OAK-B135 Advanced tokamak research in DIII-D seeks to optimize the tokamak approach for fusion energy production, leading to a compact, steady state power source. High power density implies operation at high toroidal beta, {beta}{sub T}=&lt;p&gt;2{micro}{sub 0}/B{sub T}{sup 2}, since fusion power density increases roughly as the square of the plasma pressure. Steady-state operation with low recirculating power for current drive implies operation at high poloidal beta, {beta}{sub P} = &lt;p&gt;2{micro}{sub 0}/&lt;B{sub P}&gt;{sup 2}, in order to maximize the fraction of self-generated bootstrap current. Together, these lead to a requirement of operation at high normalized beta, {beta}{sub N} = {beta}{sub T}(aB/I), since {beta}{sub P}{beta}{sub T} {approx} 25[(1+{kappa}{sup 2})/2] ({beta}{sub N}/100){sup 2}. Plasmas with high normalized beta are likely to operate near one or more stability limits, so control of MHD stability in such plasmas is crucial

Bialek, J.

Chance, M. S.

Chu, M. S.

Edgell, D. H.

Ferron, J. R.

Garofalo, A. M.

Greenfield, C. M.

Humphreys, D. A.

Jackson, G. L.

Jayakumar, R. J.

Jernigan, T. C.

Kim, J. S.

La Haye, R. J.

Lao, L. L.

Luce, T. C.

Makowski, M. A.

Murