic balance point (-1 1)The H mode power threshold varies strongly with drsep. Both DIII-D, ASDEX, as well as MAST have observed this in DN. In MAST the H mode can only be achieved in a narrow range around the balance point. 2) Strong changes in the ELM behavior and the core density have been observed to depend on the magnetic balance in this range. The strongest behavioral changes occur in the transition region between +1 and -1 in the drsep parameter. The structure of this type of SOL has potentially important properties that have never been studied in detail before.
Resource Requirements:	DN patch panel
Diagnostic Requirements:	mid plane probe, xpt probe, IRTV, floor probes, filterscopes, visible cameras
Analysis Requirements:	--
Other Requirements:	--

Title	365: Suppression of first ELM following the L-H transition
Name:	Evans evans@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	The goal of this experiment is to suppress the first ELM following the L-H transition by applying n=3 RMPs from the I-coil and C-coil starting in the L-mode phase without increasing the LH power threshold (or possibly reducing it). In ITER the first ELM must be suppressed to prevent premature eroding of the divertor target plates. Since the L-H power threshold is expected to be close to the additional heating power available in ITER, if the RMP increases the LH threshold it may not be possible to access H-mode operations. An ideal solution would be to reduce the LH power threshold with the RMP followed by full ELM suppression after the usual ELM-free density increase phase. This may be possible by combining RMPs for both the I-coil and C-coil. Since the q95 resonant window for ELM suppression must be obtained and held constant during the L-mode and L-H transition it will be necessary to establish a 1 s stationary L-mode with q95 =3.6 before increasing the NBI power to trigger the L-H transition.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Using a standard ISS RMP H-mode shape (ref 129750) with Bt = 2.15 T and grad_Bt away from the x-point, proceed as follows: Step 1 - establish low power L-mode plasmas using 10 on 10 off 330L beam modulations with n=3 RMPs from both the I-coil and the C-coil. Use the edge fluctuation diagnostics to characterize the plasma with and without the RMPs. Do an upward RMP amplitude scan starting with an I-coil current of 4.5 kA and a C-coil current of 2 kA. Step 2. Set the RMP perturbation parameters for the best value obtained in step 1 (field amplitude and relative toroidal phase of the two coils) and signatures in the fluctuations, Te and ne profiles, divertor recycling patterns, divertor heat flux patterns). Starting with a 10 on 10 off 330L beam, increase the NBI power by adding the 330R beam. If an L-H transition os obtained proceed to step 3 if not add the 150L beam on the next shot. Continue to add one beam at a time until an L-H transition is achieved and then proceed to step 3. Step 3. Using the parameters for Step 2, reduce the NBI power by reducing the voltage on one of the beams in steps of 2 kV until the L-H transition is not obtained (e.g. for the 330R beam this gives 2.412 MW at 75 kV, 2.291 MW at 73 kV, 2.171 MW at 71 kV, etc.). When the NBI power drops below the LH threshold increase the power by increasing the duty cycle of the 330L beam in small steps until the LH transition returns (this could also be achieved by adding ECH instead of increasing the duty cycle of the 330L beam). Step 4. Increase the RMP amplitude in small steps from shot-to-shot to see if either the properties of the LH transition change or the ELMs are suppressed (including the first ELM). Step 5. Document the best case that has full ELM suppression (including the first ELM).
Background:	Experiments done in 2007 demonstrated that the density could be controlled during the ELM-free phase in ISS plasmas by turning on the n=3 I-coil perturbation immediately after the L-H transition in discharges with Pinj = 7.4 MW (ref. 129750). In this discharge the L-H transition was triggered by stepping up Pinj from 2.6 MW to 7.4 MW at 1100 ms and turning on the I-coil (4.2 kA) at 1220 ms and didn't enter the q95 resonant window for ELM suppression until 1315 ms. Nevertheless, we observed a change in the ELM-free density rise and in the properties of the first ELM. The hypothesis is that if we tune the front end of the discharge to get a steady L-mode at q95 = 3.6 we will get better ELM suppression. At this point we can look for changes in the LH power threshold due to the RMP.
Resource Requirements:
Diagnostic Requirements:	Full RMP H-mode diagnostics including edge fluctuations and L-H transitions diagnostics. High time resolution CER data for studying changes in the carbon rotation and impurities across the L-H transition and during the ELM-free phase, fast IR camera, fast CCD imaging.
Analysis Requirements:	TRIP3D, profile analysis, fluctuations, flows and divertor heat flux.
Other Requirements:

Title	366: Target plate Vf evolution as an indicator of ELM suppression
Name:	Watkins watkins@fusion.gat.com	Affiliation:	Sandia National Lab
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Not requested
Co-Author(s):	Evans	ITPA Joint Experiment :	No
Description:	This relationship of large negative Vf values at the target plate strike point can be used to study the transition to ELM suppressed operation.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Place the strike point just inside one of the Langmuir probe tips and push the plasma across one of the thresholds to ELM suppression. Thresholds that can be studied with this technique are: I-coil current, density, the q95 window, power level, rotation (using counter beams), and toroidal phase.
Background:	There are several parameters that exhibit threshold behavior at the transition to ELM suppressed operation using RMP. We see the density has an upper limit around n= 0.3ng. The edge q must be within a certain range of values (depending on I-coil current) to get ELM suppression. Conversely, the I-coil current threshold depends on q95 and plasma rotation. The target plate Vf near the strike point has been observed to drop to large negative values as the ELM suppression threshold is crossed. These large negative values occur at points along the target plate that are predicted (TRIP3D) to connect to points deep (rho=0.9) within the core plasma. The large negative Vf indicates an upstream electric potential change which we think is related to the formation of an electric field structure around the islands that is driving ExB radial transport related to the ELM suppression. Since the Vf signal is sensitive to the onset of ELM suppressed operation, we would like to use it as a tool to study the transition.
Resource Requirements:	4 co and 2 counter beams
Diagnostic Requirements:	Langmuir probes
Analysis Requirements:	I/V analysis
Other Requirements:	good strike point control

Title	367: Electron heat transport in 3rd harmonic-ECH-heated hybrid plasmas
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	E.D. Doyle, T.L. Rhodes, W.A. Peebles, C.C. Petty, G.R. McKee, G. Wang	ITPA Joint Experiment :	No
Description:	Investigate electron heat transport scaling with Te/Ti and underlying intermediate/small scale turbulence in 3rd harmonic ECH-heated Hybrid plasmas. Identify optimum conditions for increasing Te/Ti ratio	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Based on previous 3rd harmonic ECH Hybrid discharges (129282) reduce NBI heating and increase ECH in steps to vary Te/Ti ratio. Strongly co-rotating plasmas will be used initially to obtain suppression of long wavelength (ITG) turbulence and achieve high confinement. This will allow us to achieve the maximum possible increase in Te/Ti. Strong rotation also simplifies Doppler Backscattering (DBS) measurements due to the larger ExB Doppler shift. 1) Obtain core fluctuation levels (k rho_s >2) vs. Te/Ti using Doppler Backscattering and high-k microwave backscattering. The 3rd harmonic hybrid scenario with Bt=1.3 T is well suited for Doppler Backscattering access, and core fluctuations in the wavenumber range 2 < k rho_s < 5 (for r/a >0.5) can be measured. DBS can also follow changes of the fluctuation level and ExB poloidal flow throughout the ELM cycle. 2)Vary beam momentum input at constant power to investigate scaling of intermediate/small scale turbulence vs. toroidal momentum input.
Background:	Hybrid discharges are very attractive for burning plasma regimes. Scaling hybrid performance to a regime with Te/Ti > 1 is therefore important, as is the understanding of confinement changes as Te/Ti is increased. The achievable Te/Ti ratio may be linked to understanding electron heat transport in these plasmas. approach burning plasma conditions Recent gyrokinetic calculations indicate that electron transport in hybrid plasmas in dominated by high-k turbulence (J. Kinsey, PoP 2008). Preliminary measurements of smaller-scale fluctuation levels have been made in DIII-D by Doppler backscattering, These measurements have indicated increased TEM/ETG-scale fluctuation levels and reduced confinement during ECH.
Resource Requirements:	7 beams, 6 gyrotrons
Diagnostic Requirements:	DBS, high/intermediate-k scattering, FIR scattering, BES, MSE, PCI
Analysis Requirements:	TGLF/XPTOR runs for preparation/optimization of experiment and for data analysis
Other Requirements:	--

Title	368: Dynamic D retention studies
Name:	Pigarov none	Affiliation:	UCSD
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	W.P. West	ITPA Joint Experiment :	No
Description:	Wall pumping is transitional effect. In long pulse tokamaks, when wall saturates it becomes a dominant source of gas resulting in uncontrollable density raise, MARFE, and degradation of plasma confinement. So far only STATIC (very long-term, a campaign) retention of D was measured on tokamaks (e.g. on DIII-D showing [D]/[c]~0.1 of strongly bonded D). DYNAMIC retention on short time scales ~1s is practically unexplored and the concentrations of weakly bonded, mobile D in wall are not known. However, it is DYNAMIC retention that is responsible for transition from wall pumping to outgassing and that can affect the plasma performance and, say, generation of MARFE. Notice, dynamic retention is an important issue for plasma performance in ITER, whereas static rention is more the safety related issue. In early experiments on wall pumping, for example, [S.A. Cohen, PPCF 29 (1987) 1205] on JET when plasma colomn was positioned closer to inner wall the plasma density decreases on a second time scale, whereas plasma positioning closer to outer wall recovered the density level, thus highlighting the importance of dynamic retention. DIII-D capabilities are suitable for more detailed studies of this effect. The featured experiments will provide input to and will be modeled with WallPSI code for hydrogen transport and retention in wall materials.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	several consequitive shots, probably L-mode
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:	WALLPSI+UEDGE
Other Requirements:

Title	369: Investigations of ITER-like castellation: castellation shaping to reduce fuel inventory in the gaps
Name:	Litnovsky a.litnovsky@fz-juelich.de	Affiliation:	Juelich
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	D. Rudakov (UCSD), V. Philipps (FZJ), C. Wong (GA), W. West (GA), R. Boivin(GA), N. Brooks (GA), P. Wienhold (FZJ), O. Schmitz (FZJ), R. Bastasz (SNL), J. Whaley (SNL), W. Wampler (SNL), J. Watkins (SNL), J. Brooks (ANL), T. Evans (GA), D. Whyte (UW), P. Stangeby (Univ. of Toronto), A.Mclean (Univ. of Toronto), J. Boedo (UCSD), R. Moyer (UCSD)	ITPA Joint Experiment :	Yes
Description:	The aim of this experiment is to mitigate carbon transport and fuel deposition in the gaps of ITER-like castellated structures. Castellation cells having special roof-like shape will be used in this experiment. Shaping of castellation cells should provide significant difficulties for impurities and fuel particles to penetrate and accumulate inside the gaps.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Previous experiments performed in DIII-D in 2005 with heated gap samples demonstrated the positive effect of reduction of carbon deposition and corresponding fuel inventory in the gaps at elevated temperatures. As a next step the role of shaping of the castellation in reduction of impurity and fuel transport into gaps was studied in 2008 using specially prepared castellated cells installed on DiMES transport system. The dedicated experiment had indeed demonstrated the positive effect of castellation shaping. However, because of limited exposure time, the detailed characterization and quantification of the deposition in the gaps is accompanied with significant difficulties and uncertainties. It is planned to elaborate the experiment with castellation shaping by making the long-term piggyback exposure using DiMES system. Castellation samples of different shapes will be exposed simultaneously with ones having the conventional (rectangular) cells. Tungsten castellation will be used for the experiment, since tungsten is planned to be used in ITER divertor. Piggyback exposure during approximately 1 week of plasma operations, preferably in LSN configuration, is requested for castellation to obtain the representative deposition patterns in the gaps.
Background:	In ITER, the castellated armor of the first wall and divertor will be used to maintain the durability of the machine under the thermal excursions during plasma operation. There are concerns about the impurity deposition and fuel accumulation in the gaps of castellated structures, representing safety issue for ITER operation. Past and present research demonstrated that the fuel inventory in the gaps of castellated structures is significant and there are essential difficulties in fuel removal. To address this problem, dedicated investigations are ongoing on several tokamaks worldwide: DIII-D, TEXTOR, Tore Supra, MAST and JET. Investigations on fuel retention in the gaps of castellated structures are presently recognized as an important activity of the ITPA Topical Group on Divertor and SOL and are the subject of IEA-ITPA Joint Experiments Program (Task DSOL 13). The optimization of the shape of castellation is the relatively natural way towards the minimization of the fuel retention in gaps. This investigation was already started at DIII-D, the aim of the present exposure is to obtain the deposition pattern allowing robust, direct and precise interpretation of results. Understanding of the physical processes behind the transport into gaps is of significant importance. Flexible design allows for a direct comparison of conventional and optimized shaping within the same experiment. Simultaneously there is a possibility to study both poloidal and toroidal gaps under ITER-relevant conditions. An essential advantage of this experiment is that the exposure of castellated samples will be performed at shallow angle with respect to magnetic field, similarly as expected in ITER.
Resource Requirements:
Diagnostic Requirements:	DiMES TV, floor Langmuir probes, in particular the probe at the DiMES radial location, MDS chord looking at DiMES.
Analysis Requirements:	Assistance in analysis of data from edge diagnostics is desirable
Other Requirements:	One week piggyback exposure using DiMES manipulator system, preferably LSN operation, NBI-heated ELMy H-mode.

Title	370: Mirror tests for ITER diagnostics: impact of wall conditioning by oxidation on mirror properties
Name:	Litnovsky a.litnovsky@fz-juelich.de	Affiliation:	Juelich
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	D. Rudakov (UCSD), V. Philipps (FZJ), C. Wong (GA), W. West (GA), R. Boivin(GA), N. Brooks (GA), P. Wienhold (FZJ), M. Matveeva (FZJ), M. Schaefer (FZJ), R. Bastasz (SNL), J. Whaley (SNL), W. Wampler (SNL), J. Watkins (SNL), J. Brooks (ANL), T. Evans (GA), D. Whyte (UW), P. Stangeby (Univ. of Toronto), A.Mclean (Univ. of Toronto), J. Boedo (UCSD), R. Moyer (UCSD)	ITPA Joint Experiment :	Yes
Description:	Oxidation is presently proposed as one of the candidate techniques for wall conditioning in ITER. There is a known concern, that application of oxidation may adversely affect the properties of in-vessel diagnostic components and in particular, the mirrors	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Oxidation experiment is planned on DIII-D shortly before shutdown. This is a good opportunity to reveal the effect of oxidation on the performance of diagnostic mirrors. In this experiment, several sets of mirrors made from various ITER-candidate materials: Mo, Rh (coated) and Cu will be placed in the diagnostic ports of DIII-D -to study the effect which oxidation causes on their optical characteristics: total and diffuse reflectivity and reflectivity of polarized light. Before undergoing an oxidation treatment in DIII D, optical and surface properties of all mirrors will be pre-characterized. Some of these mirrors will be pre-deposited with an amorphous carbon film to study the cleaning efficiency of oxidation on contaminated diagnostic components. After oxidation the mirrors will be analyzed at FZJ and in other partner laboratories. After finishing of measurements, an attempt will be given to recover the optical properties of mirrors by exposing them in cleaning plasma discharge in the laboratory at FZJ.
Background:	In ITER, all laser and optical diagnostic systems will implement metallic mirrors as their first plasma-viewing optical components. Mirrors will be installed in the upper, midplane and divertor diagnostic ports, under the divertor dome and mounted on the inner wall. Mirrors will suffer from erosion, deposition and particle implantation leading to a degradation of their properties and causing an adverse effect on the performance of entire respective diagnostic systems in ITER. Special care must be given to preserve optical properties of mirrors during the wall conditioning discharges. The robust solution is needed to ensure the optimal performance of diagnostic mirrors in ITER. The investigations on first mirrors are presently recognized as High Priority Task of the ITPA Topical Group on Diagnostics and are the subject of IEA-ITPA Joint Experiments Program (Task DIAG 2). The importance of the R&D program on first mirrors was outlined in the recent ITER Diagnostics review, carried out in Cadarache, France in July 2007. Presently, the Work Plan of the R&D program on diagnostic mirrors is under development and work packages on mitigation of erosion and deposition play an important role within this program. In the present experiment, several sets of pre-characterized diagnostic mirrors made from ITER-candidate materials and supplied by FZJ, will be exposed during the oxidation wall conditioning in DIII D. After exposure all mirrors will be characterized again and the conclusion can be given, if such a wall conditioning technique compatible in its present form with diagnostic components.
Resource Requirements:	Part of the oxygen bake experiment
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:	Extreme care during installation and dismounting: prevention of the mechanical contacts with polished surfaces, mounting only in gloves, avoidance or reduction of air storage of the mirrors â?? sealing in argon.

Title	371: Mirror tests for ITER diagnostics: active control over the deposition on mirrors by the gas feed
Name:	Litnovsky a.litnovsky@fz-juelich.de	Affiliation:	Juelich
Research Area:	General ITER Physics	Presentation time:	Not requested
Co-Author(s):	D. Rudakov (UCSD), V. Philipps (FZJ), C. Wong (GA), W. West (GA), R. Boivin(GA), N. Brooks (GA), P. Wienhold (FZJ), M. Matveeva (FZJ), M. Schaefer (FZJ), R. Bastasz (SNL), J. Whaley (SNL), W. Wampler (SNL), J. Watkins (SNL), J. Brooks (ANL), T. Evans (GA), D. Whyte (UW), P. Stangeby (Univ. of Toronto), A.Mclean (Univ. of Toronto), J. Boedo (UCSD), R. Moyer (UCSD)	ITPA Joint Experiment :	Yes
Description:	This experiment is a continuation of mirror tests for ITER diagnostics. The goal is to investigate the efficiency and applicability of gas feed near the diagnostic mirrors as an active technique to prevent deposition.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Tests of the gas feeding near the diagnostic mirrors were started in DIII D in 2008. The first results of pilot studies revealed the positive effect of gas feeding: the deposition on exposed mirrors was essentially minimized as compared with deposition patterns formed without gas feeding. These results providing a good basis for future dedicated investigations. In the new experiment, it is planned to expose a set of mirrors in the private flux region for series of identical ELMy H-Mode discharges similar to those used in 2005-2008 experiments. The exposure is to be made using the DiMES Mirror holder. Molybdenum mirrors will be used in experiment. Parts of mirrors will be pre-coated with amorphous carbon film, to study the possible cleaning effect. Unlike in the previous exposure, deuterium should be fed directly into the DiMES channel. This will help in the interpretation of the results and possibly, in the subsequent modeling of mirror performance. Total exposure time of >40 seconds (>8 plasma discharges) is requested.
Background:	All laser and optical diagnostic systems in ITER will implement metallic mirrors as their first plasma-viewing optical components. Diagnostic mirrors will suffer from erosion, deposition and particle implantation leading to a degradation of their properties and impacting the performance of entire respective diagnostic systems in ITER. The robust solution is needed to ensure the optimal performance of diagnostic mirrors in ITER. The development of the deposition mitigation techniques is of crucial importance on the way towards such a solution The investigations on first mirrors are presently recognized as High Priority Task of the ITPA Topical Group on Diagnostics and are the subject of IEA-ITPA Joint Experiments Program (Task DIAG 2). The importance of the R&D program on first mirrors was outlined in the recent ITER Diagnostics review, carried out in Cadarache, France in July 2007. Presently, the Work Plan of the R&D program on diagnostic mirrors is under development and work packages on mitigation of deposition play an important role within research on diagnostic mirrors. The dedicated experiments with ITER-candidate mirror materials under ITER-relevant conditions delivered important information on the active deposition mitigation on divertor mirrors by elevated temperature and demonstrated a capability to prevent the carbon deposition and degradation of optical properties by heating the mirrors. Gas feeding in the vicinity of first mirrors is another attractive option to gain an active control over deposition. In experiments performed in TEXTOR it was demonstrated that gas feed is capable to prevent the carbon deposition on the mirrors directly exposed in plasma. First experiments with a gas feed near the mirrors in DIII-D showed significant mitigation of deposition. In the proposed experiment the optimized way of gas feed will be used: D2 will be fed directly into the DiMES channel to ease the evaluation of neutral pressure and to provide advantages for possible modeling.
Resource Requirements:	1/2 day experiment for 2010 (when the capillary is supposed to be installed in DiMES channel). Number of neutral beam sources: 3
Diagnostic Requirements:	All SOL and lower divertor diagnostics, DiMES TV, core Thomson scattering, CER.
Analysis Requirements:	Assistance in analysis of the diagnostic data is desirable.
Other Requirements:	Plasma shape and discharge parameters similar to those of experiment with molybdenum mirrors performed on September 8, 2006.

Title	372: Measurement of In-Plasma Neutral Beam Atom Excited State Lifetime
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	M. Shafer, D. Schlossberg, C. Holland	ITPA Joint Experiment :	No
Description:	Indirectly measure the effective lifetime in plasma of the n=3 excited state of beam atoms as a function of density. This is to provide important atomic physics information to assist with calculations of the point spread functions for beam emission spectroscopy measurements of turbulence. This information is used in calculations of measured turbulence characteristics (wavenumber spectra, amplitudes) and spatial transfer functions which are used for synthetic diagnostics for simulation comparisons.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Run low power, high field, low-temperature L-mode plasmas and measure the turbulence eddy structure with the 2D configuration of BES. Locate the BES array at the radial position with best optical resolution (near r/a=0.75), and at a safety factor profile (q95 ~ 4) that most closely aligns the local pitch angle with the sightline angle. Minimize the turbulence structure size, which has been shown to be proportional to ion gyroradius size, requiring low ion temperature and high field. Vary the 150L neutral beam acceleration voltage over as wide a range as feasible (e.g, 45-85 keV) and then separately vary the density over as wide a range as feasible, while maintaining the ion temperature (and gyroradius) nearly constant (power/density scan), staying in L-mode throughout. The goal is to minimize the turbulence structure size and optimize spatial resolution so that the finite lifetime effects of beam atoms will be most accute. Measure turbulence radial (and poloidal) correlation lengths as a function of beam voltage (velocity) and density. Radial correlation length is of key importance here. The excited state lifetime will affect the measured radial correlation and variation should be discernible over the range tested. Basically, the variation of the radial correlations will be related to a spatial smearing that arises from the finite lifetime effects of the beam atoms, allowing for the inference of the effective lifetime.
Background:	The spatial resolution of BES plays an important role in the wavenumber sensitivity of the diagnostic and for discerning spatial characteristics of turbulence. This resolution is calculated from viewing optics, neutral beam geometry, magnetic field pitch angle as well as on the finite lifetime of the collisionally-excited atoms in the n=3 state (BES views emission from the n=3-2 D-alpha transition). This natural (in vacuum) excited state lifetime is near tau_l=10 ns; this is reduced in plasma to an effective lifetime of tau = 2-5 ns as a result of collisional excitation/ionization processes [I. Hutchinson, PPCF (2002), fig. 2(c)]. This lifetime is calculated from theoretical atomic physics excitation rate calculations and therefore subject to uncertainties in those calculations. An 80 keV Deuterium beam atom has a line-of-sight velocity of v_b=2.8x10^6 m/s, so the spatial spread from from this finite lifetime is about L=v_b * tau=1 cm., comparable to the optical resolution and turbulence correlation lengths, so has a non-negligible impact on the effective spatial resolution of the diagnostic. There is a concern from the turbulence imaging measurements obtained with BES that the actual spatial resolution is slightly better than that calculated using these theoretical rates. One possible explanation is that the effective lifetimes are shorter than calculated. Given the importance of these lifetime for calculating the effective point spread function (spatial transfer function), for use in unfolding the turbulence characteristics and performing experimental validation of simulation codes via synthetic diagnostics, an experimental test is warranted. Note that the power and density scans requested will be of direct value to transport model validation since this will vary turbulence amplitudes, as well as temperature and density profiles, providing useful scaling data for use in the TMV effort.
Resource Requirements:	150LT neutral beam and the 30/330L neutral beams; USN or IWL plasma
Diagnostic Requirements:	BES & CER are essential; all fluctuation diagnostics for TMV application.
Analysis Requirements:	TGLF/GYRO; fluctuation analysis
Other Requirements:

Title	373: Excitation of the Geodesic Acoustic Mode via Radial Field Oscillation
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	G. McKee, A. Garofalo, C. Holland, G. Jackson, M. Shafer, D. Schlossberg	ITPA Joint Experiment :	No
Description:	Amplify the inherent Geodesic Acoustic Mode, a coherent electrostatic zonal flow oscillation, using the high-frequency capability of the DIII-D I-Coils. Measure the turbulence and GAM response to this radial field perturbation. The aim is to enhance the GAM shearing of turbulence and thereby reduce turbulent transport, enhancing global energy confinement.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish plasma conditions were the GAM has been observed: USN plasmas at moderate power (1-2 sources, co-injected, including 150L (steady) and 30L and 330L modulated out of phase). q-scaling experiments indicate that the GAM oscillation is stronger in higher q-discharges, so pick a higher q95 condition, e.g., Ip=1.0 MA, B_t=2.0 MA (119526). The I-Coil is setup in an n=0, m=1 configuration (upper and lower coils 180 out of phase) and run near 15 kHz (SPAs operate at up to 40-100 kHz, so this is technically feasible). Establish basic plasma condition and benchmark GAM parameters with radial scan of BES. Turn on I-Coil in above configuration at near 15 kHz, the known GAM frequency. Scan different frequencies in the expected GAM range (14-18 kHz).
Background:	The Geodesic Acoustic Mode (GAM), a high frequency zonal flow, has been observed at DIII-D in the outer regions of L-mode discharges. It has been measured via high-frequency poloidal velocity analysis of the turbulence obtained with BES and also observed with the Doppler Backscattering diagnostic. It is predicted theoretically to be radially localized, but is poloidally and azimuthally symmetric (m=0, n=0), consistent with the flow measurements obtained at the outboard midplane. Theoretically, it is predicted to have an m=1, n=0 pressure sideband as a result of the non-uniform ExB flow on a flux surface, which has been observed in some experiments. The pressure oscillation, nominally at the "top" and "bottom" of the plasma, relaxes via a radial drift current which gives rise to the very coherent GAM oscillation under some plasma conditions. Typically, the GAM is observed near 15 kHz, peaking near r/a = 0.85-0.95, consistent with its predicted frequency of omega=c_s/R. The GAM is capable of shearing turbulent eddies, and thus controlling and mitigating the saturated level of turbulence and resulting transport. It has also been shown to interact nonlinearly with the turbulence, driving a forward transfer of internal energy [C. Holland, PoP (2007)]. Estimates of time-varying GAM velocity gradient suggest that its shearing rate is comparable to the turbulence decorrelation rate and thus should play a role in turbulence saturation. If it were possible to amplify the GAM, it might be feasible to control and perhaps further reduce turbulence and resulting transport. The question is whether there is a feasible method of tweaking/perturbing/amplifying the GAM. The high frequency I-Coil and audio amplifiers implemented at DIII-D potentially offers such a mechanism. The concept would be to generate an n=0, m=~1 radial magnetic field perturbation at or near the GAM frequency with the I-Coil. The I-Coils produce a radial magnetic field. This may interact with the GAM in one of two (or more?) ways: 1) by producing a radial field that amplifies the radial drift current and thus the pressure relaxation, and 2) by creating a small pressure perturbation through small but finite equilibrium shape modification that enhances the pressure sideband. Quantitative estimates of the radial magnetic field should be performed to assess the feasibility of this.
Resource Requirements:	I-coils setup in n=0, m=1 configuration, connected to SPAs operating at high frequency.
Diagnostic Requirements:	BES, Doppler-reflectometer, CECE
Analysis Requirements:
Other Requirements:

Title	374: Exploration of better feedback toward q=3, 2 and usage of hybrid current- / pressure- driven RWM b
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	Y. In, H. Reimerdes, T. Strait	ITPA Joint Experiment :	No
Description:	This proposal is to explore the RWM suppression toward q = 3, 2 with the emphasis of non-ideal environment for feedback such as uncorrected error field, noise and other MHD activity. The injection of balanced NBI is also proposed to investigate rather poor mode amplification factor of current-driven mode. (1) Inclusion of slow-ramp period in the discharge time sequence The 2008 run, the time interval between the q=5 and q=3 is about 100ms and the growth time ~ 30-50ms. After the q=4 reached, it is hard to separate the dynamic error field correction behavior and the onset of q=3 RWM. Here, we will add slow ramp period every time after integer q_95 is passed. This should isolate current-driven RWM modes at q=4,3,2 and make the optimization of the dynamic error field correction easier. (2) Autonomous upper/lower independent control including the simultaneous n=1/n=3 control The 2008 results also indicated the sensor signals used for feedback (nrsmpid) have large component of n=3 around q=3 period. Preliminary analysis suggested that the large portion is attributable to the compensation process and the incremental change before /after mode onset has much less n=3 component. By using n=1/n=3 simultaneous control, we may be identify the cause of n=3 component (3) The gain scan and time derivative constant scan, while keeping the product of Gd*tau_d same. The key factor is the relationship between tau_d and tau_w (4) Poor gain amplification factor of current-driven RWM and the relation to pressure-driven RWM The gain range for suppressing the current-driven RWM (Gp=40-160) is rather high judging from the typical gain setting for the pressure-driven RWM (typically set as Gp=40 and Gp=60 starts to force the feedback oscillatory). The poor performance with Gp=20 (the mode was converted into rotating mode) is also unique for current-driven mode. This may imply that the amplification by the current-driven RWM mode itself is rather smaller (by a large margin like 5) than the pressure driven mode. In order to understand the poor amplification factor, the transition to the pressure driven-RWM can be investigated by applying balanced NBI with minimum momentum input. - This also provides the full capability of documenting the rotation and q-profile. These studies will be useful to design the ITER dynamic error field correction system	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:
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Title	375: Error field threshold in counter rotation
Name:	Garofalo garofalo@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	Because of the existence of a neoclassical offset rotation in the counter-Ip direction, nonresonant fields applied to a counter rotating plasma are expected to increase the plasma resilience to penetration of resonant fields. This experiment will seek to confirm this never-before observed effect of nonresonant non-axisymmetric magnetic fields.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Prepare discharge with counter-Ip toroidal rotation close to neoclassical offset rotation rate (such as reversed-Ip discharge 131408). Apply strong n=1 braking by ramping down C-coil error correction field. If necessary, extend ramp-down into a ramp-up (at opposite toroidal phase) in order to apply an n=1 error with the C-coil, rather than correct. Look for catastrophic rotation collapse (error field penetration) at rotation = 1/2 of initial rotation. Next, apply large n=3 field using odd-parity I-coil, while keeping n=1 correction. Look for "nothing" happening (because plasma is already at offset rotation. Next, apply strong n=1 braking by repeating C-coil field ramp-down and up (at opposite phase), this time in presence of large n=3 I-coil field. Look for reduction in effect of n=1 error field, compared to case without n=3 field.
Background:	The torque from resonant n=1 magnetic fields can drags the plasma rotation toward zero. Sufficient amplitude of the n=1 field can cause magnetic reconnection and catastrophic locked mode growth. The torque from nonresonant magnetic fields drags the plasma toroidal rotation toward an offset rotation rate in counter-Ip direction. If the initial plasma rotation is already in the counter-Ip direction, a large-enough nonresonant field torque can therefore impede the penetration of n=1 resonant magnetic fields.
Resource Requirements:	Same as DIII-D discharge 131408. This a reversed-Ip discharge.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:	This is a 1/2 day experiment

Title	376: Steady-state high beta with NCS and qmin>2
Name:	Garofalo garofalo@fusion.gat.com	Affiliation:	GA
Research Area:	Assess Steady-State Current Profiles for Optimum Performance	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	This is a re-entering of 2005 ROF proposal #1118. Utilize new current drive capabilities and new lower divertor to extend duration of discharge 122959, which achieved betan>4 with qmin>2 and surface voltage=0 during plasma current flattop. This discharge used a Bt ramp-down technique for off-axis current drive. With 6 gyrotrons this year, we may be able to replace the Bt ramp-down with steady-state compatible ECCD.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Restore discharge 122959. Test effect of new lower divertor on the ability to control the density. Investigate effect of varying the density on the evolution of the q-profile and the pressure profile (lower density is favorable for better ECCD efficiency). Use ECCD for co-Ip current drive at radius larger than rho(qmin). Use FW for counter-Ip current drive at radius smaller than rho(qmin). Some counter-NBI may be used for counter-current drive at small radius, as long as lower rotation does not reduce confinement severely. Initially, use Bt ramp down as in 122959. Work toward reduced or no Bt ramp rate. Initially, keep betaN from exceeding 4 by using NBI feedback control. Work toward higher betan with improved noninductive current profile alignment.
Background:	Analysis of discharge 122959 shows that the high beta phase is terminated by a kink mode destabilized by qmin dropping below 2 because of noninductive current overdrive near rho of qmin. With the new lower divertor we should be able to keep the density low, slow down the qmin evolution and improve the ECCD efficiency. With higher available ECCD power and FW power we should be able to improve the noninductive current alignment, avoid the rapid qmin drop observed in 122959, and extend the duration at betaN=4 and qmin>2. The availabily of counter-Ip NBI provides an additional tool to oppose the noninductive current overdrive inside rho of qmin, and improve the noninductive current alignment.
Resource Requirements:	5-6 gyrotrons, 6-7 neutral beam sources, ICRF. 2 run days
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	377: Baffled-probe measurements of plasma properties at the divertor
Name:	Raitses none	Affiliation:	PPPL
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Y. Raitses, V. Demidov (WVU), A. Pigarov and P. West ?	ITPA Joint Experiment :	No
Description:	Simultaneous measurements of steady-state and fluctuating values of plasma potential, electron and ion temperatures in the divertor using floating baffled-probe technique.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	An ideal scenario is to build and install an array of simple, compact, and inexpensive flush-mounted baffled-probes placed at different oblique magnetic fields on the divertor plate. An alternative option is to modify one or two existing divertor probes. Plasma potential, electron and ion temperatures will be determined from floating measurements.
Background:	The principle of operation of the baffled-probe is based on the dependence of the sheath voltage drop on the local direction of the magnetic field [1,2]. Compared to conventional electrostatic probes, the baffled-probe offers the advantages of direct measurements of plasma potential and electron temperature, while being non-emitting and electrically floating. When the ion saturation current is much greater than the electron saturation current, the baffled probe can potentially provide a measure of the ion temperature. In addition, the simplicity of the probe design and the use of the electric circuitry without bias power supplies are favorable features of the baffled-probe with respect to the other ion-sensitive probes. For the use on the divertor, we developed a flush-mounted baffled-probe design, which can be easy manufactured or modified from the existing flat probes. 1. V. I. Demidov et al., Rev. Sci. Instrum. 74 4558 (2003) 2. V. I. Demidov et al., Contributions to Plasma Physics 44, 689 (2004)
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	378: He/D2 plasma exposure of VPS-W for EAST
Name:	Luo none	Affiliation:	ASIPP
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	Clement Wong, Dmitry Rudakov, Phil West, Karl Umstadter, Russ Doerner	ITPA Joint Experiment :	No
Description:	W plasma facing material (PFM) is proposed to be used for EAST, ITER and DEMO. We would like to make use of the DiMES and MiMES system to test thick W buttons prepared by vacuum plasma spraying (VPS), with exposure to helium or deuterium plasmas in DIII-D. Comparison can be made with the results from a mid-plane material and plasma evaluation system (MAPES) on EAST.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	VPS-W button samples will be loaded onto DiMES and MiMES. The samples will be exposed to as many detached He or D2 plasma discharges as possible, and at different surface temperature provided by resistive heating or by plasma contact.
Background:	ASIPP is now developing VPS-W PFM for the second R & D phase of the EAST PFMC under a new project for a VPS-W/Cu divertor which may be installed in 3-5 years. However, there are still many PWI issues to be studied, e.g., surface erosion, deuterium retention, bubble and blistering formation, disruption tolerance, especially, helium irradiation damage at/near surface at different temperatures. It is thus important to acquire integrated and controlled experimental data in operation tokamaks like DIII-D and EAST to understand better the effects for the design and fabrication of advanced limiter and divertor in EAST, ITER and DEMO.
Resource Requirements:	DIII-D operation and DiMES and MiMES system with material button samples.
Diagnostic Requirements:	All the outboard chamber and lower divertor diagnostics and the core performance diagnostics.
Analysis Requirements:	Surface and retention analysis from EAST, US and Japan laboratories like PISCES, SNL, Kyushu-U and Shizuoka-U. (K. Umstadter, R Doerner, W Wampler, N. Yoshida, K. Okuno)
Other Requirements:	--

Title	379: Investigation of hybrid scenario access conditions with ï? rho*,beta, q and rotation
Name:	Joffrin none	Affiliation:	JET-EFDA-CSU
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	This experiment has the goal to understand the source of improved confinement taking place in hybrid discharges and cross-check the conditions for accessing the improved confinement in hybrid discharges between DIII-D and JET. There is a variety of physics parameter that is thought to improve the confinement in hybrid discharge such as the magnetic shear, rotational shear or Ti/Te, q95, shape and density. Experimental observations on DIII-D and JET on the hybrid scenario are indicating: - Operationally, high initial internal inductance appears as a necessary condition to produce improved confinement. - During the main heating the power dependence of confinement is less than in typical H-mode. (~P-0.3) - Balanced beam in DIII-D or EFCC rotation breaking at JET appears to reduce the confinement. - Low density (low ï?®*) is more favourable to improved confinement. - The increased confinement appears to originate from the outer half of the plasma.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The strategy of the experiment would be as follow: - Establish DIII-D hybrid scenario with a similar plasma shape as JET in order to prepare for an identity experiment. - Test the access condition with DIII-D recipe at a lower toroidal field with the JET shape to have another ï?²* point. - Test the power access condition using a coarse power (or beta scan using DIII-D recipe - Establish hybrid scenario in DIII-D using JET recipe with current overshot and repeat beta scan at two different fields. - Vary the proportion of co- and counter beam in this recipe to test its sensitivity to rotation profile at both toroidal field. This set of data should provide: - connection with the JET data at lower ï?²* at identical ï?¢. - provide identity discharge in terms of q profile and rotation. - help in the cross check of the recipes to obtain the hybrid regime. - cover the space in terms of plasma rotation. In this experiment the documentation of the pedestal is crucial, since the modification of the edge magnetic shear is a potential candidate to explain the improved confinement.
Background:	JET has recently found a route to reach high confinement in hybrid scenario (with central q profile close to 1) by tuning the q profile using current ramp down technique. This has the effect to increase the internal inductance and develop a very broad q profile close to unity in the plasma core. With this technique H>1.4 have been routinely produced. Other experiments indicate that using this q profile the confinement time has a different dependence with power than normally expected from the IPB98y2 scaling. In DIII-D the internal inductance is also increased prior to the main heating using a different technique relying on the shape modification. [1] P. Politzer, Nuc Fus 2008 [2] E. Joffrin, 22nd FEC conference in Geneva 2008 [3] C. Petty, 22nd FEC conference in Geneva 2008
Resource Requirements:	Full NBI power (co- and counter-NBI) and ECCD/ECRH power. Real-time betaN control using NBI power
Diagnostic Requirements:	Plasma pressure and rotation diagnostics (including charge-exchange, poloidal rotation, Thomson scattering, electron cyclotron emission, interferometer). Visible spectroscopy. Turbulence measurements q-profile diagnostics (motional Stark effect). Magnetics.
Analysis Requirements:	Analysis is desirable using tools present in DIII-D and JET (TRANSP, CRONOS)
Other Requirements:

Title	380: Driven Open Loop Stabilization of Vertical Instability
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	Nick Eidietis, Mike Walker	ITPA Joint Experiment :	No
Description:	The goals of this 1/2-day experiment are to study the possibility of open loop driven stabilization of vertical instability. If vertical instability can be stabilized robustly by open loop drive using nonlinear aspects of the vacuum fields, it may provide a way to keep post-TQ plasmas centered, including RE channels, when feedback control is impossible.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Starting with ITER similarity plasmas, apply open loop coil voltage waveforms to produce optimal â??orbitâ?? motion calculated to stabilize (with vertical feedback disabled). If stabilized for > 20 growth times, increase elongation to determine controllability limit.
Background:	A nonlinear description of the vacuum field for elongated equilibria can be cast as a form of Mathieu equation which can be stabilized by Floquet-like solutions (much like an inverted pendulum).
Resource Requirements:	0-4 beams (co), PCS modification
Diagnostic Requirements:	MSE, 5 kHz magnetics sampling
Analysis Requirements:	standard EFITs, TokSys, Corsica
Other Requirements:

Title	381: Role of MHD in Disruption Mitigation
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	T. Evans, E. Hollman, A. James, T. Jernigan, P. Parks, J. Wesley, J. Yu	ITPA Joint Experiment :	No
Description:	The goals of this 1-day experiment are to study the role MHD activity plays in impurity assimilation and RE suppression/deconfinement, and to identify scenarios for producing effective levels of MHD for disruption mitigation for rapid plasma shutdown.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Using standard disruption LSN target plasmas, various stability thresholds will be lowered and raised by applying co- or ctr-ECCD, by varying edge q95, beta, and rotation. Then MGI will be applied to observe the impact on standard mitigation metrics, including current quench rate, halo currents, and RE suppression.
Background:	Experiments in recent years (and careful analysis of data from previous years, primarily by Hollmann and Groth) have suggested that impurity penetration of the plasma is not ballistic, and the fraction of impurity ions assimilated in the plasma core in MGI disruption mitigation does not increase linearly with injected quantity. One strong possibility is that the mixing that results in core assimilation is mediated by the turbulent reconnection activity that accompanies a thermal quench. These processes may also serve to suppress the generation of RE, or to deconfine the RE that are generated. Thus, the amplitude of MHD activity and the spatial size of the largest (global) eigenmodes may provide effective knobs to enhance the effectiveness of disruption mitigation.
Resource Requirements:	Massive gas jet hardware (MEDUSA valve), Ar, D+Ar gases
Diagnostic Requirements:	Usual disruption diagnostics: 5 kHz magnetics, Thomson, new IR camera strongly desirable, fast cameras (LLNL and UCSD), UCSD scintillators for RE detection, FPLASTIC
Analysis Requirements:	JFIT; NIMROD analysis desirable
Other Requirements:

Title	382: Role of Kappa and Growth Rate in RE Suppression
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	Disruption Characterization	Presentation time:	Requested
Co-Author(s):	T. Evans, E. Hollman, A. James, T. Jernigan, P. Parks, J. Wesley, J. Yu	ITPA Joint Experiment :	Yes
Description:	The goals of this 1-day experiment are to study the role elongation and relative vertical stability play in RE suppression/deconfinement, and to help identify scenarios for exploiting this for rapid plasma shutdown.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Starting with standard disruption LSN target plasmas, MGI will be applied to produce reproducible RE channels, visible by their photoneutron and gamma ray signatures as they are deconfined. Then the elongation and vertical stability thresholds will be varied separately to study the effect on RE production and deconfinement. Limited vs diverted plasmas will be studied, and H-mode vs L-mode vs ohmic (to vary growth rate at same elongation).
Background:	Granetz has suggested (ITPA Stability TG, IAEA FEC, APS 2008) that the presence of RE channels among devices worldwide seems to be correlated with low elongation, limited vs diverted plasmas, and/or relative vertical stability. It is not clear why relative vertical stability itself would contribute to RE formation/suppression. However, if the strength of MHD activity plays a role, lower nonaxisymmetric stability thresholds that may result from higher elongation and diverted (less wall stabilized) plasmas could explain the observation. The ITPA Stability TG has recommended this study.
Resource Requirements:	Massive gas jet hardware (MEDUSA valve), Ar, D+Ar gases, 0-4 beams
Diagnostic Requirements:	Usual disruption diagnostics: 5 kHz magnetics, DISRAD, Thomson, new IR camera strongly desirable, fast cameras (LLNL and UCSD), UCSD scintillators for RE detection, FPLASTIC
Analysis Requirements:
Other Requirements:

Title	383: Demonstrate Integrated Active Control Methods for Optimal Rapid Shutdown
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Requested
Co-Author(s):	T. Evans, E. Hollman, A. James, T. Jernigan, P. Parks, J. Wesley, J. Yu	ITPA Joint Experiment :	No
Description:	The goal of this 1-day experiment is to demonstrate integrated active control methods along with impurity gas injection to produce optimal disruption mitigation for rapid plasma shutdown.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Using standard disruption LSN target plasmas, various elements contributing to good rapid shutdown with effective mitigation of disruption effects will be applied and then varied. Elements include moving the plasma to the neutral point, dropping kappa, start Ip rampdown, drive co/ctr-ECCD to modify MHD characteristics, apply RMP, regulate plasma position with closed or open loop drive, move plasma inboard possibly limiting. Optimal gas jet species or other impurity injection method (derived from results of other experiments) will be applied as well. These elements will be varied to assess their relative role in an optimal shutdown scenario for ITER.
Background:	The window for informing key aspects of the ITER design to provide necessary hardware for effective rapid shutdown/disruption mitigation capability is closing. The ITER IO reports that design-modifying information must be supplied within roughly 9-12 months in order to avoid adversely impacting the schedule (or avoid the input being ignored). This experiment will integrate the results of several focused experiments to be performed this year under the Rapid Shutdown for ITER TF, in order to demonstrate and assess the combined effects of many key elements of a candidate scenario for ITER.
Resource Requirements:	Massive gas jet hardware (MEDUSA valve), Ar, D+Ar gases
Diagnostic Requirements:	Usual disruption diagnostics: 5 kHz magnetics, DISRAD, Thomson, new IR camera strongly desirable, fast cameras (LLNL and UCSD), UCSD scintillators for RE detection, FPLASTIC
Analysis Requirements:
Other Requirements:

Title	384: Vertical Stability Controllability Physics for ITER
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Requested
Co-Author(s):	M. Cavinato, Yu. Gribov, G. Jackson, A. Portone, M. Walker	ITPA Joint Experiment :	Yes
Description:	The goals of this 1-day experiment are to complete the quantification of fundamental limits of vertical controllability of DIII-D, validating theoretical analysis, and identifying the exact role of noise and disturbances. Robustness of operation near control limits will be studied, key disturbances will be triggered and characterized, and the control responses to noise and disturbances will be quantified. This falls under an ITPA joint experiment topic and contributes to critical ITER R&D needs providing input to the ITER design over the next 9 months.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Using ITER similarity plasmas, increase elongation in solw ramps to reconfirm control limits in flattop. Then operate at fixed elongation for hundreds of growth times for varying proximity to the control limits. Produce disturbances near control limits by dropping beams, either staying in H-mode or dropping to L-mode. Inject artificial Z-measurement noise of varying amplitude to study quantitative response and impact on controllability. Produce Ip rampdown trajectories that cross the controllability boundary, and compare with tailored rampdown trajectory to remain close to but within controllable region of kappa and li. Some of these studies could potentially be done in piggyback on other ITER Startup/Rampdown experiments, especially in making use of the flattop. However, some discharges must necessarily disrupt the plasma, limiting the ability to do all cases in piggyback.
Background:	The ITER vertical control system was substantially modified in 2007-08, in part due to successful experiments on DIII-D done in that period as well as theoretical analysis identifying key metrics for vertical control performance. The present performance requirement specification guiding new ITER in-vessel coil designs is based on certain assumptions about the expected noise/disturbance environment in ITER, and theories predicting the consequences for controllability. These assumptions and the theoretical basis for the resulting design guidance must be confirmed by experiment to place the present design proposals on a firm footing (and possibly changing them) and quantify the actual likelihood of loss of control for licensing purposes.
Resource Requirements:	0-4 beams (co)
Diagnostic Requirements:	MSE, 2-5 kHz magnetics sampling
Analysis Requirements:	standard EFITs, TokSys, Corsica
Other Requirements:

Title	385: Improved Gas Jet Disruption Mitigation by I-coil-Enhanced Impurity Transport
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Not requested
Co-Author(s):	T. Evans, E. Hollman, A. James, T. Jernigan, P. Parks, J. Wesley, J. Yu	ITPA Joint Experiment :	No
Description:	Continue study of possible improvements to disruption mitigation effects by massive gas injection using the I-coil to impose high Chirikov parameter fields on pre-disruption and plasmas and during injection. Develop potential key element in optimal rapid shutdown scenario for ITER.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Standard disruption target plasmas in close proximity of I-coil and gas jet (R+1) for good penetration of I-coil field (high Chirikov parameter). Re-establish baseline jet injection pre-emptive disruption mitigation without I-coil field. Use Ar MGI to produce RE channel at time of TQ. Re-establish n=3 suppression, test n=2, n=1 at different coil amplitudes.
Background:	Experiments last year showed apparent suppression of RE with n=3 at full I-coil current in several discharges, but later attempts to reproduce were unsuccessful. A more consistent RE target case produced with Ar MGI was developed, and will be used for the present experiment. The possibility of suppression RE production or enhancing RE deconfinement during the TQ is important because achieving the Rosenbluth density required to suppress RE through collisional damping has been challenging in experiments to date. RMP fields have been successful in suppressing RE production (or enhancing deconfinement) in devices such as TEXTOR.
Resource Requirements:	Massive gas jet hardware (MEDUSA valve), Ar, D+Ar gases
Diagnostic Requirements:	Usual disruption diagnostics: 5 kHz magnetics, DISRAD, Thomson, new IR camera strongly desirable, fast cameras (LLNL and UCSD), scintillators for RE detection, FPLASTIC
Analysis Requirements:
Other Requirements:

Title	386: Vertical Stability Control Using C-coil and I-coil
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Control/Operations	Presentation time:	Not requested
Co-Author(s):	D. Gates, S. Sabbagh, J. Leuer	ITPA Joint Experiment :	No
Description:	The goals of this ½ -day experiment are to study the potential for augmenting vertical control capability using the C-coil or the I-coil.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Using ITER similarity plasmas, perturb vertical positions with usual (F6,7,2) control, â??ITER-likeâ?? control (F6,7), and â??ITER-likeâ?? control + C-coil or + I-coil, or + both. Increase elongation in each case in steps, holding for periods > 10-20 growth times. Determine highest growth rate sustainable with no VDE onset, and growth rate beyond which VDE is guaranteed (or extremely likely). Using targets near maximum growth rate, freeze coil commands to disable vertical/shape control for varying lengths of time, allowing VDE to acquire growth rate data.
Background:	Various modeling results suggest that the vertical control system in the present ITER design will not provide sufficiently capable and robust control to maintain vertical stability during several points in the ITER nominal scenarios. While DIII-D does not need the extra capability, the added capability potentially made available to ITER could be very important. The C-coil can apply a true n=0 field, but is rather distant from the plasma and is shielded by the vessel. The I-coil array produces a fairly small radial field when run in n=0 mode, but is very close to the plasma and the shielding principally produces a decrease in amplitude, not in phase. The two coils thus represent two extrema in possible geometry of such coils, roughly corresponding to the present ITER midplane EFCC array and the â??Blanket-Vessel Interfaceâ?? coils respectively. A similar experiment was proposed to be executed in NSTX, allowing comparison and validation of modeling results and general predictions for ITER.
Resource Requirements:	0-4 beams (co), PCS modification
Diagnostic Requirements:	MSE, 5 kHz magnetics sampling
Analysis Requirements:	standard EFITs, TokSys, Corsica
Other Requirements:

Title	387: Application of DEFC from t=0, considering the application to ITER ohmic period
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The initial breakdown and establishing ohmic current process is still not well understood in the startup process. However, it is important to reduce the ohmic flux consumption as well as to produce discharges in a reproducible manner. This process is likely governed by variety of non-axi-symmetric global MHD activity. These MHD are expected to be highly resistive, but some of mode structures extend to the plasma boundary. If we can influence the mode boundary condition, it is possible to modify ( may not reach to control) the initial ohmic period and to make the discharge less-interacted with the wall, and less-impurity or gas inflow from the periphery. Here, it is proposed to apply the n=1/n=3 feedback from t=0. The feedback will be carried out in a dynamic error field correction mode. In the experiment of the current-driven RWM which was carried out by large Ip ramp, the feedback coil current and sensor signal behavior is consistent with the one in the dynamic error field correction mode, although we do not know what kind of mode was interacted through the feedback.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Here, it is proposed to study (1) The currently-proposed ITER internal coil arrangement by using I_coil and to determine the most effective time constant of DEFC in relation to (1) the wall time constant, (2) Ip-ramp rate, (3) the plasma shaping changing rate, and (4) vertical plasma displacement rate. The most interesting time period may be the period where the internal inductance varies in a few times wall time for shaping shape (2) Impact of the thick blanket penetration by using C-coil as the feedback coils. It is also possible to use the I-coils as short-circuited so that these coils serve as a variable helical flux dissipation material.
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	388: Resolve interior of separatirx lobes by very slow q_95 sweeps
Name:	Schmitz oschmitz@wisc.edu	Affiliation:	U of Wisconsin
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):	T. Evans (GA), M. Fenstermacher (LLNL), M. Jakubowski (MPI), R. Moyer (UCSD), J. Wattkins (SNL), P. West (GA), A. Wingen (U Duesseldorf),	ITPA Joint Experiment :	Yes
Description:	Resolution of the interior of the lobes of the perturbed separatrix during application of n=3 RMP fields for ELM control shall allow to determine the level of stochasticity achieved. It was measured experimentally [O. Schmitz et al., PPCF 50 (2008) 124029] that during q_95 scans these lobes rotate and spiral along poloidally and toroidally fixed diagnostics. We suggest to place lobes on selected target Langmuir porbes and sweep q_95 very slowly up and down by 0.05-0.1 within the ELM suppression window. This technique will shift the internal structure of the lobe along the probe and should exhibit the different radial connection points of the interior field lines.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	This experiment will have two sections. First, we want to place one lobe on one of the lower divertor shelf target Langmuir probes. The advantage is that for this setup we know how to get ELM suppression and therefore investigate the field structure in the desired working scenario and without ELMs, having the advantage of quiet Hmode plasmas facilitating the Langmuir probe and camera measurements. Second, we want to place the lobes on top of the outer divertor shelf and identify the structure there. Here a much better resolution can be achieved having more probes and a wide spread out of the lobes. The topology will be adapted such that the lobes hit as many as possible probes and then again slight q_95 ramping up and down in the ELM suppression window will be performed. For both parts a change of the toroidal phase is foreseen which will place different lobes on different probes.
Background:	One crucial and very generic question in ELM suppression experiments is the penetration depth of the external field and the width and structure of the stochastic layer induced. Based on a technique successfully applied at TEXTOR [M. Jakubwoski et al., 363 (2007) 371], [A. Wingen et al., PoP 14 (2007) 042502] the resolution of the tangles of the tangles of the resonant island chains allows to identify directly the penetration depth of the RMP field and the connected stochastic structure. Recent analysis of the internal structure of the perturbed separatrix lobes during RMP application at DIII-D [A. Wingen, PoP (2007), submitted], [O. Schmitz et al., PPCF 50 (2008) 124029] showed that the connections to the different resonant surfaces are embedded into a fine mesh of boomerangs inside of the separatrix lobes on the target and therefore identificantion of the single resonance layer getting stochastic is possible in principle. However, resolution of the fine structure is challenging but seems to be possible based on existing data. It was shown experimentally by measurements with an IR and filtered CCD cameras, that the lobes spiral along the target during q_95 ramp down [O. Schmitz et al., PPCF 50 (2008) 124029]. This means that the interior of each lobe can be moved along a fixed target Langmuir probe by very small q_95 sweeps, i.e. q_95 ramp down by 0.05 and ramp up again. This shall allow to measure electron density, temperature and in particular the floating potential inside of the lobes yielding to information about the penetration depth.
Resource Requirements:	standard ISS plasmas used for ELM suppression experiments (e.g. #132741), SPA supplies for I-coil, C-coil in standard n=1 EFC setting (same as #132741), 1.0ITER04 f-coil patch panel, adaptation needed for setting of fb09 which is in open loops setting for 1.0ITER04 but does nor allow to put outer strike point on top of the shelf.
Diagnostic Requirements:	ECE, Thomson Scattering (core, tangential and divertor systems would be desirable), BES, reflectometer, target Langmuir probes, DiMES_TV with C filters in filter wheel, tanTV system with C filters, IRTV (both cameras / LLNL+TEXTOR), MSE measurement combined with fast Li beam would be beneficial for edge current evolution during heat pulses, CER needed for rotation measurement
Analysis Requirements:	TRIP3D, manifold topology by A. Wingen, equilibrum from kinetic EFIT, EMC3/EIRENE for 3D fluid modeling of stationary heat and particle fluxes
Other Requirements:	This experiment directly transfers a technique from TEXTOR to DIII-D and will allow direct comparison as intended in ITPA task PEP19.

Title	389: A possibility of modifying the VDE process by the DEFC application
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	In the process of the VDE, the wall eddy current dissipates the toroidal flux and the simultaneous reduction of the plasma current makes the poloidal flux loss. As the consequence, helical flux loss process takes place. The pattern of wall current was reported to include n=0, as well as n=1. However, it is not well understood yet how the process takes place, although some MHD activity has been reported together with VDE process. The final VDE reaches huge current beyond any control. However, the application of DEFC could be able to influence the initial stage of VDE and modify the current path leading to a different type of VDE or weakening the VDE process. This can be studied by applying the DEFC at the anticipated VDE time period. In particular, when some MHD is involved in the initial VDE process, the final outcome may depend on the initial DEFC even with 100 gauss level with I-coils connected to the SPA. If the DEFC is found to create some impact, various pattern of preprogrammed coil current on I-coil with maximum current with SPA should be explored. A typical vertical positional instability under DEFC may provide the essence of physical process. This study will help for understanding the VDE process	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	390: Study of High Temperature Pedestals in VH-mode
Name:	Solano emilia.solano@ciemat.es	Affiliation:	Ciemat
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	JET: Peter J. Lomas, Barry Alper, Vassili Parail DIII-D: Alan Turnbull, Tom Osborne, Gary Jackson, Ed Lazarus, others?	ITPA Joint Experiment :	No
Description:	Use the VH-mode as a vehicle for studying a pedestal as close to ITERâ??s as it is possible in DIII-D and JET. In terms of both resistivity and collisionalities VH-mode pedestals can be very ITER relevant. Note that the present plan is for ITER to access the H-mode with full heating at low density, and then increasing the density until a steady ELMy H-mode regime (without ELMs) is attained. According to TRANSP simulations by Bob Budny, initially the plasma could be in a Hot Ion regime. We can not predict recycling with the ITER wall, possible Be in ITER may be equivalent to Boron in DIII-D. The low density access is necessary because of the high L to H power threshold expected. Note that the natural route to obtain the VH mode automatically produces such a pedestal evolution. We believe that the characterization and understanding of pedestal evolution and MHD instabilities present in a VH plasma before the first ELM provide a good benchmark of ELM models to be applied to ITER prediction, as well as pedestal scaling (heights, widths). Further, such a plasma is a good target to test ELM control or suppression techniques.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	revisit the DIII-D VH regime to study a high temperature low density pedestal.
Background:	In JET the Hot Ion H-mode regime is very similar to the DIII-D VH-mode. Low recycling and sawteeth control are essential ingredients to obtain high Te and Ti pedestals. In the past pedestal Te as high as 3.5 keV were obtained in JET in Hot-Ion H-modes. Recently the regime was recovered, and we obtained up to Te_ped= 3 keV during the ELM-free phase, but edge MHD (â??outer modesâ??) reduced the Te to 2.7 keV just before the ELM. Ti_ped is always higher than Te_ped. Outer modes are described in (Nave, Nucl. Fusion 37 (1997) 809-24): they are low frequency low n modes, localized to the pedestal region, and were identified (Huysmans et al, Nucl. Fusion 38 (1998) 179-187) as external (current-driven) kink-modes. We have various recent examples with similar temperatures, and have not had opportunity to fully analyze the data yet (experiments were carried out on 29/10/2008 and 25/11/2008). We clearly see outer modes, very distinct from the high n ELM precursors observed more typically in the ELMy H-mode regime. In some pulses we observe both types of modes, either simultaneously (bursts of high n modes amongst low n kinks), or at different times. Washboard modes [C P Perez et al 2004 Plasma Phys. Control. Fusion 46 61-87] are seen in many of these plasmas, but appear to be weaker when outer modes are present. The DIII-D VH mode is best obtained with boronization and high shape. Comaprison of pedestal characteristics of DIII-D's VH mode and JET's hot ion H-mode can shed light on pedestal MHD stability and evolution.
Resource Requirements:	High beam power, boronization, high triangularity. ECH for NTM control? Experts on VH-mode.
Diagnostic Requirements:	All pedestal and divertor diagnostics. Edge current measurements (especially simultaneously) with the Li-beam and co- plus counter-beam MSE would be highly desirable as would fast divertor IRTV.
Analysis Requirements:	MDetailed kinetic equilibrium reconstruction, MHD identification and modelling. Transport modelling.
Other Requirements:

Title	391: Coupling between Peering mode driven RWM and ELM and possible issues of non-helicity-preferred DEFC
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	At IAEA 2008, it was reported that the observation of peaking of RFA below no-wall limit, in addition to the main RFA at no-wall limit by Liu. The response was interpreted as a RFA driven by the response to a pealing mode. In the DIII-D, we sometimes observe the RFA bellow no-wall limit. One example was reported at â??Post-APS 3rd Error/Non-Axisymmetric Magnetic Field Workshopâ?? organized by Dr. L. Lao. The plasma condition was bellow no-wall limit, betaN~ 3li (no wall limit of 4li plasma). Series of type 1 ELMs were excited extremely in a reproducible manner. The time traces of ELM excitation / decay observed by Mirnov coils were extremely reproducible at any toroidal location. Also RFA was excited reproducibly as a precursor event to each ELM event. This was carried out with DEFC operation using C-coils. There could a possibility that the DEFC with non-helical preference like C-coil allows the n=1 RFA below no-wall limit through the harmonics of a marginally-stable mode like the peering mode. Here, it is proposed to study the RFA below no-wall limit with non-helicity-preferred DEFC C-coil operation. The comparison with I-coil operation will provide the requirement of DEFC on the tolerance of harmonics in the ITER operation.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	392: Investigate physics of rotation modification of EHO-induced edge transport
Name:	Burrell Burrell@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T.H. Osborne, P.B. Snyder	ITPA Joint Experiment :	No
Description:	Perform a systematic investigation of the effect of theoretically important stability parameters (safety factor, outer gap, squareness, etc.) on the changes in the EHO-induced transport caused by decrease in the edge toroidal rotation. In addition, input power will be varied from 2 to 5 sources to see if the heat flux through the edge has an effect.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Basic discharge will be a reverse Ip QH-mode similar to shots 131920-22. Beam torque will be reduced in the middle of the shot to change the edge rotation. Systematic scans of safety factor, outer gap, squareness and total beam power will be made to see how the EHO-induced transport changes with these parameters
Background:	QH-mode experiments in 2006-2008 have shown that the edge particle transport caused by the EHO can be modified by changing the edge toroidal rotation. However, the physics mechanism behind this is not yet understood. The best plasmas to study this effect are the balanced double null discharges such as 128510, 128513, 131384, 131386, and 131920-22. For these, the density limit for the QH-mode is sufficiently high that we can study the transport change without having the ELMs come back. As can be seen by the list of shot number, we only have a few shots under these conditions and they show different EHO phenomenology. What we need is a systematic study under reproducible conditions where we change the parameters that are known to affect the EHO and then seen how the transport changes. We willuse the theoretically important stability parameters such as safety factor,outer gap, squareness, etc. The effect of heat flux will also be investigated.
Resource Requirements:	Reverse Ip. 7 NBI sources
Diagnostic Requirements:	Profile diagnostics, edge fluctuation diagnostics for EHO studies
Analysis Requirements:
Other Requirements:

Title	393: Study of non-axissymmetric field coupling with rotating magnetic perturbation fields
Name:	Stoschus stoschus@fusion.gat.com	Affiliation:	Oak Ridge Associated Universities
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	Oliver Schmitz (FZJ), M.J.Schaffer(GA), T.E.Evans(GA), M.E.Fenstermacher (LLNL), B.Unterberg(FZJ)	ITPA Joint Experiment :	Yes
Description:	The experimental study of the coupling of external, resonant and non-resonant, non-axissymmetric magnetic fields and its dependence on the plasma parameters, in particular on the relative rotation between plasma and field is an important task for understanding the plasma response on external non-axissymmetric fields. In this experiment the audio amplifiers (AAs) will be used to generate a rotating, non-axissymmetric field rotating in co- and counter current direction in order to change the relative rotation between plasma and external field. This shall open a new way to study the field penetration and plasma response characteristics in poloidal divertor H-mode plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Audio Amplifiers (AAs) will be used to generate a rotating RMP field. The relative rotation frequency between the RMP field and the plasma rotation will be steered by both, change of the AA's field frequency and the toroidal rotation frequency using co- and counter current beams. The working points will be determined by a scan through the minimum of the slip frequency, i.e. the relative rotation frequency between external field and plasma rotation on different resonant surfaces. In the second part we will attempt to scan through the minimum of the relative frequency between external field and the electron diamagnetic drift frequency observing appearance and disappearance of resonant currents in particular with the fast MHD diagnostics. For diagnosis of the related effect we will focus on (a) known signatures of magnetic structures, i.e. striated target footprint pattern and pedestal profile modifications, (b) fast MHD signals and evidence for loss and appearance of resonant currents during relative frequency scans and (c) performance characteristics, such as density pump out and affect on rotation when frequency resonant effects are expected.
Background:	At TEXTOR-DED experimental evidence was found that two effects concerning the relative rotation of external field and plasma fluid determine the field coupling. First, rotational screening theories predict that the relative rotation frequency between external perturbation fields and the plasma rotation, the so-called slip frequency, determines the field penetration, i.e. the actual B_r value on given resonant surfaces. On the other hand the relative rotation frequency of the RMP field and the electron diamagnetic drift frequency shall determine tearing mode evolution and/or resonant field amplification on resonant surfaces as a form of plasma response. These two important concepts need to be resolved experimentally as they are important for both, basic physics understanding and application of RMP in low rotation plasmas for ELM control as planned for ITER. These concepts are also studied in detail with rotating resonant magnetic perturbation (RMP) fields induced by the Dynamic Ergidic Divertor (DED) at TEXTOR. Here the relative rotation between plasma and RMP field was varied by supplying the coils with AC currents with 90 deg phase difference. This leads to a rotating magnetic structure. In addition neutral beam torque injection into co- and counter current direction was used to steer the plasma rotation and the plasma temperature at different densities, i.e. the plasma resistivity. Evidence was found recently that matching of the electron diamagnetic drift frequency leads to a spontaneous formation of magnetic islands with an extend larger than predicted by vacuum magnetic field modeling. Below the DED current level accociated for a given rotation to this MHD threshold, the relative rotation determined the evolution of local magnetic structures used in comparision to DED-DC experiments as a proxy for the field penetration. These experiments were performed in circular, high field side limited L-mode plasmas. With this application for the "Torkil Jensen Award for Innovative Research", we want to study the same behavior in poloidally diverted H-mode discharges in two scenarios: (a) application of n=3 RMP field with optimum error field correction, similar to typicall RMP H-modes and (b) configurations with good error field compensation but strong non-resonant components (reverse I_p or B_t to get non-resonant), scanning the torque input through the slip frequency and electron diamagnetic drift frequency resonance. This will allow to widely extend the basic findings at TEXTOR towards high-performance H-mode discharges at DIII-D and answer generic questions on the coupling of external, non-axissymmetric fields. We submit this proposal for the "Torkil Jensen Award for Innovative Research" as we are attacking new physics in combining the unique DIII-D RMP and torque input capabilities combined with a collaborative experimental approach comparing directly DIII-D and TEXTOR-DED results with rotating fields
Resource Requirements:	ISS plasma used for ELM suppression experiments (e.g. #132741), audio amplifier (AA) supplies for I-coil, C-coil in standard n=1 EFC setting (same as #132741), 1.0ITER04 f-coil patch panel
Diagnostic Requirements:	fast MHD, all mirnov coils and magnetic probe, ECE, Thomson Scattering (core, tangential and divertor systems would be desirable), BES, reflectometer, target Langmuir probes, DiMES_TV with C filters in filter wheel, tanTV system with C filters, fast framing IRTVs (both cameras / LLNL+TEXTOR), MSE measurement combined with fast Li beam would be beneficial for edge current evolution during heat pulses, CER needed for rotation measurement, He gas puff imaging if established already
Analysis Requirements:	CER analysis and evaluation of other systems listed above, TRIP3D as vacuum topology to compare with, kinetic EFIT, impact of rotation on current distribution
Other Requirements:	Preparation of suited audio amplifier I-coil patch panel and discussion with engineers on maximum frequency and optimum current/frequency setting.

Title	394: Deposition and erosion studies with the DiMES PPI sample in detachment
Name:	McLean mclean@fusion.gat.com	Affiliation:	LLNL
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Requested
Co-Author(s):	P.C. Stangeby, J.W. Davis, Y. Mu, A.A. Haasz (UofT), B.D. Bray, N.H. Brooks, W.P. West, C.P.C. Wong (GA), D.G. Whyte (MIT), S. Brezinsek (Jülich), E.M. Hollmann, D.L. Rudakov (UCSD), M. Fenstermacher, M. Groth, C.J. Lasnier (LLNL), J.G. Watkins (Sandia), R. Isler (ORNL)	ITPA Joint Experiment :	Yes
Description:	This experiment will build on previous PPI run days in 2003 and 2005, but make use of significant upgrades to DIII-D diagnostics, making full use of the plasma configuration, and maximizing use of the PPI itself. Motivations for this experiment include: -Development and characterization of a sustained, steady, fully detached regime for the OSP as will be used in ITER -Use of much improved optical transmission and a potential improvement in diffraction grating efficiency on the MDS spectrometer in 2009 -Monitoring of the CD/CH B-X band at 390 nm for the first time, and comparison of its intensity with the A-X band. -Use of improvements made in sensitivity of the lowermost DTS chords in 2008/2009 -Full use of the LPs on the lower divertor shelf to better characterize the OSP, PFZ, and SOL than done previously in 2005 -Insertion and spectral characterization of both a blank graphite and the porous PPI sample head without gas injection to compare each to intrinsic emission from surrounding tiles and ensure their use as a spectral background -Injection of heavy hydrocarbons from the PPI representative of ~50% of carbon-containing molecules derived from the chemical erosion process	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	CH4/C2H4 will be injected into the OSP whose position is fixed over DiMES from 1.0-5.0 seconds in each data shot. All shots will be SAPP L-mode, beginning with a 20 eV, 2.5E13 /m3 attached plasma at the OSP. At 1.5 seconds, additional D2 will be injected in order to detach the OSP, leading to stable strike point conditions from 2.0-5.0 seconds. Repeat shots will be made both for spectroscopic coverage, and to compare puff vs. no puff emission characteristics. The outer target will be swept in selected shots to characterize the OSP.
Background:	In its first two applications in 2003 and 2005, data from the PPI have directly and indirectly helped advance our understanding of chemical sputtering and solve longstanding mysteries regarding measurements of chemical erosion yield in DIII-D. The PPI has helped demonstrate the importance of injection at intrinsic release rates for measurement of chemical erosion yield and photon efficiencies, and remains the only artificial injector in tokamaks worldwide which replicates release of chemically eroded molecules from the surface over a distributed, porous area. The PPI has also revealed that mechanisms for excitation of chemically eroded molecules may be at work for intrinsically released hydrocarbon fragments, potentially with great influence on measurements of photon efficiency and the relative contribution of chemical vs. physical sputtering to the C-atom and C+ sources in the divertor. Significant improvements in our understanding of the processes involved, the diagnostics available for study of emission from the PPI, and simulation tools available to model its result make a new experiment an exciting proposition to boundary physics.
Resource Requirements:	2.0 day experiment -0.5 day for characterization of spectroscopic signals from a blank graphite sample -0.5 day for characterization of spectroscopic signals from the PPI sample head without gas injection -0.5 day for characterization of spectroscopic signals from the PPI using CH4 in a fully detached OSP -0.5 day for characterization of spectroscopic signals from the PPI using C2H4 in a fully detached OSP Number of neutral beam sources: 2 No requirement for ECH or ICH
Diagnostic Requirements:	PPI DiMES sample, MDS, lower divertor IR camera, DiMES TV with light-limiting mask, divertor Thomson scattering, fixed floor Langmuir probes, midplane and X-point reciprocating Langmuir probes, filterscopes, CER, tangential divertor cameras, Ocean Optics spectrometer, RGA, SPRED.
Analysis Requirements:	DIVIMP-HC 3D Monte Carlo impurity simulation code (Y. Mu), MDS-SIM synthetic diagnostic simulation code for MDS (McLean, Brooks, Isler)
Other Requirements:

Title	395: Tearing mode stability of high-beta discharges
Name:	Turco turcof@fusion.gat.com	Affiliation:	Columbia U
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):	Luce, Ferron, Petty, Turnbull, Brennan	ITPA Joint Experiment :	No
Description:		ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A series of high-beta discharges is proposed, with and without ECCD, with ECCD location scans in rho and varying the EC current deposition shape.
Background:	Tearing mode stability is crucial for high-performance scenarios intended for steady-state operation. The appearance of tearing modes in DIII-D discharges leads to loss of energy confinement, but more importantly to redistribution of the current profile that is not recoverable with the available non-inductive current drive sources. Tearing modes can appear after 1-2 s at constant pressure (i.e., on the resistive evolution time scale). The stability is strongly affected by the location and distribution of the applied electron cyclotron current drive (ECCD), but not through direct interaction with the mode rational surface.
Resource Requirements:	Full power NBI and all the available gyrotrons.
Diagnostic Requirements:	MSE, magnetics and (fast-)ECE are crucial. BES desirable.
Analysis Requirements:
Other Requirements:

Title	396: Tearing mode stability in presence of pellet fuelling
Name:	Turco turcof@fusion.gat.com	Affiliation:	Columbia U
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):	Luce, Commaux, Parks	ITPA Joint Experiment :	No
Description:	Tearing mode stability is crucial for high-performance scenarios intended for steady-state operation. The appearance of tearing modes in DIII-D discharges leads to loss of energy confinement, but more importantly to redistribution of the current profile that is not recoverable with the available non-inductive current drive sources. Moreover, it has been put forward that the presence of pellet injection, which is foreseen for ITER fueling, has a negative effect on the tearing stability of the discharges. Therefore, within the frame of investigating the tearing stabilization by preemptive ECCD injection, it is important to address the same subject taking into account the use of pellet injection during the discharge.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A series of high-beta discharges with different ECCD deposition locations are required, with simultaneous pellet injection in the high-beta phase, in order to test the mechanism of tearing stabilization with the additional potentially destabilizing presence of pellets.
Background:
Resource Requirements:	Full NBI power, all the available gyrotrons, pellet injection system.
Diagnostic Requirements:	MSE, magnetics and (fast-)ECE are crucial. BES desirable.
Analysis Requirements:
Other Requirements:

Title	397: ITER Demo steady state target scenario development
Name:	Murakami masanori_murakami@att.com	Affiliation:	Retired
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	T. Luce, J.M. Park, E. Doyle, , J. Ferron, J. DeBoo, et al.	ITPA Joint Experiment :	Yes
Description:	â?¢ Ip scan between the last yearâ??s two best shots (131198 (xx MA) and 134372 (yy MA) â?¢ Optimize the best case and document the edge region (r â?¥ 0.8) â?¢ Serve as target discharge for scenario modeling study to seek optimal strategy for ITER shape plasma and DIII-D nearly fully non-inductive, high performance discharges	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Reproduce shot like 131198 2) betaN scan to see how far can be increase 3) Evaluate f_NI, f_BS and G=betaN*H/q^2 4) Scan Ip 5) Find the best f_NI, f_BS and G=betaN*H/q^2 6) Document the discharge, in particular in the edge (with ELM average)
Background:	â?¢ ITER steady state scenario aims at demonstrating full NI operation at Q=5 and fBS>0.5 â?¢ This achievement requires good confinement to satisfy both pressure and current balance â?¢ We have been working modeling using theory-based (GLF23) modeling with self-consistent sources and sink based on the day-1 hardware capabilities. Because of the stiff transport model, we used the â??edge temperatureâ?? as a confinement control knob. â?¢ What we ended up with is beta_N ~ 1.5 at r~0.8 which is only slightly higher than those achieved (134372). However within the â??edge regionâ??, the assumed pedestal temperature was too high and pedestal width too broad, compared with P. Snyderâ??s EPED1. We need more realistic edge profile including realistic pedestal. â?¢ T. Luceâ??s 0-D prediction based on ITER Demo shots indicates that there may be more promising operation region at a higher current (1.3 MA rather than 0.9 MA) â?¢ So we seek for : (1) Ip scan for optimization of fNI and fusion performance G; (2) reliable edge (r = 0.8 â?? 1.0) profiles including pedestal
Resource Requirements:	NBI: >5 co sources + 210RT for balanced MSE EC: >4 gyros FW: (90 MHz and 60 MHz) >2 MW desirable
Diagnostic Requirements:	Fast ion diagnostic (UC, Irvine), MSE (LLNL), edge reflectometer (UCLA
Analysis Requirements:	Scenario modeling with GLF23/ONETWO, TRANSP/GLF23, ONETWO analysis; CURRAY/ONETWO
Other Requirements:	Multiple days

Title	398: Extend fully non-inductive high-beta operation
Name:	Murakami masanori_murakami@att.com	Affiliation:	Retired
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):	T. Luce, J. Ferron, JM Park, C. Greenfield, E. Doyle, et al.	ITPA Joint Experiment :	No
Description:	Use 30RT to extend 100% noninductive, high performance for 2*tauR (~4s) With ptimized ECCD ; add FWH&CD	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Start with 133103 (or 129830) 2) Bring the start of the high power phase earlier to optimize q-profile and save NB energy 3) Work on PCS control of substituting the low power phase NB power 4) Optimize ECCD for 2/1 stability 5) Add FW heating and CD
Background:	Sustainment of 100% NI, high-beta operation has been limited by co NB deliverable energy, not EC energy in 2007 and 2008 â?¢ Hopefully toward end of the 2009 campaing, the No. 8 source will be available for the experiment to make optimized use of NB I for this purpose. Optimization of PCS control is needed * Optimization of ECCD location and distribution leads to operation at higher sustained betaN â?¢ Plasma shape has been optimized for the best confinement in 2008 operation â?¢ Further improvement in error correction may need.
Resource Requirements:	NBI: all co-sources plus 210RT for balanced MSE EC: Maximum number (5-6) of gyrotrons FW: fast wave (both 60 and 90 MHz) is desirable
Diagnostic Requirements:	Fast ion diagnostic (UC, Irvine), MSE(LLNL), edge reflectometer (UCLA)
Analysis Requirements:	Scinario modeling with GLF23/ONETWO, TSC/TRANSP, ONETWO analysis; RF codes â?? CURRAY/ONETWO, AORSA, RANT3D;
Other Requirements:	Multiple days

Title	399: Complete prototype off-axis NBCD
Name:	Murakami masanori_murakami@att.com	Affiliation:	Retired
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	J.M. Park, C. Petty,W. Heidbrink, T. Osborne, etc	ITPA Joint Experiment :	Yes
Description:	â?¢ Off-axis NBCD with small, up-shifted plasma, reverse BT â?¢ Vertical position scan â?¢ Gather sawtooth behavior â?¢ Take normal BT data if an opportunity arrives	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	â?¢ Continuation of the last yearâ??s experiments of off-axis NBCD using a small, upward-shifted plasma in reverse BT direction â?¢ Vertical position scan â?¢ Take data for NBCD and sawtooth behavior â?¢ Look for opportunity in the normal BT direction
Background:	the last yearâ??s protoff-axis NBCD experiment demonstrated that: (1) the existence of off-axis NBCD (2) the predictions of sensitivity to the alignment of NBI to the magnetic field line pitch by reversing BT or moving the plasma from down-shifted to up=shifted poition (3) measured off-axis NBCD is approximately linear to the injected co-power and only a modest ad-hoc fast ion diffusivity (<0.5 m2/s) is needed to explain high pwer CD and FIDA data â?¢ Since ITER off-axis NBCD system relay on combination of a vertical plasma shift and an oblique (â??downwardâ??) injection. However, ITER operated in the normal BT (CW) direction, the planned off-axis NBCD is in the â??unfavorableâ?? alignment configuration. â?¢ A limited number of experiments would be worth doing in better understanding the role of the vertical shift in the off-axis NBCD. â?¢ Since the, we plan to start of with the up-shifted plasma with reverse BT for better diagnostics. â?¢ Take data at a few number of intermediate vertical positions. This vertical position scan provide us with additional ITER-relevant data â?? sawtooth control (or destabilization) using off-axis NBCD (JET, IAEA08, I. Chapman)
Resource Requirements:	NBI: 3-LT-co sources + 210RT counter EC:: 4 gyros FW: No Reverse BT, up-shifted plasm operation
Diagnostic Requirements:	FIDA (UC, Irvine), MSE with counter-system (LLNL) Soft X-ray data (with tomography) critical; ECE, including oblique ECE
Analysis Requirements:	Scenario modeling with TRANSP, NUBEAM/GLF23/ONETWO
Other Requirements:

Title	400: ITER accessibility to hybrid regime using RF
Name:	Murakami masanori_murakami@att.com	Affiliation:	Retired
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	R. Pinsker, C. Petty, W. Baity, A. Negy, D. Rasmussen, J. Hosea, J.M. Park, et al.,	ITPA Joint Experiment :	Yes
Description:	1) Study the accessibility of hybrid regime using substantial RF power in ITER-shape plasmas 2) Sustain q(0)>1 with FWCD while 3/2 or 4/3 tearing modes feedback stabilized 3) Gain experience of FW operation with ELMing H-mode with a relatively large gap 4) RF edge diagnostic development â?? RF reflectometer	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Scenario modeling study to seek optimal strategy for accessibility and sustainment of hybrid scenario in ITER-shaped plasma in DIII-D 2) Reproduce shot like 119733, and apply counter-FWCD 3) Adjust power to find L-H threshold power, and lowest betaN for hybrid operation 4) Settle at a baseline (ITER-like) condition and characterize RF operation (optimal coupling, etc) Study maintenance of q(0)>1 with FWCD while ECCD stabilize 3/2 or 4/3 NTM
Background:	* Important question for ITER is its accessibility to the hybrid regime with the Day-1 hardware capability (33MW NB and 20MW IC and EC each). P_loss/P_threshold; and betaN are not necessarily in the experimentally proven ranges for hybrid. For example, recent ITER simulation by C. Kessel indicates that fluence limit falls into betaN <2 which is outside the present experimental 2.5â?¢ Sustainment of high fusion performance relies on q(0)>1 with tearing modes. The stabilzation of tearing modes turned into sawtoothing discharges. Can we use central FWCD or ECCD to replace tearing modes for prevention of sawteeth? â?¢ ITER Hybrid simulations by J.M. Park shows counter-FWCD capability in Day-1 is enough to raise q(0) above 1. Similarly simulations with GLF23/ONETWO modeling using counter-FWCD showed qmin (= q0) can be sustained above 1. â?¢ ELMing H-mode with relatively large outer gap (8-9 cm) in ITER is a challenge for ICRF operation. Characterization of the hybrid edge and optimal coupling of RF power would help the ICRF technology and simulation
Resource Requirements:	NBI: 7 EC: 4 FW: fast wave (both 90 and 60 MHz) -- >2 MW coupled
Diagnostic Requirements:	Fast ion diagnostic (UC, Irvine) edge reflectometer (ORNL, UCLA )
Analysis Requirements:	Scenario modeling with GLF23/ONETWO, TSC/TRANSP, ONETWO analysis; RF codes â?? CURRAY/ONETWO, AORSA, RANT3D;
Other Requirements:

Title	401: The Role of Fast Ion Loss in ELM-Free Plasmas
Name:	Zhu none	Affiliation:	UC, Irvine
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	W.W. Heidbrink	ITPA Joint Experiment :	No
Description:	To further investigate the roles, e.g., appearance conditions, characteristics and control techniques, of fast ion loss in the ELM-free plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Piggy-back with ELM suppression/pacing experiments 2) Some extra plasma shots, if possible
Background:	A new type of high frequency, broadband burst of fast ion loss is recently observed during counter-QH-mode experiments (APS2008, Y. Zhu et al). This kind of non-prompt loss usually accompanied with EHO and broadband MHD. It is more sensitive to perpendicular than tangential beams. It has been observed in counter-QH and RMP ELM-suppressed plasmas, but has never been observed in co-QH-mode plasmas.
Resource Requirements:	ELM suppression/pacing plasmas
Diagnostic Requirements:	Fast ion loss collectors, FIDAs, neutrons and fast magnetic pickup coils
Analysis Requirements:	ORBIT, PENCIL code ......
Other Requirements:

Title	402: Fast wave heating and current drive in AT plasmas
Name:	Murakami masanori_murakami@att.com	Affiliation:	Retired
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	C. Petty, R. Pinsker, W. Heidbrink, W. Baity, A. Negy, G. Hanson, T. Bigelow, and Adv. Scenario Develop Team	ITPA Joint Experiment :	Yes
Description:	Robust coupling to AT plasma (2) Characterize beam ion absorption of FW power at 90HZ and 60MHz (3) Validate scenario modeling of FWCD in AT (4) RF edge diagnostic development -- reflectometer	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Reproduce shot #123159 and/or baseline NBI-plasmas ) [BT=1.85T, Ip=1.2 MA,nebar=3.5e19, NB=9.3 MW] 2) Study q0 control using co- and counter-FWCD in the AT plasmas. 3) Use PCS control of q0 if available. The push-pull operation with two ICRF systems is highly desirable.
Background:	There was significant improvement in the FW 90-MHz (ABB1 and 2) systems, achieving 2.5 MW FW operation in L-mode plasmas. High (6-th) harmonic beam ion absorption experiments with L-mode plasmas showed that primary absorption of 90 MHz was via electron damping, and very little higher harmonic absorption of beam ions were observed (in ECE, stored energy, neutron and FIDA data.). This is potentially excellent news for application of FWCD in AT regime. This is a very active area of investigation in RF modeling. There are considerable differences in results from Monte-Carlo ORBIT-RF codes, and the full-wave AORSA code combined with the CQL3D Fokker-Planck code â?¢ FW coupling and heating experiments in H-mode with RMP-stabilized ELMs showed central electron heating by FW with 5 cm outer gap. â?¢ There was a less than one-day experiment for the AT regime in 2007. Unfortunately the ABB1 high voltage power supply failed prior to the higher power FW operation, so that only the ABB2 system was available. In addition, because of the constraints due to the interim arc protection system, substantial portion of the power is routed out to the dummy load during the ELMs, and effectively only about a half of the source power was estimated to be delivered to AT plasma. There was a good AT discharge with RF (ABB2 source power of 0.75 MW) with 9 cm out gap. It is not clear the FW contribution to the excellent performance of the shot. â?¢ The most important issue for FW experiments is robust FW heating for ELMing H-mode and AT discharges. Existing FW antenna design does not couple effectively to plasmas with large outer gaps and ELMs. We need to devise more efficient FW operation scheme. â?¢ Scenario simulations promise potentials of the FW in steady state scenario: central heating to improve off-axis CD efficiency; increase in bootstrap current fraction; and better control of q-profile near the axis. Modeling incorporated with feedback algorithms for NBI and FW power showed that modulating input NB power and FW co- and counter CD fraction allowed sustaining favorable q profiles with controlled central shear (q0 â?? qmin) equal to â??0.5.
Resource Requirements:	NBI: >= 4 co-beams + 210RT EC: 4 gyrotrons FW: 60MHz + 90 MHz, >2MW
Diagnostic Requirements:	Fast ion diagnostic (UC, Irvine) edge reflectometer (ORNL, UCLA)
Analysis Requirements:	CURRAY/ONETWO, TRANSP, ONETWO; SciDAC-RF [AORSA/CQL3D; ORBIT-RF/TORIC, etc]; RANT3D
Other Requirements:

Title	403: Modulation of ECCD for 2/1 NTM suppression
Name:	Welander welander@fusion.gat.com	Affiliation:	GA
Research Area:	NTM Stabilization including Rotation Dependence	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 the benefit of modulating the electron cyclotron current drive (ECCD) for suppression of the m/n = 2/1 NTM and to scan how ECCD suppression rate 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. The plan is to restore shot 132574. 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 Î?Ï? = 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 realtime 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 will be readily extendable 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	404: Effect of rotation, nonresonant field perturbation, betap, and triangularity on ELM size
Name:	Osborne osborne@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	N. Oyama (JT60-U), M. Fenstermacher, A. Leonard	ITPA Joint Experiment :	No
Description:	This will look at the effect of rotation and non resonant field perturbation on ELM size, particularly in relation to the grassy ELM regime which was obtained on JT60-U.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Starting with discharge similar to 128464 (q=7.5 dicharge which showed large reduction in ELM size with large non-resonant I-coil field) explore the effect of rotation and nonresonant field perturbation by varying CO/CNTR bream mix and I-coil current. Effect of betap and triangularity/closeness to double null on ELM size would also be explored. It would be highly desirable to obtain data, possibly from the fast UCSD camera to identify any difference in ELM structure through the hoped for variation in ELM size.
Background:	ELM size has been tied to plasma rotation, betap and triangularity at low collisionality on JT60-U. In particular a regime of very small grassy ELMs was obtained with rotation low CO injection rotation, higher betap, higher triangularity discharges. Also on JT60-U, adding the ferritic inserts to reduce toroidal field ripple increased ELM size. THe grassy ELM regime is the only small ELM regime which offers high pedestal energy at low collisionality and is therefore attractive for future tokamaks. A search of the DIII-D archive did not reveal the presence of this small ELM regime at high betap and triangularity, with ELM size for low collisionality DIII-D dischages genarally > 10% of the pedestal energy. In RMP discharges with large resonant field component (q=3.7) when ELMs are not completely eliminated the ELM size is not reduced as a fraction of pedestal energy. However in a few discharges at high q (7.5) which used a secondary maxima in the resonant field to reduce the pedestal pressure there was a strong reduction in ELM size as a fraction of pedestal energy. The I-coil was producing a strong nonresont field in these discharges, the betap was somewhat higher due to the higher q value, and the rotation was reduced. The requirements of very small ELMs should make study of small ELM regimes a high priority.
Resource Requirements:	I-coil probably in odd parity. CO and CNTR NBI sources.
Diagnostic Requirements:	CER and TS essential. UCSD fast camera, BES.
Analysis Requirements:	Profiles, kinetic EFIT, ELITE.
Other Requirements:	N. Oyama would visit GA to participate in the experiment which may affect scheduling

Title	405: Experimental investigations on Current Driven RWM feedback control
Name:	Marrelli lionello.marrelli@igi.cnr.it	Affiliation:	Consorzio RFX
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	P.Martin, P.Piovesan, L.Piron	ITPA Joint Experiment :	No
Description:	A) Changing feedback control parameters in time windows during the discharge In current driven RWM stabilization experiments, I-coils are both correcting error fields and stabilizing the modes. In order to better distinguish the two mechanisms it is proposed to modify the feedback law in a short time window around periods in which RWM are unstable. 1) complete switch-off of the I-coils currents 2) Freezing of the I-coils during the same unstable period 3) I-coils currents are feedback controlled A reproducible RWM growth is expected in experiments 1). If RWM modes are not observed in 2) then the error field correction is the dominant stabilization mechanism. On the other hand, if only in 3) RWM are stabilized, then feedback is the dominant mechanism. B) Experiments with independent power supplies for upper and lower I-coils (n=1 only) Previous RWM stabilization experiments were performed by connecting in series upper I coils and lower I-coils, in order to use the same power supply units and drive more current. This approach limits the flexibility in setting the relative phase between the n=1 fields produced by the upper and lower set of coils. It is proposed to drive independently the upper and lower I-coils, so that the relative phase can be prescribed more precisely and possibly matched to the value of q95. 1) experiments may be performed with a pre-programmed phase delay between the upper and lower I-coils commands. 2) if possible, it may be interesting to set different phases delays in different time windows, in order to match the expected q95 evolution. If the matching of the phase delay to q95 will allow a more efficient stabilization (with lower I-coil currents, e.g.) this may motivate the development of a feedback algorithm using q95 in real time. 3) The phase delay may be estimated in real time, by using different set of sensors (for example mid plane poloidal sensors MPID) and upper poloidal sensors (UMPID)). This may be an alternative strategy to explore, in order to better react to changes in the mode poloidal structure when q95 changes. 4) these experiments may be useful to deal with violation of "mode rigidity", i.e. the helicity switch as predicted by the NMA code [Okabayashi, IAEA 2008]. C) Experiments with "complex gain" It has been found that complex gains in the Co-Ip direction were effective for RWM stabilization. It is proposed to apply complex gains only during a prescribed time window, as proposed in A) with the upper and lower I-coils independently fed, as in B). The main questions to be addressed are: Is the stabilization linked to the RWM mode rotation? Are complex gains stabilizing the RWM or improving the error field correction? Is there an optimal combination of derivative gains and complex gains? If total stabilization does not occur, is the phase velocity of the mode affected by the complex gain? Does the effect on mode phase changes when an island appear?	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:	Plasma current ramp experiment allows the reproducible excitation of current driven RWM, allowing to investigate different feedback stabilization algorithms.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	406: ECRH at the 3rd Harmonic
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	R.Prater, J.Lohr	ITPA Joint Experiment :	No
Description:	Demonstrate efficient 3rd harmonic X-mode (X3) heating, both for fundamental and applicative reasons (heating at high beta).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Start with old discharge with 40% damping at 3rd harmonic and 60% at 2nd. Reduce TF to move X3 inward. The higher Te and ne will result in higher optical thickness tau and therefore stronger absorption at X3. The enhanced local heating will further, non-linearly increase tau (which for X3 scales like Te^2 rather than Te) and therefore the absorption coefficient. Increasing ne would also be beneficial. The goal is to reach 80% X3 absorption at mid-radius. This can be evaluated by ECH modulation and heat-pulse analysis. ECE signals might be difficult to interpret because of harmonic overlap, with the outer harmonic being grey. SXR will be diagnostic of choice. The measured absorption bell will be translated in quantitative estimates of tau, to be compared with toray calculations.
Background:	Up to 40% of ECRH power has been damped in the past on the 3rd harmonic at the edge of DIII-D. This is generally considered a nuisance, which damps at the edge energy meant for X2 heating in the core. For this reason, X3 heating is generally minimized by adjustments of the TF, plasma shape and launch direction. The proposal here is to apply the opposite adjustments, in order to maximize the X3 heating. These and other changes would further increase the X3 optical thickness tau, which is already fairly high. As a result of high tau, there is no need for shallow injection toward the resonance to lengthen the absorption path like in TCV. X3 might represent a valuable heating method at low TF and high Te, accessing high densities at which X2 is cut off. The combination of low TF and high pressure makes it appealing for applications to high beta. Note that the 2nd harmonic would act as a view damper for non-absorbed power. Hence, stray ECRH would be low and not endanger diagnostics. Pure X3 heating and CD can also be tried, though: at low TF and for strongly oblique (toroidal) launch the gyrotron beams would cross the X3 layer twice, one from the outside, the other from the inside. The double pass has a potential for nearly complete absorption.
Resource Requirements:	4 gyrotrons
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	407: Electron Bernstein Wave Studies
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Demonstrate Electron Bernstein Wave (EBW) emission, heating and current drive, for the first time in a big tokamak, utilizing the Ordinary-eXtraordinay-Bernstein (OXB) mode conversion.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) angular scan of EBW emission with the oblique ECE 108-112GHz radiometer, as piggyback on pellet or other high density discharges; 2) dedicated H&CD attempt at 110GHz, assisted by pellet or other means of achieving ne>1.5e20m-3 in the core: use max ECH power but modulated with low duty-cycle, to avoid excessive stray and damages to diagnostics. Perform a shot-to-shot angular scan in steps of 2deg around calculated optimum. Heat-pulse-analyze SXR and stored diamagnetic energy (ECE will be in cutoff) looking for evidence of EBW-generated heat waves.
Background:	At DIII-D, conventional 2nd harmonic X-mode (X2) ECH and ECCD at 110GHz, as well as a considerable number of channels of the ECE radiometer, go in cut-off at ne=7.5e19m-3. If the local density could non-disruptively, e.g. by pellets, be raised up to ne>1.5e20m-3, an additional microwave heating, CD and diagnostic scheme would become available. This makes use of OXB-converted EBWs, requiring the presence of the O-mode cut-off in the plasma, steep ne gradients (making the ne scale-length comparable with the wave-length) and a special view/launch direction, accessible to the DIII-D launchers. EBWs were used with success in stellarators and spherical tokamaks but, so far, only in one conventional tokamak: TCV. It is the aim of the present proposal to extend it to larger tokamaks such as DIII-D, also in view of testing the feasibility of EBW-based divertor diagnostics being considered for ITER. The high densities required for this experiment are a consequence of exploiting the existing hardware at 110GHz, i.e. at the second harmonic. In case of encouraging results, one might consider radiometric measurements and eventually heating at 60GHz, i.e. at the first harmonic. The requested density in that case would be 4.5e19 m-3, which is routinely accessed at DIII-D.
Resource Requirements:	pellet, 5 gyrotrons
Diagnostic Requirements:	Oblique ECE
Analysis Requirements:	toray, with modifications
Other Requirements:

Title	408: ECH effects on pedestal and ELMs
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	J. Callen, J. Lohr, T. Osborne, R. Prater et al.	ITPA Joint Experiment :	No
Description:	Understand/characterize density pump-out, its balance with heating effects and their combined effect on the height of the pedestal shoulder and on the steepness of the pressure gradient. Mix these effects in a stabilizing/destabilizing way on marginally ballooning-unstable/stable plasmas, respectively. Repeat for type-I, type-II and type-III ELMs.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Deposit ECH on low field side (LFS). Scan rho~0.85-1.05 in small steps of Drho=0.02-0.05. Effect on ELM frequency and amplitude should be immediately visible. For its interpretation in terms of heating, pump-out ad ballooning stability, diagnose edge density, temperature and current.
Background:	A similar experiment was carried out with success at DIII-D by J.Lohr et al. in the late 80's and early 90's (Stambaugh et al., PPCF 1988; T.Luce et al., IAEA 1990). 1.2MW of ECH at 60GHz were launched from the high field side (HFS) and deposited on the LFS. Deposition inside/outside the separatrix was observed to halve/double the ELM period; no ECCD was attempted at that time. The idea would be to repeat the experiment with the new 110GHz system and with the improved edge diagnostics (improved TS, to measure the edge pressure, and MSE, to measure the edge current, among others) and codes (ELITE) that became available in the meantime. These offer the perspective of a deeper understanding of peeling-ballooning physics and of density pump-out. From the point of view of MHD control, the minimum goal would be to reproduce and possibly improve past results, thanks to the increased ECRH power, the improved focusing and the fact that gyrotron beams would be launched from the LFS and absorbed on the LFS, thus they would broaden less. Also, ELMs develop predominantly on the LFS, so intuitively they should be attacked on the LFS. Additional reasons to prefer LFS deposition come from the scenario development: the scenario for LFS deposition was found to be more resilient to mode locking and shape control issues, in a more recent attempt at DIII-D, on April 4, 2008. Those ECH experiments at rho=0.82-0.94 saw increased ELM frequency but also increased energy-loss per ELM. The balance between pump-out and heating at various radii might reconcile these results with earlier Lohr's experiments, but need further experimental tests. There are 4 radii of interest for a scan of ECH deposition: the pedestal shoulder, the centre of pedestal region, the separatrix and the "origin" of ELMs. Hence, time permitting, 12 radii should be considered for a comprehensive scan: the 4 mentioned before, 4 positions shifted upstream, and 4 downstream. TORAY calculations suggest that the deposition will be sufficiently localized for this fine scan to make sense. Fast diagnostics should monitor what happens to the ELM as it passes through the ECH deposition region. Finally, April 2008 experiments serendipitously found a QH-mode in full co-injection that needs better characterization and optimization and, at the same time, might represent an easier, more quiescent ELM-free target where to isolate and contrast heating and pump-out effects of ECH without the complication of ELMs and associated changes of plasma parameters.
Resource Requirements:	5-6 gyrotrons
Diagnostic Requirements:	edge TS, MSE
Analysis Requirements:	ELITE
Other Requirements:

Title	409: Oblique ECE for radial alignment during NTM suppression
Name:	Welander welander@fusion.gat.com	Affiliation:	GA
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):		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 based on Mirnov data.	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 two frequencies, one directly above and one directly below the ECH frequency. This allows detection of the temperature fluctuation 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 effecient 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 both ECE channels are on one side of the island and also that the Mirnov data has much less noise.
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 very desirable to improve the signal-to-noise-ratio of the oblique ECE diagnostic. For step 3: 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	410: ECCD and Ohkawa CD stabilization of marginally peeling-unstable ELMs
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	J. Callen, J. Lohr, T. Osborne, R. Prater et al.	ITPA Joint Experiment :	No
Description:	Drive small edge currents both via the conventional Fisch-Boozer mechanism and the less exploited, for ECCD, Ohkawa effect, which is expected to dominate at rho>0.95. Demonstrate and characterize the effect of edge currents (both co- and ctr-) on the pedestal and on ELMs. Exploit small edge CD as a fine knob to investigate peeling stability and marginal stability.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The shot plan and diagnostic requirements are very similar to ECH Proposal No.408, except that the launch will be oblique, for ECCD. Compare co/ctr, expected destabilizing/stabilizing, respectively.
Background:	Unlike ECH, the use of ECCD for controlling ELMs has never been systematically experimented, except for sporadic attempts at AUG and JT-60U. Reasons include unfavourable CD estimates. At DIII-D, for example, it is estimated that 2MW can drive 1-10A/cm2 at rho>0.9. Typical bootstrap (BS) current densities are much higher. However, plasmas were identified with edge BS current densities of 50A/cm2, in which the ECCD might act as a "dwarf on giant's shoulders", the giant being the BS current and, we hope to demonstrate, ECH control. ECCD control could add fine adjustments at zero cost: once ECH is used, a toroidal tilt of launch would result in ECCD too, with no deterioration of ECH heating and, presumably, pump-out. Note that although ECH effects on ne and Te at the edge are significant, their product -the effect on pressure- is modest. For this reason it will be worth assessing ECCD effects on ELMs: they might play a role comparable with or not-much-smaller than ECH. Besides control, ECCD might find application as a tool for fundamental studies of peeling stability and marginal stability. For instance, for strong plasma shaping the peeling-ballooning stability boundary should "bulge" and give rise to a bifurcation and a second instability region at low currents and high pressure-gradients which hasn't been proved experimentally yet, even because the two requirements are in contradiction: high pressure gradient is coupled with high (BS) currents. Ctr-ECCD, however, could be used to reduce the latter and provide 2D experimental flexibility. The experiment should begin with a marginally (un)stable plasma, that a small EC-driven current can (de)stabilize. For this reason, the experiment could be combined with experiments aimed at tracing the peeling-ballooning stability boundary. Finally, it is expected that significant currents can be driven at extreme radii, rho>0.95, by the Ohkawa CD mechanism, based on asymmetric de-trapping in the velocity space. Its experimental identification is made possible by the fact that Ohkawa CD's direction is opposite to conventional Fisch-Boozer ECCD. The (de) stabilization of ELMs would be a simple and spectacular evidence of this change of CD direction.
Resource Requirements:	5-6 gyrotrons
Diagnostic Requirements:	MSE
Analysis Requirements:	ELITE
Other Requirements:

Title	411: Filling with ECCD an unstable ELM current-hole
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	J. Callen, R. Groebner, T. Osborne, R. Prater et al.	ITPA Joint Experiment :	No
Description:	Use ECCD to modify a special feature of the edge current profile and make it less ELM-unstable.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Restore an ELMy shot with modest bootstrap (BS) current at the edge. Use MSE data and control-room analysis to radially localize the "hole" in the current profile between the bulk core current and the edge BS peak. Repeat with co-ECCD at that location, and scan around it. Repeat with ctr-ECCD for comparison. Look for changes in the inter-ELM period in D_alpha.
Background:	Li-beam measurements, diagnostic-constrained EFIT reconstructions and ELITE modelling of the current profile exhibit a â??holeâ?? between the core and the edge peak. A rapid transition to a smoother profile is observed at ELM crashes. Later, the previous profile recovers, slowly, and eventually crashes again. It is proposed here that co-ECCD at rho=0.8-0.9 might fill or reduce such a hole, and the current profile be more stable as a result. The increased stability should manifest itself as a delayed crash, or possibly complete ELM avoidance.
Resource Requirements:	5-6 gyrotrons
Diagnostic Requirements:	MSE
Analysis Requirements:	ELITE
Other Requirements:

Title	412: ELM-pacing by modulated ECH/ECCD
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):	J.Lohr, R.Prater	ITPA Joint Experiment :	No
Description:	Demonstrate that modulated ECH/ECCD can periodically destabilize the edge pressure and gradient, and so trigger ELMs	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Begin by setting a value of rho which is known or expected to maximize ECH/ECCD effect on ELMs. Modulate ECH/ECCD 10% slower/faster and a factor of 2 slower/faster than the natural ELM frequency. Compare perpendicular, co- and ctr- launch. Initial tests can piggyback on RMP control of ELMs. ECH pulses of 200ms towards the end of the discharges are brief enough not to damage diagnostics and other equipment in case absorption is only partial and stray is high. At the same time, they are long enough to see effects on ELMs, if any, including slow changes following transport and relaxation phenomena. Dedicated run-time could be shared with continuous ECH/ECCD proposals no.408 and 410, by adding modulation at the end of the continuous pulses. The advantage of exclusively dedicated time, not shared with others, is that different ECH frequencies can be tested in the same discharge, at different times. Finally, if ELM-frequency control succeeds, the other important parameter to control is the energy loss per ELM. The knobs in this case are the radius of ECH deposition and the amount of driven current. This is why, as a final experiment we propose to fix the ECH frequency and scan, on a shot-to-shot basis, the radius of deposition and the toroidal direction of launch.
Background:	In 2004, AUG modulated ECH at 100Hz at the edge of an ELM-ing plasma. The ELM frequency, initially of 150Hz, changed accordingly. Also, as expected, the ELMs slightly intensified. The opposite change, i.e. making the ELMs smaller and more frequent, has not been demonstrated yet, and would be highly desirable for ITER. Moreover, in AUG the ECH effects were suspected to dominate over ECCD, but the two were never really disentangled. Although seminal, AUG results leave much room for improvement and for the first demonstration of ELM-pacing by modulated ECH/ECCD at HIGHER frequencies.
Resource Requirements:	5-6 gyrotrons
Diagnostic Requirements:
Analysis Requirements:	ELITE
Other Requirements:

Title	413: ECH/ECCD modulated in the rotating ELM filament
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	M. Henderson (ITER)	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 doesn'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:	5-6 gyrotrons
Diagnostic Requirements:	Modified D_alpha diagnostic. Modified oblique-ECE "box" in the annex.
Analysis Requirements:	ELITE
Other Requirements:	--

Title	414: The full Monty: A complete ITER ohmic discharge
Name:	Jackson jackson@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):	T.C. Luce, D.A. Humphreys	ITPA Joint Experiment :	No
Description:	Simulate experimentally an entire ITER Ohmic discharge, using large bore startup	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	With 3v ohmic startup, obtain an ohmic flattop. Use tools available (e.g. gas puffing) to maintain a sufficiently low li to avoid VDE, then rampdown
Background:	The initial campaigns in ITER may be done with minimal or no auxiliary heating. Hence it is important to demonstrate that ITER can operate ohmically without disruptions or VDEs
Resource Requirements:	NB "beamlets" for mse and CER, ECH on standby if problems develop during startup. PCS li feedback
Diagnostic Requirements:
Analysis Requirements:	JFIT. A realtime diagnosis of n=0 stability is desireable. Corsica, EFIT, TOKSYS modeling of the discharge
Other Requirements:

Title	415: H-mode pedestal width in ITER demonstration Discharges
Name:	Osborne osborne@fusion.gat.com	Affiliation:	GA
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	R. Groebner, A. Leonard, P. Snyder	ITPA Joint Experiment :	No
Description:	Using the fact that the edge transport barrier width is independent of rhostar, operating at the pedestal collisionalities and perhaps Mach numbers of ITER should should reveal the pedestal structure to be expected in ITER.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Guided by the EPED1 model prediction for pedestal beta we would attempt to obtain data in each of the 4 ITER scenario demonstration discharges at consistent collisionality with the pedestal density in ITER set by the required fusion power output. The experiment would produce a set pedestal widths vs pedestal nustar, beta, and perhaps Mach number in the range of what is required for ITER without recourse to further scaling. Given the possible dependence of pedestal structure on rotation and the uncertainty of expected rotation in ITER a reange of rotations will be produced varying Co/Counter beam mix.
Background:	Experiments on DIII-D and JET indicate that the edge transport barrier width is independent of rhostar. Given this result ITER demonstration discharges in DIII-D should exist with H-mode pedestals matching all other dimensionless parameters with ITER pedestals.
Resource Requirements:	ITER demonstration discharges with divertor pumping to low density to reach ITER pedestal collisionality. 210 beam sources
Diagnostic Requirements:	TS and CER at least
Analysis Requirements:	Profile, kinetic EFIT, ELITE, EPED1 modeling
Other Requirements:

Title	416: Reynolds stress peak in momentum counter-injection H-modes
Name:	Muller mullersh@fusion.gat.com	Affiliation:	UCSD
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	J. Boedo, R. Moyer, D. Rudakov, G. Tynan, J. DeGrassie, W. Solomon	ITPA Joint Experiment :	No
Description:	In the 2008 runs, the Reynolds stress head of the midplane probe was used successfully to measure Reynolds stresses and flows across the SOL up to 1 cm inside the separatrix in H-mode discharges with co-, counter- and balanced momentum input. The results indicate that the flow profile remains mostly unchanged throughout the SOL and changes only inside the LCFS. Similar results of a mostly unchanged SOL with strongly different core rotation profiles have been reported from the TCV tokamak in L-mode. While the SOL flows are mostly unaffected by the change in momentum input, the measured Reynolds stresses indicate a strong peak inside the LCFS in the counter-injection case. Such a signature would be expected if turbulent momentum transport was responsible to remove the excess momentum from the core plasma. As only 3 shots were available for this experiment in 2008, it was not possible to reach a conclusive result. The present proposal aims at complementing the 2008 experiments to verify the reproducibility and parameter dependence of the 2008 preliminary results.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use experimental conditions as in shot 133053. Two ECH sources bring the lower-single null configuration into H-mode. Two derated NBI sources are used in co-, counter and balanced configuration. In these conditions, the midplane probe can breach the separatrix by up to 1 cm. Verify that peak in Reynolds stress in counter-injection case can be reproduced. Change NBI to ECH power balance and study behavior of Reynolds stress and SOL flows. Attempt same experiments with NBI only. Vary plasma parameters and study behavior of Reynolds stress profiles.
Background:	--
Resource Requirements:	ECH (2 sources), NBI (co, counter and balanced)
Diagnostic Requirements:	All edge and SOL diagnostics. In particular midplane probe, X-point probe, CER, Edge-Thomson.
Analysis Requirements:	CER
Other Requirements:	--

Title	417: ECE Imaging of ELM-NTM coupling
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The ECE imaging camera being developed by U.C. Davis and collaborators, and scheduled to be installed in August of 2009, 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:
Diagnostic Requirements:	ECE imaging diagnostic, to be installed in August of 2009, is required.
Analysis Requirements:
Other Requirements:

Title	418: Dimensionless parameter scaling of the H-mode edge transport barrier width.
Name:	Osborne osborne@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	M. Beurskens (JET), L. Horton(AUG), R. Groebner, A. Leonard, P. Snyder	ITPA Joint Experiment :	Yes
Description:	In this experiment we will carry out dimensionless parameter scans of the H-mode pedestal rhostar, q, nustar, and beta to determine the scaling of the edge transport barrier width.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This experiment completes and extends an experiment in the 2008 campaign in which H-mode pedestal rhostar scans were carried out in DIII-D and JET with the high rhostar point in DIII-D matching all the dimensionless parameters of the low rhostar point in JET. Part of this experiment will be to complete the 2008 experiment by obtaining points at intermediate rhostar and obtaining better quality data at the large rhostar point in low triangularity discharges. JET has also now completed a rhostar scan at high triangularity which we would also match with a rhostar scan in DIII-D in this experiment. Both DIII-D and JET independently and in last year's comparison of the two machines find no dependence of the ETB width on rhostar. With this result in hand scans can be devised to test beta, nustar, q, and Mach number scaling of the pedestal width.
Background:	Results of independent experiments on DIII-D and JET indicate no dependence of the H-mode edge transport barrier width on rhostar. In experiments last year a rhostar scan was carried out in JET with the low rhostar point in JET matched to a high rhostar point in DIII-D. We obtained data at large rhostar in DIII-D at B=0.75T to produce a factor of 4 change in rhostar with the two machines combined. Only 2 discharges were obtained under otherwise matched conditions and the quality of the data was rather poor. As part of this experiment we would return to this lower triangularity discharge and again obtain data at 0.75T but also data at B=1T for an intermediate rhostar value. We would also repeat the rhostar scan matching a JET scan at higher triangularity which has now been completed. Given a lack of dependence on rhostar it is possible to scan the other dimensionless parameters in a single machine. Indeed a scan of collisionality occurred as part of the attempt to obtained matched discharges in last years experiment. There was not obvious dependence of the ETB width on collisionality based on this roughly factor of 2 variation in collisionality. In the new experiments we would extend the collisionality range and also test scaling with betaped where previous results have indicated a betapped**0.5 scaling and also q.
Resource Requirements:	Discharges matched to JET at high and low triangularity.
Diagnostic Requirements:	TS and CER at a minimum.
Analysis Requirements:	Profiles, kinetic EFIT, ELITE. EPED1
Other Requirements:	Marc Beurskens may visit from JET to participate. Lorne Horton may visit from AUG to participate.

Title	419: Effect of ECH/ECCD on H-mode Pedestal Characteristics and ELMs
Name:	Osborne osborne@fusion.gat.com	Affiliation:	GA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	J. Callen, J. deGrassie, R. Groebner, A. Leonard, J. Lohr, G. McKee, R. Prater, T. Rhodes, P. Snyder, F. Volpe	ITPA Joint Experiment :	No
Description:	In this experiment we will exploit the localized transport, heating, and current drive effects of ECH/ECCD to modify the H-mode pedestal structure and ELM characteristics. The results of this experiment will help to improve our understanding of H-mode pedestal structure and the factors controlling ELM characteristics. The experiment may also suggest an avenue toward pedestal and ELM control with the potential of providing an additional technique for ELM suppression which is a prime requirement for ITER. The experiment could also provide data for testing models for ECCD.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	ECH will be used to heat and drive current in the H-mode pedestal. We will look for effects on the pedestal structure and ELMs which might result from breaking the usual connection between edge bootstrap current and pressure gradient or from ECH density pumpout. We will scan the ECH/ECCD through the pedestal region, change the direction of the current drive, and drive current in region near the X-point and far from the X-point.
Background:	ECH has been previously demonstrated to have interesting effects on the H-mode pedestal and ELMs. Under some conditions ECH results in a significant reduction in the plasma density including in the H-mode pedestal region. It is conceivable that there is a regime where ECH induced pedestal particle transport would also result in steady state ELM free discharges. Previous results on DIII-D [J. Lohr, et al.] have shown that the ELM frequency can be decreased by applying ECH heating outside the separatrix. In many cases on DIII-D and other tokamaks it appears that the pedestal pressure gradient saturates well before an ELM. Modeling and limited Lithum beam measurements [D. Thomas et al.] suggest that the pedestal current density continues to increase probably through resistive relaxation of the Ohmic component after the pressure gradient saturates. This suggests that under some conditions the natural pedestal transport is sufficient to keep the pressure gradient below the ballooning limit but that the edge current density eventually reaches the peeling limit, triggering an ELM. Under these conditions we propose using ECCD to break the connection between the edge pressure gradient and current density through the bootstrap current. Here we would look for changes in the ELM frequency with co versus counter ECCD. It is possible there is a regime where counter ECCD would result in an ELM free steady state discharge. Magnetic shear may play a role in setting the pedestal width leading to an unstable condition where more edge current allows a larger ETB width resulting in even more edge current and so on until a peeling mode is triggered. We will be looking for effects of co versus counter ECCD on the ETB width. Results of this study could have an impact on models for the ETB width which is also of great significance in predicting ITER performance. Finally ECH heating might trigger ELMs with higher n number a thus reduced size along the ballooning boundary possible similar to the effect of pellet injection providing another possible ELM mitigation technique.
Resource Requirements:	All gyrotrons operational. ECH/ECCD aiming directed to pedestal. Density and temperature of pedestal optomized for ECCD. Upper and lower single null dischrages.
Diagnostic Requirements:	Profile diagnostics.
Analysis Requirements:	Profile, kinetic efit, ELITE analysis
Other Requirements:

Title	420: Use of non-axisymmetric fields to reduce requirements for access to H-mode
Name:	Solomon solomon@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	A.M. Garofalo, H. Reimerdes, K.H. Burrell	ITPA Joint Experiment :	No
Description:	The basic goal is to use non-axisymmetric fields as a way to lower the power threshold requirement for access to H-mode, which may benefit ITER operations.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This experiment should be conducted in an L-mode plasma at injected power just below the nominal threshold power, using co-NBI injection. Attempt to reduce the rotation (or if possible drive counter rotation) using non-resonant braking to induce an H-mode at the reduced power threshold associated with less co-rotation. Alternatively, apply resonant braking to attempt to create regions of high ExB shear in the plasma in an attempt to suppress turbulence and perhaps trigger an H-mode.
Background:	Non-axisymmetric fields have been shown to affect plasma rotation. Non-resonant magnetic fields tend to drag the rotation toward the neoclassical offset rotation, which is in the counter direction and where the power threshold has been reduced to be lower on DIII-D. Resonant magnetic fields are believed to apply a localized torque on the plasma, which might create regions of high ExB shear suitable for triggering H-mode. Both effects, however, are likely to be significantly reduced at lower beta in L-mode, and as such it would probably be desirable to have some demonstration discharges of the magnitude of the braking in L-mode before running the experiment.
Resource Requirements:	1 day experiment
Diagnostic Requirements:	Standard profile and fluctuation diagnostics
Analysis Requirements:
Other Requirements:

Title	421: Invariance of SOL flows of plasma rotation in H-mode
Name:	Muller mullersh@fusion.gat.com	Affiliation:	UCSD
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	J. Boedo, R. Moyer, D. Rudakov, G. Tynan, J. DeGrassie, W. Solomon	ITPA Joint Experiment :	No
Description:	In the 2008 runs, the Reynolds stress head of the midplane probe was used to measure SOL-flow profiles during plasma-rotation-scaling experiments in H-mode. While significantly different core rotation profiles were achieved by varying the external momentum input, the flow profiles in the SOL seemed entirely unaffected, showing the same shapes and magnitudes throughout the experimental day. Similar results of a mostly unchanged SOL with strongly different core rotation profiles have been reported from the TCV tokamak in L-mode. This proposal aims to identify to what extent this invariance of the SOL flows of the core rotation holds. We will attempt to trigger a measurable change in the SOL flow profiles by varying rotation-relevant, but not primarily shape-related parameters. This will yield insight on the core-SOL coupling mechanisms for momentum transport.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use low-power lower-single-null plasmas with lowest possible L-H transition threshold. Use ECH and NBI to heat plasma and vary momentum input to achieve different core rotation profiles. Vary parameters such as B-field and density, in large exploratory steps, attempting to maintain the same plasma shape, until a change in the SOL flow profiles can be observed. If yes, move on to attempt to identify the systematic behind the variation.
Background:
Resource Requirements:	ECH (2 sources), NBI (co, counter and balanced)
Diagnostic Requirements:	All edge and SOL diagnostics. In particular midplane probe, X-point probe, CER, Edge-Thomson.
Analysis Requirements:
Other Requirements:

Title	422: Control of the heat and particle flux in radiative divertors
Name:	Welander welander@fusion.gat.com	Affiliation:	GA
Research Area:	General Plasma Control/Operations	Presentation time:	Not requested
Co-Author(s):	Mathias Groth, the Boundary Physics, Plasma Control, and Core-Edge Integration groups	ITPA Joint Experiment :	No
Description:	Develop and demonstrate stable, feedback-control, radiative divertor scenarios. Identify suitable sensor (e.g., bolometer channel, filterscope, Langmuir probe, IRTV) and actuators (gas valves in the divertor). Test feedback in several operating regimes (e.g., with type-I ELMing toward detached divertor)	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Select one of our established radiative divertor scenarios as starting point. Identify sensors and actuators. Run this scenario in feedback control. Scan collisionality (density and power) to test feedback. The sensors will be bolometer and/or filterscope chords. The actuator will be gas puffing. A new code will be implemented in the PCS.
Background:	Control of the detachment (ionization) front is instrumental in future devices (=ITER!) with strong heatfluxes to the target plate. To reduce the heat flux in ITER, the currently envisioned scenario foresees radiative divertors using Ne and Ar seeding. However, stable divertor operation must be demonstrated in present devices, including the feedback-controlled position of the detachment front. Large, type-I ELMs may still burn through the detachment front, requiring a novel feedback approach.
Resource Requirements:	As part of radiative divertor program (core-edge integration) develop feedback control. Dedicate run time when feedback loop is programmed in PCS.
Diagnostic Requirements:	Bolometer, floor Langmuir probes, filterscope, IRTV -> sensors must be accessible by PCS
Analysis Requirements:
Other Requirements:

Title	423: Study of termination event of pure co-injected QH-mode
Name:	Osborne osborne@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	K. Burrell, B. Hudson	ITPA Joint Experiment :	No
Description:	We will use current ramps, beam torque mix, gap scans, and density variation to try to determine the reason all pure co-injected QH-mode discharges terminate in a large ELM.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Last year's results suggest that the termination of QH-mode might be coming from either a gradual increase in the edge current density or gradual reduction in the rotational shear. This experiment would would try to sort out the cause of the termination and possible mitigate it. ELITE studies in preparation for this experiment might indicate a discharge regime which is more susceptible to rotational destabilization and therefore more robust.
Background:	Although QH-mode clearly occurs in purely co-injected discharges discovered in last year's campaign, the longest duration QH-mode phase was about 1 sec. All the pure CO QH-mode phases where terminated in a large ELM. A relatively small reduction in rotational shear is observed in the QH-mode phase along with a reduction in li suggesting that the ELM might be triggered by either a loss of rotational shear drive or increase in the edge current density.
Resource Requirements:	Start with last years pure CO QH-mode discharges
Diagnostic Requirements:	Profile diagnostics along with BES for EHO characterization
Analysis Requirements:	Profile, kinetic EFIT and ELITE
Other Requirements:

Title	424: Role of flows and multi-scale fluctuations in RMP induced ELM-free operation
Name:	Zeng zeng@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T.E. Evans, UCLA group, R.A. Moyer	ITPA Joint Experiment :	Yes
Description:	The physical mechanism of ELM suppression by RMPs will be addressed via measurement of fluctuation flows, multi-scale and multi-field turbulence, and fast profile measurements. Plasmas will be specifically designed to take advantage of the high time and space resolutions possible with the current Doppler backscattering (DBS), CECE, BES, correlation reflectometry (ITG and TEM scales), and reflectometer profile systems on DIII-D. Spatially resolved high k (ETG scale) backscattering will also be used. The dynamic variation leading up to the last ELM like event will be followed. Comparison will be made between RMP application before the H-mode transition to RMP application after the H-mode transition, as well as to RMP ELM suppression in low rotation, balanced NBI H-modes.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Standard RMP ELM suppressed discharges will be utilized. These plasmas must be designed for optimum diagnostic access. Note that we desire high time/space resolution ECE measurements that may require higher Bt or modification of the UCLA CECE system at 60º. The RMP will be applied both before and after the H-mode transition and results compared. In addition, ELM suppression during low rotation, balanced NBI H-mode will be of interest.
Background:	Although ELMs have been successfully suppressed in DIII-D by using n=3 resonant magnetic perturbations which is generated by I-coil, the mechanism(s) leading up to the change in transport is not yet understood. We propose to investigate the physics of the suppression using the full range of diagnostics available on DIII-D. For example, previous work has shown interesting changes in the poloidal flow of fluctuations and fluctuation amplitude during the onset of ELM operation. However, these plasmas were not optimal for diagnostic access.
Resource Requirements:	I coil Co and counter NBI
Diagnostic Requirements:	all fluctuation diagnostic, profile reflectometer,ECE
Analysis Requirements:
Other Requirements:

Title	425: Transition from drift to BAAE instabilities
Name:	Gorelenkov none	Affiliation:	Princeton U, PPPL
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	Michael Van Zeeland	ITPA Joint Experiment :	No
Description:	We are aiming at exploring the excitation threshold of BAAEs and a smooth transition from KBM drift frequency branch to BAAE branch, which is predicteby by theory.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Create similar plasma to 132710 shot but vary ECH power and NBI power keeping total power constant. One of the important subjects to study is the beam ion redistribution from a localized BAAE modes.
Background:	Recent theoretical study predicts that away from the rational surface KBM branch is transformed to BAAE mode. At transition the instability frequency is characterized by the up sweep. It was also predicted that KBM should be unstable due to ITB in DIII-D and has ion propagation direction, but were not observed in a dedicated experiments. Theory suggests that near the rational surface global KBM like solutions should experience strong continuum damping if shear is finite. In addition if Te/Ti is not large strong ion Landau damping may contribute to the stabilization. The mode, therefor, is observed when it becomes a gap mode and decouples from the continuum. The excitation of gap modes is primarily due to fast ions from NBI. To maintain gap mode instability we are planning to lower the drive and damping at the same time. But because KBMs are more sensitive to Te/Ti than to fast ion drive we expect that KBM will be more unstable if ECH powere increases and beam ion power decreases keeping total power fixed. Employ new diagnostic capabilities to measure fast ion effects due to BAAEs.
Resource Requirements:	ECH and NBI with co and counter injections.
Diagnostic Requirements:	FIDA, NPAs, BES, ECE
Analysis Requirements:	TRANSP, FIDA imaging
Other Requirements:

Title	426: Reynolds stress measurements deep inside LCFS in rotation experiments in L-mode
Name:	Muller mullersh@fusion.gat.com	Affiliation:	UCSD
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	J. Boedo, R. Moyer, D. Rudakov, G. Tynan, J. DeGrassie, W. Solomon	ITPA Joint Experiment :	No
Description:	In 2008, the Reynolds stress head of the midplane probe was demonstrated to be capable of measuring Reynolds stress and flow profiles up to 1 cm inside the separatrix in not-so-low powered H-mode discharges. While this represents the limit in H-mode, penetrations of the order of 5 cm would be possible in L-mode discharges. A variety of exciting results from rotation measurements in the TCV tokamak show that L-mode is no less relevant to experimentally test theoretical models of turbulent momentum transport, while being significantly more accessible to fluctuation measurements by Langmuir probes. This proposal aims at measuring the Reynolds stress as far inside the separatrix as possible in low-power limited L-mode discharges, aiming at a penetration depth of at least 5 cm. We will vary the external momentum input via co-, counter- and balanced NBI. This will yield much more information on the momentum transport mechanisms active in the core plasma than possible in H-mode and allow for a comparison of the Reynolds stress and flow profiles across the SOL between the L- and H-mode configuration.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use low-power limited L-mode discharges, ohmic and/or with low-power ECH. Use co, counter and balanced NBI to vary momentum input to achieve different core rotation profiles. Measure full Reynolds stress and flow profiles with midplane probe as far inside the separatrix as possible (goal: more than 5 cm). Vary plasma parameters to study parametric dependences.
Background:	--
Resource Requirements:	ECH (2 sources), NBI (co, counter and balanced)
Diagnostic Requirements:	All edge and SOL diagnostics. In particular midplane probe, X-point probe, CER, Edge-Thomson, fast camera.
Analysis Requirements:	--
Other Requirements:	--

Title	427: Investigation of ZMF zonal flows in plasmas with and without RMP
Name:	Krämer-Flecken none	Affiliation:	Juelich
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	Yes
Description:	Produce L-mode plasmas with ZMF zonal flows. Study the velocity shear and turbulent transport properties in the radial region of the zonal flow. Vary plasma parameters to see how the ZMF zonal flow can be influenced by monitoring the poloidal velocity oscillations. Study the effects on turbulent transport by measuring the turbulence de-correlation time and compare it to the shearing rate. Switch on the RMP to measure its influence on ZMF zonal flows. Repeat experiments on TEXTOR to have intermachine comparison. Look for methods to enhance the ZMF zonal flow.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Regarding the plasmas scenario I think that L-mode plasmas could be a good starting point. As far as I know the results on ZMF zonal flows are performed in L-mode. In a further step the measurements should be repeated in diverted plasmas. In the last years the understanding of the interaction between the background turbulence and the zonal flows in general has increased. However there are still open issues. Up to now the suppression of the background turbulence by zonal flow shear is not yet demonstrated. Experiments at DIII-D (A. Gupta et al., PRL 97 125002) report on the ZMF zonal flow but a significant influence on the turbulent transport is not mentioned. Furthermore the observation of the ZMF-zonal flows more deep in the plasma (r/a=0.8) suggests that the ZMF-zonal flows are not that much of importance for the L-H transition as it is expected. For a better understanding of the turbulence suppression it is therefore of interest to understand interaction of ambient turbulence with zonal flows and how they can be used in future devices as ITER to suppress turbulent transport.
Background:	In the last years the understanding of the interaction between the background turbulence and the zonal flows in general has increased. However there are still open issues. Up to now the suppression of the background turbulence by zonal flow shear is not yet demonstrated. Experiments at DIII-D (A. Gupta et al., PRL 97 125002) report on the ZMF zonal flow but a significant influence on the turbulent transport is not mentioned. Furthermore the observation of the ZMF-zonal flows more deep in the plasma (r/a=0.8) suggests that the ZMF-zonal flows are not that much of importance for the L-H transition as it is expected. For a better understanding of the turbulence suppression it is therefore of interest to understand interaction of ambient turbulence with zonal flows and how they can be used in future devices as ITER to suppress turbulent transport.
Resource Requirements:
Diagnostic Requirements:	Beside standard diagnostic BES, Reflectometry, ...
Analysis Requirements:
Other Requirements:

Title	428: Transition from drift to BAAE instabilities
Name:	Gorelenkov none	Affiliation:	PPPL, Princeton U
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	Michael Van Zeeland	ITPA Joint Experiment :	No
Description:	We are aiming at exploring the excitation threshold of BAAEs and a smooth transition from KBM drift frequency branch to BAAE branch, which is predicteby by theory.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Create similar plasma to 132710 shot but vary ECH power and NBI power keeping total power constant. One of the important subjects to study is the beam ion redistribution from a localized BAAE modes.
Background:	Recent theoretical study predicts that away from the rational surface KBM branch is transformed to BAAE mode. At transition the instability frequency is characterized by the up sweep. It was also predicted that KBM should be unstable due to ITB in DIII-D and has ion propagation direction, but were not observed in a dedicated experiments. Theory suggests that near the rational surface global KBM like solutions should experience strong continuum damping if shear is finite. In addition if Te/Ti is not large strong ion Landau damping may contribute to the stabilization. The mode, therefor, is observed when it becomes a gap mode and decouples from the continuum. The excitation of gap modes is primarily due to fast ions from NBI. To maintain gap mode instability we are planning to lower the drive and damping at the same time. But because KBMs are more sensitive to Te/Ti than to fast ion drive we expect that KBM will be more unstable if ECH powere increases and beam ion power decreases keeping total power fixed. Employ new diagnostic capabilities to measure fast ion effects due to BAAEs.
Resource Requirements:	ECH and NBI with co and counter injection capabilities
Diagnostic Requirements:	FIDA, NPA, BES, ECE
Analysis Requirements:	TRANSP
Other Requirements:

Title	429: Active disruption avoidance
Name:	Strait strait@fusion.gat.com	Affiliation:	GA
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	Yes
Description:	The goal is to develop methods of postponing a disruption, allowing time to bring the plasma thermal energy and current down in a controlled way. In ITER, a soft landing is greatly preferable to a fast shutdown with massive gas injection, because of concerns over local heating of the first wall and loading of the vacuum system. In the proposed DIII-D experiment, the approach is electron cyclotron heating of a tearing mode island, triggered by detection of the island growth. Additional methods of slowing the growth of an island could also be tried, including rotating RMP to avoid locking of the island (as also proposed by others). In any case, here the goal is not to achieve full stabilization, but to buy time for a controlled shutdown.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	In a low beta, L-mode plasma, reduce the density until a locked mode appears and leads to a disruption. Use the dud detector to trigger ECH heating near the q=2 surface, followed by a current rampdown. Vary the location and power of the ECH to determine the requirements for disruption postponement. Repeat in a high beta H-mode plasma, triggering the ECH and rampdown on the growth of a rotating 2/1 mode. At the same time, bring down the NBI power and then the plasma current.
Background:	As reported at the 2008 IAEA conference, FTU and ASDEX-Upgrade have demonstrated disruption avoidance using ECH near the q=2 surface. These results were obtained in ohmic or L-mode plasmas, and need to be extended to elongated, H-mode plasmas. The requirements for ECH width, location, and power need to be determined. This is the topic of a new ITPA joint experiment.
Resource Requirements:	ECH, 4 gyrotrons minimum.
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	430: Reynolds stress and flow evolution during an L-H transition
Name:	Muller mullersh@fusion.gat.com	Affiliation:	UCSD
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	J. Boedo, R. Moyer, D. Rudakov, G. Tynan, J. DeGrassie, W. Solomon	ITPA Joint Experiment :	No
Description:	In the 2008 runs, it was observed that in some configurations the L-H transition could be made to occur very reproducibly at a specific time. In such configurations, it would be possible to time the plunge of the midplane probe such that the probe is close to its maximum penetration depth when the transition occurs. This would allow one to observe the change in Reynolds stress and flows directly when the plasma goes from L- to H-mode. If the integral heat load can be kept acceptably small, it is technically possible to increase the dwell time such that the chances of "catching" the L-H transition are ameliorated.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use lower-single-null discharges with lowest L-H transition threshold (2 ECH sources) and highest reproducibility of the transition time. If reproducible L-H transition time can be obtained with any combination of NBI try to obtain data for different values of momentum input.
Background:	--
Resource Requirements:	ECH (2 sources), NBI (co, counter and balanced)
Diagnostic Requirements:	All edge and SOL diagnostics. In particular midplane probe, X-point probe, CER, Edge-Thomson.
Analysis Requirements:	--
Other Requirements:	--

Title	431: Multi-experiment validation of edge-turbulence codes against basic experiments and DIII-D
Name:	Muller mullersh@fusion.gat.com	Affiliation:	UCSD
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	J. Boedo, C. Holland, R. Moyer, D. Rudakov, G. Tynan	ITPA Joint Experiment :	No
Description:	As the plasma conditions in the tokamak edge and scrape-off layer are generally comparable to those encountered in basic experiments, the numerical simulations intended to model the tokamak edge-SOL system are generally capable of running also in the much simpler geometry of basic toroidal and linear experiments. These can provide a much higher accessibility for fluctuation diagnostics, such as spatially resolved Langmuir probe arrays and fast framing cameras, which can be used to setup stringent validation scenarios for the models behind these simulation codes. UCSD collaborates with several authors of edge turbulence codes, who have provided or agreed to provide adapted versions of their codes for the linear and/or basic toroidal geometry to undergo validation scenarios against the CSDX and VINETA experiments (linear) as well as the TORPEX and HELIMAK devices (simple magnetized tori). As success in the modeling/simulation of a simpler device is a necessary but not sufficient condition for successful validation, validation against actual tokamak data is indispensable. This proposal aims at setting up a well-defined validation scenario for these codes in the DIII-D tokamak, emphasizing compatibility with the respective codes' capabilities and the scenarios that can be achieved in basic experiments.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Fluid models and codes are most applicable to cold, collisional plasmas, so we would try lowest-power, limited L-modes first. The plasma geometry should be as simple as possible and the boundary conditions should be as well defined as possible. There should not be MHD activity or any kind of other effects that is not captured by the simulations models. As most edge-turbulence codes simulate the edge-SOL region only and assume a certain flux drive from the core region, a scenario should be picked for which these fluxes are experimentally as well known as possible.
Background:	--
Resource Requirements:	ECH (2-4 sources)
Diagnostic Requirements:	All available edge and SOL diagnostics, in particular all fluctuation diagnostics.
Analysis Requirements:	--
Other Requirements:	--

Title	432: Preferential Locking - Complete stabilization and avoidance of locked modes
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	R. La Haye, R. Prater, E.J. Strait, A. Welander	ITPA Joint Experiment :	No
Description:	Use RMPs to bring locked mode in view of gyrotrons and completely stabilize it by ECCD.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Repeat with more ECCD power encouraging experiments of 2006-07. If not sufficient for complete stabilization, optimize radial alignment. Acquire reference shots for comparison: deliberately misaligned, ctr-ECCD, pure ECH. In lieu of 1-2 complete revolutions where the mode was dragged past the optimal position, arrest the mode at optimum. If time permits, optimize repositioning: i.e. try to bring mode from natural locking to forcefully locked position as quickly as possible. In principle there is no impediment to a nearly instantaneous transition, if designed properly, i.e. avoiding some dangerous (disruptive) configurations in between. A simple theory involving mode rotation in the presence of the static residual error field and of the applied rotating RMP will help designing this transition. Instantaneous repositioning means, in fact, that the mode is directly forced to lock in the desired position, with little or no time spent in the "natural" one. Hence, ECCD stabilization will begin immediately.
Background:	In 2006-07 a combination of n=1 RMPs and ECCD was successful in controlling the toroidal phase and amplitude, respectively, of a 2/1 island initially locked to the wall and/or residual error field. Mitigation by up to a factor 2 was demonstrated with 1.3MW ECCD in the O-point. The primary goal of the present proposal is to achieve complete stabilization with the increased ECCD power which will be available in the next campaign. A corollary of this is that the mode, shrinked by the ECCD, should naturally unlock and spin up. The secondary goal is to provide reference shots which would make the evidence for RMP+ECCD control even more compelling. One of these shots should have the same global ECH/ECCD effects (thus, same shape, ne and Te profile, etc.), but no local ECCD inside the island, on the net of current diffusion. For this purpose, 4-6 gyrotrons will be radially misaligned in opposite directions: 2-3 inwards, 2-3 outwards. The other reference shots will have ctr-ECCD injected in the island (the effect on the island should be nearly equal and opposite to that of co-ECCD) and pure ECH in the island (the effect should approximately equal the asymmetry between co- and ctr- effects). A comparison of the three cases, along with a thorough diagnostic characterization, will help theorists to formulate the yet-to-be-developed modified Rutherford equation for locked modes, which will then be used as a predictive tool in experiments.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	"active tracking" algorithm in PCS

Title	433: Locked mode AVOIDANCE by "catching" its precursor with a rotating field
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	R. La Haye, R. Prater, E.J. Strait	ITPA Joint Experiment :	No
Description:	Apply rotating resonant magnetic perturbations (RMPs) to the rotating precursor of a locked mode to "catch it" and entrain it while it slows down, and so avoid locking, which is one of the main causes of disruptions.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	As soon as Mirnov probes detect a mode slowing down below 1kHz, apply intense n=1 I-coil travelling wave, initially at 0.9kHz, then slowing down at the same rate as the mode, as inferred from real-time spectral and mode analysis from the newspec code, recently become available in real-time. Alternatively, use magnetic feedback, for the I-coils to feed back on Mirnov. While the mode slows down, reduce its amplitude accordingly, under PCS: high RMP intensity was required at the beginning to overcome rotational shielding at ~1kHz, but becomes less and less necessary, and possibly detrimental, as the mode slows down. If evidence of coupling between the RMP and mode rotation is found at a certain time t, repeat with pre-programmed changes in the RMP rotation after t, for example keep the rotation steady, or accelerate it again.
Background:	So far, RMPs successfully controlled initially locked modes at DIII-D. Here we propose to pre-emptively apply rotating RMPs and avoid locking altogether. Applying the RMP while the mode is still rotating, though, introduces the difficulty of adapting the travelling wave to the mode rotation, in order for the former to entrain the latter. By contrast, if the differential rotation was excessive, shielding would be excessive too, and the mode wouldn't "feel" the rotating perturbation and wouldn't lock to it. MHD spectrograms recently made available in real-time (every 2ms) will help in this difficult adaptation, which has the promise for "catching" the rotating precursor of a locked mode before it locks to the wall or residual error field. Instead, the mode will lock to a rotating perturbation. At that point, the rotation can either be kept constant, at a safely high level, or accelerated. As a result the mode will accelerate too, and be mitigated by rotation, both because of shielding and because of flow shear effects
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	PCS changes involving real-time newspec and/or adaptation of magnetic feedback to NTMs

Title	434: Pellet in a locked-mode
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Vary from shot to shot the toroidal position of a locked mode and drop D and impurity pellets to evaluate their effect -through resistivity and radiative losses- on the O- and the X-point. Possible mitigation/stabilization.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Either deliberately cause locking with a certain toroidal phase by ramping the error field, or, after "spontaneous" locking, reposition the mode by means of RMPs. Drop pellet in locked mode of known toroidal phase. Repeat for various phases. The pellet will cause different effects, depending whether it's injected in the O- or the X-point and whether it's a deuterium or impurity pellet. Deuterium will increase ne, drop Te, increase the resistivity, reduce the local current. If done in the X-point, this would equalize the (higher) local current with the (smaller) current in the O-point, which suffers from a bootstrap (BS) current deficit. Re-establishing the poloidal (and, de facto, toroidal) symmetry of current would suppress the mode. A possible limit is represented by the effect of resistivity on reconnection, which might make the mode worse rather than milder. The other approach is to launch an impurity pellet at the radial location of the locked mode, and cause it to radiatively dissipate its magnetic energy. It's not clear yet -and it will be interesting to determine experimentally- whether the O-point will radiate or "leak" more than the X-point, and whether and how the magnetic topology in the vicinity of the X-point might be altered and possibly ameliorated by radiative losses. Might require some q profile tailoring to adjust size and position of q=2 surface underneath the pellet dropper. Try dropping the pellet both tangentially, for maximum interaction length, and orthogonally, for comparison.
Background:	There isnâ??t much background. To authorâ??s knowledge, controlled injection of pellets in locked islands (or in islands, in general), has never been tried. However, there are hints of impurity effects on island stability both at DIII-D and NSTX, e.g. in relation with wall conditioning.
Resource Requirements:	pellet, I-coils
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	435: NBCD in a locked mode
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	R. La Haye, M. Murakami, Y. Park, C.Hegna	ITPA Joint Experiment :	No
Description:	Magnetically steer an initially locked mode to align NBCD to its O- or X-point. A new stabilization/mitigation tool?	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Similar to ECCD experiments on an initially locked 2/1 island, slowly dragged in the toroidal direction by an n=1 I-coil travelling wave. However, use NBCD as a source of CD. This requires a different scenario, with a vertically shifted LSN or USN plasmas. In order to separate NBCD from momentum injection effects on the mode, repeat for various co/ctr mixes. Generating fast particles just outside the rational surface is also expected to be stabilizing, according to C.Hegna [PRL 1989]. A vertical scan of the plasma might help isolating this effect.
Background:	NBCD is not as localized as ECCD, but is an important source of CD at mid-radius which will become increasingly important in future DIII-D operations. Its effectiveness in controlling small rotating islands is limited by its excessive radial width, especially as far as replacing the missing bootstrap current is concerned. However, the radially broad, relatively intense NBCD can have an important effect on big locked islands. NBI at the island location can affect its stability also by injecting momentum and by generating fast particles.
Resource Requirements:	I-coils
Diagnostic Requirements:	FIDA and other fast particle diagnostics? MSE.
Analysis Requirements:
Other Requirements:

Title	436: Unlocking by NBI Torque and Locking/Unlocking Hysteresis
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	J. Ferron, C. Hegna, R. La Haye, H. Reimerdes et al.	ITPA Joint Experiment :	No
Description:	At locking, change the mix of co/ctr NBI but not the power, in order to unlock the mode without sacrificing beta_N.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Ramp NBI and thus beta to trigger a 2/1 NTM. Balanced injection will keep rotation low and make the mode prone to locking. When dud detects locking, transit to new NBI phase (see details below). The NBI power and co/ctr mix in the post-locking phase are chosen before the shot. Repeat for different choices of these parameters. J. Ferron's model for simultaneous, independent beta and torque control will initially be used off-line, and possibly in real-time later, if encouraging and if time permits. Requires preparatory work in the PCS. Time permitting, finer torque scans will be performed, shot-by-shot or in a single discharge, by simply ramping the torque down and up again, in order to 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 "dud detector" 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. A different real-time change is proposed here: it is proposed to 1) only drop the ctr-NBI power, 2) add some co-NBI, if it helps to spin the plasma more, and if the resulting beta is not too high. Instead of keeping the mode locked and mitigating its effects, we want to unlock it, even at the cost of slightly increasing its size, initially. Later it will shrink again as a result of rotation. This work will take advantage on one hand of J. Ferron's simultaneous control of beta and torque by acting on the total NBI power and the co/ctr share and on the other hand of a real-time solver of the torque balance equation for rotating modes including the Neoclassical Toroidal Viscosity, to be developed 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 by coincidence in #128903 during other locked mode control experiments. The scope of this proposal is to assess the pros and the limits of this technique and to "automate" it, for future dial-in.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	torque balance equation solver to be developed at UW-Madison
Other Requirements:	--

Title	437: Modulate I-coils to induce edge currents and affect/study ELMs
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Use ac currents in the I-coils to induce currents in the plasma edge and so perturb the edge current above/below the peeling boundary. NB: this is NOT a proposal of modulated RMPs.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Starting with a marginally peeling-stable plasma, modulate current in the I-coils on a time-scale comparable with the current-diffusion time. Several kA of current might be necessary.
Background:	External coils were used in COMPASS-D to induce currents in the plasma edge. Indirectly, they also affected the edge pressure gradient. Although not measured directly (COMPASS-D was not equipped with MSE, Li-beam or edge Thomson scattering), these perturbations were strong enough, according to EFIT, to cause a modulation across the peeling limit, which, in fact resulted in recursive stabilization/destabilization of ELMs [S.J. Fielding et al., EPS 2001, P5.014, Sec.2].
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	438: Modulate Ip to modulate edge current above/below peeling limit
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Modulate the plasma current Ip to indirectly modulate the edge current above/below the peeling limit.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Pre-program oscillating plasma current.
Background:	Similar to proposal #437, except that the modulation in the edge current is not induced by the I-coils, but by the E-coil. This will cause a modulation of Ip and of the whole current profile. The shape of the current profile will also be modulated. The effect at the edge will be an oscillation of the local current density. Probably the edge pressure gradient will fluctuate too.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	439: NTMs "on demand", by ECH
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Requested
Co-Author(s):	R. La Haye	ITPA Joint Experiment :	No
Description:	ECRH slightly outside q=3/2 to provoke a local flattening of pressure, thus a Bootstrap deficit and therefore a 3/2 NTM when desired, for example in hybrid discharges.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Very simple, with ECRH, perpendicular launch and deposition slightly outside q=3/2. A toroidal field scan (within the shot, or from shot to shot) will allow to find the best location.
Background:	NTMs are not always undesired instabilities: small 3/2 NTMs, for example, are desirable in hybrid discharges, where they help to prevent sawteeth. Occasional difficulties were encountered recently (March 2007) in reproducing hybrid scenarios with "natural" 3/2 NTMs. The purpose of this proposal is to develop a tool to trigger these modes on demand, when required. It might also shed light on the physics of seeding, and permit accurate measurement of the NTM growth rate, although under "artificial" conditions, which can be contrasted with predictions from the Rutherford equation. In particular, the flattening of the pressure profile is expected to affect the Bootstrap term in the Rutherford equation and, to a higher order, to make Delta' less negative. The presence of other NTMs in the plasma (typically 4/3, in the absence of 3/2) is not expected to constitute a problem, as ECRH will be deposited outside the 3/2 surface and thus even farther from the 4/3 one.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	440: NTMs "on demand", by modulated ECCD
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Requested
Co-Author(s):	R. La Haye	ITPA Joint Experiment :	No
Description:	Drive current filament on the q=3/2 surface (around which island will form), by means of ECCD modulated at twice its CER rotation frequency.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Measure with CER the toroidal rotation velocity of the 3/2 surface (identified/localized through MSE and EFIT). Repeat the shot with modulated ECCD at twice that frequency (because n=2). In case of excessive variation from-shot-to-shot or within-the-shot, some work on the PCS might allow to respond in real-time to the changes of CER frequency. Compare co- and ctr-CD, expected to drive classical and neoclassical TMs, respectively.
Background:	NTMs are not always undesired instabilities: small 3/2 NTMs are desirable in â??hybridâ?? discharges, where they help to prevent sawteeth. Despite the existence of consolidated experimental recipes, difficulties were recently encountered (March 2007) in developing hybrid scenarios with â??naturalâ?? 3/2 NTMs. The purpose of this proposal and of #371 is to develop a tool to trigger these modes on demand, when required. The basic idea is that, in the absence of mode and therefore of filamentation, the 3/2 surface is â??smoothâ?? and consists of identical 3/2 current filaments, all carrying the same current. Artificially increasing or decreasing the current in one of them by means of co- or ctr-CD would break the axisymmetry and introduce the helical current perturbation around which an island would form. If sufficiently big (wider than â??marginalâ??), this island would evolve, grow and saturate. Tailoring the island to the hybrid discharge needs would then be a question of setting beta â??or otherwise modifying the Rutherford equation- in such a way that the saturated width is tolerable
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	441: Test of causality: mode rotation vs. plasma rotation
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	In locking at low rotation, mode slows down the plasma? or plasma slows the mode down? And in rotational mitigation, it is well-known that the rotating plasma drags the mode, but can torque imparted to the mode drag the plasma?	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use co/ctr NBI mixture to control plasma rotation, I-coils to control mode rotation, CER to measure plasma rotation and Mirnov spectrograms (newspec) to measure mode rotation. Then use an approach similar to K.Burrell's H-mode studies (PoP 1999): modulate the mode (plasma) rotation and measure the delay of the plasma (mode) response. Hysteresis curves will be obtained.
Background:	NTMs are approximately "frozen" in the plasma and spinning the latter (by momentum injection with NBI) also spins the former. Various proposals and some experimental evidence exists, suggesting that the opposite is also true, i.e. that torque imparted to the mode also spins the plasma. If confirmed, plasma rotation (and not just mode rotation) would represent one more application for internal coils in devices with low NBI momentum injection like ITER. Similar considerations apply to plasma and mode braking, and how they affect each other. For example, rotating modes are known to lock even in rotating plasmas which keep rotating. On the other hand, slow plasma rotation encourages mode locking.
Resource Requirements:
Diagnostic Requirements:	CER
Analysis Requirements:
Other Requirements:

Title	442: Disruption Mitigation with Large, Shattered Pellets
Name:	Jernigan jernigan@fusion.gat.com	Affiliation:	ORNL
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Not requested
Co-Author(s):	L.R. Baylor, S.K. Combs, S.J. Meitner	ITPA Joint Experiment :	No
Description:	Basic idea is to inject a frozen deuterium or neon pellet into DIII-D that is large enough to approach or exceed the Rosenbluth density for runaway electron suppression.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Use the new ORNL Large pipe gun (~16 mm dia) to form and accelerate a frozen D2 or Ne pellet into DIII-D, striking a target to shatter the pellet and and direct the debris toward the center of the plasma. Using the UCSD fast framing camera and the other existing disruption diagnostics, we want demonstrate improved penetration and assimilation of the injected material relative to the Massive Gas Injection used so far in DIII-D. It is expected that a significant fraction of the pellet mass will penetrate beyond the q=2 surface and produce a much higher assimilation fraction than the ~10% seen in MGI experiments. Assuming 100% assimilation, the D2 pellets should reach ~15-20% of the Rosenbluth density and the Ne pellet should reach ~100% counting both free and bound electrons. The purpose of shattering the pellets is to protect the first wall from damage if the pellet is not fully ablated.
Background:	Even though MGI experiments have been very successful in dramatically reducing the heat and mechanical loads during disruptions in DIII-D and other experiments, they have not been able to reach the densities expected to be necessary to prevent the production of large runaway electron currents expected to be produced during disruptions by the avalanche process in ITER and larger tokamaks. Even in very low pressure plasmas the injected gas stopped at the edge of the plasma. The ionized particles then produce a radiation layer which diffuses into the plasma. When this layer reaches approximately the q=2 surface, an MHD event rapidly transports a portion of the ionized material into the center. Unfortunately, only about 10% of the injected electrons reach the central portion of plasma before the current has decayed significantly. If the theories of electron acceleration and deceleration are correct, not enough cold electrons are available in the plasma interior to prevent the formation of large runaway currents.
Resource Requirements:	The ORNL Large Pellet Injector is expected to be ready for initial testing in DIII-D for the June 2009 operating period. Ohmic plasmas are acceptable for initial tests, but ECH and/or beam heated plasmas will be welcome.
Diagnostic Requirements:	Critical: UCSD fast camera and fast bolometer and X-ray arrays and high speed interferometers for density measurements. Desirable: Thomson scattering in burst mode and fast magnetic probes.
Analysis Requirements:	fast camera penetration analysis 0-d radiation analysis 2-d bolometric reconstruction
Other Requirements:	Probably 2 half day experiments with perhaps a Thursday night 2 hour initial test for timing setup, etc.

Title	443: Effect of impurities and wall conditioning on NTMs
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Stability	Presentation time:	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 by density control and fuelling (pumping, puffing) or deliberate impurity contamination (impurity pellet, puffing, laser blow-off). There are various ways of building the database. Approach 1: run reference shot every time one of the above conditions changes, within reason, and only if session leader allows so. Approach 2: if Phil West permits, 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 3: try building the database in a single session or half session. 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:	Post-disruption data at DIII-D and lithium wall conditioning data at NSTX suggest that good wall conditioning and low impurity content might help avoiding NTMs. 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. A similar survey of the interplay between NTMs, impurities and wall conditioning has been approved for a half day experiment at NSTX. It would be interesting to compare the two machines.
Resource Requirements:	pellet, impurity pellet, laser blow-off?
Diagnostic Requirements:	edge diagnostics, spectroscopic diagnostics
Analysis Requirements:
Other Requirements:

Title	444: ECH Target Development for Disruption Mitigation Experiments
Name:	Jernigan jernigan@fusion.gat.com	Affiliation:	ORNL
Research Area:	Rapid Shutdown Schemes for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Develop a target plasma for disruption mitigation experiments using ECH heating.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use ECH to heat a plasma which is then used as a target for massive gas injection with neon gas (~1000 Torr-liters). Continue with several identical shots to verify that the ECH system is not adversely affected by the injected gas. Increase the neon amount to ~2000-3000 torr-liters and repeat. If time remains and the neon tests are successful, turn off the cryopumps and try argon at a fairly low level (500-1000 torr-liters)
Background:	Previous experiments with massive gas injection into NBI heated discharges have had deleterious effects on the NBI systems at high levels of injected gas. It is hoped the the ECH system will be immune to the injected gas.
Resource Requirements:	Critical: ECH (2 or more gyrotrons) and the Medusa MGI array with a fixed pc trigger system or using the old CAMAC system. Desired: Cryopumps for neon, but warmed for argon
Diagnostic Requirements:	Desired: usual disruption diagnostics
Analysis Requirements:
Other Requirements:

Title	445: Onset condition on anomaly in off-axis NBCD
Name:	Park parkjm@fusion.gat.com	Affiliation:	ORNL
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	M. Murakami, C. Petty, B. Heidbrink	ITPA Joint Experiment :	Yes
Description:	Determine dependency of anomaly in off-axis NBCD on NB injection power. More specifically, determine if there exists threshold power for anomaly. If not, determine if anomaly increases with NB injection power.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use small, up-shifted plasma with reverse BT (like 134426). Avoid MHD and fast-ion driven instabilities (except ELM) at high NB power. Vary NB injection power (2 co + 0.2 counter < P_NB < 3 co + 0.2 counter). Vary ECH power at fixed P_NB = 2 co + 0.2 counter. Use co and balanced MSE to measure NBCD profile. Use FIDA to measure fast ion density and energy profile
Background:	The 2008 experiments demonstrated successfully robust off-axis NBCD in a wide range of co-injection power using vertically shifted small plasma. The measured off-axis NBCD increases approximately linearly with the injection power. The measured NBCD fits the best with the theoretical calculation using classical beam slowing down model up to P_NB ~ 6.3 MW. However, modest anomaly in the radial profiles of NBCD and fast ion density was observed at P_NB ~ 7.2 MW, which might be related to anomalous fast ion transport. The primary objective of this experiment is to find out onset condition on anomaly in NBCD and/or beam ion transport for off-axis injection by a detailed NB/ECH power scan.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	446: Fully non-inductive operation using off-axis NBCD
Name:	Park parkjm@fusion.gat.com	Affiliation:	ORNL
Research Area:	General SSI	Presentation time:	Not requested
Co-Author(s):	M. Murakami	ITPA Joint Experiment :	No
Description:	Demonstrate fully non-inductive operation with off-axis NBCD and central ECCD using vertically shifted small plasma	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use small, up-shifted plasma with reverse BT (like 134426). Reduce toroidal field (2.1 T -> 1.95 T) and plasma current (0.9 MA -> 0.8 MA) to locate EC resonance position at rho < 0.2 keeping q95~5. Optimize particle control to maximize current drive efficiency. Vary Z0 (vertical shift position) to modify target q profile
Background:	The 2008 experiments demonstrated successfully robust off-axis NBCD in a wide range of co-injection power using vertically shifted small plasma. Off-axis NBCD can provide most of external current needed at half the plasma radius for broad q > 1.5 for steady-state operation. Onetwo predicts that full non-inductive operation can be achieved by adding central ECCD to the existing high power off-axis NBCD discharge (like 134426), if the efficiency of off-axis NBCD is not degraded at high injection power. The primary objective of this experiments is to control current profile for steady state scenario by using combination of off-axis NBCD and ECCD.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	447: Simultaneous control of ELMs and RWMs
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Simultaneously use I-coils for RMP control of ELMs and for RWM control. Look for conflicts or synergies.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Simple superposition of AC, <1kA RWM control waveform, whatever algorithm or method it is based on (DEFC, ) on DC, 6kA baseline for ELM control. Ramp down q95 in initial discharges, then continue at optimal values for ELM control. Try to obtain first ELM suppression above no-wall limit (and, for comparison, shots with ELM control only, RWM control only, and no control at all).
Background:	I-coils have been used with success at DIII-D to control ELMs or RWMs. It is important to assess conflicts or synergies between the two control techniques, on the way to an integrated coil system for ITER, which will have to fulfil these and other tasks, such as error field correction. The concern is that optimal current and helicity settings for RWMs might not be optimal for ELMs and viceversa. For example, the n=1 dynamic correction for the RWMs might change the optimal q95 window for ELM control. On the other hand, from a RWM perspective, ELM-suppressing RMPs are a static, typically n=3 error field affecting the plasma rotation and thus the RWM stability. Hints of simultaneous RWM suppression and ELM mitigation can be found in A.Garofalo�??s shots 122591-594, where an n=3 magnetic braking field was applied with the C-coils, during n=1 RWM feed-back with the I-coils. Although the real goal of those experiments was to study RWM control in the presence of n=3 magnetic braking, they provide useful information on the interplay with RMPs for ELMs. For instance, they seem to confirm the above speculation on the modified q95 window. Although encouraging, the experiments will need to be repeated at increased n=3 current (6kA, as 3kA were marginal for ELM suppression). Moreover, the I-coils can be wired as to simultaneously apply an n=1 and n=3 perturbation. Finally it will be beneficial to drastically reduce the gas puff and move the strike point to improve the pumping and limit the collisionality, and to reduce the triangularity (was 0.6). All these modifications go in the direction of facilitating the ELM suppression. It will also be important to avoid ELM-free H-modes or temporary losses of H-mode, in order to isolate real ELM-suppression evidence. n=3 RMPs were proposed here because they are the best known and most successful RMPs. However, if successful, the same experiment could be repeated with n=1 or n=2 RMPs, which showed promising results in recent ELM control experiments by R.Buttery.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	448: Hot electron target plasma for runaway experiments
Name:	Jernigan jernigan@fusion.gat.com	Affiliation:	ORNL
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Develop hot electron target plasma runaway characterization.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Create a low density, hot discharge heated only with ECH and/or FW. Use neon pellets or small puffs of neon to produce runaway discharges during the current decay. If FW is used, either use upper single null plasmas or limiter discharges to keep the plasma in L-mode for favorable coupling. ECH/ECCD during the breakdown phase may be explored.
Background:	Experiments in 2008 which needed runaway electrons for testing spent much of their run days searching for repeatable runaway production. Most ended up using argon puffs to produce the desired runaways. The problem with argon, is that it builds up on the cryopumps since they don't get warm enough during the glow discharge to dump the argon. This produces significant changes in the discharges as the day proceeds. In addition, the NBI systems are particularly sensitive to the argon used. The desire here is to produce a reproducible target for runaway production that can be used for runaway characterization experiments. Hopefully, if such a target is available, the runaway experiments will be much more productive.
Resource Requirements:	Low density, pumped discharges ECH (as many gyrotrons as possible) and FW. Neon pellets and gas puff (argon as a fallback if available)
Diagnostic Requirements:	All available runaway diagnostics as well as Thomson and ECE (for slideaway detection)
Analysis Requirements:	Check reproducibility of runaway generation throughout the day.
Other Requirements:

Title	449: Impact of RMP on L-H Transition Power Threshold
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	This proposal addresses an urgent ITER need; ITER will likely enter H-mode during the current ramp phase. It is important to eliminate even the first ELMs in this H-mode, which will require application of the RMP in the L-mode phase during current ramp-up. ITER needs to know how the L-H transition threshold will be affected by the RMP in this type of scenario. This experiment will also provide information on the physics of the L to H transition.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	- LSN ISS plasmas at low collisionality with cryopumping - establish L-mode, turn on RMP, then ramp Pinj - measure Pthres in steady state first: step neutral beams to get close; then modulate to close in on value with constant power - repeat for even and odd parity; both phasings at each parity - because ITER is expecting the L-H transition during startup in the current ramp phase, and we know that rising current inhibits the transition (raises the L-H transition threshold) repeat with current ramp up. - time permitting, repeat with both parities and both phasings
Background:
Resource Requirements:	cryopumps co and counter NBI to vary rotation of target I coils
Diagnostic Requirements:	- document with full pedestal profile, rotation and Er, and fluctuation diagnostics
Analysis Requirements:
Other Requirements:

Title	450: Poloidal asymmetry of density profile in core region due to large poloidal rotation
Name:	Park parkjm@fusion.gat.com	Affiliation:	ORNL
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	M. Murakami	ITPA Joint Experiment :	No
Description:	Determine dependency of poloidal asymmetry of plasma density on poloidal rotation using up-shifted small plasma.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use small, up-shifted plasma with normal BT (poor alignment of beam injection to local magnetic filed pitch). Vary NB injection power using RT sources (more perpendicular injection). Avoid L-H transition at high injection power if possible. Compare core and tangential Thomson data at the same radial location.
Background:	It has been assumed in modeling of transport and stability that the plasma density profile is constant on magnetic surface. Ideal MHD predicts that a significant poloidal asymmetry of plasma density can be induced if there is rapid poloidal rotation. If plasma is moved upward, core electron density can be measured at different poloidal locations but at the same minor radius by comparing core and tangential Thomson data. Poloidal momentum input can be maximized by using RT sources in up-shifted/normal BT configuration.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	451: the q<1 regime
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	For toroidal magnetic confinement devices, the regime of 0.1The principal difficulty in considering exploration of the q<1 regime on DIII-D is access. The DIII-D transformer is limited to supplying a loop voltage of the order of 10 V. Thus it is likely that, if transition through q=2 (not to mention q=1) is attempted with a fully developed plasma, disruptive kink instability is inevitable. The only possibility would seem to be to establish the q<1 regime very early in the discharge formation, while the toroidal magnetic field is low and the plasma is still very resistive. Given the very speculative nature of this proposal, the initial time committment should be a half day -- just to test various approaches to cracking the q=2 and q=1 barriers. Should there be signs of success, more experiment time should be considered.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	452: Quantify transport changes during RMP in H-modes
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Understanding the physics behind the tranpsort changes in RMP H-modes is complicated by the typical few tens to hundreds of ms of rapid, low amplitude ELMing prior to complete suppression. To simplifiy measuring how the RMP changes the rotation, Er, and turbulent transport in H-mode pedestals (as opposed to similar measurements in L-modes in e.g. Tore Supra and TEXTOR DED), this proposal is to make these measurements in H-modes with long ELM-free phases. We will use lower single null ISS plasma at low collisionality with ECH heating to obtain a long, ELM-free H-mode as obtained in 2006 for turbulent transport change measurements with the reciprocating probes. Understanding the coupling among rotation, pressure profile, fluctuations and transport is complicated by the persistence of the rapid ELMs which may dominate the tranpsort changes in the early phase of the RMP. This proposal will provide a means to look at early changes induced by the RMP without the complication of modifications to ELMs.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	LSN ISS plasmas at low collisionality from cryopumping; use ECH for L-H transition and to generate long, ELM-free phase (up to 1 second already achieved perviously) add short 10-20 ms neutral beam blips for CER measurements apply RMP in the ELM-free phase access the separatrix region with the midplane RCP; the auxiliary heating scheme here should allow access to the H-mode shear layer, but in any event these conditions should allow maximum penetration by the probes for direct measurement of transport changes when the RMP is applied.
Background:
Resource Requirements:	cryopumping I-coil CER and neutral beams for diagnostics 4 gyrotrons
Diagnostic Requirements:	reciprocating probes, fluctuation diagnostics (except for BES due to limited NBI pulses, UCSD fast imaging camera, full boundary diagnostics for RMP ELM control (DiMES TV, floor LPs, fast IRTV)
Analysis Requirements:
Other Requirements:

Title	453: Collisionality effect of RMP on ELM control and density pump-out in 2 collisionality regimes
Name:	Mordijck smordijck@wm.edu	Affiliation:	The college of William and Mary
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	In this experiment we investigate the influence of RMP on collisionality. Previous RMP experiments at high collisionality mitigated ELMS without density pump-out. In low collisionality, density pump out was always observed. However, the shape and the distance from the plasma to the wall was very different. To eliminate these effects, we propose to run the same experiments, once with the pump pumping and once with the pump off.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Establish a good ELM free RMP H-mode in low collisionality in ISS. Vary parameters like q95 and Beta (no scans) and keep coil pattern and currents the same. Change coil pattern from even to odd parity and repeat. Then repeat the experiment without the pumping capability to have a good comparison, which will help identify physic different due to collisionality.
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	454: Assess optimal Error Field Correction by modulating I-coils at incommensurable frequencies
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Perform several non-destructive Error Field Correction (EFC) tests within a single discharge, including non-resonant components.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Feed AC currents of incommensurable frequencies to each pair of I-coils in order to generate error fields that are different at every instant. Infer from plasma rotation or other indicator what set of currents gives best error correction.
Background:	Most of the times, optimal EFCs are assessed by â??trial-and-errorâ??, with a new EFC being tested in each shot. One of the reasons for this is that the current indicator of optimal EFC is the lowest density at which the plasma can be ramped down without locking. This usually results in a disruption and can thus be considered a â??destructive testâ?? (destructive of the plasma). Utilizing a non-destructive indicator such as the plasma rotation (faster or slower, depending on how strong the magnetic braking from the residual error field is) would allow multiple EFC tests within a single discharge. The limit on how many EFC configurations can be tested is set by the plasma rotation response time. At this point, there are various choices on how to conduct the EFC scan during the discharge. For example, one can fix the phases between the I-coil circuits and scan the currents, while keeping their ratios fixed. This would fix the 3D geometry of the EFC field and scan the overall strength of the correction. Alternatively, one can fix the strength and vary the phases so as to â??rigidlyâ?? rotate the EFC. In reality, to maximize the number of configurations, one can change the amplitude and the phases, as well as the topology, including non-resonant components. The latter can be achieved by individually modulating the I-coils, which is technically possible. In particular, modulating them at different frequencies would permit to test various strengths, topology and directions of the resulting EFC. To maximise the number of configurations, distinct coils should be modulated at incommensurable frequencies (incommensurable over the duration of a discharge, or a number of discharges).
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	455: Pair formation during disruptions
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Requested
Co-Author(s):	Alex James (UCSD)	ITPA Joint Experiment :	No
Description:	Provide first evidence of disruption-generated positrons	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Perform microwave and gamma-ray measurements during disruptions. Cause of disruption is not important. Can be combined with any experiment with intentional or probable disruptions, especially if copious runaway electrons are expected. 1. Microwaves: discriminate between clockwise and counter-clockwise elliptical polarization at the X2 harmonic in the oblique ECE radiometer. Because positrons gyrate opposite to electrons, polarization also change. Counter-clockwise polarization would be a signature of "Positron Cyclotron Emission". 2. Gamma-rays: Use scintillators to compare forward and backward emission either within the same discharge, with two scintillators looking in the co- and ctr-Ip direction respectively, or with the same scintillator in two discharges of opposite Ip and BT. Emission will be in one case the sum of backward emission from electrons (e-) and forward emission from positrons (e+). Because forward and backward emission of each specie are related to each other, it will be easy to recognize anomalies in the comparison with the forward emission from e- summed to the backward emission from e+. 3. Finally, coincidence counters of 511keV photons, of the the type used in accelerators or in positron emission tomography might also be utilized.
Background:	P. Helander (IPP Greifswald) predicted that, during disruptions, collisions between runaway electrons and thermal ions generate electron-positron pairs [PRL 2003]. Here it is suggested that channelling disruption energy in these electron-positron pairs might constitute an innovative mitigation scheme: pairs eventually annihilate in 511keV photons, relatively innocuous for the tokamak walls, certainly much less harmful than massive particles. First, however, we need to prove the formation of positrons in a tokamak, for the first time.
Resource Requirements:	Li-pellet desirable but not indispensable
Diagnostic Requirements:	UCSD scintillators. Oblique ECE. New coincidence counters.
Analysis Requirements:	--
Other Requirements:	--

Title	456: Study of resonant window for ELM suppression in q95 at Ip=const
Name:	Marina none	Affiliation:	CEA Cadarache
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	A. Loarte, T. Evans	ITPA Joint Experiment :	No
Description:	Study of resonant window in safety factor q95 for ELM suppression using I-coils at constant Ip.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Next step in understanding of resonant window in q95 for ELM suppression which appears to be narrower than predicted by vacuum modelling : scan of q95 by changing Btor, and not plasma current Ip ramp.
Background:	Present understanding relay on the fact that ELMs are supressed when Chirikov parameter >1 for r>~0.9. This criterion is used for ITER coils design for all scenarios for a range of q95=3-5. However on DIII-D the resonant window is reported to be narrow compared to vacuum modelling criterion. These experiments were done with Ip current ramps and hence there is a possibility that the edge current which appears in the case of Ip ramp influenced edge stability. To avoid this possible confusion q95 resonant window scan should be done at constant Ip as proposed here.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	457: Hybrid Confinement and m/n=3/2 MHD
Name:	Crisanti none	Affiliation:	Euratom-ENEA Fusion Association
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):	P. Buratti, V. Pericoli, G. Calabroâ??	ITPA Joint Experiment :	No
Description:	To investigate the role of the m/n=3/2 mode in the achievement of enhanced confinement (H98y>1.2) under different plasma rotation condition and with different q95. The ECRH will be used, with different q profiles, using different q95, either to replace the co-beam power or to affect the MHD behaviour by some localized Current Drive.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Perform ECRH deposition scans in Hybrid discharges at different q95, in order to change 1) the q-profile, 2) the pressure gradient at the q=1.5 surface and 3) the degree of local ECCD stabilisation. This should produce large variations of 3/2 mode amplitude, and possibly access regimes with fishbone activity instead of the 3/2 mode. The MHD conditions that give optimum confinement will be identified and compared to the ones found in other tokamaks.
Background:	In the last IAEA conference DIII-D has shown experiments (at q95=4) with H98~1.4 with a very low level of 3/2 mode activity, in presence of large plasma rotation. When replacing part of the co-beam injection with counter beam (reducing the momentum input) the mode was increasing in amplitude (up to 9 G rms) and the confinement was decreasing down to 1.1. Meanwhile at q95~6 the confinement remained good (H98y~1.4) with a mode amplitude of the order of 7 G (rms) and in presence of some ECRH power. In quite recent JET experiments it has been observed that the very high quality confinement (H98y~1.4) was linked with the complete absence of the 3/2 mode activity and with the presence of fishbones and of tiny sawteeth.
Resource Requirements:	NBI: all sources required. EC: minimum of 3 gyrotrons.
Diagnostic Requirements:	Standard diagnostics
Analysis Requirements:	Transport analysis, MHD analysis.
Other Requirements:

Title	458: ELM suppression at counter low plasma rotation
Name:	Marina none	Affiliation:	CEA Cadarache
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	A. Loarte, T. Evans, A. Garofalo	ITPA Joint Experiment :	No
Description:	Demonstrate ELM suppression at low counter rotation.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Establishe a referense case of ELM suppression at counter rotation and keeping beta_n=const change from counter to co rotation. Test if ELMs are suppressed and how confinement changes depending on the rotation direction.
Background:	Due to the strong Neoclassical Toroidal Viscosity (NTV) it is highly possible that with RMP coils for ELM suppression ITER plasma will rotate in counter to Ip direction with the frequency of order of diamagnetic one (Becoulet IAEA 2008). The conter rotation with I-coils was demonstrated already experimentally on DIII-D (Garofalo PRL2008), but in the case of odd parity and without ELM suppression. These shots can be used but with even parity I-coils. To test more ITER relevant situation we propose to test this more ITER-like conditions : ELM suppression with even parity I-coils at counter low rotation.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	459: Pedestal scaling between DIII-D, AUG and JET (ITPA-PEP-2)
Name:	Beurskens none	Affiliation:	UKAEA
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	Tom Oborne. Rich Groebner, Tony Leonard, Alberto Loarte, Lorne Horton	ITPA Joint Experiment :	Yes
Description:	ITPA PEP-2: The purpose of the experiment is to extend the range of the studies to investigate the physics mechanisms that determine edge pedestal gradients by comparing discharges with similar pedestal plasmas in dimensionless parameters in devices of various sizes. The experiments in 2009 will continue the 2008 experiments and investigate the scaling of the pedestal gradients with rho* and the effect of plasma triangularity on pedestal gradients. This proposal is directly linked to ITPA DSOL-1 in which the aim is to investigate the scaling of ELM energy losses in dimensionless similar discharges and their dependence on rho* and nu*.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	The experiments proposed aim at finalising the study of the influence of triangularity on pedestal gradients for discharges with similar pedestal parameters. This will be done for delta ~ 0.25 and ~ 0.4, q95 ~3.6, which corresponds to Ip ~1 MA and BT ~ 2 T in ASDEX-U and DIII-D and ~ 1T in JET. For such configurations the scaling with rho* will be assessed by discharges at ~ 0.5 MA/1T and intermediate points in DIII-D and AUG, and by discharges at ~ 1.7MA/1.8T and ~ 2.5MA/2.7T and higher in JET. Pedestal parameters will be measured by ECE, Thomson scattering, Edge Lidar diagnostic, Lithium beam, reflectometer, interferometer and edge charge-exchange measurements. In addition a q95 scan and a beta scan at the identiy point will be carried out.
Background:	In 2008 experiments were carried out in all three devices. JET has finished the low triangularity (~0.25) part of the experiment and is to carry out the high triangularity (~0.4) part of the experiments in November-December of 2008 or in early 2009. DIII-D carried out a two point rho* scan at low triangularity in 2008. AUG has carried out initial experiments to match the JET plasma shape and dimensionless parameters. The preliminary result is that no significant variation in the temperature pedestal width was found while ï?²* was varied by a factor of four across JET and DIII-D. However the density pedestal did not follow this trend; at the identity point the DIII-D density pedestal was two times narrower than the JET density pedestal. These results are in agreement with an earlier study presented by Tom Osborne at the 2004 APS and by Max Fenstermacher in a 2005 Nuclear fusion paper.
Resource Requirements:	3 session
Diagnostic Requirements:	Edge pedestal and ELMpower load diagnostics
Analysis Requirements:	Pedestal strucuture, transport heat load modeling of edge stability etc.
Other Requirements:	Teleconferncing wth JET and AUG

Title	460: Characteristics of m/n=2/1 modes at high beta
Name:	Buratti none	Affiliation:	Euratom-ENEA Fusion Association
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	F. Crisanti, V. Pericoli, G. Calabroâ??	ITPA Joint Experiment :	No
Description:	To measure the degree of magnetic reconnection at the onset of beta-limiting modes in order to investigate whether they start as NTM modes, i.e. with a magnetic island from the very beginning, or if they are initiated by the growth of a reconnection-free mode, which does not require any externally produced seed island.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use Hybrid discharges at high beta that develop the 2/1 mode. Analyse the radial phase profile of plasma oscillations by means of ECE, X-rays, BES; pi-phase jumps allow identifying full reconnection, while smooth profiles are characteristic of ideal-like modes. Vary q0 and pressure peaking to explore the stability domain.
Background:	Tearing modes with m/n=2/1 limit the normalised beta in the DIII-D hybrid scenario. The trigger mechanism of these modes is currently an open issue. Similar modes have been studied in JET; the main result was that the 2/1 modes at high beta do not have tearing parity at their onset, i.e. no magnetic island is detected at first (with already large ECE and soft-x ray signals), while reconnection takes place subsequently.
Resource Requirements:	NBI: all sources required.
Diagnostic Requirements:	Fast MHD diagnostics, Resonant Field Amplification.
Analysis Requirements:	Dedicated analysis of MHD signals. MHD stability analysis
Other Requirements:

Title	461: Summary of ORNL Proposals on Diagnostics for RF
Name:	Hillis HillisDL@ornl.gov	Affiliation:	ORNL
Research Area:	Heating & Current Drive	Presentation time:	Requested
Co-Author(s):	Baity, Hanson, Rasmussen, Ryan, Wilgen, Hosea, Nagy, Pinsker	ITPA Joint Experiment :	No
Description:	This proposal provides diagnostic support for the Fast Wave experiments proposed (52, 53, 54, 109,158, 190). The diagnostic support will include: (1) the re-establishment of the microwave reflectometer to provide edge density profile measurements in the SOL near the fast wave antennas, (2) equipping the MDS spectrometer with new fibers that view the front of the FW antennas for measurement of the rf electric fields via Stark broadening, (3) infrared observation and identification of hot spots during RF heating (in collaboration with LLNL), (4) fast visible imaging of the SOL and antenna structure to observe rf/plasma interactions and power loss mechanisms to the divertor and wall.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	These diagnostics will be installed on DIII-D by mid-May to provide support for RF experiments. In particular, the edge density profile (reflectometer) and electric field (Stark broadening) diagnostics will be important for the gas puffing (Pinsker) and edge loss (Hosea) experiments.
Background:	See Hosea and Pinsker RF experimental proposals
Resource Requirements:	These experiments will be an integral part of the RF proposals ((52, 53, 54, 109,158, 190).
Diagnostic Requirements:	UCSD Fast Visible camera, LLNL IR camera, MDS (ORNL)
Analysis Requirements:
Other Requirements:

Title	462: ELM suppression with RMP in high performance AI plasmas
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Using the now well developed techniques for ELM suppression using RMP with the I-coils, assess the impact of ELM suppression on the confinement and beta limits in high performance AI and hybrid plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	For both AI and hybrid plasmas in an ITER-like shape, scan q to find the resonant windows of optimal ELM suppression. With q at the best value(s) determine the limiting beta with and without ELM suppression. Also determine the lowest accessible rotation with and without ELM suppression.
Background:	ELMs are a serious threat to the integrity of the ITER divertor. A major effort is being devoted to the development of ELM suppression methods. The Advanced Inductive plasma scenario is the leading candidate for the optimization of the high performance, high Q mission of ITER. Thus, an assessment of the impact of ELM suppression with RMP on this scenario is essential.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to ITER Physics -- ELM Control for ITER (#463)

Title	463: ELM suppression with RMP in high performance AI plasmas
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Using the now well developed techniques for ELM suppression using RMP with the I-coils, assess the impact of ELM suppression on the confinement and beta limits in high performance AI and hybrid plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	For both AI and hybrid plasmas in an ITER-like shape, scan q to find the resonant windows of optimal ELM suppression. With q at the best value(s) determine the limiting beta with and without ELM suppression. Also determine the lowest accessible rotation with and without ELM suppression.
Background:	ELMs are a serious threat to the integrity of the ITER divertor. A major effort is being devoted to the development of ELM suppression methods. The Advanced Inductive plasma scenario is the leading candidate for the optimization of the high performance, high Q mission of ITER. Thus, an assessment of the impact of ELM suppression with RMP on this scenario is essential.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to Physics of Nonaxisymmetric Field Effects in Support of ITER (#462)

Title	464: Transport of current and poloidal flux with voltage modulation
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Investigate the transport of poloidal flux and the evolution and sustainment of stationary q profiles in quiescent and high performance plasmas, using modulation of the applied loop voltage as the principal actuator.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply modulation of the surface loop voltage (by E-coil modulation) at a few Hz and measure the amplitude and phase of the synchronous current density perturbation across the plasma, using MSE as the primary diagnostic. From these measurements the flux diffusivity (the effective resistivity) can be determined. Assess the impact of steady MHD (e.g., the 3/2 NTM in AI and hybrid plasmas) on flux transport.
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	465: Comparison of rotation in ECCD plasmas to C-Mod LHCD
Name:	Rice none	Affiliation:	MIT PSFC
Research Area:	Rotation Physics (2009)	Presentation time:	Requested
Co-Author(s):	Nat Fisch Matt Reinke Ron Parker John Wright John deGrassie Wayne Solomon	ITPA Joint Experiment :	Yes
Description:	The purpose of this experiment is to compare toroidal rotation velocity profiles in DIII-D ECCD discharges to those in C-Mod with LHCD. Parameter scans of electron density, ECCD power and deposition location will allow a comparison of C-Mod results, and to a model based on an inward pinch of energetic trapped electrons.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The approach will be to vary ECCD power and deposition location under different operating conditions, and to measure the toroidal rotation velocity profiles.
Background:	Rotation velocity profile control without neutral beam injection is desirable for ITER. In C-Mod LHCD plasmas, strong counter-current toroidal rotation has been observed, which is in contrast to the co-current rotation in ICRF heated discharges. Parameter scans of electron density and waveguide phasing have revealed a strong correlation of the rotation to the plasma internal inductance. Modeling indicates that this may be due to an inward pinch of energetic trapped electrons. Previous DIII-D results have shown a variation of rotation velocity profiles with ECCD deposition location. A comparison between the two methods will help reveal the underlying physics and lead to a powerful rotation velocity profile control technique.
Resource Requirements:	ECCD power and deposition location control.
Diagnostic Requirements:	CXRS and MSE from beam blips.
Analysis Requirements:	Transport and wave codes.
Other Requirements:

Title	466: MHD and confinement in fully noninductive high beta_p plasmas
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine the nature of the broadband MHD turbulence that occurs in high beta_p noninductive plasmas and that appears to limit confinement.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish an optimized high beta_p noninductive discharge. During the early part of the noninductive phase, when the MHD is active, use the density and Te fluctuation diagnostics in conjunction with magnetics (Mirnov and MSE) to characterize the spatial location, mode structure, and spectral properties of these fluctuations. Correlate the modes with variations in current and pressure profile widths.
Background:	In high beta_p noninductive discharges, the plasma profile evolution in the first second or two of the noninductive phase is slow, and beta rises slowly as well. During this phase broadband MHD turbulence is seen (noticed after the experiment). After a period of evolution, the MHD turbulence disappears, the confinement improves noticably, and beta rises more rapidly. Understanding this process will help to eliminate it and improve the ultimate performance of these plasmas.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	467: Fast ions, NTMs, and the current profile in AI plasmas
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Energetic Particles (2010)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Using the recent new and upgraded fast ion diagnostics (FIDA, the Da camera, FILD), probe the interaction in Advanced Inductive plasmas between fast ions, the 3/2 NTM, and the current profile.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	--
Background:	In hybrid and AI plasmas in DIII-D the presence of a n/m = 3/2 NTM is known to modify the current profile in a way that is favorable for stability and (probably) for confinement. Several hypotheses have been put forward for the mechanism of this interaction. One is that the NTM modifies the profile of fast ions, and so of the NBCD. Direct observation of the evolution of the fast ion distribution as the NTM develops will help test this hypothesis and will provide basic data on the interaction between fast particles and magnetic islands.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to Steady State Integration -- Core Integration (Advanced Inductive) (#468)

Title	468: Fast ions, NTMs, and the current profile in AI plasmas
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Using the recent new and upgraded fast ion diagnostics (FIDA, the Da camera, FILD), probe the interaction in Advanced Inductive plasmas between fast ions, the 3/2 NTM, and the current profile.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	--
Background:	In hybrid and AI plasmas in DIII-D the presence of a n/m = 3/2 NTM is known to modify the current profile in a way that is favorable for stability and (probably) for confinement. Several hypotheses have been put forward for the mechanism of this interaction. One is that the NTM modifies the profile of fast ions, and so of the NBCD. Direct observation of the evolution of the fast ion distribution as the NTM develops will help test this hypothesis and will provide basic data on the interaction between fast particles and magnetic islands.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to Fusion Science -- Energetic Particles (2010) (#467)

Title	469: Density dependence of n=1 error field tolerance in NBI heated H-modes
Name:	Reimerdes reimerdes@fusion.gat.com	Affiliation:	CRPP-EPFL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Test whether the good agreement of the tolerable total resonant field found in H-mode discharges with nearly balanced NBI heating with the linear density scaling found in low beta L-mode plasmas is fortuitous or evidence of the same physics governing both thresholds. In addition the experiment should confirm the role of resonant and non-resonant braking as well as the intrinsic rotation in determining the error field tolerance.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We plan to apply a slowly rotating n=1 field (I-coils with 240Deg phasing) in SND counter-rotating plasmas, similar to the targets used in the 2007 (MP 2007-04-02/04) and 2008 (MP 2008-02-05) n=1 magnetic braking experiments, but with varying density ne, in order to measure the tolerable n=1 field amplitude. In addition to ne we plan to vary q95 and BT in order to test whether the error field tolerance is insensitive to these parameters as found in the low beta experiments [J.-K. Park et al, Phys. Rev. Lett. 99 (2007) 195003]. If this is indeed the case, these additional scans would show how intrinsic rotation, magnetic braking strength and/or plasma response must change to keep the error field tolerance constant. The value of βN has to be chosen to avoid the onset of a 2/1 NTM but still allow sufficient amplification to cause the rotation collapse. Since the beta dependence of the error field tolerance has been explained by amplification it is not necessary to carry out the TNBI scan at constant βN.
Background:	n=1 braking experiment in 2007 and 2008 showed that the tolerable plasma response to externally applied n=1 fields is consistent with a δBcritâ??sqrt(TNBI+T0) scaling, where TNBI is the NBI torque and T0 an estimate of the intrinsic torque. For low values of TNBI the tolerable total resonant perturbation calculated with IPEC [J.-K. Park et al., Phys. Plasmas 14 (2007) 052110], is consistent with the linear density scaling of the Ohmic locked mode threshold, suggesting that both, low density n=1 locked mode threshold and high beta n=1 error field tolerance are caused by the same physics [H. Reimerdes, et al., IAEA FEC (2008)].
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	470: AI and hybrid plasmas rotating in the counter-Ip direction
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine the performance limits for counter-rotating hybrid and advanced inductive plasmas. Find out whether low rotation conditions can be more readily accessed from the counter-Ip direction.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply the standard AI/hybrid startup sequence in a negative Ip discharge. Compare confinement and stability characteristics with previous co-Ip operation.
Background:	Determine whether behavior at low rotation depends on co- versus counter-rotation. Extend our understanding of the low rotation behavior of hybrids and AI plasmas.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	471: Optimized AI operation for ITER (low rotation, Te=Ti)
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Establish the limits to operation (beta, density) at the low rotation operating boundary for AI plasmas, with Te �?? Ti.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In the ITER shape, use counter- as well as co-NBI and distributed ECH to equalize Te and Ti and to help stabilize the 2/1 mode. Determine the low rotation limit and the beta and density achievable near that limit.
Background:	As ITER is expected to operate with low rotation, and with approximately equal electron and ion temperatures, it is important to validate the AI scenario under these conditions.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to Steady State Integration -- Core Integration (Advanced Inductive) (#472)

Title	472: Optimized AI operation for ITER (low rotation, Te=Ti)
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Establish the limits to operation (beta, density) at the low rotation operating boundary for AI plasmas, with Te �?? Ti.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In the ITER shape, use counter- as well as co-NBI and distributed ECH to equalize Te and Ti and to help stabilize the 2/1 mode. Determine the low rotation limit and the beta and density achievable near that limit.
Background:	As ITER is expected to operate with low rotation, and with approximately equal electron and ion temperatures, it is important to validate the AI scenario under these conditions.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to ITER Physics -- ITER Demonstration Discharges (#471)

Title	473: Oblique-ECE-assisted MECCD suppression of 2/1 NTM
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	M. Austin, R. La Haye, R. Prater, E.J. Strait, A.Welander	ITPA Joint Experiment :	No
Description:	Use oblique ECE to check radial alignment of ECCD to 2/1 island and determine the best frequency and phase of modulation for its suppression. Stagger poloidally the launching directions of the various gyrotrons. This will reproduce ITER-like conditions of broad deposition and further enhance the benefits of modulation.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	All the hardware is ready, although it might be desirable to improve the SNR for channel 2 of the oblique ECE radiometer, which in the past represented the limiting factor for the quality of modulation and for how well the ECCD correlated, for example, with Mirnov signals. All the preparatory work is ready from the 3/2 experiment. The increased ECCD power available this year will help against the 2/1, which is harder to suppress. Other than that, the main experimental effort on the day of the experiment will consist in developing a low rotation 2/1 target. This is because the system was designed for 1-10kHz NTMs, but works best at f<7kHz and the gyrotron power supplies work best below 5kHz. Achieving an n=1 frequency f<5kHz is perfectly doable: similar frequencies were reached even for the 3/2 mode, which in this sense is more challenging, as it lies deeper in the plasma, where rotation is higher, not to mention that the effective modulation frequency is multiplied by n=2. We intend to compare narrow vs. broad ECCD; modulated vs. continuous ; deposition in the O-point, X-point and in between. A new hardware ECE-Mirnov correlator, to be constructed, will help extracting from the oblique ECE only that temperature fluctuation which is relevant to the NTM (as inferred from Mirnov). This correlation signal will have the advantage, over a Mirnov-only signal, of possessing the correct phase.
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 never been stabilized by modulated ECCD, neither using Mirnov drive, nor oblique ECE. It will be important to do this for the first time, on the way to ITER.
Resource Requirements:	5 gyrotrons
Diagnostic Requirements:	Oblique ECE
Analysis Requirements:	--
Other Requirements:	--

Title	474: Compare co-/ctr-ECCD in O-/X-point (4 cases)
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Requested
Co-Author(s):	R. La Haye, A. Welander	ITPA Joint Experiment :	No
Description:	Drive current in the O-point or X-point of a neoclassical island in the co- or ctr-direction (4 cases). Compare NTM stabilization efficiency and characterize (de)stabilization mechanisms.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Modulate co-ECCD. Decide whether to modulate after oblique ECE (analogically) or Mirnov signals (analogically or digitally) depending on whatâ??s available and most reliable at the moment of experiment. Compare O- and X-point phasing. Tilt launchers toroidally and repeat for ctr-ECCD.
Background:	: It has already been shown, both at AUG and DIII-D, that co-ECCD in the O-point is more effective than in the X-point (although X-point deposition is still better than doing nothing). The aim of this proposal is to add two more cases, namely ctr-CD in the O- and in the X-point. These are similar but not identical to co-CD in the X- and O-point, respectively. One possible difference is that ctr-CD might give rise to a 4/2 component that would make the 2/1 island narrower and thus easier to stabilize. Similar considerations apply to a 6/4 distortion of the 3/2 island. Further differences between co- and ctr-CD stabilization might be unveiled by the experiment, in particular by magnetic probe and ECE measurements of the island width evolution and of the poloidal and toroidal mode numbers. MSE measurements of the total local current and ONETWO and TORAY calculations of the Bootstrap and EC-driven currents will also be essential in the analysis.
Resource Requirements:	5 gyrotrons
Diagnostic Requirements:	Oblique ECE
Analysis Requirements:	onetwo
Other Requirements:

Title	475: Search for core plasma H-mode trigger
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	T.L. Rhodes, W.A. Peebles, A.E. White, J. DeBoo, G.R. McKee, J.C. Hillesheim, E.D. Doyle, G. Wang, L. Zeng	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to obtain simultaneous, local measurements of ITG-scale and intermediate-scale fluctuation levels across the L-H transition with improved time resolution (100 microseconds) to determine the relative timing of flow shear increase and fluctuation reduction in the core and pedestal.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The goal of this experiment is to obtain simultaneous, local measurements of ITG-scale and intermediate-scale fluctuation levels across the L-H transition with improved time resolution (100 microseconds). A multi-channel DBS system will simultaneously provide local ExB flow and flow shear measurements with similar time resolution. #131912 can be used as reference shot but a vertical shift is required to allow DBS to access ITG scale wavenumbers.
Background:	Recent Measurements by Doppler Backscattering and BES indicate "prompt" reduction of core fluctuation at the L-H transition in some plasmas. The time resolution of these measurements and of CER ExB flow measurements has not allowed a direct comparison of pedestal and core fluctuation suppression to date. Core fluctuation reduction across the L-H transition has been most pronounced in low density QH-mode plasmas.
Resource Requirements:	Beams, discharge development for vertically shifted equilibrium
Diagnostic Requirements:	DBS,BES,MSE,CER
Analysis Requirements:
Other Requirements:

Title	476: ITG-scale and intermediate-scale H-mode core turbulence vs shear
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	T.L. Rhodes, A.E. White, W.A. Peebles, G.R. McKee, E.J.Doyle, G. Wang, J.C. Hillesheim, K.H. Burrell, W. M. Solomon,	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to measure the core turbulence level vs. ExB shearing rate in H-mode for a wide range of poloidal wavenumbers (0.1 < k rho_s < 2). The shearing rate will be changed by modifying the beam torque in QH-mode.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Measure turbulence amplitude by DBS and BES in QH-mode during a slow torque scan from counter-to predominantly co-injection (Beam torque changed in steps while keeping beam power constant. Doppler Backscattering can access 0.5 < k_rho_s < 2 to map turbulence amplitude vs. time at r/a ~0.6-0.7. The measured wavenumber depends on launch angle at the LCFS (variable 3-15ª by changing angle of parabolic mirror and vertical plasma position). BES can access k rho_s < 0.5. Low density QH-mode allows DBS core access and absence of ELMs simplifies measurements.
Background:	No systematic comparison of H-mode fluctuation levels with ExB shearing rates has been made in the plasma core. The unique combination of diagnostic capabilities at DIII-D and the QH-mode (which now can be extended into a regime with predominant co-injection) allow investigating turbulence in H-mode across a wide range of poloidal wavenumbers. An H-mode shearing rate scan may also be of interest for GYRO benchmarking.
Resource Requirements:	7 beams
Diagnostic Requirements:	DBS, BES, MSE, , PCI, CER
Analysis Requirements:	--
Other Requirements:	Discharge development for vertically shifted equilibrium

Title	477: Outer PF only start-up on DIII-D
Name:	Mueller none	Affiliation:	PPPL
Research Area:	Torkil Jensen Award for Innovative Research	Presentation time:	Not requested
Co-Author(s):	J. Menard, M. Ono, D. Gates - PPPL, R. Raman - UW, D. Humphreys, J. Leuer, J. Lohr, M. Walker, G. Jackson, A Hyatt - GA and the NSTX Research Team	ITPA Joint Experiment :	No
Description:	Non-solenoid start-up and ramp-up is essential for an ST based CTF and could help simplify future tokamak designs. At present only 4 techniques show substantial promise: Coaxial Helicity Injection, Plasma Guns, outer poloidal field coils and iron core. It is proposed to perform outer PF coil start-up experiments on DIII-D. DIII-D is uniquely suited for this, it has appropriately placed PF coils that are well-controlled, has the modeling tools needed to setup the initial conditions, sufficient ECH power to assist breakdown and minimize resistive losses, as well as adequate NBI power to ramp-up Ip beyond 200 kA.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The Outer PF coils (6,7,8,9) will be energized so as to form an adequate field null and provide the needed Vâ?¢s required to drive Ip inductively. ECH power (up to maximum available) will used to provide plasma initiation, to heat the plasma, and reduce resistive Vâ?¢s consumption. The plasma will be fueled by gas puffing to increase ne and at about Ip=200 kA, NBI will be used to ramp-up and sustain Ip. It is expected that the early phase of the discharge (0 to 100 kA) will be the most difficult to control and will require some trial and error to settle on the best trade-off between forming a good field null and maximizing the Vâ?¢s available. It would be best if this experiment could receive its runtime in short (one or two hour) chunks spread over several days to allow time think about the results and refine what to do next.
Background:	NSTX has achieved Ip=20kA with very modest auxiliary heating power in the presence of vessel currents that were about 10X Ip. JT60 has achieved about 100 kA with 1 MW ECH and report that increased power decreases the requirement for a good field null. Both machines would have benefitted from additional heating power and better PF coil configurations.
Resource Requirements:	This requires the basic operations setup, but only the outer PF coils will be used. It would be best performed when DII-D is ready to perform with moderately good wall conditions, the ECH power available exceeds 2 MW (higher referred) and could benefit from ECCD after initiation. Also to ramp to higher Ip using NBI, high power beams (10MW) need to be available, but the requirement for NBI is only after Ip=200 kA has been demonstrated.
Diagnostic Requirements:	Magnetics, Thomson Scattering, density measurements, fast plasma TV, impurity measurements
Analysis Requirements:	Minimal, achievement of high Ip will be obvious from magnetics, temperature, density and impurity measurements will confirm the quality of the plasma.
Other Requirements:

Title	478: High bootstrap fraction noninductive operation at high G
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Core Integration (Steady State)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Expand the regime of high beta-p noninductive plasmas to lower q95. Begin studies of the operating limits and control issues for self-organized plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Previous operation of noninductive, high beta-p discharges was restricted to low levels of EC power and to a single value of the magnetic field (1.9 T). Using the expanded ECH/ECCD capability, and working at lower magnetic field, this work will extend the operating range of these plasmas to lower q95 and higher beta, yielding more interesting and relevant values of G (beta-N*H/q^2).
Background:	Eventually, steady state AT plasmas will have to rely on high bootstrap fraction operation. In such plasmas the self-organization of pressure and current profiles is a dominant effect. The operating limits and control issues for such plasmas have not been studied. To reach this regime, magnetic field strength is less important in maintaining confinement and noninductive current fraction than is the total plasma current. Reducing the magnetic field will expand the operating range for the high beta-p noninductive regime and make these plasmas more interesting for future applications and will allow the study of self-organized tokamak plasmas.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	479: Burn control simulation
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	General Plasma Control/Operations	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Study the evolution and stationary state of a plasma with power input dependent on plasma parameters (e.g., beta^2 to simulate an alpha particle heat source). Develop methods to control the operating point.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Regulate a portion of the power input to the DIII-D plasma to be proportional to the equivalent fusion power, P_alpha ~ n^2*f(T_i). The remainder of the power is used in part for steady auxiliary heating and in part for feedback control of beta and of the operating point (i.e., maintain constant P_alpha) in the presence of perturbations such as ELMs , sawteeth, MHD instabilities, etc. Also examine burn control using fueling and pumping to modify the fuel ion mix, impurity radiation, and average density.
Background:	It may be desirable to operate a burning plasma at a sub-ignited operating point, which will probably be unstable to temperature and power excursions. DIII-D now has the capability to start work on the control of such an operating point.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	480: Control requirements for self-organized, high beta-p, noninductive plasma operation
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	General Plasma Control/Operations	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Begin studies of the operating limits and control issues for mostly self-organized plasmas. By increasing the ECH/ECCD and operation at lower magnetic field, make high beta-p, noninductive target plasmas with more interesting and relevant values of G (beta-N*H/q^2).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Previous operation of noninductive, high beta-p discharges has shown that operation at high (> 3) values of beta-p and beta-N without inductive current drive is possible. The EC power on DIII-D and the capability of controlling rotation have been added since these plasmas were last studied. These provide new actuators for the control of high f_bs, high beta operation. Using the expanded ECH/ECCD capability, and working at lower magnetic field, this work will extend the operating range of these plasmas to lower q95 and higher beta, yielding more interesting and relevant values of G (beta-N*H/q^2).
Background:	Eventually, steady state AT plasmas will have to rely on high bootstrap fraction operation. In such plasmas the self-organization of pressure and current profiles is a dominant effect. The operating limits and control issues for such plasmas have not been studied. To reach this regime, magnetic field strength is less important in maintaining confinement and noninductive current fraction than is the total plasma current. Reducing the magnetic field will expand the operating range for the high beta-p noninductive regime and make these plasmas more interesting for future applications and will allow the study of self-organized tokamak plasmas.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	481: Pedestal modification and ELM effects via loop voltage (E_phi) variation
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Determine the extent to which the pedestal profiles can be modified by controlled variation of the E-coil loop voltage, and also determine whether this technique provides a useful means for modification of ELMs.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use controlled modulation (and filtering of noise) of the surface loop voltage to modify radial ExB drifts in the pedestal region. Look for changes in the pedestal profile and in ELM characteristics.
Background:	The steady applied toroidal electric field (via the E-coil) causes an inward radial drift (pinch) of the plasma. Ordinarily, this is balanced by outward transport and only stationary profiles are observed. Controlled modulation of the pinch by modulating the loop voltage should have observable effects on the pedestal profiles and may provide an additional actuator for modification and control of ELMs.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to Integrated Modeling -- Pedestal Structure (#482)

Title	482: Pedestal modification and ELM effects via loop voltage (E_phi) variation
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Pedestal Structure	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Determine the extent to which the pedestal profiles can be modified by controlled variation of the E-coil loop voltage, and also determine whether this technique provides a useful means for modification of ELMs.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use controlled modulation (and filtering of noise) of the surface loop voltage to modify radial ExB drifts in the pedestal region. Look for changes in the pedestal profile and in ELM characteristics.
Background:	The steady applied toroidal electric field (via the E-coil) causes an inward radial drift (pinch) of the plasma. Ordinarily, this is balanced by outward transport and only stationary profiles are observed. Controlled modulation of the pinch by modulating the loop voltage should have observable effects on the pedestal profiles and may provide an additional actuator for modification and control of ELMs.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to ITER Physics -- ELM Control for ITER (#481)

Title	483: ELM synchronization (pacing) via pulsed ECH
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine whether the ELM frequency can be modified or controlled by application of pulsed ECH power to the pedestal.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	In a standard H-mode plasma, apply maximum ECH power to the pedestal. Scan the modulation frequency and duty cycle. Also scan the deposition location across the pedestal to the separatrix. Look for synchronous modification of the ELM frequency and/or amplitude.
Background:	ELMs are a nonlinear evolution of the peeling-ballooning mode at the H-mode pedestal. Local modulation of the pedestal pressure and pressure gradient with modulated ECH should change the stability conditions and thus entrain the ELMs, at least over some range in applied EC power, location, and modulation rate.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	484: Advanced inductive performance comparison in H and He plasmas
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Repeat in hydrogen and helium plasmas the hybrid and AI discharges that were used to test the impact of wall conditioning in deuterium in DIII-D. Compare performance characteristics, particularly confinement.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The discharge set used over the past two years to assess the impact of wall conditioning on DIII-D operation included standard hybrid and AI plasmas. These are well documented and very reproducible. Repeating these discharges in hydrogen and helium with the same control settings (field, current, density, beta, â?¦) will provide an excellent assessment of the importance of ion mass and Z/A ratio in plasma performance.
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	485: Elimination of ELMs from SNs Using the RMP coils with B x gradB Away From the Divertor
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment is the first necessary step for determining whether the ELM suppression method developed here at DIII-D is compatible with radiating divertor scenarios. Previous DIII-D studies focused on the effect that particle drifts in the SOL/divertor had on fueling, particle pumping, and radiating divertor behavior. We concluded that the most promising (only?) way to successfully employ a radiating divertor in order to reduce heat flux at the divertor targets with a minimal cost to plasma core H-mode properties was to use a SN plasma characterized by having the gradB ion drift directed OUT of the divertor. Presently, however, it is unclear whether ELM suppression in SNs using the RMP coils is attainable, if the gradB ion drift is directed out of the divertor. In this experiment, we investigate if it is possible to suppress ELMs of a SN plasma with the gradB drift direction out of the divertor. Once demonstrating the feasibility of eliminating ELMs under these conditions, we are then ready to examine the behavior of trace injected impurities in an RMP ELM-suppressed environment and ultimately to demonstrate the feasibility of RMP ELM suppression in a radiating divertor environment in subsequent experiments.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The upper SN plasma is maintained in a standard ELMing H-mode regime (i.e., Ip =1.2 MA, Bt = -1.75 T, dRsep = +1.0 cm, q95=4.2, and Pinj = 6 MW). These parameters yielded the best of the radiating divertor results, but the resulting q95 may (or may not) be optimal for ELM suppression with the I-coil. To identify the range in q95 that yields the best prospects for ELM suppression, q95 is reduced during the shot by reducing Bt while the I-coil current is set to maximum. Once this q95 range is identified, choose value of q95 in the middle of that range and run successive shots with increasingly lower I-coil current. This is done to identify the minimum coil current, so as to minimize the perturbing effect of the RMP on the pedestal region.
Background:	Eliminating ELMs from H-mode plasmas using the I-coil approach presents an interesting possibility for resolving the ELM-issue in ITER and future highly powered tokamaks. Yet, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, whereby an impurity gas is injected into a pumped divertor with simultaneous deuterium gas puffing upstream of the divertor, has shown promise as a way to reduce peak power loading at the divertor targets without concomitant degradation of the ELMing H-mode plasma properties [IAEA 2006, PSI2006]. However, in combining the I-coil approach with such â??puff and pumpâ?? scenarios while still maintaining favorable H-mode operation, the injected impurity must still be prevented from escaping the divertor and contaminating the main plasma. The most promising radiating divertor scenario involves using a SN divertor with the gradB directed out of the divertor. However, it has not been demonstrated that a SN with the gradB out of the divertor is itself compatible to ELM suppression with the RMP coils. We suspect it is, because SN plasmas run at typically lower density than corresponding plasmas with the gradB drift directed into the divertor. This should result in lower collisionality in the pedestal in the gradB OUT case and thus better ELM suppression. Lower collisionality in the pedestal is helpful in ELM suppression with the RMP coils. As a result, the collisionality in the gradB out of the divertor cases will be lower than in the more standard gradB into cases, and the former would be expected to tolerate the higher gas puff rates needed to impede the escape of the impurities from the divertor.
Resource Requirements:	Machine time â?¤ 0.5 day (in forward Bt), I-coil, dome- and upper baffle cryo-pumps cold, minimum 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, dome and upper baffle locations), core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER.
Analysis Requirements:	UEDGE, with ONETWO runs.
Other Requirements:

Title	486: Impurity Screening Comparison Between ELMing and ELM-suppressed Plasmas
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment presents the second step in determining whether the ELM suppression method developed here at DIII-D is compatible with radiating divertor scenarios. This experiment, which uses non-perturbing (trace) argon under puff-and-pump scenarios, provides a side-by-side comparison of how well the injected impurity is screened in ELMing H-mode plasmas and in ELM-free H-mode plasma (with I-coil). DIII-D IS UNIQUELY CONFIGURED TO DO THIS EXPERIMENT. We focus on addressing the following questions: (1) Is there a significant difference in the argon accumulation in the plasma core under I-coil operation? (2) How does the exhaust enrichment change between the ELMing- and the ELM-free (I-coil) cases?	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The plasma is maintained in a â??standardâ?? ELMing H-mode regime under the conditions established in the lead-in experiment (e.g., Ip =1.2 MA, Bt = -1.75 T, dRsep = +1.0 cm, and upper SN). A trace amount of argon is injected into the private flux region of the upper divertor, while deuterium plasma flow toward the divertor is enhanced by a combination of deuterium gas injected from the bottom of the vessel and active cryo-pumping from both upper divertor locations. Trace argon is injected at a steady but trace level from t = 3.0 s to 7.0 s. A standard ELMing H-mode regime is established previous to t = 4.5 s of the discharge. At t = 4.5 s, the I-coil is activated and the ELMs are eliminated. This provides a direct comparison of the trace argon behavior between ELMing H-mode and ELM-free H-mode (I-coil) under similar conditions. [Note that the q95 and the coil current selected have been determined from the lead-in experiment: Can the RMP coils eliminate ELMs from SNs with the gradB out of the divertor?] The deuterium injection rate is scanned. The key measurables are the accumulation of argon in the core and divertor plasmas.
Background:	Eliminating ELMs from H-mode plasmas using the I-coil approach presents an interesting possibility for resolving the ELM-issue in ITER. Yet, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, whereby an impurity gas is injected into a pumped divertor with simultaneous deuterium gas puffing upstream of the divertor, has shown promise as a way to reduce peak power loading at the divertor targets without concomitant degradation of the ELMing H-mode plasma properties [IAEA 2006, IAEA2008, NF2008]. However, in combining the I-coil approach with such puff and pump scenarios while maintaining favorable H-mode operation, the injected impurity must still be prevented from contaminating the main plasma. This is by no means assured, due to the ergodic nature of the pedestal. In this experiment, we compare the dynamics of impurity screening between ELMing H-mode plasmas and ELM-free H-mode (plus I-coil) plasmas.
Resource Requirements:	Machine time 0.5 day (forward Bt), I-coil, dome- and upper baffle cryo-pumps cold, 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, dome and upper baffle locations), Penning gauge, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER.
Analysis Requirements:	UEDGE, with ONETWO and MIST runs.
Other Requirements:

Title	487: Compatibility of ELM Suppression with Radiating Divertor Scenarios
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment is the third in a series experiments for determining whether the I-coil method of ELM suppression is compatible with radiating divertor (i.e., puff and pump) scenarios. This experiment provides a side-by-side comparison of how a standard ELMing plasma and an ELM-free plasma (with I-coil) respond to a â??puff-and-pumpâ?? scenario with a perturbing amount of argon as the injected impurity. DIII-D IS UNIQUELY CONFIGURED TO DO THIS EXPERIMENT. We focus on addressing the following questions: (1) Is there a significant change in the argon accumulation in the plasma core under I-coil operation? (2) How does argon entrainment in the divertor change when the I-coil is activated? (3) How does the ratio of divertor-to-core radiated power change when the I-coil is activated?	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The plasma is maintained in a standard ELMing H-mode regime (e.g., Ip =1.2 MA, Bt = +1.7 T, dRsep = -1.0 cm, and lower SN). This setup is different from the previous two proposed experiments in that we use lower SN plasmas, as opposed to upper SN. This is due to need to have an operating IR camera measuring the changes in heat flux between RMP and non-RMP conditions. Significant deuterium gas puffing, which is needed to raise the SOL plasma flow into the divertor, will raise the density. A previous experiment will have shown that ELMs are suppressed under these conditions by a pre-selected I-coil current: Comparison of impurity screening between ELMing and ELM suppressed plasmas. Hence, it is possible that adjustments to the I-coil settings may be necessary if the D2 injection rate is changed, as it is in this experiment. After this prep work is done, argon is injected into the private flux region of the lower divertor, while deuterium plasma flow toward the divertor is enhanced by a combination of deuterium gas injected from the top of the vessel and active cryo-pumping from the lower outer divertor location. A radiating divertor plasma in an ELMing H-mode regime is established previous to t = 4.5 s of the discharge. At t = 4.5 s, the I-coil is activated and the ELMs are eliminated. This provides a direct comparison between ELMing H-mode and ELM-free H-mode under similar plasma conditions. The argon injection rate with a fixed D2 injection rate is scanned; then the D2 injection rate with a fixed argon injection rate is scanned. Important measurables are the changes in the radiated power distribution and heat flux values, the accumulation of argon in the core and divertor plasmas, and the density and temperature conditions at both divertor targets.
Background:	Background: Eliminating ELMs from H-mode plasmas using the I-coil approach presents an interesting possibility for resolving the ELM-issue in ITER. Yet, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, whereby an impurity gas is injected into a pumped divertor with simultaneous deuterium gas puffing upstream of the divertor, has shown promise as a way to reduce the peak power loading at the divertor targets without concomitant degradation of the ELMing H-mode plasma properties [IAEA2006, NF2008, and IAEA2008]. However, in combining the I-coil approach with such puff and pump scenarios, it is by no means clear that the injected impurities can be prevented from building up the main plasma as effectively as in the ELMing H-mode cases.
Resource Requirements:	Resource Requirements: Machine time 0.5 day, I-coil, lower baffle cryo-pumps cold, 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, in the dome and upper baffle locations), Penning gauge, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, lower divertor IR camera, and CER.
Analysis Requirements:	UEDGE, ONETWO, MIST.
Other Requirements:

Title	488: Compatibility of ELM Suppression with the Radiating Divertor in Hybrid Mode
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	N. Brooks, T. Evans, M. Fenstermacher, R. Pitts, and M. Schaffer	ITPA Joint Experiment :	No
Description:	This experiment is the fourth in a series of experiments for determining if the ELM suppression method using the I-coil is compatible with radiating divertor scenarios in the hybrid regime. This experiment provides a side-by-side comparison of how an ELMing hybrid plasma and an ELM-free hybrid plasma with I-coil perform under puff-and-pump scenarios with argon as the injected impurity. DIII-D WOULD BE UNIQUELY CONFIGURED TO DO THIS EXPERIMENT. We focus on addressing the following questions: * Is there a significant change in the argon accumulation in the plasma core under I-coil operation? * How does argon entrainment in the divertor change when the I-coil is activated? * How does the ratio of divertor-to-core radiated power change when the I-coil is activated? This experiment will be attempted only after it has been demonstrated that RMP ELM suppression has been demonstrated in the hybrid regime, where the gradB ion drift is directed OUT of the divertor.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The plasma is maintained in a ELMing hybrid H-mode regime (e.g., Ip =1.2 MA, Bt = +1.7 T, dRsep = -1.0 cm). Argon is injected into the private flux region of the lower divertor, while deuterium plasma flow toward the divertor is enhanced by a combination of D2 gas injected from upstream of the lower divertor targets and active cryo-pumping from the lower divertor location. A radiating divertor plasma in the ELMing hybrid regime is established previous to t = 4.5 s of the discharge. At t = 4.5 s, the I-coil is activated and the ELMs are eliminated. This provides a side-by-side comparison between ELMing hybrid and ELM-free hybrid. The argon injection rate and the deuterium injection rate are separately scanned. Key measurables are the changes in the radiated power distribution and heat flux values, the accumulation of argon in the core and divertor plasmas, and the density and temperature conditions at both divertor targets.
Background:	Plasma operation in the hybrid regime has been put forward as a paradigm for ITER, and would become particularly attractive if ELMing could be eliminated. Suppressing ELMs from hybrid H-mode plasmas using the I-coil approach is an intriguing possibility for resolving the ELM-issue in ITER. However, even if the damage to the divertor structure from ELMs pulses were eliminated via the I-coil approach, steady peak power loading at the divertor targets could still be unacceptably high. A radiating divertor solution, in this case, puff and pump, has shown promise as effective way of reducing peak power loading in ELMing hybrid plasmas, while at the same time maintaining good hybrid properties, e.g., energy confinement time. Our experiment will address whether the puff and pump radiating divertor concept is also as effective for hybrid ELM-free plasmas during I-coil operation. At this juncture, the I-coil method has not been completely successful in eliminating ELMs for plasmas in the hybrid regime. Clearly, this experiment should be attempted after ELM suppression in hybrid is successfully demonstrated.
Resource Requirements:	Machine time 0.5-1.0 day, I-coil, lower baffle cryo-pumps cold, 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, dome and upper baffle locations), Penning gauge, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, lower divertor IR camera, and CER.
Analysis Requirements:	UEDGE, ONETWO, and MIST
Other Requirements:

Title	489: Aerogel targets to study velocity, size and composition of dust particles in DIII-D SOL
Name:	Rudakov rudakov@fusion.gat.com	Affiliation:	UCSD
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	S. Ratynskaia (Royal Institute of Technology, Stockholm), C. Castaldo (Euratom ENEA Association, Frascati, Italy)	ITPA Joint Experiment :	No
Description:	Targets made of silica Aerogel â?? highly porous material composed of clusters of 2-5 nm solid silica spheres with up to 95 % empty space â?? will be used to study velocity, size and composition of dust particles present in the outboard SOL of DIII-D during plasma discharges.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Aerogel target will be installed in Midplane Material Evaluation Station (MiMES) and kept just outside of the wall tile radius in 240R0 port during plasma discharges for a few days. Then the target will be removed and analyzed for captured dust.
Background:	Dust penetration of the core plasma in ITER can cause unacceptably high impurity concentration and degrade performance. Therefore, knowledge of the dust transport and dynamics is important. Detection of dust present in the plasma during discharges is non-trivial. The main parameters of interest, apart from the particle material, are dust velocity, size and number density. Due to the uncertainties in the present estimates of the dust parameters it is important that diagnostics cover the maximum possible range of these parameters in order not to overlook some dust populations. A new method for dust collection has recently been proposed and is based on the use of aerogel - a highly porous, very low density material. Aerogel collectors can capture dust grains without destroying them, even in the high velocity range. Analysis of the tracks of captured particles allows to evaluate the dust velocity and the dust composition can be deduced upon particle extraction.
Resource Requirements:	MiMES with a slot for an Aerogel target. Piggyback experiments, no machine time requested.
Diagnostic Requirements:	Fast camera viewing MiMES from 135T0 port highly desirable.
Analysis Requirements:
Other Requirements:

Title	490: Compare AI and hybrid discharges with fishbones and with 3/2 NTMs
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Prepare advanced inductive and hybrid scenario discharges with 3/2 NTMs (DIII-D style) and with fishbones (AUG style). Determine whether and how the eventual performance is differerent after a stationary state is reached.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Using the AUG initiation prescription for hybrid and AI discharge formation (late heating, fishbones) make plasmas with the same control parameters (shape, field, current, density beta, â?¦) as the standard DIII-D versions. Compare confinement characteristics, profiles, and beta limits.
Background:	DIII-D and AUG use different prescriptions to make hybrid and advanced inductive plasmas. However, in a statistical sense these appear to have very similar properties once the final quasi-stationary state is reached. It would be valuable for eventual implementation of these scenarios on ITER to understand whether the resulting states are really the same, and to expand the range of initiation techniques.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	491: Possible control of NTM mode in AI plasmas by shaping
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Determine whether the occurrence of n/m = 4/3 NTM dominated advanced inductive plasmas (versus 3/2 NTM dominated) depends on shape.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	For otherwise standard conditions for advanced inductive plasmas vary the shape from high triangularity quasi-double-null to ITER-like single null. Determine whether and how the rate of incidence of 4/3 mode versus 3/2 mode dominated plasmas changes.
Background:	Usually, in discharges with a shape close to double null and high triangularity, 3/2 mode dominated plasmas occur much more often than 4/3 mode plasmas. The latter are of interest because the NTM has less of a deleterious effect on plasma performance. In the few AI and hybrid discharges produced for the ITER demonstration discharge study, 4/3 mode discharges were more frequent. This experiment is intended to follow up on this observation and to verify (or exclude) the possibility that the dominant NTM mode number depends on shape.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	492: Access to advanced inductive and hybrid scenario plasmas in ITER
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Develop discharge initiation, rampup, and heating trajectories that lead to optimized advanced inductive and hybrid scenario plasmas, and that are consistent with the limitations of the ITER systems.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Extend the work done on ITER startup to the development of high performance AI and hybrid plasmas. Vary the heating power time profile, the density trajectory, shape, and other control parameters to optimize G in the eventual quasi-stationary state.
Background:	This task integrates work done on plasma initiation and startup, demonstration discharges, and the various methods for initiating the AI regime developed on DIII-D and elsewhere (AUG, JET). It should be directed toward providing a reliable prescription for forming reproducible AI and hybrid plasmas in ITER.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to Plasma Control and Operations -- ITER Scenario Access, Startup and Ramp Down (#493)

Title	493: Access to advanced inductive and hybrid scenario plasmas in ITER
Name:	Politzer pete.politzer@gmail.com	Affiliation:	Retired
Research Area:	ITER Scenario Access, Startup and Ramp Down	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	Develop discharge initiation, rampup, and heating trajectories that lead to optimized advanced inductive and hybrid scenario plasmas, and that are consistent with the limitations of the ITER systems.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Extend the work done on ITER startup to the development of high performance AI and hybrid plasmas. Vary the heating power time profile, the density trajectory, shape, and other control parameters to optimize G in the eventual quasi-stationary state.
Background:	This task integrates work done on plasma initiation and startup, demonstration discharges, and the various methods for initiating the AI regime developed on DIII-D and elsewhere (AUG, JET). It should be directed toward providing a reliable prescription for forming reproducible AI and hybrid plasmas in ITER.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	Also submitted to ITER Physics -- ITER Demonstration Discharges (#492)

Title	494: Performance optimisation of steady-state plasmas with ITER shape and low torque
Name:	Challis clive.challis@ukaea.uk	Affiliation:	CCFE
Research Area:	ITER Demonstration Discharges	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	A complimentary approach is suggested to proposal 269. In the case of 269 it is envisaged to re-establish the ITER demonstration scenarios developed in 2008, and use the co-/counter-beam mix to scan rotation down to the ITER relevant values and determine the effect on confinement and fusion performance. In this parallel proposal the possibility is considered that the performance optimisation in terms of q-profile shape, heating waveforms, current drive, etc., developed in the presence of strong rotational shear and, to some extent, high plasma shaping, may not be fully optimised for operation with low applied torque and with the ITER plasma shape. It is proposed, therefore, to revisit this performance optimisation in condition closely approaching ITER for normalised rotation and plasma shape.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The proposed approach would use a steady-state scenario ITER demonstration pulse as a starting point. After first limiting the NBI power to near-balanced to achieve the ITER relevant normalised rotation, the q-profile shape would be varied (both the value of q-minimum and the level of magnetic shear in the plasma core) by varying the NBI power waveform, initial current ramp-up rate, prelude density and L-H transition time. Avoidance of performance degrading NTMs would be attempted using ECCD
Background:	The development of ITER demonstration discharges in DIII-D in the steady-state relevant domain has been a valuable step towards qualifying these plasma scenarios for future application to ITER. The capability of DIII-D to vary the applied NBI torque provides the important possibility to further evaluate the performance of the present scenarios with ITER relevant normalised rotation (see proposal 269 by Ed Doyle). However, the possibility exists that the steady-state scenario development already performed in the presence of high plasma rotation may not be fully optimised for operation at low torque. For example, the relative importance of magnetic shear and flow shear in providing the required transport levels for steady-state operation may be different at low rotation. In this case, the optimisation of the q-profile could be revisited with the aim to provide good confinement in the absence of strong flow shear. The same logic could also apply to the step towards ITER-like plasma shapes, where the contribution of the plasma shape in optimising confinement and stability may be affected. This could result in a shift in the relative importance of the performance optimisation in the core and edge regions of the plasma, which, it turn, may affect the criteria for the q-profile optimisation.
Resource Requirements:	Full NBI and ECCD power
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	495: Lower-B ELM-Control Target Plasma
Name:	Schaffer schaffer411@gmail.com	Affiliation:	Self-Employed
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	TBD	ITPA Joint Experiment :	No
Description:	Develop lower B target plasmas than 2T for ELM control and suppression experiments, e.g. at Bo = 1.4 and 1T. At Bo = 1T the available dB/Bo within I-coil and power supply limits is doubled. This will make it possible to explore both ELM suppression and RMP physics at lower beta, different collisionalities, different rotations, etc, than has been possible with the Bo = 2T that has been used in the past.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce a past ITER-similar-shape, 2T target plasma. Reduce Bo to 1.4T and then to 1T at constant q95 ~3.6. Beta match is not important at first, so develop the target plasmas first at e.g. beta_N ~1.5. Expect to finish in 0.5 day.
Background:	Most ELM suppression experiments in DIII-D have been done with Bo near 2T. Successful suppression seems to be facilitated by high beta_N, but these discharges are near stability boundaries and many shots are lost. The high bets_N is actually above ITER's main Scenario 2 mission, but at Bo=2T in D3 one frequently runs into the SPA current limit before either suppressing ELMs at lower beta. One also runs out of current while varying other parameters to study their effect on ELMs. While the C-supplies can deliver more I-coil current than SPAs, they lack experimental flexibility. Many of the ELM suppression experiments can be done at lower Bo and beta [if not we need to find this out and the reason behind such a surprise!] Lower Bo target plasmas should improve ELM control experimental productivity. It should also open access to experiments with larger dB/Bo, which would allow us to explore ELM control at new plasma parameters. It would be possible to apply both n=3 RMP and n=1 error correction with the I-coils alone.
Resource Requirements:	Usual setup for H-mode plasmas. NBI. I-coils for RMP, C-coils for error correction.
Diagnostic Requirements:	For this initial plasma development, we only need diagnostics for reliable operation. But it would be interesting to collect CER data as well.
Analysis Requirements:	surfmn, trip3d
Other Requirements:

Title	496: Reproducibility of q95 RMP ELM suppression widow
Name:	Evans evans@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to assess the reproducibility of the q95 RMP ELM suppression widow using slow q95 ramps. We will start by reproducing an established discharge (128473) that has BT=1.9 T and increasing Ip. We will attempt to reproduce the ELM suppression window with the same RMP coil currents on 2 successive discharges and fully document the diverrtor, SOL and pedestal parameters (including edge fluctuations). We will repeat these discharges with BT=2.15 T while increasing the Ip ramp down offset 13% in an attempt to reproduce the same ELM suppression window (the same center position and width). Next we will repeat the q95 scan with fixed Ip and a downward BT ramp. After this sequence, we will select the best conditions from above and do a 3 point NBI power scan starting with 5 co-NBI sources, then 4 and 3 from one shot to the next. The final step is to select the conditions with the largest q95 ELM suppression window and vary the RMP coil spectrum by turning off the C-coil and changing the toroidal phase of the I-coil for 0º to 60º with at least one repeat discharge for each condition. The objective this approach is to assess the reproducibility of each configuration and to acquire a full set of RMP/transport data which can be compared to the width of the ELM suppression window and TRIP3D field line loss calculations. This data is needed in order to develop techniques for increasing the width of the q95 ELM suppression window which is important for improving the performance of the ITER RMP coils since these coils need to be capable of suppressing ELMs under various operating conditions.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	1. Reproduce reference discharge and document. 2. Increase BT to 2.15T from 1.9 T with a 13% increase in the Ip ramp offset, reproduce the q95 RMP ELM suppression window and document. 3. Fix Ip at 20% above the top of q95 widow found during part 2 and ramp BT down slowly. Repeat and document. 4. Three point co-NBI power scan using the conditions from parts 1 through 3 with the largest q95 ELM suppression window. 5. Select the best Pinj from part 4 and vary the RMP coil spectrum. 6. Obtain full data documentations discharges for the best conditions from above on at least 3 repeat discharges (more if time permits).
Background:	Previous experiments with constant Bt and upward Ip ramps have resulted in approximately a factor of 2 variation in the width of the ELM suppression q95 window with similar Chirikov overlap parameters (e.g. 132741 and 128473). It is essential to know if the q95 ELM suppression window is highly reproducible or if the width of the window varies with parameters that are not measured in these experiments (e.g., wall conditions, ect.). Additionally, it is important to fully document any observable parameter variations, such as edge fluctuation levels or core MHD activity, that may lead to changes in the width of the suppression window since these may have indirect effects on the performance that are not yet understood. This experiment will focus on determining whether the suppression window is reproducible over multiple repeat discharges and if not what may be contributing to any variations observed.
Resource Requirements:	n=3 I-coil, 7 kA, C-coil, 5 co and 2 counter NBI sources
Diagnostic Requirements:	full RMP and fluctuation diagnostics.
Analysis Requirements:	TRIP3D, EMC3, fluctuations, profiles.
Other Requirements:	--

Title	497: The co- and counter- neutral beam switching experiment
Name:	Ida none	Affiliation:	National Institute for Fusion Science
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	Punit Gohil, Wayne Solomon	ITPA Joint Experiment :	No
Description:	The beam modulation is useful tools to study the non-diffusive term of momentum transport, which is identified as a spontaneous rotation or spontaneous torque. However, the simple modulation of the neutral beam causes the modulation of ion temperature gradient which affects the spontaneous torque. There are two approaches to minimize the modulation of temperature gradient. One is fast modulation and the other is switching the beam direction. If the modulation is fast compared with the slowing down time there is little effect on the heating and the torque can be modulated by the prompt JxB torque not by momentum input trough the slowing down process. When the co-injection beam and counter-injection beam is switched with the identical heating power, the modulation of the ion temperature can be minimized even slow modulation comparable to or longer than the slowing down time. The switching beam experiment was done in JFT-2M by switching the co-injection neutral beam to counter-injection neutral beam of counter-injection to co-injection (reference: K.Ida et al., PRL 74 (1995) 1990). This experiment was done in the L-mode plasma and only one switching in a discharge. In D-IIID, we can modulate at constant power using co/ctr beams which simplifies the matter and improve signal to noise ratio considerably by increasing the number of modulation. The multi modulation would contribute the improvement of quality in evaluating quantitatively the non-diffusive term of momentum transport.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The ion temperature gradient of ion pressure gradient is one of the candidate of driving mechanism of spontaneous rotation. In order to eliminate the change of ion temperature gradient during the modulation, I would like to propose a beam modulation experiment by switching co-NBI and ctr-NBI with keeping the total power to be constant in time. Furthermore, I would like to propose the power modulation with balanced injection to modulate the ion temperature gradient at constant torque. Although there is a modulation component due to the prompt jxB torque, this can be estimated. The modulation frequency depends on the energy or momentum confinement time in the plasma interested and time period of each beam should be long enough, 2-3 times of the energy and momentum confinement time. The target plasma would be L-mode, H-mode and VH-mode ot ITB plasma if possible. If one discharge can cover these three modes, it would be ideal.
Background:	I have been interested in the non-diffusive term of momentum transport since I have done the beam switching experiment in the L-mode in JFT-2M in 1995. Recent beam modulation experiments done in various tokamaks encourage me to do the beam switching experiment in the plasma with improved mode. From the experience in JFT-2M experiment, I feel that the beam switching is useful technique to study momentum transport. The DIII-D tokamak has a capability to switch the beam direction since co and counter beams have been installed. When my proposal is accepted in the research opportunities forum for the 2009 experimental campaign and scheduled in March to mid-September (not conflict to the LHD experiment in October to February), I intend to visit GA and participate the experiments, but at the moment I can not guaranty my visit.
Resource Requirements:	Co-injection and counter-injection neutral beam More details will be discussed later.
Diagnostic Requirements:	Time evolution of the toroidal rotation velocity profiles with charge exchange spectroscopy. More details will be discussed later.
Analysis Requirements:	Calculation of radial profiles of momentum torque input from the neutral beam.
Other Requirements:	It would be nice if the proposed experiment would be scheduled in mid-May to mid-September 2009 or March to mid-September 2010, which can avoid the conflict to the LHD experiment at NIFS.

Title	498: High central fast wave current drive efficiency at high electon beta with ECH preheating
Name:	Pinsker pinsker@fusion.gat.com	Affiliation:	GA
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	F.W. Baity, J.C. Hosea, M. Porkolab, A. Nagy, J.M Lohr, R. Prater	ITPA Joint Experiment :	No
Description:	Combine 6 gyrotrons-worth of central 110 GHz ECH (all launchers aimed at or near the center of the discharge without driving toroidal current) with the combined 60 MHz and 90 MHz fast wave power. We would use the minimum neutral power necessary to create the sawtooth-free discharge in which the driven currents can be accurately measured, and to make the (MSE) measurement. Both L-mode and H-mode target discharges would be tried, although at full rf power, one would expect difficulty in keeping the discharge in L-mode. The basic scan would be a density scan, at each case obtaining at least a matching pair of discharges with co- and counter-current FW phasing. The object of the exercise would be to extend the range of central electron beta values, and hence of single-pass absorption of the FW power, considerably beyond what was possible without high power ECH. If time permits, comparison of the current drive efficiency of the 60 MHz and 90 MHz systems could be performed - as the single-pass absorption increases, we expect at some point to observe more efficient current drive at higher launched parallel phase velocity (the higher frequency case). The experiment would seem to be the logical precursor to full utilization of the combined rf systems for AT work involving tailoring of the current profile.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	See description.
Background:	This experiment was tried on two days in the 2004-2005 campaign. However, technical problems prevented any useful data to be obtained. The 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	499: Where are the &%$#! antenna arcs occuring?
Name:	Pinsker pinsker@fusion.gat.com	Affiliation:	GA
Research Area:	General ITER Physics	Presentation time:	Not requested
Co-Author(s):	F.W. Baity, A. Nagy, J. Yu, J.C. Hosea, A.G. Kellman, D. Rasmussen, D.W. Swain, J. Caughmann	ITPA Joint Experiment :	No
Description:	Essentially all experiments that use high-power ICRF heating are limited in the power that can be coupled to a given plasma by high-voltage antenna arcs/breakdown. Yet in no experiment has the actual location in the antenna structure where the breakdown occurs been unambiguously determined, except in pathological cases. This data is absolutely needed to inform any reasonable design for a significantly improved antenna. In the 2008 campaign, DIII-D commissioned a data acquisition system of unprecedentedly high time resolution (up to 1 GHz sampling rate) to study the electric signatures of antenna breakdown at the rf timescale. The UCSD fast camera was also used to look at the 0 deg FWCD antenna for the first time. What is needed is to combine these diagnostics to attempt to image an arc on the 0 deg antenna while at the same time acquiring fast electrical data on the same event.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	What needs to be done is to trigger both the camera running at its maximum frame rate and the fast digitizer system with the same event trigger that tells the transmitter to blank the rf power, and to acquire both pre-trigger and post-trigger data. It may prove to be necessary to intentionally slow down the response of the blanking system to allow enough energy to be deposited in the arc to produce a visible image on the camera. The usual system shuts off the rf within about 20 microseconds of the arc detection. Perhaps an L-mode plasma with a reasonably large outer gap, without NBI, would be the simplest target plasma to start with.
Background:	The ITER ICRF antenna system will be required to operate at higher rf voltage levels than have ever been obtained in a reliable, long-pulse way in any previous experiment. The limitation has always been electrical breakdown in the antenna system, yet nobody has been able to actually determine the location of the arcs in existing antennas, presumably due to the successful fast shut-off of the rf power when an arc is detected. This is clearly important for learning how to achieve the needed reliability for ITER.
Resource Requirements:	Fast framing camera looking at an optimized view of the 0 deg antenna, fast digitizer system connected to raw rf samples on the same antenna system, triggering set up specially as required.
Diagnostic Requirements:	Special diagnostics set up as listed above in resource requirements.
Analysis Requirements:
Other Requirements:

Title	500: QH mode with NRMF driven rotation
Name:	Garofalo garofalo@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Solomon	ITPA Joint Experiment :	No
Description:	Goal of this experiment is to test access to QH mode operation with counter rotation driven by static nonresonant magnetic fields	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Start from reversed-Ip QH-mode plasma. This will have a large net counter-Ip NBI torque. Maximize betan. Optimize error field correction using RWM feedback driven C-coil. Apply large n=3 field using I-coil (~7 kA). This may have small effect on rotation, if (1) rotation is close to the neoclassical offset, or (2) if betan is still too low. Reduce NBI torque at constant betan, by adding co-Ip sources. In case (1), the rotation will change little. Look for effect on QH mode characteristics. In case (2), the rotation will drop significantly. Look for QH mode changes.
Background:	QH mode is a very desirable mode of operation for a fusion reactor or other burning plasma experiment. Doing away with neutral beam injection is also desirable. Without NBI, heating and current drive can be provided by ECH and ECCD or LHCD. Momentum injection can be provided by static, non-axisymmetric fields. Does the rotation profile driven by non-axisymmetric fields have the edge shear that seems to be necessary for QH mode operation?
Resource Requirements:	Reversed Ip. All NBI sources.
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	501: Characterize non-resonant magnetic torque offset rotation
Name:	Solomon solomon@fusion.gat.com	Affiliation:	GA
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	A.M. Garofalo, H. Reimerdes	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to characterize the so-called offset rotation associated with non-resonant magnetic fields to benchmark with theory.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Since the offset rotation is theoretically expected to depend on the Ti gradient, this will be the key parameter scan. Practically, this will likely entail a combination of betaN and density scans. The offset rotation was determined in 2008 by adjusting the input NBI torque to scan the initial rotation, until a case was found where essentially no change in the rotation was observed. Of course, this is a relatively expensive way (in terms of number of shots) to find the offset rotation for many condition. To maximize the amount of relevant data for characterization of the offset, we will instead use a "train" of I-coil pulses as the torque is slowly ramped down. The effect on the rotation from the NRMF should decrease as one approaches the offset, and change direction as we pass through it. Modulation techniques might allow this to be done even more efficiently; however, the consequent pacing of the ELMs makes such an approach problematic.
Background:	Experiments have indicated that the torque exerted by non-resonant magnetic fields (NRMF) tends to drag the plasma to a finite toroidal rotation in the counter Ip direction. Some estimates of the NRMF torque indicate that it may be two orders of magnitude larger than the NB-driven torque in ITER during RMP ELM suppression. Under such conditions, the rotation on ITER will rapidly evolve toward the offset rotation, and therefore careful characterization of the offset rotation is critical for rotation predictions on ITER.
Resource Requirements:	n=3 I-coil, reverse Ip
Diagnostic Requirements:	standard profile diagnostics
Analysis Requirements:	TRANSP
Other Requirements:

Title	502: RMP ELM Suppression Dependencies
Name:	Schaffer schaffer411@gmail.com	Affiliation:	Self-Employed
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Expect to work with Evans, Moyer, Fenstermacher and others TBD	ITPA Joint Experiment :	No
Description:	Using reduced magnetic field ELMing discharges, e.g. as developed in ROF 495, "Lower-B ELM-Control Target Plasma," determine dependencies of RWM ELM suppression on as many parameters as possible in 2009-10 runs. Highest priority are beta, density and rotation dependencies. If at all possible, the scans should attempt to reach a large, infrequent ELM regime like the 2008 ITER Demonstration shot. Operation at reduced field allows one to apply larger dB/Bo with SPAs and thus explore plasmas away from "sweet spot" operation. Approximate internal magnetic fields (in non-island internal regions) will be computed by IPEC and correlated with ELM suppression and other experimental outcomes.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Run at reduced field, to apply larger dB/Bo with SPAs and expand possible ELM suppression operating space. Vary beta and rotation by the established NBI techniques. Run outer divertor strike point for good exhaust pumping, and then vary density. Because recent IPEC results show that n=3 I-coil RMP fields ARE affected significantly by plasma response, the data from the new experiments with new equilibria and new parameters will also be analyzed by IPEC and the results correlated with ELM suppression and/or other experimental outcomes.
Background:	Because of the urgency of establishing and studying RMP ELM suppression in ITER-like plasmas, only a minimum of plasma development was done. These target plasmas have tended to favor ELM suppression. ELM suppression at other parameters appears to be more difficult, but it is not know by how much. The practicality of RMP depends on this knowledge. The previous target plasmas run near stability limits and offer limited possibilities for parameter scans. Also, almost all the target plasmas have been at 2 tesla, but then the I-coil power supplies can not give enough dB/Bo for some experiments. The proposed experiment will start from lower beta and Bo, to have more stability margin and greater dB/Bo, and thus better possibility of meaningful parameter scans.
Resource Requirements:	ITER-similar shape. Standard Bo and Ip directions. NBI heating.
Diagnostic Requirements:	Thomson, CER, MSE, Reflectometry, other selected diagnostics
Analysis Requirements:	Kinetic EFITs. IPEC analysis.
Other Requirements:	--

Title	503: Single-Row ELM Suppression vs q95
Name:	Schaffer schaffer411@gmail.com	Affiliation:	Self-Employed
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	Expect: Evans, Moyer, Fenstermacher, others TBD	ITPA Joint Experiment :	No
Description:	Test the hypothesis (see Background below) that the n=3 magnetic field from a single I-coil row should suppress ELMs over a wide range of q95. Also, complete the search for ELM suppression by the C-coil with a q95 scan. The results will help us understand and formulate requirements for RMP ELM suppression.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce conditions close to the 2008 single-row ELM suppression experiments. Scan q95 while applying single-row n=3 I-coil field (upper, then lower) and then C-coil n=3 field.
Background:	ELMS were successfully suppressed in 2008 by just a single row of I-coils with n=3 currents, rather than the usual two. The SURFMN-computed magnetic spectrum had just two very broad peaks; in particular, it had no narrow peak aligned with the resonance locus m=nq. The usual two-row I-coil even parity current distribution makes a magnetic field with a well defined spectral peak, and experimental ELM suppression is correlated with resonance or near resonance between m=nq and this peak. These observations motivate the hypothesis: A "non resonant" single-row I-coil field should suppress ELMs over a wide range of q95. In contrast, the single-row C-coil has essentially a single-peak n=3 magnetic spectrum, but so far it has not produced ELM suppression.
Resource Requirements:	Standard ITER-similar shape, RMP ELM suppression target plasma. NBI.
Diagnostic Requirements:	Thomson scattering, MSE, CER, reflectometry, other selected diagnostics.
Analysis Requirements:	SURFMN harmonic analysis
Other Requirements:

Title	504: Phirsch-Schluter current in spiral footprints
Name:	Schaffer schaffer411@gmail.com	Affiliation:	Self-Employed
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Do a preliminary piggyback search for signs of Phirsch-Schluter currents in spiral divertor footprints produced by resonant magnetic field (RMP) application. During q95 sweeps, the spirals sweep across the fixed divertor Langmuir probes. The probes will be operated at constant voltage = local vessel voltage, in order to measure the natural current from plasma to tiles. This initial experiments should be treated as "discovery science", after which more rigorous experiments can be formulated, if warranted.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Piggyback on q95 sweeps, which cause the spirals to sweep across the fixed divertor Langmuir probes. The probes will be operated at constant voltage = local vessel voltage, in order to measure the natural current from plasma to tiles.
Background:	The natural current in the diverted scrape-off layer (SOL) has thermoelectric and a Phirsch-Schluter (PS) components. Under some circumstances in axisymmetric SOLs the two can be separated. The PS current was shown to be both theoretically and experimentally proportional to the pressure gradient p' in the SOL. It is not clear that a similar simple interpretation will be possible in non axisymmetric spiral divertor footprints, but it is worth taking a quick piggyback look for signs of Pfirsch-Schluter currents in the spirals.
Resource Requirements:	Piggyback on q95 sweeps during RMP experiments.
Diagnostic Requirements:	Fixed divertor Langmuir probes running at constant voltage = local vessel voltage, in order to measure current from plasma to tiles.
Analysis Requirements:	Initially, look for radially narrow features in the profile of J_plasma plotted as a function of a radial coordinate. If something interesting is seen, and depending on its nature, we can try to develop analysis for the new non axisymmetric system.
Other Requirements:

Title	505: Suppression of first ELM at zero net torque
Name:	Evans evans@fusion.gat.com	Affiliation:	GA
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The goal of this experiment is to produce L-H transitions with balanced co- and counter NBI (i.e., zero net torque) during the application of n=3 RMPs and to suppress the first ELM. This experiment will provide key information for ITER operations during low torque L-H transitions with various levels of applied RMPs including the influence of RMPs on the properties of the L-H transition, the ELM-free phase and the resulting ELM suppressed state when the required q95 resonance conditions are satisfied.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	1. Obtain a reference ohmic discharge with q95 = 3.5 during the current plateau (e.g., RMP H-mode discharge 126006 with no NBI heating) and document edge fluctuations, Thomson and ECE profiles, divertor conditions and radiation properties. 2. Repeat with n=3 I-coil, even parity, RMP steps from 4->6 kA and document any changes observed. 3. Add modulated 210 R and 330 R beams starting 500 ms after the Ip plateau with low duty cycle (i.e., 5 ms on and 95 ms off) while maintaining zero net torque and/or rotation across the edge of the plasma (Psi_N=0.85-1.0). Repeat with 6 kA of n=3, even parity, I-coil current. 4. Increase NBI duty cycle in steps of 5 ms with and without n=3 RMPs until an L-H transition is obtained. Fully document the properties of the L-H transition, ELM-free phase and ELM suppressed phase with and without the n=3 RMP. 5. Using the best condition found in step 4, reduce the I-coil current until no effects is seen.
Background:	L-H transitions in ITER will be obtained at relatively low toroidal rotation (low NBI torque). Theory predicts that the penetration of the RMP field is affected by the plasma rotation. It is important to suppress the first ELM following the L-H transition in ITER. Thus, experiments are needed to better understand the effects of RMPs on the L-H transition, the ELM-free phase and the ELM suppressed phase at low torque (rotation). This experiment will provide answers to an number of questions needed to plan for ITER RMP H-mode operations.
Resource Requirements:	Standard RMP H-mode operations and diagnostics (I-coil, C-coil, SPAs, C-supplies). 210 and 330 beams.
Diagnostic Requirements:	Divertor, fluctuation and profile diagnostics.
Analysis Requirements:	SOLPS5, TRIP3D, EMC3-EIRENE, profile analysis.
Other Requirements:

Title	506: D injection for quantification of the recycling flux in the detached outer divertor of DIII-
Name:	Brezinsek none	Affiliation:	FZJ/Germany
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	N. Brooks, A. McLean, M. Groth, E. Hollmann	ITPA Joint Experiment :	Yes
Description:	The goal of this experiment is to determine the deuterium flux at the outer strike point under detached plasma conditions. The photon efficiencies of different molecular and atomic transitions, as well as the recycling flux are quantified by in-situ calibration of spectroscopic signals using deuterium injections through the DiMES Porous Plug Injector (PPI). Additional information about the atomic and molecular physics in recombining plasmas like e.g. the occurrence and strength of MAR will be obtained. Note that the quantification of chemical erosion yields of carbon relies on the determination of the incident ion flux on the graphite surface. The recycling flux is equivalent to incident ion flux from the spectroscopic point of view, and thus, the knowledge of the recycling flux is necessary to quantify the erosion yield.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The general configuration is an L-mode discharge in almost pure deuterium (hydrogen) with lower single null configuration, high electron density and long current flat-top phase. â?¢ Set-up shots with the outer strike point fixed on the PPI cap. The electron density is increased by applying additional fuel. The outer divertor plasma changes from the attached to the detached regime. A reference shot without deuterium injection and a shot with constant deuterium puffing through the PPI are used to determine the signal strength. â?¢ One diagnostic optimised shot in the attached divertor with the strike-point fixed on the PPI cap. A high injection rate of deuterium is used to obtain a best as possible spectroscopic signal of the molecular Fulcher band recorded with the Mechelle system in high spectral resolution. â?¢ A set of about six identical plasma discharges with two strike point sweeps over the PPI cap in the attached outer divertor. The first sweep is used as the reference for the recycling case whereas the second one used for the in-situ calibration with a moderate local deuterium injection through the PPI. Both sweeps are performed stepwise with three constant plateaus to allow an injection in the PFR, near and far SOL. Different vibrational transitions of the D2 Fulcher-band (v=0..4) as well as atomic transitions (Balmer lines) are recorded shot by shot with the change of the interference filters applied in the tangential divertor camera as well as the setting of the MDS system. â?¢ A set of about six identical plasma discharges with two strike point sweeps over the PPI cap in the detached outer divertor. The first sweep is used as the reference for the recycling case whereas the second one used for the in-situ calibration with a moderate local deuterium injection through the PPI. Both sweeps are performed stepwise with three constant plateaus to allow an injection in the PFR, near and far SOL. Different vibrational transitions of the D2 Fulcher-band (v=0..4) as well as atomic transitions (Balmer lines) are recorded shot by shot with the change of the interference filters applied in the tangential divertor camera as well as the setting of the MDS system.
Background:	The DIII-D tokamak has the unique feature of the DiMES Porous Plug Injector (PPI) which is able to inject different gaseous and simulate wall sources. In combination with the good diagnostic access in the outer divertor complementary experiments to TEXTOR (ionising plasma regime) can be performed. The use of a high resolution overview spectrometer (possible loan from the FZJ) will provide additional information about the ro-vibrational population of the D2 Fulcher band. The ratio of atoms to molecules as well as the absolute photon and particle fluxes of molecular and atomic transitions can be deduced from the measurements. Modelling of the background plasma with plasma edge codes including the neutral particle code EIRENE can be benchmarked with the injection.
Resource Requirements:	Machine Time: 1/2 experimental day Number of gyrotrons: none Number of neutral beam sources: one Other requirements: none
Diagnostic Requirements:	MDS, lower divertor IR camera, DiMES TV, divertor Thomson scattering, fixed floor Langmuir probes, midplane and X-point reciprocating probes, filterscopes, CER, tangential divertor camera, ocean optics overview spectrometer, Spectrelle spectrometer from FZJ (high resolution overview spectrometer)
Analysis Requirements:
Other Requirements:	Part of the DSOL-2 ITPA task

Title	507: Measurement of Thermo-electrically Driven Scrape-Off-Layer Current (SOLC) in DIII-D Discharges
Name:	Takahashi htakahashi@pppl.gov	Affiliation:	PPPL
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	H. Takahashi, E. Fredrickson, M. Schafer, J. Watkins, and Others	ITPA Joint Experiment :	No
Description:	Measure a radial distribution of current that flows along open field lines in the SOL, together with the electron temperature and density at plasma sheath on the divertor tile surface, during a quiescent phase of NBI-heated H-mode tokamak discharge, and relate characteristics of the observed distribution of SOLC to its possible origins, including thermo-electric potential and Pfirsch-Schlueter effect.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Create a steadily ELMing moderate betaN LSN discharge, similar to those used in the past basic ELM studies, but with a somewhat lower x-point height (with a resulting smaller private flux region) and ample room above the plasma top. This discharge configuration should allow connection between inner and outer divertors, including far SOL regions, along flux surfaces uninterrupted by the top outboard divertor baffle. Langmuir probe measurement is made in ion-saturation current, zero-bias, and swept-bias modes in three successive discharges. Inner and outer strike point locations are slowly swept during a several hundred ms period over a modest distance comparable to inter-probe separations. Repeat the measurement in discharges at three NBI power (betaN) levels to vary sheath conditions. A minimum of about 12 discharges, including set-up shots, is anticipated to fully accomplish the proposed experiment.
Background:	Current has been observed to flow in the SOL in NBI-heated tokamak discharges, both during quiescent and MHD-active periods. Candidate mechanisms for driving SOLC include thermo-electric potential difference between the two ends of open field lines in the SOL, Pfirsch-Schlueter effect in the SOL plasma, and time-varying flux coupling to MHD instability and plasma current. Flux coupling is an intuitively plausible mechanism responsible for at least parts of SOLC observed during MHD activity. Thermo-electric potential and Pfirsch-Schlueter effect are theoretically considered candidates for SOLC observed during quiescent discharge periods, but not yet clearly established experimentally. Abrupt changes in SOLC characteristics are observed at the onset of MHD activity. Elucidating origins of SOLC would contribute importantly to understanding whether it is involved in physical processes that trigger MHD activity.
Resource Requirements:	None
Diagnostic Requirements:	Essential diagnostic is the Langmuir probe arrays, and desirable diagnostic is Thomson scattering sampled at short intervals.
Analysis Requirements:	Standard and specialized Langmuir probe signal analyses
Other Requirements:	None

Title	508: Core turbulence evolution and transport in RMP H-modes
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	T.L. Rhodes, L. Zeng, G.R. McKee, R. Moyer, W.A. Peebles, E.J. Doyle, J.C. Hillesheim, G. Wang	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to map out H-mode core turbulence and correlate turbulence changes with radial particle and momentum transport in RMP H-modes. This experiment may provide insight into the physics of anomalous momentum transport and density pump-out in the presence of resonant non-axisymmetric fields.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Recent diagnostic improvements and the implementation of a four-channel DBS system will allow us to follow the evolution of density fluctuations in the core of H-modes as the applied non-axisymmetric field is increased. Turbulence measurements can be made across a range of poloidal wavenumbers (2cm^-1< k_perp < 8 cm^-1) by DBS and BES. Radial mapping of core turbulence for r/a > 0.3 can be achieved (ref. shot 129193), using multi-channel DBS systems.
Background:	Substantial effects of RMP fields on core toroidal rotation and core transport have been observed. Very preliminary fluctuation measurements indicate changes in intermediate scale turbulence inside the H-mode pedestal during I-coil application (#129193)
Resource Requirements:	7 Beams, I-coil
Diagnostic Requirements:	DBS, BES, MSE, FIR, high-k scattering, PCI
Analysis Requirements:
Other Requirements:

Title	509: ETG turbulence scaling in RMP H-modes
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	T.L. Rhodes, E.J. Doyle, L. Zeng, G.R. McKee, W.A. Peebles, J.C. Hillesheim, G. Wang	ITPA Joint Experiment :	No
Description:	The goal of this experiment is the investigation of ETG turbulence and electron transport in ELM-controlled RMP plasmas where the critical gradient is varied and the direct comparison with Gyro-calculated fluctuation levels and transport fluxes. Scaling with Te/Ti will be investigated by using off-axis ECH to control the local density and temperature gradient for 0.5 < r/a < 0.8. Measurements will be done for two different values of q-edge using different RMP resonant windows.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	ITG turbulence is expected to be quenched in strongly rotating H-mode plasmas with RMP ELM control. RMP plasmas offer a unique opportunity to vary the ETG critical gradient in ITG-stable plasmas via the density gradient which is can be controlled by the I-coil current and local ECH deposition for 0.5 < r/a < 0.8. Recent diagnostic improvements and the implementation of a four-channel DBS system will allow us to map intermediate scale and ETG-scale density fluctuations in the core of RMP H-modes. Turbulence measurements will be made across a range of poloidal mode numbers (0.1< k_perp rho_s< 5 cm^-1) by DBS and BES, and k_perp rho_s ~ 10 by high-k scattering. In the first part of the experiment we will attempt to vary the ETG critical gradient and record fluctuation levels at two radial locations. This data will be used together with transport analysis to compare with GYRO-calculated ocal fluctuation levels and transport fluxes. In the second part of the experiment Te/Ti and q_edge will be varied to obtain scaling data (2 values of q_edge and three values of Te/Ti). Off-axis ECH using six gyrotrons will be employed to vary Te/Ti.
Background:	Recent DBS measurements have identified regimes where ITG-scale and intermediate turbulence is quenched or greatly reduced in H-mode and QH-mode plasmas. TGLF calculations have helped in identifying a similar regime in RMP H-modes. Absence of ITG scale turbulence will enable investigations of the ETG threshold and ETG fluctuation levels in these plasmas, and the scaling of ETG-related electron transport vs. Te/Ti and local safety factor.
Resource Requirements:	7 Beams, I-coil, ECH 6 gyrotrons
Diagnostic Requirements:	DBS, high-k scattering, BES, PCI, MSE
Analysis Requirements:	Local GYRO high-k runs for preparation/optimization of experiment and data analysis
Other Requirements:	--

Title	510: Avoidance of disruptions due to the NTM-locking with feedback application
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Outline The disruption due to the tearing-mode-locking has been considered as a potentially-serious obstacle for reactors. The mode-locking is governed by the external non-axisymmetric field environment, which can be, in principle, correctable and at least controllable. The forced-mode rotation by the feedback field is one approach to overcome the problem, if the applied field can synchronize and sustain the mode rotation during the shutdown period. The objective is to reduce plasma pressure and plasma current while the mode is sustained rotating.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Approach In the DIIID experiment a few years ago, we observed an example of synchronization between the NTM and external rotating field using feedback. A shot (#127927) showed the application of feedback could shift the mode rotating frequency from > 1 kHz to 10 Hz. The shift of the feedback toroidal phase forward from -30 deg to +30 deg including the polarity change could redirect the mode rotation from -10 Hz to + 50 Hz without losing the stored energy (according to the diamagnetic loop signal). The proper choice of toroidal phase shift was found to be important. The successibility of the frequency conversion and the mode rotation sustainment depends on factors such as the rotation profile, NBI torque input, magnetic shear profile, and the size of magnetic island. This experiment: (1) To synchronize the NTM and the feedback field and to optimize the mode rotation with feedback toroidal phase forward as was #127927 (2) To clarify the critical factors for mode synchronization such as the rotation profile, NBI torque input, magnetic shear profile, and the size of magnetic island (3) To explore the discharge shutdown by lowering the plasma current and plasma pressure while sustaining /increasing the mode rotation frequency.
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	511: Active MHD spectroscopy under feedback stabilized plasmas
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	RWM Physics including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	APPROACH step-1. To suppress the current driven RWM by feedback on with proportional only modest gain step-2. To reduce the error field by C-coil DEFC step-3. To add extra offset rotating field step-4. To scan the frequency with a fixed feedback gain, step-5. To observe the request of coil current δI with modest feedback gain. The ratio of δI and preprogrammed offset current Io is KG/(1-KG), where G is the PCS gain and K is the plasma RFA factor. Using the frequency dependence of this ratio, it should be possible to determine whether the K as a complex value reflecting some dissipation process Comments - The toroidal phase shift of K is expected to be quite different from the one obtained by rotationally stabilized plasma. - If large phase shift exists, other possibility like edge-phenomena should be investigated Overall objective of this experiment is to investigate the similarity of RWM stabilized by dissipation and by feedback and to assess the possibility of other stabilizing mechanisms. It was shown that the feedback can suppress current-driven RWM. The high betan plasmas above the no-wall limit can be stabilized by to-be-identified dissipation mechanism, but so far, the stabilizing mechanism has not been well understood yet. Here, it is intended to apply the active MHD spectroscopy to the feedback stabilized plasmas and to assess the plasma response, in particular, toroidal phase shift to the extra applied rotating field.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	512: Determination of the Intrinsic Turbulence Dispersion Properties in the Plasma Reference Frame
Name:	Hillesheim jon.hillesheim@ukaea.uk	Affiliation:	CCFE
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	W.A. Peebles, T.L. Rhodes, L. Schmitz, T.A. Carter	ITPA Joint Experiment :	No
Description:	Use Doppler Backscattering (DBS) to measure the dispersion properties of turbulence in the plasma frame of reference and compare these measurements, along with multiscale, multifield turbulence measurements from other diagnostics, to simulation results from TGLF and GYRO. DBS measures the Doppler frequency shift associated with the propagation velocity of turbulence in the laboratory frame at a well-defined wavevector. This velocity includes a contribution from both the ExB background flow of the plasma and from the intrinsic phase velocity of the turbulence. When the local electric field is zero, which can be determined directly from DBS measurements, only the intrinsic turbulence phase velocity contributes to the measured frequency spectra. Under these conditions a direct measurement of the dispersion properties of the turbulence is possible in the rest frame of the plasma. From DBS it is also possible to obtain radial profiles of the radial correlation length, autocorrelation time, and relative fluctuation level at intermediate wavenumbers. These local measurements along with conventional reflectometry, BES, CECE, and other fluctuation diagnostics would be integrated to provide a detailed picture of the intrinsic turbulence properties measured in the plasma frame. This measurement approach removes a major source of uncertainty (knowledge of Er and potentially Er shear) for comparison with TGLF and GYRO computations. Since DBS measures the turbulence frequency spectra at well-defined intermediate-k wavenumbers, the proposed experiment will provide the first measurement of the intrinsic turbulence dispersion properties in the core of a high-performance fusion plasma.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish a stationary L-mode plasma with one counter-directed neutral beam source such that the rotation is negative and under conditions such that additional neutral beam blips do not cause a transition to H-mode. Co-directed neutral beam blips are then applied to the stationary target plasma with a sufficient momentum input to cause the local electric field to move slowly through zero. As the background electric field moves through zero the measured frequency spectra is then determined by the intrinsic phase velocity of the turbulence at the chosen intermediate-k. The point at which the electric field becomes zero is determined by identifying the time period when the measured frequency broadening is minimized, and not the time when the mean frequency equals zero. During this time period, the measured offset from zero and any residual broadening is determined solely by the dispersion properties of the underlying turbulence. The neutral beam blips would be applied in increasing duration (10 ms, then 20 ms, then 30 ms, etc.) in an attempt to produce cases where the zero radial electric field condition occurs under slightly different transient conditions so that it will be possible to estimate any changes to the dispersion from the beam blips, or at least to be aware of any such modifications. Sufficient time between beam blips would be allowed for the beam induced rotation to damp, so that the rotation is not slowly increasing throughout the shot. DBS measurements would be taken simultaneously at several radial locations during each shot. On subsequent shots the same plasma would be run, but the DBS launch angle would be adjusted so that different wavenumbers are probed. From these measurements a radial profile of the dispersion relation, omega(k), would be built up. The measured dispersion properties, along with data from the full complement of DIII-D fluctuation measurements, would then be compared to TGLF and GYRO results.
Background:	During 2008 a handful of isolated cases have been found where DBS measurements of the turbulence dispersion properties in the plasma frame are possible.
Resource Requirements:	Machine time: 1 day, might be able to be combined with other proposals Neutral beams
Diagnostic Requirements:	DBS. Profile reflectometer. BES, FIR, high-k backscattering, other fluctuation diagnostic. ECE, TS, etc.
Analysis Requirements:	TGLF, GYRO
Other Requirements:

Title	513: Flow damping in response to transient momentum input
Name:	Hillesheim jon.hillesheim@ukaea.uk	Affiliation:	CCFE
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	W.A. Peebles, T.L. Rhodes, L. Schmitz, T.A. Carter	ITPA Joint Experiment :	No
Description:	Doppler backscattering (DBS) measures the propagation velocity of turbulence in the laboratory frame, which is usually dominated by the plasma ExB flow. This allows highly localized (� 1 cm) and temporally resolved measurements of the ExB flow and the radial electric field. Excellent signal to noise ratios can be obtained under favorable operating conditions enabling a time resolution of ��10 us. It is proposed to use this capability to study the radial profile of the response to transient momentum input from neutral beam blips and the associated damping of the induced flow, even in the absence of any external background momentum input e.g. beam blips into Ohmic, ECH, balanced NBI plasmas.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Establish a stationary, diverted Ohmic plasma. Apply a sequence of ~ 1-10 ms duration neutral beam blips which are long enough in duration to produce a sufficiently large rotation perturbation to allow the subsequent damping rate to be measured, but short enough to minimize any background temperature/density changes that might result from the beams. The sequence would include both co- and counter- directed blips. A sufficiently long period between beam blips would be allowed for the flow to be completely damped. This would be the baseline measurement of the ExB flow damping in the absence of any continuous background external momentum input. The same sequence of beam blips in the same plasma shape would then be run, but with the addition of a co-directed beam for the entire shot. This would be repeated using a counter-directed beam. Finally the case of balanced injection would be studied. To avoid H-mode, where DBS measurements are problematic due to the flattening of the density profile, either a plasma which would remain in L-mode would have to be used, or the beam power would have to be reduced. Finally, two cases with background ECH instead of beam heating would be studied, one deposited near axis, the other off axis. The flow damping rate together with the amplitude of intermediate-k density turbulence would be measured via DBS at several radial locations across the minor radius for each shot with DBS. Other fluctuation diagnostics (FIR scattering, high-k backscattering, BES, PCI, CECE, etc.) would also be used to measure modifications to the background turbulence. The measured damping rates would then be compared to neoclassical predictions. Any anomalous viscosity could then be quantified under varying background operating conditions. Previous experiments have shown significant differences in the turbulence characteristics for ECH heated and NBI heated plasmas. Previous experiments have also shown modifications to the electric field and to the electric field shear during ECH. It would be expected that ECH off-axis heating would result in a larger modification to the electric field near the DBS measurement locations. The turbulence measurements would be used to identify the mechanisms responsible for any differences between neoclassical predictions and measurements of the viscosity.
Background:	During 2008, initial data of flow damping following beam blips was acquired for sporadic cases in the course of other experiments.
Resource Requirements:	Machine time: 1 day, or combined with other experiments where DBS accessibility requirements are taken into consideration
Diagnostic Requirements:	DBS. Profile reflectometer. FIR scattering, high-k backscattering, BES, PCI, CECE, etc. ECE, TS, etc.
Analysis Requirements:
Other Requirements:

Title	514: Measurement of Neoclassical Edge Current by Lithium Beam Spectroscopy
Name:	Hudson bfhudson@ucsd.edu	Affiliation:	UCSD
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	C. C. Petty, D. M. Thomas, C. Holcomb, M. Makowski, T. Osborne, L. Lao	ITPA Joint Experiment :	No
Description:	Use the lithium beam diagnostic to measure the neoclassical current in the plasma edge.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	There are two target plasmas for this experiment. An ELM-free H-Mode, at high Bt, with as much beam power as possible would maximize the edge pedestal gradient and thus the edge current. The ELM-free period would have to be at least 300 ms in duration for the expected lithium beam S/N to be sufficient. The other target plasma would be a QH-mode plasma, with the same plasma parameters as the ELM-free H-Mode if possible. The QH-mode plasma may have a longer time window for collecting lithium beam data, and if plasma parameters can be matched, a comparison to the ELM-free H-Mode discharge could be made.
Background:	A reliable measurement of the edge current is critical for testing of the peeling-ballooning stability model. MSE and Lithium beam diagnostics provide a set of internal magnetic field constraints used in equilibrium reconstruction. The edge current resulting from the reconstructions presently have little constraint in the edge, i.e. FFâ?? is not determined in the Grad-Shafranov equation, which is what is solved to obtain the current profile. The lithium beam is a strong, local, measurement of the poloidal field and compliments the MSE data in the edge. If lithium beam data of sufficient resolution in field pitch angle is obtained, direct calculation of the edge current profile can be made as has been successfully done in ECCD experiments. This experiment is predicated on a successful recommission of the lithium beam and likely would not occur until the 2010 run campaign.
Resource Requirements:	ECE, NBI (210 RT for MSE, 30 LT for MSE, 330 RT for CER). Machine time: 1 day for a combination of ELM-free H-Mode and QH-mode discharges.
Diagnostic Requirements:	Thomson, CER, MSE, lithium beam
Analysis Requirements:
Other Requirements:

Title	515: DiMES test of in situ carbon coating for FDF, reactor PFCs
Name:	Stangeby peter.stangeby@utoronto.ca	Affiliation:	U of Toronto
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):	--	ITPA Joint Experiment :	No
Description:	--	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	--
Background:	Tungsten PFCs may not be compatible with AT operation since AT tends to involve high plasma temperature at the divertor targets, resulting in high sputtering rates. Concentrations of high-Z impurities in the confined plasma must be kept to very low levels. Much higher concentrations are permissible for the low-Z refractory carbon; however, graphite PFCs may not be compatible with the high neutron fluence of devices like FDF due to swelling damage. It has been recently discovered that tungsten also suffers from an enhanced erosion effect due to He bombardment (simultaneous D + T + 5-10% He, as will occur in burning plasmas), creating a surface â??fuzzâ??. A solution to the latter problem has been proposed: coating the W with a low-Z material such as carbon or boron. Coating W with a low-Z material could also solve the problem of unacceptable high concentrations of W in the confined plasma; if the W substrate is always covered by the low-Z coating then only low-Z impurities would be present in the plasma. The low-Z coatings would be consumed and replenished continuously and so neutron damage would not be an issue. In a high duty cycle device plasma erosion also creates the basic problem of loss of structure, requiring replacement of the substance of the PFC. Continuous refurbishment of the plasma-facing surfaces with coatings could also solve this problem. Reactors will operate with hot walls, e.g. 700C, for reasons of thermal efficiency. At such high temperatures there is very little tritium retention in any materials, including carbon. It is also proposed that FDF operate with hot walls. Low-Z coatings in hot wall devices will result in little tritium retention by the co-deposition process, which is a serious issue for cold (200-300C) wall devices like ITER. In devices with high wall temperatures the tritium problem is less likely to be one of retention than of permeation â?? into the cooling system. Low-Z coatings on a W-substrate may also act as a permeation barrier. The continual replenishing of the low-Z coatings will inevitably result in thick deposits accumulating at some locations inside the vessel which, even if they contain very little tritium, will cause other problems, including dust formation. It will therefore be as important to be able to remove, in situ, the low-Z material as to carry out the in situ coating in the first place. Carbon would appear to be unique in this regard since it is possible both to introduce it into the vessel as a gas, e.g. as methane, and to remove it as a gas â?? as CO and CO2 using oxygen baking. Since little tritium would be contained in the carbon deposits, there would be little creation of tritiated water, which is problematic to handle and process. For locations experiencing low plasma fluxes the coatings might last long enough that their refurbishment could be carried out during â??downâ?? periods. For other locations such as the divertor strike point region, however, the coatings would probably have to be continually refurbished, i.e. during the normal plasma operation. This might be done by puffing methane or other hydrocarbon gases into the tile gaps in such regions during plasma operation. It is difficult/impossible to reliably predict what injection rate would be required to keep the coating in such locations at constant thickness. Experiments are required to provide an empirical basis for projection. The DIII-D DiMES facility provides an excellent opportunity for establishing the empirical basis for assessing this potential solution for FDF and reactors.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	516: 4/3 NTM as an n=3 RMP suppressing ELMs
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Check whether "internal" n=3 perturbations exerted by a 4/3 NTM are as effective as external n=3 RMPs in suppressing ELMs, in certain ranges of q95. If it works, exciting a 4/3 NTM and tuning q95 might represent a new, relatively simple recipe for an ELM-free or small-ELMs scenario, with no need for external coils.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	For the sake of comparison, start with a discharge normally used for RMP control of ELMs, including the q95 ramp. Turn off the external n=3 RMPs and try to excite a 4/3 NTM, e.g. by increasing beta. The simultaneous presence of 2/1 or 3/2 NTMs is permitted, but is preferable to suppress them by means of ECCD in order to make free energy available and ease the excitation of the 4/3. In case of difficulties in adapting ELM-control shots to 4/3 NTMs, try the opposite approach: begin with a discharge which is known to develop 4/3 NTMs, ad a q95 ramp-down and gradually make the shot as similar to ELM-control discharges as possible, but without RMPs. If difficulties persist, follow ECH/ECCD strategy of proposals 439 and 440, "NTMs on demand".
Background:	n=3 externally generated RMPs are known to suppress ELMs in DIII-D. Our aim is to test whether "internal" n=3 perturbations, exerted for instance by a 4/3 NTM, can be equally effective in stabilizing ELMs.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	517: ECH/ECCD modulated in the rotating ELM filament
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	M. Henderson (ITER)	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 doesn'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:	5-6 gyrotrons
Diagnostic Requirements:	Modified D_alpha diagnostic. Modified oblique-ECE "box" in the annex.
Analysis Requirements:
Other Requirements:

Title	518: Improved entrainment, with diagnostic applications
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	R. La Haye, G. Jackson	ITPA Joint Experiment :	No
Description:	Apply rotating n=1 error field to unlock and spin up to 300Hz an initially locked mode. Try to reproduce rotational mitigation at 10Hz. Resolve (forcefully) rotated islands with diagnostics such as CER and MSE which normally have too little temporal resolution.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Use dud detectors as in #432 to trigger an I-coil travelling wave right after mode locking. Pre-program concomitant ramps of frequency and amplitude of travelling wave. Begin with 1-300Hz, 1-2kA, 2s linear ramps. Experimentally (and with help from calculus of variations?) make ramps slow enough to avoid slipping but fast enough to avoid that the mode grows too much and disrupts. For fixed frequency ramp, improve amplitude ramp: make it as steep as necessary to defeat shielding at high frequencies by image currents in the wall, but not too steep, as excessive amplitude results in excessive error-field penetration and possible disruption. For higher current, the SPAs will be adopted as I-coil power supplies. Even so, the amplitude limit will be hit before the frequency limit. At that time, amplitude will kept to the maximum or slowly ramped down, according to SPA capabilities, while the frequency will be further increased. The intention is to maintain the coupling between the travelling wave and the mode as long as possible, even if the travelling wave is not as intense as desired. Non-linear ramps could help in this respect. We will also try to reproduce and understand the unexpected mode mitigation obtained during the frequency ramp at about 10Hz, for which various interpretations have been formulated. If successful, we will compare post-locking intervention (present proposal) with the more difficult and ambitious pre-locking intervention (proposal 433), i.e. with the attempt to "catch the mode" while it slows down and sustain its rotation without letting it lock at all.
Background:	Sustained rotation (a.k.a. entrainment) has been already demonstrated, but at relatively modest frequencies of up to 60Hz, possibly 180Hz (unclear, waiting for confirmation). Sustained rotation at higher velocities, of the order of 200-1000Hz has a number of advantages, ranging from locking avoidance, rotational mitigation, an easier ECCD modulation, accessibility to diagnostics which are normally too slow to resolve NTMs. The frequencies considered here are accessible to most diagnostics, including MSE and CER. Much higher frequencies are impossible for the SPAs, or are possible but at too low amplitudes. When compensation from image currents in the wall is taken into account, the effect in the plasma would be far too weak.
Resource Requirements:	SPAs
Diagnostic Requirements:	CER, MSE
Analysis Requirements:	--
Other Requirements:	--

Title	519: Easier modulated ECCD on forcefully rotated mode
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	R. La Haye, A. Welander	ITPA Joint Experiment :	No
Description:	Demonstrate a new NTM stabilization technique in which, instead of adapting the ECCD modulation to the natural, non-uniform mode rotation, the mode is magnetically forced to rotate at a known frequency and with a known phase, and the ECCD is modulated accordingly.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Force a tearing mode to rotate at a prescribed frequency and with known phase. At the same time, inject ECCD and modulate it at that frequency and phase. This is easier than adapting the ECCD modulation to the spontaneous, non-uniform mode rotation. The minimum rotation frequency at which it makes sense to modulate has already been reached. The minimum requirement is that during modulation the ECCD is not kept off for too long: if off for 400-600ms, the mode would recover. With the present 60Hz rotation, the ECCD would be kept off for much less: 8ms at the time. Nevertheless, for the sake of comparison with ECCD modulation on NTMs rotating at their natural frequency, of 1kHz or more, it would be desirable, time permitting, to first reach forced mode rotation of 0.2kHz or more. For this reason, the present proposal might represent the follow-up of 432 (i.e. beginning with a mode which is initially locked to the wall or residual error field) or, more difficult, of 433 (i.e. beginning with a mode which is initially rotating at some higher frequency, and is slowing down). Reference shots: with continuous ECCD, with no ECCD, with deliberately misaligned ECCD (e.g. 2/3 gyrotrons too far in, 2/3 too far out) and with X-point phasing.
Background:	Sustained mode rotation (a.k.a. entrainment) at up to 60Hz (and possibly 180Hz, unclear, waiting for confirmation) have been already demonstrated in the absence of ECCD. Rotational mitigation has been observed, but ECCD will be necessary for full stabilization. Furthermore, because the frequency and phase of rotation will be known and controllable, the control by modulated ECCD is expected to be easier.
Resource Requirements:	SPAs, 4 gyrotrons
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	520: Benefits of ECCD modulation as functions of phase, deposition width, misalignment & duty-cycle
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	R. La Haye, A. Welander	ITPA Joint Experiment :	No
Description:	Confirm and quantify benefits of proper phase, broad deposition and good alignment in the stabilization of NTMs by modulated ECCD. Investigate duty-cycles alternative to the conventional 50/50%.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Experimental Approach/Plan: This can be an extension of any successful NTM stabilization by modulated ECCD experiment in the next campaign, regardless of the diagnostic (Mirnov, oblique ECE, horizontal ECE) and interface (digital, analogue). Simply repeat the experiment for various phasings (O-point, X-point and intermediate phases) and for various deposition widths delta_ECCD, by staggering the gyrotron launches more and more in the poloidal direction. Note that the ratio of delta_ECCD to the island width w is scanned automatically and for free in every shot, dynamically, as delta_ECCD remains fixed and w shrinks. However, various other plasma parameters change in the process, not to mention that at the end of the stabilization process w becomes small, delta_ECCD/w large, hence the modulation more beneficial and the stabilization more rapid, but, unfortunately, in coincidence with other effects leaving the same signature, of making the stabilization quicker (e.g., when w becomes comparable with the marginal island width). For this reason, it is preferred to scan delta_ECCD instead, in a dedicated series of 3-5 shots. 2-4 misaligned shots for each case should suffice to study the resilience of modulated ECCD to misalignment in 4 cases (broad/narrow deposition, O/X point phasing) and compare with formulas by Rip Perkins. A preliminary DIII-D result suggest that ECCD in the X-point might be more resilient to misalignment, but needs confirmation. Finally, in addition to the conventional 50/50% duty cycle, we intend to keep the ECCD on for 10, 30, 70 and 90% of the time.
Background:	An extensive phase scan and a qualitative (2-3 point) width scan were carried out at AUG, for a 3/2 mode. The goal at DIII-D is to repeat, confirm and improve those scans for the 3/2 mode, especially as far as delta_ECCD is concerned, and to perform them also for the 2/1 mode, for the first time. It is still unclear whether ITER really needs the ECCD to be modulated, and this scan can help answering this question. As a by-product, the phase-scan can help separate the ECCD effect on Delta', which does not depend on the phase, from the replenishment of the missing BS current, which evidently does. Should modulation be recommended for ITER, the proposed misalignment scan is of obvious importance in setting the requirements for the real-time scan of the launchers. Finally, the scan of the duty cycle is unique in its kind. Duty cycles of 50% on, 50% off time are customarily considered both in theory and experiment. However, stabilization is the resultant of the two mechanisms mentioned above, one of which does not depend on the phase and, as a matter of fact, does not improve with modulation. It would rather benefit from a longer duty cycle, or from the ECCD being on all the time. For all these considerations, it is speculated that the optimum might be other than 50/50%, and we propose to find it experimentally. Note the optimum would be machine- and scenario-dependent.
Resource Requirements:	4 gyrotrons
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	521: ECCD modulated by horizontal ECE
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	M. Austin, R. La Haye, A. Welander	ITPA Joint Experiment :	No
Description:	Show that horizontal ECE can replace oblique ECE as a driver for ECCD modulation in phase with NTM O-point. Coincidentally, no phase correction is required for the 2/1 mode, because the ECE and the ECCD happen to be about 180deg out-of-phase. Some correction â??to optimize experimentally- is required for the 3/2 mode.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Similar if not identical to shots where ECCD was modulated by oblique ECE, but with the analogue interface in the annex (â??the boxâ??) connected to horizontal ECE. Apart from some scenario development, to obtain an NTM rotating at less than 5-7kHz, the main variable to experimentally optimize will be the phase delay between the ECE and the ECCD, in the 3/2 case, but not for the 2/1 mode, for the following reasons. The horizontal ECE is located at phi=60deg. The gyrotrons are located at phi=240-270deg and drive current at approximately the same phi (about +10/-10deg apart, for typical co/ctr-ECCD settings). This toroidal displacement can be approximated with 180deg, and the poloidal displacement can be ignored, in controlling a 2/1 NTM, the O-point of which has a toroidal extent of 180deg. In brief, the horizontal ECE is measured in a position from which it can directly modulate the ECCD in phase with a rotating 2/1 mode, with no need for phase correction. The approximations cannot be repeated for the narrower (90deg) 3/2 O-point and a phase correction becomes necessary. Applying the correction is not a problem thanks to an adjustable, flat response phase-shifter. Finding the optimal correction might require some time, though.
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, 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:	4 gyrotrons
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	522: Analogue approach to Mirnov modulation of ECCD, with phase scan
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	A. Welander, R. La Haye, E.J.Strait	ITPA Joint Experiment :	No
Description:	Use Bp probes at various locations, connected to the analogue interface (a.k.a. the "box") originally developed for oblique ECE, for modulated ECCD at various phases relative to NTM.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Similar if not identical to shots where ECCD was modulated by oblique ECE, but with the analogue interface in the annex ("the box") connected to a Mirnov probe (or a combination of probes, to isolate odd/even modes or specific n numbers). Then repeat for different probes around the torus, to scan the phase of ECCD relative to the island O-point and X-point.
Background:	At DIII-D, both Mirnov signals and oblique ECE signals are being explored as drivers for ECCD modulation in synch and in phase with the O-point of rotating NTMs. Mirnov signals are digitally interfaced to the gyrotrons, whereas a fully analogue approach has been adopted for the oblique ECE. Here it is proposed to combine the best aspects of the two approaches. In the last campaign, the analogue electronics worked very reliably: it delivered to the gyrotrons the expected modulation waveforms, on the basis of the received oblique ECE signals. These signals, however, were not perfect NTM measurements, partly because of intrinsic reasons (ECE is sensitive to all Te fluctuations, not just to NTMs) and partly because of the signal-to-noise ratio of the radiometer, which needs to be improved. Mirnov signals, on the other hand, are with no doubt the best, cleanest measurements of NTMs. Their digital interface with ECCD, however, is still being commissioned. Here we propose to use Mirnov signals in combination with the analogue interface. The only drawback of this approach is the uncertainty in the phase extrapolation from the Mirnov sensor to the ECCD actuator, lying in different positions. A simple way around the problem is to utilize different sensors (from the same toroidal array) in different discharges. In this way, the phase-delay would be scanned from shot to shot, with the twofold result that a complete toroidal scan would be performed and the sensor yielding the best phase would be found.
Resource Requirements:	4 gyrotrons
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	523: Torque Waves
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Inject torque at specific radial locations. See how the CER rotation profile evolves. Repeat in a pulsed, repetitive way, to generate â??torque wavesâ??. From their propagation,infer transport of momentum.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Start with a successful example of RMP-locked-mode or RMP-NTM coupling. Repeat but turn the RMPs off after 2-3 momentum confinement times (assumed of the order of the local energy confinement time). After a comparable time, turn on again. For this technique, it is important for the mode to always be present â??and, at the same time, not to grow too large-. If the mode or the plasma do not survive to this first on-off-on sequence, either because the mode gets too small or too disruptive, change the beams and therefore beta accordingly. When successful for 1-2 pulses, repeat for several pulses within the same shot. Acquire CER.
Background:	Rotating RMPs have been successfully used to unlock initially locked modes and force their rotation. The coupling of rotating RMPs with rotating NTMs is now under investigation (see proposal 433). Either way, RMPs impart momentum to tearing modes. Indirectly, they also impart momentum to the plasma, because islands are nearly â??frozenâ?? in the plasma, therefore the rotation of the island nearly coincides with the rotation of the plasma, apart from a small offset. Here it is proposed to inject torque in the plasma at specific radial locations, by means of rotating RMPs: if there are NTMs in the plasma, the RMPs will impart momentum to them and, indirectly, to the plasma, at the rational surface locations. It will be interesting to measure, for example by means of the CER diagnostic, how the toroidal rotation profile evolves/relaxes after the perturbation. In particular, it will be interesting to time-modulate this space-localized injection of torque. â??Torque wavesâ?? will be generated as a result, similar to â??heat wavesâ?? generated by modulated, localized heating. Temperature diagnostics such as ECE allow to study the propagation of these heat waves and, from them, infer the profiles of heat transport coefficients. In the same way, it is proposed to diagnose â??torque wavesâ?? by CER, measure how they propagate (viscously or by other means) i.e. how momentum is transported.
Resource Requirements:	SPAs, I-coils
Diagnostic Requirements:	CER
Analysis Requirements:
Other Requirements:

Title	524: Current diffusion measured by MECCD and MSE
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Heating & Current Drive	Presentation time:	Not requested
Co-Author(s):	R. Prater	ITPA Joint Experiment :	No
Description:	Use ECCD pulses to periodically perturb the current profile and the MSE diagnostic to measure how these perturbations propagate.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The first goal is to demonstrate a new diagnostic technique. This goal can probably be achieved during the machine re-commissioning phase, when several discharges are run with the aim of conditioning the machine and its walls: during that phase, if the ECCD and MSE are ready, we propose to modulate ECCD in the core and at mid-radius, with maximum modulation depth (70%?). Because removing the ECCD would also remove an important source of heat, which also affects the current profile, we propose that when the ECCD is off, other gyrotrons be turned on for pure ECH, at the same power level. Also, for the effect on the current profile to be visible, the ECCD needs to be turned on (and off!) for at least a couple of local energy and particle confinement times, whichever is longer. This translates for examples in 200ms pulses in the core or 100ms pulses at mid-radius. In summary, we propose to alternatively fire 2/3 gyrotrons for pure ECH and 2/3 for co-ECCD, in periods of 200ms for deposition in the core or 100ms for deposition at mid-radius. Initially, we propose to do so during the machine re-commissioning phase, for the purpose of demonstrating the new diagnostic technique. One of the advantages of re-commissiong shots is that they are relatively long and steady-state, which is good for averaging MSE data. If the signal-to-noise ratio is good enough, these new current transport measurements will be compared with TRANSP, ONETWO and other codes. If and when the diagnostic technique will be demonstrated, we would like to deploy it in piggyback as well as dedicated experiments, under a variety of conditions (see below).
Background:	Classical current diffusion times are easy to calculate. However, there are only global, space-integrated measurements, which these estimates can be compared with. Local measurements of particle and energy exist, but, to authorâ??s knowledge, not of current transport. The current density is the sum over the species of the integrals over the distribution functions f(x,v) of contributions of type nqv, where n is the density, q the charge and v the velocity. As a result, current transport is a combination of particle transport, through n, and of energy transport, through v. As such, it is also likely to be interested by anomalous transport phenomena. Additional open questions are: how does the current â??transportâ?? through transport barriers? And inside or across magnetic islands? Do different components of the current profile (Ohmic, bootstrap, NB current drive, EC current drive, etc.) transport at different rates?
Resource Requirements:	4 gyrotrons
Diagnostic Requirements:	MSE
Analysis Requirements:
Other Requirements:

Title	525: Magnetic Transport Barriers from Coalescing Islands
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	H. Ali, A. Punjabi (Hampton University, VA)	ITPA Joint Experiment :	No
Description:	Make two large islands (either NTMs, or, better, RMP-induced islands) of adequate width and distance non-linearly interact with each other in such a manner that they generate a transport barrier in the middle of a stochastic region.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Consider a discharge with 2/1 and 3/2 NTMs. The widths and distance will fluctuate. At times when parameters are good, the magnetic barrier will form and its strength will also fluctuate. Occasionally, the island could disappear. In this discharge, modulate central ECH to generate heat pulses. Their slowing down in between the 3/2 and 2/1 NTM should reveal the presence of a transport barrier, if any. Repeat the experiment with large 3/1 and 4/1 islands. These can be generated by n=1 RMPs exerted by the I-coils, wired in a configuration such as IU60 or IU120, yielding steep pitch angles. The RMPs offer more flexibility, compared with NTMs, in that the island widths can be more easily varied. Note that magnetic barriers are only expected to form in a certain range of parameters. Another parameter to adjust will be the distance between the islands, e.g. by varying q95, which affects the edge current profile.
Background:	The overlap of large magnetic islands generates chaotic fields. These generally enhance transport, deteriorate confinement, possibly cause or contribute to the "density pump-out". These effects are well-known from the RMP control of ELMs. However, theory suggests that invariant manifolds can develop within chaotic fields, with the effect of reducing, rather than enhancing, transport. In particular, barriers should form at some special, "noble" irrational values of q in between two rational surfaces such as the 3/2 and 2/1 surfaces, or the 3/1 and 4/1. Barriers can also form at "mediant" irrational q's. Here it is proposed to experimentally test modelling by H. Ali and A. Punjabi [PPCF 2007]. If successful, it would represent the first realization and demonstration of a new kind of transport barriers. The requested mode amplitudes, dB/B=1e-4 - 1e-3, can be obtained either with sufficiently large NTMs (i.e., experimentally, by powerful enough NBI and thus high enough beta) or by error-field penetration (i.e. by applying strong enough RMPs or EFs). The simultaneous generation of two NTMs is relatively common. The excitation of two modes by RMPs is also nearly automatic, because externally generated RMPs or EFs of a certain toroidal number n typically have a spectrum of poloidal mode numbers m. Therefore, they naturally excite various modes of different m. The third requirement is for the islands to be close enough. In turn, this would require a difficult control of the q profile at the edge. A simple knob in this respect would be q95. In the difficulty of making finer adjustments, we simply suggest to make up for the excessive distance between the modes by applying a stronger RMP, if necessary (see first requirement). Ultimately this "barrier" is a flux surface. Locally, in the middle of a stochastic environment, it acts as a barrier. However, according to Ali and Punjabi, the strength of the barrier depends on the parameters of the two modes, and can be increased by optimizing the magnetic configuration (island widths, distance between rational surfaces). Even if this barrier should result too difficult to trigger and/or to control, or too weak, we think that this study will help to better understand stochastic fields and their role in ELM control. For example, it can be speculated that barriers form during ELM control experiments, but only for certain intervals of q95 they are weak enough not to inhibit the beneficial effect of the stochastic fields.
Resource Requirements:	4 gyrotrons
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	526: NTM stabilization by RMP stochastization of X-point
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Use external RMPs to generate chaotic fields that alter an NTM island (not a closed island anymore) and arrest its further growth.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Apply strong non-resonant fields, for example n=3 fields to control a 3/2 or, better, as it is closer to the edge, 2/1 NTM. Stochastic fields should result from the overlap of the pre-existing neoclassical island with the penetrated n=3 error. These chaotic fields should smear the 2/1 separatrix and X-point. This doesn't suppress the island in a strict sense, but converts it in a new magnetic structure, with a different -hopefully milder- growth rate and impact on confinement.
Background:	Stochastic magnetic fields form at the overlap of large islands generated by error-field penetration. It is thought that RMP-generated stochastic fields play a role in the stabilization of ELMs. In ELM control, the main effect is probably indirect: stochastic fields enhance transport, thus making the edge gradients milder. At the same time, however, stochastic fields obviously perturb and, if strong enough, destroy, pre-existing islands. In particular, they alter or destroy X-points. The aim of the present proposal is to perturb, possibly destroy the X-point of an NTM. The immediate effect will be transport enhancement through the X-point: basically, due to stochastization, the X-point will leak more energy and particles. The other effect, however, will be beneficial: the new perturbed structure will not be a closed island anymore and should grow more slowly, or not grow at all. The hope is that, all in all, a narrow stochastic region at the rational surface will be less detrimental, for confinement, than a large island. Certainly it will be preferable from the point of view of disruptions. To authorâ??s knowledge, this is a new idea. However, stochastic fields have been invoked to explain ELM control at DIII-D as well as incomplete reconnection during sawtooth crashes at AUG (Igotchine, Dumbrajis et al.). Hopefully ELM and sawtooth literature can be adapted to NTMs.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	527: Offset velocity of NTMs
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Establish at least the direction in which the NTM rotates, in the plasma frame of reference: whether it is the ion or the electron diamagnetic direction.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The NTM offset velocity is the small difference between two large velocities: of the fluid and of the mode, measured respectively by CER and by magnetics. We need to minimize the error on these measurements, or on their difference. For this, we will adopt two approaches. In the first one, we will keep the rotation as constant as possible for as long as possible (e.g. by means of J. Ferronâ??s algorithm), average the two velocities over that long interval, and take the difference. The second approach is a bit more risky in terms of scenario: the idea is to keep the total NBI power constant (or, even better, beta constant) but alternatively inject the beams in the co and counter direction. The consequent modulation of co/ctr-torque injection should take place on a time-scale of 2-3 momentum confinement times (200-300ms), to give the plasma the time to respond. The plasma will alternatively rotate in the co- and ctr-direction. If done carefully, the CER fluid rotation will average to zero, on a long time-scale. As a result, the time-averaged Mirnov measure of the mode rotation in the lab frame will, de facto, represent the mode rotation in the plasma frame. Both approaches should be repeated for Ohmic and NBI plasmas, where electron and ion diamagnetic effects, respectively, are expected to dominate.
Background:	The Neoclassical toroidal Velocity (NTV) theory has successfully predicted the rotation frequency of Resistive Wall Modes (RWMs) in the plasma frame of reference. This natural, â??offsetâ?? rotation frequency is a fraction of the Alfven frequency. Works by H.Wilson et al. suggest that Neoclassical Tearing Modes (NTMs) obey a different physics, related with the diamagnetic frequencies. However, there is still little agreement and experimental validation on whether and under which conditions NTMs should rotate in the ion or the electron diamagnetic direction, and at which fraction of the corresponding diamagnetic frequency.
Resource Requirements:	210 NBI
Diagnostic Requirements:	CER
Analysis Requirements:
Other Requirements:

Title	528: Imaging island formation for various NTM triggers
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Study NTM triggering by resolving â??e.g. by ECE- the formation of a neoclassical island triggered by an ELM, sawtooth, fishbone or other trigger.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Consider three types of discharges: ELMing, sawtoothing and fishboning. Slowly ramp up NBI power. NTMs are expected to develop at the first event (ELM, sawtooth or fishbone) occurring after beta has exceeded the threshold for NTM meta-stability, provided such event is intense enough to exceed the marginal stability condition. The scope is to image the formation of the island with ECE and, at the same time, acquire as many data as possible to reconstruct the d /dt vs. stability curve. This serves to determine when the plasma enters in an NTM-unstable regime, according to Rutherford's theory, and compare the measured width with the calculated marginal and saturated width.
Background:	The modified Rutherford equation satisfactorily predicts under which conditions a plasma is NTM-unstable, i.e. neoclassical islands can form. However, very little is known on when and how NTMs do form. We only know that NTMs need to be triggered by other MHD instabilities such as ELMs, sawteeth and fishbones.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	529: Locked mode AVOIDANCE by "catching" its precursor with a rotating field
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):	R. La Haye, R. Prater, E.J. Strait	ITPA Joint Experiment :	No
Description:	Apply rotating resonant magnetic perturbations (RMPs) to the rotating precursor of a locked mode to "catch it" and entrain it while it slows down, and so avoid locking, which is one of the main causes of disruptions.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	As soon as Mirnov probes detect a mode slowing down below 1kHz, apply intense n=1 I-coil travelling wave, initially at 0.9kHz, then slowing down at the same rate as the mode, as inferred from real-time spectral and mode analysis from the newspec code, recently become available in real-time. Alternatively, use magnetic feedback, for the I-coils to feed back on Mirnov. While the mode slows down, reduce its amplitude accordingly, under PCS: high RMP intensity was required at the beginning to overcome rotational shielding at ~1kHz, but becomes less and less necessary, and possibly detrimental, as the mode slows down. If evidence of coupling between the RMP and mode rotation is found at a certain time t, repeat with pre-programmed changes in the RMP rotation after t, for example keep the rotation steady, or accelerate it again.
Background:	So far, RMPs successfully controlled initially locked modes at DIII-D. Here we propose to pre-emptively apply rotating RMPs and avoid locking altogether. Applying the RMP while the mode is still rotating, though, introduces the difficulty of adapting the travelling wave to the mode rotation, in order for the former to entrain the latter. By contrast, if the differential rotation was excessive, shielding would be excessive too, and the mode wouldn't "feel" the rotating perturbation and wouldn't lock to it. MHD spectrograms recently made available in real-time (every 2ms) will help in this difficult adaptation, which has the promise for "catching" the rotating precursor of a locked mode before it locks to the wall or residual error field. Instead, the mode will lock to a rotating perturbation. At that point, the rotation can either be kept constant, at a safely high level, or accelerated. As a result the mode will accelerate too, and be mitigated by rotation, both because of shielding and because of flow shear effects.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	PCS changes involving real-time newspec and/or adaptation of magnetic feedback to NTMs

Title	530: Improved Front End for RMP H-mode Shots
Name:	Fenstermacher fenstermacher@fusion.gat.com	Affiliation:	LLNL
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Modify the temporal evolution of the Ip ramp, injected NB power, shape development and gas fuelling of discharges to be used in RMP ELM suppression experiments so that stable operation without locked modes is obtained reliably, both for I-coil on RMP Elm suppression shots and for I-coil off reference shots.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Develop two reference discharges, vis. 1) an ELMing H-mode discharge with initial beta prior to RMP application sufficiently high (eg. BetaN ~ 2.2 â?? 2.4) so that ELM suppression is obtained, and 2) a second discharge to be used as an I-coil off reference discharge which has the same beta value as achieved during the I-coil on phase of discharge (1) above. Use careful variations of Ip ramp rate, including variation of the â??flat spotâ?? (ie. Ip_dot=0) part of the ramp for triggering the H-mode transition, injected power timing, early shape development and supplemental gas fueling to find a stable discharge that will run reliably without early locked modes prior to I-coil RMP application.
Background:	I-coil n=3 RMP ELM suppression in low collisionality plasmas was first discovered (see shot 122342 and others from Feb 25, 2005) in discharges developed for RWM stability experiments, (eg. 118943). For the last 3 years many of the n=3 I-coil RMP ELM suppression experiments at low collisionality have used the temporal evolution of the first 1500 ms from these discharges developed to produce RWMs. Some of the shots develop early locked modes and a non-negligible fraction of those disrupt even before application of the RMP. This reduces the efficiency of RMP ELM suppression experiments. If a early temporal evolution could be found that was very resistant to locke modes and still allowed ELM suppression with n=3 I-coil RMPs, this would substantially increase the productivity of ELM suppression experiments. The dedicated time necessary to find such a stable shot scenario could be more than made up by the increase in productivity of future experiments.
Resource Requirements:	I-coils in n=3 with capability for maximum current (6.4 kA). C-coils in optimal n=1 error field configuration Desireable for all C-power supplies operating simultaneously at full current capability (6.4 kA). All 5 co-beams. All cryopumps LHe cold.
Diagnostic Requirements:	All pedestal, SOL and divertor diagnostics especially pedestal Thomson, pedestal CER, midplane and divertor filterscopes, fast and slow divertor IRTV and visible cameras, divertor probes, midplane recip probe. Analysis
Analysis Requirements:	Control room SURFMN of applied mode spectrum on EFIT01. Post experiment profile analysis, kinetic efits, VARYPED â?? ELITE stability analysis, TRIP3D field line loss fraction analysis, divertor strikepoint pattern analysis from cameras and particle balance analysis.
Other Requirements:

Title	531: Validation of TokSys a priori Simulations of DIII-D Plasma Control
Name:	Humphreys humphrys@fusion.gat.com	Affiliation:	GA
Research Area:	Integrated and Model-Based Control	Presentation time:	Not requested
Co-Author(s):	M. Walker, M. Wade, T. Casper	ITPA Joint Experiment :	No
Description:	The goals of this several hour to 1/2-day experiment are to produce experimental validations of a priori TokSys simulations against DIII-D plasma responses. Various position/shape control perturbations will be performed, and the results will be compared with a priori simulations using TokSys simservers. Key comparisons include large but linear equilibrium perturbations, nonrigid perturbations, and nonlinear plasma responses (which require the LLNL Corsica kernel to be implemented in TokSys simulations).	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Identify a standard control reference equilibrium from previous data in which the plasma resistivity can be determined accurately. Reproduce this equilibrium in 2-hour Thursday evening window to confirm resistivity and transport characteristics. Run Simserver(s) to predict plasma response to various programmed position and shape perturbations (R, Z triangle waveforms; X-point perturbations�?�, large and nonrigid changes in configuration to validate LLNL Corsica models). Execute experiment in second 2-hour period or ½ day and program waveforms as in simulation. Compare with predictions.
Background:	Simserver simulations constructed from TokSys objects and modeling tools with linear plasma models have been extensively validated against experimental responses and used for control design over the last decade. However, the user base for these tools is expanding, and is expected to increasingly include DIII-D physics operators this year. Updated demonstration of the usefulness and validity of these simulations will support the general use, as well as the further development and maintenance of the tools. This experiment is also expected to enable first-time validation of nonrigid models produced by the LLNL Corsica code. Validation of the Corsica nonlinear/nonrigid response model is essential for a wide range of control categories that depend irreducibly on the nonlinear and/or nonrigid plasma response. These include large changes in the plasma configuration such as moving from LSN to USN, or evolving from an initial breakdown state to a fully-formed diverted discharge.
Resource Requirements:	0-4 beams (co)
Diagnostic Requirements:	MSE, 2-5 kHz magnetics sampling
Analysis Requirements:	standard EFITs, TokSys, Corsica code, Corsica Matlab kernel in TokSys simservers
Other Requirements:	--

Title	532: Comparing static and dynamic particle balance
Name:	Whyte whyte@psfc.mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	Hydrogenic Retention (2009)	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The experiment seeks to directly compare static and dynamic global particle balance measurements in DIII-D. The experimental goal of the proposal is to provide as accurate particle balance measurements as possible, over a wide range of plasma operating conditions, by combining the standard flexible dynamic balance with the recently developed static balance, which provides high accuracy but with less flexibility to plasma operational scenarios. These results are directly applicable to the 2009 Joule milestone on particle control which is being jointly accomplished by DIII-D, C-Mod and NSTX. The experiments should provide valuable insights into how plasma conditions and operations affect global particle balance and retention in a graphite wall, which is also of critical interest to ITER, particularly if operational regimes can be validated that deplete fuel from the wall.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The experimental approach is to build on the initial experience we gained in the 2008 campaign, in particular to examine the changes in particle balance as operating regimes are changed. Furthermore, the experimental goal will be to directly compare the results of static gas balance to dynamic particle balance (available on every shot) in order to understand differences, if any, and therefore improve confidence or accuracy in the dynamic particle balance. The 2008 experiments were only in low-density ohmic plasmas, which highly limits the plasma operational regime. The key parameters to scan will be: 1. plasma density (known to strongly affect recycling) 2. RF plasma heating (presumably ECH) 3. NBI heating, with rapid TIV closure after the shot 1. The plasma density is an obvious control parameter in setting recycling and wall pumping. We should first document the effect of raising and lowering density in ohmic shots which allow for routine static particle balance. 2. If possible we would like to obtain quiescent ELMy H-mode using RF heating only, presumably ECH. With NBI gate valves closed this trivially provides static balance measurements to be compared with dynamic particle balance. We would directly compare pumped versus unpumped H-modes since previous measurements showed that pumping led to wall depletion. The pumped discharges would use the shot-to-shot regeneration technique to count the pumped particle inventory independently of the pumping speed calculations 3. Using () NBI opens up the different operating scenarios, but introduces the complexity of particle sources and sinks from the NBI itself. We'd proposed to replicate as close as possible the D3D ECH-only H-mode (same heating, density). The static balance measurement is more difficult to achieve, although we will attempt to close the beam TIVs as fast as possible following the shot. Again, the cryopump can be regenerated shot-to shot.
Background:	Obtaining accurate global particle balance is one of the key diagnostic challenges towards understanding particle control and fuel/tritium retention. Simply stated, one would like to keep track of the particles entering the tokamak, and leaving the tokamak. The differences between these indicates the particle inventory in the plasma-facing surfaces. The retention of tritium is a likely operational limit for ITER. Yet our present understanding of the complex, interlinked process that determine particle/fuel retention is highly incomplete, making design aspects of ITER, such as PFC choices, very difficult. The 2009 Joule milestone seeks to exploit the largely different features of the three major confinement devices to shed light on the controlling physics behind particle balance. DIII-D with a graphite wall, NSTX with lithium coatings of a graphite wall and C-Mod with high-Z refractory metal walls. We have cooperatively planned experiments across the devices to better understand particle control, and obtaining accurate particle balance is one of the joint goals. A challenge of the particle balance is that one is typically using the differences of large numbers since there are a variety of sinks and sources. "Dynamic" particle balance has typically been used in DIII-D; i.e. the sinks and sources are calibrated as accurately as possible off-line, and these calibrations are applied to signals obtained during the shot. This has the benefit of obtaining 'real-time' particle balance, and can be applied in any shot, however has the detriment of relying on calibrations that may not be completely accurate during the shot (e.g. the presence of plasma affects the effective pumping speed). To this end, C-Mod recently developed a "static" particle balance diagnostic technique wherein all active pumping is temporarily turned off so the that in-vessel static pressure comes to equilibrium after the shot. While this ensures very accurate particle balance, it has the detriment that other external sinks must be removed, such as NBI cryopumping. For DIII-D this limits its application. However we note that experiments in 2008 led by Phil West and myself showed that the static balance worked on DIII-D. An important development was the ability to regenerate the cryopumps after each shot with no external pumping This provided a very accurate accounting of cryopumped inventory. Preliminary results showed that the cryopumping did not affect the global particle retention rate.
Resource Requirements:	Cryopumps with between shot regeneration. "Gate-valves" closed operational capability Fast NBI TIV closing capability
Diagnostic Requirements:	Full suite of edge diagnostics including probes, IR thermography, D-alpha, etc.
Analysis Requirements:
Other Requirements:

Title	533: Sawtooth Control
Name:	Sauter none	Affiliation:	CRPP-EPFL
Research Area:	Stability	Presentation time:	Not requested
Co-Author(s):	J. Paley, F. Felici, J. Ferron, D. Humphreys, R. La Haye	ITPA Joint Experiment :	Yes
Description:	Sawtooth control: a) Implement real-time sawtooth control algorithm tested on TCV, to offer a real-time control option for the various sawtooth control experiments on DIII-D. Implement direct link between simulink developments and real-time control of the machine. For example, would be very useful for full stabilization of RF stabilized sawteeth ("monster") (needed since q=1 evolves and so does the optimal position for stabilization) b) Add co-ECCD on-axis to destabilize sawteeth stabilized by RF and check relation to NTM trigger at sawtooth crashes. c) In parallel to these, check the relation between sawtooth period and trigger of NTMs at the various beta obtained during the experiments.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:	Also related to the ITPA experiment on relation between sawteeth and NTM onsets.
Resource Requirements:	ECCD
Diagnostic Requirements:
Analysis Requirements:	Onetwo and Toray
Other Requirements:

Title	534: ECCD and Hybrid scenario control
Name:	Sauter none	Affiliation:	CRPP-EPFL
Research Area:	Core Integration (Advanced Inductive)	Presentation time:	Not requested
Co-Author(s):	T. Casper, ...	ITPA Joint Experiment :	Yes
Description:	Presently, a 3/2 NTM is used to clamp the q profile on DIII-D. Although the mode is not very large, there is some performance degradation associated with the presence of the mode. The proposal is to use combinations of co- and cntr-CD such as to keep the adequate q profile to stay with the hybrid mode without destabilizing a NTM mode. In doing so, the 4/3 mode might be destabilized in place of the 3/2 mode and therefore this proposal is complementary to proposal 47.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Study various co- and counter ECCD components which should sustain different current density profile evolutions. Compare with presence of 3/2 or 4/3 NTMs.
Background:	Relevant for ITPA experiments for prediction of hybrid scenario in ITER and test of ECCD actuator.
Resource Requirements:	max ECCD
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	535: Helium Pumping in AT-class Plasmas
Name:	Petrie petrie@fusion.gat.com	Affiliation:	GA
Research Area:	Core-Edge Integration	Presentation time:	Not requested
Co-Author(s):	T. Petrie	ITPA Joint Experiment :	No
Description:	This experiment is a first attempt to evaluate helium transport and exhaust in high performance AT-class discharges. The main objective is to obtain estimates of the ratio of the helium particle confinement time in the vacuum vessel to that of the energy confinement time in AT-class H-mode plasmas. Our base target plasmas should have betaN > 3 and H89p > 2.5, and q0 = 1.5. Hybrid cases with similar plasma parameters should also be attempted for comparison purposes.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The techniques used in this experiment follow the methodology established earlier by Hillis et al. [Plasma Phys. Control. Fusion, vol 36 (1994) A171]. To simulate the presence of helium ash, concentration of 10-20% helium relative to plasma background density are into the plasma during an otherwise steady state phase of an unpumped AT plasma. For this experiment an upper single-null divertor shape is used with Bt= -1.7 T and Ip = 1.2 MA. The transport of the helium is monitored by measuring the temporal evolution of the helium density profile in the plasma core. During the pumping phase of this experiment, we utilize the decay of helium density and evaluate the helium particle confinement time in the vacuum vessel. This experiment should be done in forward and reverse Bt.
Background:	It would be highly beneficial if future generation tokamaks, such as ITER, could be made to operate in an enhanced confinement regime typical of AT. One drawback, however, is that the particle confinement time of core helium might also be enhanced. For futuristic, high performance tokamaks, a steady purging of helium fusion ash from the core plasma is crucial. Some estimates indicate that newly created helium ash must be removed from the system within 7 to 15 energy confinement times to maintain a continuous burn. Hillis has shown that for DIII-D ELMing H-mode plasmas, this ratio is about 14. While this result might be considered favorable for ELMing H-mode plasmas, it is by no means clear that the much higher energy confinement times expected in AT plasmas will also lead to favorable results. In this experiment, we focus on helium ash accumulation in AT-class plasmas.
Resource Requirements:	Machine time: 1.0 day (forward Bt) and 1.0 day in reverse-B, dome- and upper baffle cryo-pumps cold, 5 co-beams.
Diagnostic Requirements:	Asdex gauges (in particular, dome and upper baffle locations), Penning gauges, core Thomson scattering, upper divertor and centerpost fixed Langmuir probes, and CER.
Analysis Requirements:	UEDGE, with ONETWO and MIST runs.
Other Requirements:	--

Title	536: triangularity impact on transport
Name:	Staebler none	Affiliation:	GA
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Recent experiments on TCV show that for high Te/Ti plasmas that triagularity can have a strong impact on electron thermal transport. DIII-D has the ability to explore this physics	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce as closely as possible the conditions of the TCV experiment. Compare these results with DIII-D discharges with the same shape but different Te/Ti by adding more ion heating.
Background:	The strong triangualrity dependence may be due to the high Te/Ti since the TEM mode is dominant in this case. Changing Te/Ti close to one could eliminate most of the triangularity effect.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	537: TEM dominated transport
Name:	Staebler none	Affiliation:	GA
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The character of the transport in a plasma dominated by TEM turbulence is predicted by theory to be different than ITG modes are active. One way to produce such a plasma is to peak the density profile to stabilize ITG modes. Another way would be to heat only electrons at low density (cold ions). Both of these methods have problems of implimentation on DIII-D. I propose that the TMV task force debate the merits of each and choose one method to produce a plasma with TEM as the main instability.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	To be determined by the TMV task force
Background:	It is expected that the density scaling of energy confinement will be neo-alcator when TEM modes are dominant. The ratio of ion energy diffusivity to electron energy diffusivity should be less than 1 in TEM plasmas. For low density electron heated plasmas the density profile should be peaked by the TEM driven pinch.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	538: low collision transport
Name:	Staebler none	Affiliation:	GA
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Driftwave turbulence theory predicts that at very low electron-ion collision frequency that the thermal transport increases sharply. This could have implications for ITER since the ITER core is at a lower collision parameter (nuei*Rq/cs) than present tokmaks. With the excellent error field control on DIII-D and pumping very low density operation is possible without locked modes. This makes DIII-D an ideal machine to explore the low collisionality transport.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	produce low density ELMing H-mode discharges and scan the density at fixed beta to observe how the transport changes. Starting at high density the power requrired to hold beta fixed should increase non-linearly as density is lowered.
Background:
Resource Requirements:	medium balance beam power, pumping, error field correction
Diagnostic Requirements:	fluctuation data (FIR, BES, Te-fluctuations etc)
Analysis Requirements:	TGLF and GYRO simulations of the density scan
Other Requirements:

Title	539: Dependence of momentum and particle pinch on collisionality
Name:	Tala none	Affiliation:	Euratom-Tekes, VTT, Finland
Research Area:	Rotation Physics (2009)	Presentation time:	Not requested
Co-Author(s):	W. Solomon, S. Kaye, G. Tardini	ITPA Joint Experiment :	Yes
Description:		ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	The main idea of the experiment is to make a scan of collisionality. Reaching a large factor of more than a factor 10 in collisionality is relatively straightforward to carry out for example by varying the level of gas puffing. The magnitude of the momentum pinch is determined either by modulating NBI, NBI blips or braking the plasma. The particle pinch is to be determined for exactly the same plasmas (same collisionality scan as for momentum pinch). The perturbation in density can be induced either by pellets or gas puffing. During the collisionality scan, it is important to try to keep other local plasma parameters as unchanged as possible. In particular, a sensitivity of the momentum and/or particle pinch to following quantities is expected: q, s, R/LTi, R/LTe, vtor, Ti/Te, and the variations in these profiles should be kept at minimum while varying the collisionality. This implies that either H-mode plasmas with small ELMs (type III would be feasible) or L-mode plasmas should be used in this experiment. The main interest is in the core region at 0.2 < r/a < 0.8 where ITG turbulence dominates. There are other common parametric dependencies (in theory) of momentum and particle pinch that could be studied on top of the collsionality scan in this experiment. In particular, the dependence of both the momentum and particle pinch on q and s is well documented in theory but not verified experimentally. In addition, a dependence of momentum pinch on R/Ln is expected to be relatively strong. The outcome of the experiment is to confirm (or not) the theory and gyro-kinetic predictions for the dependence of particle and momentum pinch on collisionality and possibly also q and s (and R/Ln in the case of momentum). Carrying out the same experiment in several machines, also in spherical tokamaks where trapped particle response may be different, increase the confidence in the results achieved only on one machine.
Background:	Recently, several tokamaks have shown that a significant inward momentum pinch exists. NBI modulation technique or NBI blips has been used on JET, JT-60U and DIII-D while plasma braking has been used on DIII-D and NSTX. Now, when the existence of the inward momentum pinch has been established, in order to be able to extrapolate its significance in ITER prediction for rotation, the parametric dependencies of the pinch must be clarified. The recent theory papers and gyro-kinetic simulations have shown that the most important parametric dependence of the momentum pinch is the density gradient length R/Ln. R/Ln (measure of density peaking factor), on the other hand, depends strongly on the collisionality, reported in several tokamaks. The electron trapping is expected to play an important role in the generation of the momentum pinch, indicating why collisions are important in determining the momentum pinch. It is possible that a higher collisionality will make the electrons more adiabatic and will remove the pinch effect. Here, similar physics is expected for the particle transport where the reduction of the trapped electron response removes the inward particle pinch. The central of trapping shows the importance to include spherical tokamaks in the joint experiment where trapping is different. However, no experimental data exist to verify the dependence of the momentum pinch on either R/Ln or collisionality. Also, while the density peaking dependence on collisionality is reported on many tokamaks, the direct dependence of the particle pinch vpinch,part on collisionality has not been studied (partly due to diagnostics difficulties to measure the density evolution accurately enough and partly due to challenges in inducing suitable density perturbation). Furthermore, no attempts have been made to compare the similarity of the momentum and particle pinch although according to the theory they are linked. Quantifying the parametric dependence of the momentum and particle pinch on collisionality (and possibly some other parameters as well) consolidates the extrapolation of both the toroidal rotation and density profiles for ITER.
Resource Requirements:	Gas puff modulation and/or pellets, NBI modulation or blips (co-ctr beams)
Diagnostic Requirements:	CER, electron density measurements at good time and spatial resolution, fluctuation measurements desirable, q-profile, Zeff
Analysis Requirements:	Accurate time-dependent torque calculation (TRANSP), accurate analysis of the time evolution of the density perturbation to deduce particle pinch and diffusivity, In the case of NBI modulation, Fourier analysis of the modulated toroidal rotation
Other Requirements:

Title	540: Experimentally Relevant Benchmarks for Gyrokinetic Microstability Codes
Name:	Bravenec rvbravenec@4th-state.com	Affiliation:	Fourth State Research
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	J. Candy, R. Waltz, G. Staebler, W. Dorland, D. Ernst	ITPA Joint Experiment :	No
Description:	A few nonlinear gyrokinetic microstability codes are now capable of simulating tokamak plasmas to an unprecedented level of complexity. Verification of these "experimentally relevant" simulations is difficult, however, because no benchmarks exist with which the codes can compare. We propose to develop such benchmarks through "apples-to-apples" comparisons among codes, i.e., comparisons for the same plasma containing the same physics and having sufficient temporal, spatial, pitch-angle, and energy resolutions. The underlying assumption is that the codes must be correct if they all agree. We propose to compare the GYRO, GS2, and GKS continuum codes, and perhaps the GEM particle code. A single utility code is used to extract DIII-D data from analysis by ONETWO or TRANSP and to produce input files for all the codes. The codes are first run linearly and, if differences in the mode frequencies are found, the computations are simplified by removing shaping, collisions, etc., one at a time, until agreement is reached. This process pinpoints the source(s) of the disagreement which the code developers attempt to resolve. Next, nonlinear runs are undertaken for the same cases and the procedure is repeated. The final results are both linear and non-linear benchmarks at various levels of complexity by which other codes may be verified. This verification scheme has the additional benefits of providing analysis in support of experiment as well as validation of the codes since real experimental data are used.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	We will consult with GA experimentalists to determine DIII-D discharges which would test the codes most fully, and whose analysis and simulation would be of general interest. This may not require running new discharges -- existing data may be sufficient.
Background:
Resource Requirements:	N/A
Diagnostic Requirements:	N/A
Analysis Requirements:	Computation resources at NERSC for GYRO and GS2 and at GA for GKS.
Other Requirements:

Title	541: Carbon erosion in impurity seeded plasmas
Name:	Brezinsek none	Affiliation:	FZJ/Germany
Research Area:	General Plasma Boundary Interfaces	Presentation time:	Not requested
Co-Author(s):	N.Brooks,A.Mclean,P.Stangeby,D.Rudakov,M.Groth	ITPA Joint Experiment :	Yes
Description:	Determine the carbon erosion in N2 impurity seeded plasmas in attached and detached divertor conditions. Local injection of methane is used to calibrate in-situ the remaining hydrocarbon flux. Investigate the chemical reactivity of nitrogen radicals.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Provide the unseeded target plasma in deuterium. Inject nitrogen in the divertor to obtain a radiating mantle in the divertor. Vary on a shot to shot basis the ratio of D2 and N2 injection to swap from attached to detached divertor conditions and characterise the plasma. 2) The outer strike point is fixed on the location of the PPI and methane is injected to quantify the carbon flux. 3) Repeat discharges with different diagnostic settings to discriminate the chemical and physical sputtering and the dissociation chain of methane. Quantify the carbon sputtering induced by nitrogen.
Background:	Impurity seeding is a well known methode to radiate in the divertor and to reduce the power and particle flux to the target. In ITER with C/W divertor impurity seeding is foreseen to be able to operate in semi-detached conditions. Nitrogen is a possible candidate because its radiation characteristic is comparable to carbon. However, there is a competition between the seeding impurity which enhances the sputtering and the effective plasma cooling which reduces the sputtering. Moreover, nitrogen is chemical reactive and can cause additional sputtering. In AUG and JET good experiences with N2 seeded plasmas have been made and a reduction of the intrinsic carbon photon flux was observed. But only DIII-D offers the possibility to calibrate in-situ the hydrocarbon flux under the combined bombardment of deuterium, carbon and nitrogen ions. The experiment fits well in the DSOL-2 proposal which deals with the carbon sputtering quantification with local injection.
Resource Requirements:	units: 1.0 day experiment -0.5 day for characterization of spectroscopic signals from the PPI using CH4 in a fully detached OSP -0.5 day for characterization of spectroscopic signals from the PPI using CH4 in a attached OSP Number of neutral beam sources: 2 No requirement for ECH or ICH
Diagnostic Requirements:	PPI DiMES sample, MDS, lower divertor IR camera, DiMES TV with light-limiting mask, divertor Thomson scattering, fixed floor Langmuir probes, midplane and X-point reciprocating Langmuir probes, filterscopes, CER, tangential divertor cameras, Ocean Optics spectrometer, RGA, SPRED.
Analysis Requirements:	DIVIMP-HC 3D Monte Carlo impurity simulation code (Y. Mu), MDS-SIM synthetic diagnostic simulation code for MDS (McLean, Brooks, Isler), Please note that a high resolution overview spectrometer will be provided as loan from FZJ. The system is capable at once to record at once CD, CN, CI, CII, CIII.
Other Requirements:

Title	542: Pair formation during disruptions
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):	Alex James (UCSD)	ITPA Joint Experiment :	No
Description:	Provide first evidence of disruption-generated positrons	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Perform microwave and gamma-ray measurements during disruptions. Cause of disruption is not important. Can be combined with any experiment with intentional or probable disruptions, especially if copious runaway electrons are expected. 1. Microwaves: discriminate between clockwise and counter-clockwise elliptical polarization at the X2 harmonic in the oblique ECE radiometer. Because positrons gyrate opposite to electrons, polarization also change. Counter-clockwise polarization would be a signature of "Positron Cyclotron Emission". 2. Gamma-rays: Use scintillators to compare forward and backward emission either within the same discharge, with two scintillators looking in the co- and ctr-Ip direction respectively, or with the same scintillator in two discharges of opposite Ip and BT. Emission will be in one case the sum of backward emission from electrons (e-) and forward emission from positrons (e+). Because forward and backward emission of each specie are related to each other, it will be easy to recognize anomalies in the comparison with the forward emission from e- summed to the backward emission from e+. 3. Finally, coincidence counters of 511keV photons, of the the type used in accelerators or in positron emission tomography might also be utilized.
Background:	P. Helander (IPP Greifswald) predicted that, during disruptions, collisions between runaway electrons and thermal ions generate electron-positron pairs [PRL 2003]. Here it is suggested that channelling disruption energy in these electron-positron pairs might constitute an innovative mitigation scheme: pairs eventually annihilate in 511keV photons, relatively innocuous for the tokamak walls, certainly much less harmful than massive particles. First, however, we need to prove the formation of positrons in a tokamak, for the first time.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	543: Compare ECCD+RMP with ECH-only control of disruptions
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Try a simpler approach to the control of disruptive locked modes, with ECH (no ECCD) and no rotating RMPs (rRMPs). Identify pros and cons of the two approaches.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Cause disruptions of two types: by ramping ne and hitting the density limit, for comparison with AUG and FTU, and by ramping beta, thus forming a 2/1 NTM and causing it to grow and lock, for comparison with DIII-D experiments with ECCD Apply ECH, dudding on Vloop as in AUG and/or on conventional DIII-D dud detectors. Disruption should be delayed or avoided compared to no-ECH reference. Scan radial location of ECH from shot to shot. At least 5 radii needed: <, = and > q=2 and q=3 radii. Maximum disruption delay or complete avoidance are expected for deposition at q=2. So far, max available ECH power should be used. Then try 1.2MW only, for comparison with ECCD experiments at the same power. If time, try third type of disruption, by laser-blow off or other edge impurity seeding, for comparison with FTU.
Background:	ECCD+rRMP has partly stabilized 2/1 locked modes at DIII-D. In FTU and AUG, pure ECH has postponed and, in one shot in FTU, avoided, more complicated disruptions with a locked 3/1 mode, followed by a rotating 2/1 which eventually locked too. MARFEs were also observed in those disruptions. Here the complication of MARFEs is ignored. The question is whether ECH is sufficient to postpone (like in AUG) or suppress (like in FTU) disruptive MHD activity in DIII-D. If affirmative, the second question is whether it is more efficient. The third question is whether it also benefits, and how much, from rRMP steering of the mode. It has to be said that ECH/ECCD control of TMs is always a mixture of ECH effects (on conductivity and Deltaâ??) and ECCD effects (on Deltaâ?? and compensation for missing bootstrap). Experiments show that ECH dominates in small tokamaks like TEXTOR and FTU while ECCD effects dominate in large tokamaks like DIII-D and JT-60U and are expected to dominate even more in ITER. Analysis and modelling by Westerhofâ??s group, presented at the last MHD Control Workshop, confirmed this asymmetry. It also predicted that AUG should be the intermediate size tokamak where ECH and ECCD effects are comparable. This was not confirmed by experiments yet. However, the fact that ECH disruption control was less effective in AUG than in FTU goes in the right direction. Modelling and experiments suggest that resistive modes should be more efficiently controlled by ECCD than by ECH in the next bigger tokamak in line, which is DIII-D. Partial ECCD stabilization of locked modes was already demonstrated at DIII-D (incomplete suppression was ascribed to insufficient ECCD power, 1.2MW). ECH alone hasnâ??t been tried yet.
Resource Requirements:	4-5 gyrotrons
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	544: "Spiralling" RMPs to find EFC
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Let a mode lock to the resultant of the static machine EF and a rotating, growing RMP. The mode rotation will initially be intermittent, then complete but non-uniform, then more and more uniform. From these behaviours and from the known amplitude and phase of the RMP, one can infer the amplitude and phase of the EF.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Generate a non-disruptive locked mode. Apply growing, rotating (spiralling) RMPs. Infer the EF from the mode response (amplitude and phase) measured via internal saddle loops.
Background:	In the past, at DIII-D, modes initially locked to the wall or machine EF were forced to rotate by applied rotating RMPs. The mode locked in reality to the resultant of the static EF and the rotating RMP. This was a nuisance from the standpoint of driven rotation in that it caused the island to â??slipâ??, change direction or rotate non-uniformly. However, it might turn useful from an EFC perspective, to find an unknown EF as the difference between a measured EF+RMP and a known (pre-programmed) RMP. A standard EFC method consists in fixing the RMP phase phi_RMP and ramping the amplitude A_RMP up (or n_e down) until locking. Then phi_RMP is scanned shot-by-shot. The EF is inferred from the values of A_RMP (or n_e) at locking. Note that ramps are pre-programmed, they do not stop at locking and often terminate with disruptions. Moreover, 3-4 discharges of this kind are needed for every new EFC, i.e. in principle for every new scenario. The approach proposed here, by contrast, consists in scanning both phi_RMP and, more slowly, A_RMP, within the same shot. The resulting RMP rotates and grows, i.e. it spirals out. In doing so, it scans the A_RMP/A_EF vs. phi_RMP-phi_EF, where A_RMP and phi_RMP are known, and the EF amplitude and phase, A_EF and phi_EF, are the unknown. The behaviour of a mode locked to and dragged by the EF+RMP resultant changes as different regions of the A_RMP/A_EF, phi_RMP-phi_EF plane are explored and A_EF and phi_EF can be indirectly inferred. For example A_EF coincides with the smallest A_RMP for which the mode is successfully dragged for a complete toroidal rotation. Before then, rotation will be incomplete, and phi_EF will be the mid-phase of the arcs. All this requires the presence of a non-disruptive mode in the plasma. This mode can either be pre-existing, seeded by EF-penetration, e.g. by an earlier density ramp-down, or, inevitably, it will automatically be generated during the RMP â??spiralâ??, as soon as the total amplitude becomes high enough. It is understood that the RMP of choice has the same n as the dominant EF component. However, once a certain n-component has been identified and corrected, one can repeat the process for the second most important component. This method might represent a non-disruptive, ITER-relevant, quicker (as it requires one shot instead of four) alternative to the conventional technique mentioned above.
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	545: Integrated Disruption Control
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Disruption Characterization	Presentation time:	Not requested
Co-Author(s):	D. Humphreys, E.J. Strait, M. Walker	ITPA Joint Experiment :	No
Description:	Develop and test an integrated system for disruption avoidance/mitigation.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Technical tests and calibrations (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 what is most efficient among the following: (i) drop NBI to make mode less disruptive, (ii) maximize NBI torque to unlock the mode, (iii) apply static RMPs and continuous ECCD or (iv) rotating RMPs 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, liquid jet (if available), ECH ('a la FTU and AUG). This is an ambitious, long-term project. Some parts could be developed and tested in 2009. Some other parts need other experiments, such as 432-436, to be completed and well-understood, first. The integration is unlikely to begin before 2010, but preliminary tests could begin on that very year.
Resource Requirements:	several 2-hour slots on Thursday evening
Diagnostic Requirements:	--
Analysis Requirements:	real-time DCON
Other Requirements:	A lot of PCS preparatory work

Title	546: Scaling of transport, turbulence and zonal flow/GAM damping with collisionality
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	C. Holland, T. Luce, C. Petty, T. Rhodes, L. Schmitz, M. Shafer, G. Tynan, A. White, Z. Yan	ITPA Joint Experiment :	No
Description:	Determine dependence of turbulence and transport in L-mode and H-mode as a function of collisionality, while examining zonal flow/GAM characteristics (amplitude, structure, etc.) to test TGLF predictions and expectations for ZF/GAM damping physics.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Vary the collisionality over the maximum extent possible, from low density, high temperature, to high density, low-temperature plasmas. Perform a nu* scan a low to moderate rotation (balanced beams). Maintain other non-dimensional variable nearly constant. Run at relatively high-q, where GAM damping is minimized (USN L-mode conditions). The upper divertor pumps will be used to control density as much as possible. A combination of gas puffing, pumping, neutral beam power and ECH will be used to adjust density-temperature to vary the collisionality. Utilize BES and DBS to measure and examine zonal flow/GAM scaling behavior.
Background:	Collisionality varies significantly radially within a given plasma, as well as in different plasma modes, and can have a strong effect on transport. Past experiments suggest the L-modes transport scales weakly with collisionality, while H-mode mode transport increases more strongly (Petty, PoP, 1999). Since ITER will operate in a very low collisionality regime, it is crucial to test models and in particular their collisionality scaling. In addition, zonal flows (including GAMs) are expected to be collisionally damped by ion-ion collisions. This is founded in the theoretical understanding of zonal flows and has been borne out in simulations. Damping of zonal flows can in turn reduce zonal flow shearing, resulting in higher turbulence and associated transport levels. GAMs have been robustly observed near the outer regions of L-mode discharges on DIII-D using the multipoint BES system, and recent measurements with the upgraded, high-sensitivity BES show features of the lower-frequency residual, or Zero-Mean-Frequency zonal flow deeper in the core. Thus we have the diagnostic capability to examine zonal flows and their characteristics. Experimental determination of the role of collisionality on zonal flows will help validate turbulence models and thereby increase our overall physics understanding as well as predictive capability. This data set would allow simultaneous measurement of zonal flows/GAMs as well as the ambient turbulence as a function of collisionality. Simulations of these plasmas with GYRO and a comparison of the resulting turbulence/zonal flow characteristics with measurements will help challenge and validate the code. An added benefit to this experiment will be to obtain measurements of general turbulence characteristics as a function of collisionality, thus continuing a program of nondimensional scaling of turbulence characteristics.
Resource Requirements:	Beams, ECH
Diagnostic Requirements:	BES, DBS, CECE, FIR, Corr. Refl., PCI, reciprocating probes
Analysis Requirements:	TRANSP, TGLF, GYRO
Other Requirements:

Title	547: Database of quasi-vacuum shots, plasma-response at low-beta and AA and SPA limits
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	Build a database of quasi-vacuum shots, plasma-response at low-beta and AA and SPA limits, for MHD experiments later in the campaign.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Add various I- and C-coils to machine- and wall-conditioning shots, during reconditioning of the machine.
Background:	Reference vacuum shots are usually acquired to subtract the direct effect of perturbing I- and C-coils on the Mirnov, ESLD and ISLD diagnostics. "Net" MHD measurements are obtained as differences between plasma shots and vacuum shots. Here it is suggested that "vacuum" shots do not need to be taken in vacuum. They could be plasma shots with negligible MHD activity. In particular, it is important that the MHD phenomenon of interest is absent from the reference shot. Good candidates in this sense are the MHD-quiescent discharges repeatedly run during the re-commissioning of the machine. It is proposed to add various I- and/or C-coil waveforms to those shots, measure their effect on magnetics and so build a database of "vacuum" shots. These would be invoked for subtractions later in the campaign. Secondly, the plasma response at low beta would automatically be measured with this method. For a better understanding, it would be desirable to take a couple of shots at two different betas, for each waveform in the database. Alternatively, beta could be varied in the same shot. Third, the current limits of the AAs and SPAs at various frequencies would be explored and documented with this database.
Resource Requirements:	I- and C-coils, AAs and SPAs
Diagnostic Requirements:	all magnetics
Analysis Requirements:
Other Requirements:

Title	548: Locked mode avoidance by magnetic f/back on the saddle loops
Name:	Volpe fv2168@columbia.edu	Affiliation:	Columbia U
Research Area:	NTM Stabilization including Rotation Dependence	Presentation time:	Not requested
Co-Author(s):	M. Okabayashi, E.J. Strait	ITPA Joint Experiment :	No
Description:	Variant of mode locking avoidance 'a la Okabayashi, invoking saddle loops when rotation gets too slow for Bp probes.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Reproduce #127927, feeding back on Mirnov. Repeat feeding back on saddle loops. Repeat with both diagnostic sets enabled, for f> and <100Hz respectively. Scan phase and gains. Find conditions such that the mode, after avoiding locking, spins up rapidly (which should also rotationally mitigate it) or keeps rotating at frequencies, f<1kHz, amenable by BES, CER and MSE diagnostics, or f<5kHz, optimal for modulated ECCD. For large enough gains, the â??phase flip instabilityâ?? predicted by E. Lazzaro and R. Coelho should be observed.
Background:	Although initially conceived for RWM control, M. Okabayashi's magnetic feedback was successful in avoiding locking of a rotating, slowing down NTM in #127927. In that case, the I-coils were feeding back on magnetic probe measurements. The probes worked remarkably well even at the low rotation frequencies (tens of Hz) reached by the mode before spinning up again. As a variant of M. Okabayashi's proposal 510, here it is proposed to feed back on Bp probes when f>100Hz and on BR loops when f<100Hz. Also related to E.J. Strait's proposal 182.
Resource Requirements:	--
Diagnostic Requirements:	--
Analysis Requirements:	--
Other Requirements:	--

Title	549: ELM-driven RWM onset and the transport properties with/without suppressing ELM-driven RWM
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	In the ELMy discharges, it was reported that ion temperature at the top of the H-mode pedestal was influenced by the RFA due to ELM-driven RWM (ref 1). The ion temperature was fluctuated with Î?Ti ~ 0.5â??1.0 keV near the edge of the plasma due to 15â??20 Hz ELM events and its formation of ELM-driven RWMs ~ 3â??5 Gauss. The RFA-suppressed condition by feedback showed Î?Ti ~ 0.2 keV by reducing the mode amplitude reduction to ~1 Gauss. The fluctuation of Ti at pedestal was significantly reduced around the plasma edge corresponding to q > 4. Here, it is proposed to document the ion transport property in relation to the RFA amplitude of which magnitude can be varied with the feedback.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	This experiment is useful to conceptualize the influence of error field on the ELM onset and to optimize the edge property in a consistent manner with other requirements near the core plasmas. The comparison with IPEC analysis will lead to develop better understanding of ergotarized boundary region. Ref 1 M. Okabayashi et al., 22nd IAEA Fusion Energy Conference Geneva, Switzerland, October 2008
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	550: Scan to find (qmin, q95) that maximize fBS and evaluate the steady-state potential of this profile
Name:	Holcomb holcomb@fusion.gat.com	Affiliation:	LLNL
Research Area:	Assess Steady-State Current Profiles for Optimum Performance	Presentation time:	Not requested
Co-Author(s):	J.R. Ferron	ITPA Joint Experiment :	No
Description:	This is for DIII-D milestone 170. Perform a deadicated scan of qmin and q95 at moderate betaN and evaluate the bootstrap current fraction and relative distribution of JBS. 3 day experiment. Day 1 performs scans in standard AT DN shape. Day 2 performs scans in LSN ITER-SSI scenario shape. Day 3 we take the one or two q-profile and shape combinations with max fBS and try to push to high betaN. Evaluate stability and ability to sustain this current profile.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Pick 3 q95 values: 3.5, 5.0, 6.5. Use now standard wide-ECCD in each, so vary q95 by fixing Bt and changing Ip to avoid having to reaim gyrotrons. At each q95, use PCS qmin control or open loop programming to obtain at least a second of conditions with betaN~3 and i) 1 < qmin <1.5, ii) 1.5 < qmin < 2.0, and iii) qmin > 2 in successive shots. Repeat at other q95, and mix scan order to reduce trends introduced by changing wall conditions.
Background:	See proposal 167 for background.
Resource Requirements:
Diagnostic Requirements:	MSE, Thomson, CO2, CER
Analysis Requirements:	Control room auto-ONETWO.
Other Requirements:

Title	551: High pedestal density/collisionality RMP ELM suppressed discharges in Helium
Name:	Schmitz oschmitz@wisc.edu	Affiliation:	U of Wisconsin
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):	E.A. Unterberg, M.E. Fenstermacher, T.E.Evans, R.A. Moyer	ITPA Joint Experiment :	Yes
Description:	ELM suppression appeared to be possible with rather low pitch-resonant field amplitudes at high pedestal electron collisionality (>1) but showed a dependence on the pedestal electron density. In this experiment helium plasmas will be used to study particle transport and exhaust during RMP application. The expected good density control with He plasmas will be used to resolve in more detail the density and or collisionality dependence of ELM suppression by RMP at high collisionality with odd partity phasing for n=3 I-coil fields. This study in He plasmas facilitates the particle exhaust analysis as the wall retention time of the He is low. Direct comparissons to TEXTOR circular limter plasmas in He and with RMP application are possible as collisionalities match in this regime. These topics are content of ITPA-PEP and ITPA-DIV/SOL tasks. This proposal represents a high collisionality, odd parity comparisson to proposal #108. However, it does not rely on cryo pumps frosted in Argon as reference discharges were done w/o strong pumping capabilities.	ITER IO Urgent Research Task :	Yes
Experimental Approach/Plan:	Repeat well studied deuterium samples at high collisonality with odd parity, n=3 RMP field. Reference shot is #119390 in high triangularity or #119690, which was in an ITER similar shape. We rely on establishement of an ELMy H-mode with heating and torque input scenario similar to the reference discharges. ELM suppression appeared to be dependent on plasma beta and ECH application might be needed to achieve ELM suppression. Once ELM suppression is established we aim at a pedestal electron density scan in order to resemble results in deuterium where ELM suppression was lost with increasing density.
Background:	In application of RMP for ELM suppression the role of wall pumping, recycling and mass dependence of the particle transport are important questions to be studied. Using He as main species allows to change the recycling drastically compared to deuterium as main species from close to 1 to practically 0 recycling. Wall pumping and edge parameter dependent fuelling characteristics are more easy to analyse. In addition density control is better in He plasmas as completely determined by the external source. Particle transport and exhaust during RMP application was studied at Tore Supra and very recently at TEXTOR. Within ITPA task PEP19, direct comparisson to DIII-D diverted H-mode plasmas are suggested and this experiment gives unique opportunity for that with direct overlap in collisionality. The comparison between results on both maschines shall contribute to the assessement of a density dependence of particle transport and the implication for the application towards ITER.
Resource Requirements:	He plasmas with high He purity (<2% risidual D), plasma shapes as in #119390 and #119690, cryo pumps frosted in Ar would be advantegeous but this is not essential, ECH with 6 gyrotrons ready, SPA supplies for I-coil in n=3, C-coil in standard n=1 EFC setting
Diagnostic Requirements:	ECE, Thomson Scattering (core, tangential and divertor systems would be desirable), BES, reflectometer, target Langmuir probes, DiMES_TV with C filters in filter wheel, tanTV system with C filters, IRTV (both cameras / LLNL+TEXTOR), MSE measurement combined with fast Li beam would be beneficial for edge current evolution during heat pulses, CER needed for rotation measurement
Analysis Requirements:	single reservoir particle balance, CER analysis, TRIP3D, kineticEFIT, EMC3/EIRENE modeling of He plasmas (code benchmark is enabled with this experiments)
Other Requirements:	this is in combination with #108 a full day experiment

Title	552: H-mode core turbulence and transport scaling with RMP ELM suppression
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	T.L. Rhodes, E.D. Doyle, W.A. Peebles, L. Zeng, T. Evans	ITPA Joint Experiment :	No
Description:	The goal of this experiment is to map out core turbulence vs. r/a as a function of I-coil current and correlate turbulence changes with radial particle and momentum transport in ITER-relevant ELM-suppressed RMP H-mode plasmas. The scaling of core turbulence and core transport with triangularity and elongation will be explored.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1) Establish RMP discharges with low triangularity, which show modest effects of the RMP field on central toroidal rotation and have only moderate density pumpout. Obtain a radial scan of fluctuation levels with DBS for 0.4 < r/a < 0.8 (3-4 shots). Data will be obtained throughout the discharge from L-mode to Elm-free/Elming H-mode and across the I-coil switch-on time into the Elm-suppressed phase. BES would be used for fluctuation measurements in addition to DBS if useful signals can be obtained in the H-mode core plasma. Obtain shifted equilibrium (plasma shifted downwards by 7 cm). Obtain radial scan of fluctuation level by DBS (3-4 shots) for lower k range (k rho_s ~0.3-0.5). 2) Establish one or two higher tringularity equilibria where RMP has a stronger effect on momentum and particle transport. Repeat fluctuation measurements. DBS data will provide local measurements of the radial electric field and electric field shear in addition to fluctuation levels which can be used in conjunction with or alternatively to CER data. In addition to triangularity, the elongation dependence of turbulence could also be explored if previous experiments indicate a strong dependence of momentum transport on kappa.
Background:	Recent diagnostic improvements and the implementation of a four-channel DBS system will allow us to follow the evolution of density fluctuations in the core of RMP H-modes as the I-coil current is increased. Turbulence measurements can now be made across a range of poloidal modenumbers (0.1 < k_perp rho_s < 4) by DBS and BES. Core toroidal rotation reduction of ~40% has been observed in some RMP discharges (#129193), and preliminary BES/DBS fluctuation measurements indicate changes in intermediate scale turbulence inside the H-mode pedestal during I-coil application.
Resource Requirements:	7 beams, I-coil
Diagnostic Requirements:	DBS,BES, FIR, MSE DBS access to the H-mode core plasma is best for central densities â?¤2.3 10^13 cm^-3 in X-mode (for B =1.95 T). Slightly higher densities can be accommodated at 1.8T. O-mode backscattering can also be used with slightly decreased DBS performance which allows core access at substantially higher density. The best way to proceed in planning would be if I could look at candidate shots to evaluate accessibility and /or suggest minor modifications in n and B.
Analysis Requirements:	The TGLF code would be used for calculations of linear stability and quasilinear fluctuation levels and for comparisons with the experimentally obtained fluctuation characteristics. CER and DBS data will be used to calculate shearing rates to compare to linear growth rates. The analysis goal would be to assess whether the changes in particle and momentum confinement observed with RMP application are linked to increased turbulent transport.
Other Requirements:

Title	553: Investigate field effects on stability versus profile evolution
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):	W. Solomon, M. Fenstermacher, T.E. Evans	ITPA Joint Experiment :	No
Description:	The object of this experiment is to use the intrinsic timescales for the RMP penetration (treconn and tR = 100 microsec), stability physics (tA, etc. ) versus pedestal profile changes (5-50 ms) to try to isolate direct 3D field effects distinct from equilibrium profile changes. This separation of timescales has been seen e.g. in XGC0 simulations of pedestal evolution in realistic geometry with self-consistent Er evolution, where the Er and plasma potential change within 60 ion transit times (about 4 ms in the experiment) while the pedestal profiles continue to evolve.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1. Establish a good reference ELMing H-mode. 2. Get good RMP ELM suppression 3. Find the threshold I-coil current needed for ELM suppression 4. Add a.c oscillation to the d.c. level to oscillation RMP below threshold to above and back at about 50 Hz. 5. Use "standard" profile, ELM , and fluctuation diagnostics to characterize the plasma response 6. Attempt phase-coherent detection with UCSD fast camera to try to image the boundary.
Background:	Both experiments and XGC0 modeling indicate that there are prompt changes (e.g. to toroidal rotation and Er) which occur faster than the ramp-up time of the I-coil current (usually 20-50 ms depending on power supplies used), while pedestal profiles take somewhat longer to evolve (several ms to 50 ms; longer than this in the core). Use this separation of time scales to investigate direct 3D field effects on momentum, non-ambipolar losses, and stability independent of pumping and wall-retention effects.
Resource Requirements:	I-coils (offset a.c. even parity with 0 degree phasing already developed in CY08) cryopumps neutral beams
Diagnostic Requirements:	standard pedestal and core profile diagnostics fluctuation diagnostics boundary heat and particle flux diagnostics (IRTVs, floor and reciprocating LPs, DiMES TV, UCSD fast camera)
Analysis Requirements:
Other Requirements:

Title	554: Toroidal phase convergence on the DEFC iterations
Name:	Okabayashi mokabaya@pppl.gov	Affiliation:	PPPL
Research Area:	Physics of Non-Axisymmetric Field Effects in Support of ITER	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	When sufficient iterations are made, the DEFC can reduce the remaining error to minimum both in amplitude and torodial phase. However, required numbers of iterations for conversing to the optimum amplitude and toroidal phase direction can differ significantly . In addition, the required accuracy on the toroidal phase and the absolute amplitude may depend upon specific experimental subjects. Simple iteration formulation predicts that the toroidal phase convergence needs more iteration. Approach By applying different type of simulating error field with C-coil and individual I-coils, the error field correction, the DEFC with various I-coil and c-coil combination will be documented with several iterations. Better understanding of convergence in the toroidal phase will help to estimate the location of uncorrected error field	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements:

Title	555: Investigate field effects on stability versus profile evolution
Name:	Moyer moyer@fusion.gat.com	Affiliation:	UCSD
Research Area:	ELM Control for ITER	Presentation time:	Not requested
Co-Author(s):	W. Solomon, M. Fenstermacher, T.E. Evans	ITPA Joint Experiment :	No
Description:	This proposal is a duplicate submission to ROF 553 submitted to the Physics of Non-Axisymmetric Fields Task Force. The object of this experiment is to use the intrinsic timescales for the RMP penetration (treconn and tR = 100 microsecs) stability (tA, etc.), versus pedestal profile evolution (5-50 ms) to try to isolate direct 3D field effects distinct from equilibrium profile changes (due to density pumpout). This separatrion of timescales has been seen e.g. in XGC0 simulations of pedestal evolution in realistic geometry with self-consistent Er evolution, where the Er and plasma lpotential change within 60 ion transit times (4 ms in the experiment) while the pedestal profiles continue to evolve.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	1. Establish a good reference ELMing H-mode 2. Get good RMP ELM suppression (even parity, 0 deg) 3. Find the threshold I-coil current for suppression 4. Add a.c. oscillation to the d.c. level to oscillate the RMP from below this threshold to above at about 50 Hz 5. use "standard" profile, ELM, and fluctuation diagnostics to characterize the plasma response 6. use phase-coherent detection with the UCSD fast camera to try to image the boundary structures.
Background:	Both experiments and XGC0 modeling indicate that there are lprompt changes (.e.g. to toroidal rotation and Er) which occur faster than the ramp-up time of the I-coil current (usually 20-50 ms depending on the power supplies used), while pedestal profiles take somewhat longer to evolve (several ms to 50 ms; longer than this in the core). Use this separation of time scales to investigate direct 3D field effects on momentum, non-ambipolar losses, and stability independent of pumping and wall retention effects.
Resource Requirements:	I-coils (offset a.c. even parity with 0 degree phasing already developed in CY08) cryopumps neutral beams
Diagnostic Requirements:	standard core and pedestal profile diagnostics fluctuation diagnostics boundary heat and particle flux diagnostics (IRTVs, floor and reciprocating LPs, DiMES TV, UCSD fast camera, etc.)
Analysis Requirements:
Other Requirements:

Title	556: Simulating ITER startup, Scenario Access, and Rampdown in H2 and He
Name:	Jackson jackson@fusion.gat.com	Affiliation:	GA
Research Area:	Hydrogen and Helium Plasmas	Presentation time:	Not requested
Co-Author(s):	T. Casper, A. Hyatt	ITPA Joint Experiment :	No
Description:	Demonstrate a complete ITER-like discharge including low voltage breakdown (with ECH), rampup, H-mode in flattop, and rampdown in helium and hydrogren discharges	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Compare a hydrogen and helium discharges to deuterium discharges using the ITER large bore startup and baseline H-mode in flattop to demonstrate that all elements of an ITER discharge can be achieved in a single shot. Compare with 133176(D2). Estimate 0.5 days each (H2 and He). Depending upon conditions for other experiments, this work might become a piggyback, though it requires the 'front end' of the plasma discharges and a specific F-coil patch panel for LFS startup.
Background:	ITER will begin operations with either H2 or He in order to minimize machine activation. It is important to understand the differences between H2,D2, and He. A significant data base now exists for ITER startup scenarios scaled to DIII-D, including LFS initiation, Ip ramp up to q95 ~ 3, and a flattop phase. This should be repeated in H2 and He to provide an existence proof that it will be possible in ITER. Differences between D2 and H2/He will be noted
Resource Requirements:	ECH and NB for heating
Diagnostic Requirements:	usual diagnostics
Analysis Requirements:	--
Other Requirements:	--

Title	557: Turbulence and Transport dependence on Mach number in Hybrid discharges
Name:	McKee mckee@fusion.gat.com	Affiliation:	U of Wisconsin
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	C. Petty, T. Rhodes, L. Schmitz, A. White, Z. Yan	ITPA Joint Experiment :	No
Description:	Study 2D turbulence structure and suppression as a function of rotationally varied ExB shear in long-duration, high beta hybrid H-mode plasmas via co- and counter current 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 recently expanded upgraded and expanded 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 difficult, 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 very 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).
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 (Petty, 200?). This has been shown to be consistent with the a reduction in ExB shearing at lower Mach number from GLF23 modeling. Previous measurements of turbulence characteristics in hybrid discharges (McKee, APS-2005) with the upgraded BES diagnostic, showed that turbulent eddies exhibit a strongly tilted structure. This is in sharp contrast to the more radially-poloidally symmetric eddy structure typically observed in the core of L-mode discharges. The direction of this tilted eddy structure is consistent with the ExB shear flow in these plasmas, although it was questionable as to whether the shear magnitude could bring about such a strong eddy tilt.
Resource Requirements:	Co/Ctr NBI, hybrid plasmas
Diagnostic Requirements:	BES, DBS
Analysis Requirements:	TGLF
Other Requirements:

Title	558: Ip scan: Testing whether TGLF and GYRO results are consistent with multi-machine scaling laws
Name:	White whitea@mit.edu	Affiliation:	Massachusetts Institute of Technology
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	R. Prater, T. L. Rhodes	ITPA Joint Experiment :	No
Description:	In this experiment TGLF and GYRO predictions for transport will be tested against multi-machine scalinq laws via a traditional plasma current scan. Starting from a well-modeled discharge (TGLF/GYRO used for sawtooth-free L-mode plasma, 128913, t = 1500 ms, Ip = 1 MA) we will perform a traditional plasma current scan. Based on multi-machine scaling laws, as the current is increased the confinement time should increase in this L-mode plasma. It is expected that turbulent fluctuations will be reduced as Ip is increased as the transport is reduced. The two goals of the experiment are to (1) investigate the scaling of turbulence and transport with plasma current, Ip, in experimental conditions suitable for modeling with theory-based transport models (e.g.TGLF, TGYRO) and to (2) test the underlying drift-wave turbulence physics of those models via direct comparisons between the measured fluctuations and the GYRO code.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Recreate a discharge similar to 128913 (Bt ~ 2.14 T, Ip=1.0 MA) with a sawtooth-free L-mode plasma (t = 1300-1700 ms) that has already been extensively modeled using TGLF, TGYRO and GYRO. Perform parametric scans to investigate the scaling of turbulence and transport with plasma current. Two scans are possible: (1) Ip scan at fixed density, magnetic field and input power and (2) Ip scan at fixed q, density, input power (change magnetic field). Note that neither are a dimensionless parameter scan, but are more traditional current scans in so far as collisionality, beta, rho-star, s-hat and q are allowed to change. In scan (2), we will try to keep q at the measurement locations of interest (0.4 < r/a < 0.8) constant by changing the magnetic field as Ip changes, but again collisionality, beta, rho-star and s-hat will be allowed to vary. In the discharges, early beam heating will be used to delay the onset of sawteeth by slowing the current diffusion time, hence over the time period of interest q will be slowly evolving but q-min will be above unity. In scan (1) fluctuations will be measured across all k-ranges and low-k Te-tilde and n-tilde will be monitored simultaneously between t = 1300-1700 ms during the sawtooth-free L-mode plasma for 2-3 values of Ip. We will have 3-4 repeat shots per Ip value for profile scans of fluctuation diagnostics. In scan (2) as the magnetic field is reduced (as we would start the scan at high q, low-Ip) we will lose some use of ECE radiometers and reflectometers, so that we may only obtain fluctuation levels with CECE at one radial location. Scan 2 will still allow for BES and possibly DBS, FIR and backscattering access. Although 128913 is a the preferred reference shot, this Ip-scan could be used as a secondary scan to augment the q-scan experiment proposed by Rhodes in TMV 2009. For example, at the high-q (low Ip) point in Rhodesâ?? scan we could then vary the current to complete scan (1) as described above.
Background:	Many tokamak experiments have studied the scaling of confinement and transport on plasma current, with the result that mutli-machine empirical scaling laws show that tau_e propto Ip^x, where x is typically near unity. These types of scaling laws could also be tested against theory-based transport models (e.g. TGLF), but care must be taken that the experimental conditions are suitable for testing the physics contained in the models. For this reason, a very simple MHD free (sawtooth free) L-mode plasma will be used for a traditional plasma current scan. The dependence of confinement time (tau_e) and transport (Chi_e, Chi_i, Chi_eff) can be tested against the multi-machine scaling laws and also against predictions from theory based transport models. Ideally the experiment will also include fluctuation measurements, which can be used to further test the transport models by allowing for direct testing of the underlying physics via comparisons with GYRO.
Resource Requirements:	All available NB sources
Diagnostic Requirements:	Thomson scattering, Michelson interferometer, 40-channel ECE radiometer, CER and MSE, fast magnetics, all available fluctuation diagnostics.
Analysis Requirements:	CERFIT, EFIT, ONETWO/autoonetwo, TGLF transport model,TGLF (linear stability analysis) TGYRO, GYRO, GENRAY
Other Requirements:

Title	559: Test of Turbulence Spreading Using Turbulence Propagation
Name:	Petty petty@fusion.gat.com	Affiliation:	GA
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):		ITPA Joint Experiment :	No
Description:	The question of turbulence spreading, that is, whether turbulence is or is not a strictly local phenomenon, can be precisely tested by modulating the turbulence (and plasma profile) at a fixed location and then monitoring the propagation of the turbulence (and plasma profiles) away from this region. If the turbulence propagation speed is much faster than the temperature or density propagation speed, then this can be attributed to turbulence spreading. For this purpose it does not matter much how the turbulence is modulated; it can be a simple amplitude modulation or something more sophisticated such as a modulation of the radial correlation length. The most likely source of modulation is ECH, either as a monopolar change in the electron temperature profile or as a "swing" experiment where the ECH deposition is alternated between two (closely spaced) location. The turbulence diagnostic must be capable of covering a large radial range, so the 32 channel linear array of the BES diagnostic is ideally suited for this experiment. An 8 channel DBS diagnostic would also be useful to monitor the propagation of intermediate k turbulence.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	To minimize MHD, this experiment will use an L-mode plasma with 1-2 sources of continuous NBI for diagnostic purposes (BES, CER, MSE) and 6 gyrotrons for turbulence modulation. If the beam power needs to be limited to 1 source, repeat shots can be taken to switch between beams. The ECH modulation rate should be relatively high (~100 Hz) to allow an accurate measurement of the propagation speed. Actually it is preferable to study several different modulation rates, so repeat shots will be taken to cover the range 25-200 Hz.
Background:
Resource Requirements:	Beams: 30LT, 330LT, 150LT ECH: Six gyrotrons
Diagnostic Requirements:	BES 32 channel linear array DBS 8 channel array
Analysis Requirements:
Other Requirements:

Title	560: Toroidal/poloidal mapping of Zonal Flows and GAMs
Name:	Schmitz schmitzl@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport	Presentation time:	Not requested
Co-Author(s):	T.L. Rhodes, W.A Peebles, J. C. Hillesheim,G.R. McKee	ITPA Joint Experiment :	No
Description:	Zonal flows (both zero-mean-frequency or low frequency ZF's and GAMs) are poloidally and toroidally symmetric potential perturbations. Measurements of the toroidal coherence length of low frequency velocity fluctuations will be used to discriminate ZF's against finite parallel wavelength (n=1,2,3) modes. This technique allows the unambiguous identification of Zonal Flows levels and allows a quantitative measurement of ZF amplitude in regimes dominated by ITG and/or TEM turbulence.	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	Recent improvements in diagnostic access for Doppler Backscattering (two DBS systems located at 60 degrees and 240 degrees) will allow toroidal/poloidal mapping of ZF's and the determination of the toroidal (and poloidal) coherence length. This experiment requires two DBS system to be operated at the same frequency and a slow q-scan over a limited interval to align the diagnosed volumes in the plasma poloidally. Once the toroidal coherence length is determined, the variation of the poloidal coherence with pitch angle mismatch can be used to extract the poloidal coherence length. This experiments can be carried out in initially wall limited OH/L-mode plasmas switched to USN plasmas in order to explore GAM/ZF amplitude and frequency scaling on elongation and triangularity
Background:	DBS is a very sensitive diagnostic for the measurements of poloidal flow fluctuations, and has observed ZMF/low frequency ZF's and GAMS in the core and edge of DIII-D.
Resource Requirements:	Beams
Diagnostic Requirements:	DBS, BES, CER, MSE
Analysis Requirements:	--
Other Requirements:	--

Title	561: kappa/kappa' scaling
Name:	Rhodes rhodes@fusion.gat.com	Affiliation:	UC, Los Angeles
Research Area:	Transport Model Validation	Presentation time:	Not requested
Co-Author(s):	Kinsey	ITPA Joint Experiment :	No
Description:	This is an idea that Jon Kinsey sent to me (Rhodes) via email. I submitted it for him. Three phase kappa experiment on DIII-D -L- TEM, L- ITG, and finally H-mode. 1) L-mode ECH only w/ NBI blips for Ti measurements varying kappa from close to unity to 1.8. This would test TGLF and GYRO in the TEM/ETG dominated regime. It would also be of interest to test the collision model in TGLF in this regime. 2) Later in the discharges, add 2.5-5.0 MW of NBI. This would then test the kappa dependence in more ITG dominated L-modes. 3) Increase NBI late in the discharges in order to enter H-mode	ITER IO Urgent Research Task :	No
Experimental Approach/Plan:	A best attempt match with a priority on matching rhostar, q, and a/Ln, a/Lt. A match in nustar would be nice it is not as important given the weak dependence seen in previous DIII-D experiments and in GYRO simulations. Ideally, the NBI phases should be balanced in order to reduce the impact of variations in ExB shear.
Background:
Resource Requirements:
Diagnostic Requirements:
Analysis Requirements:
Other Requirements: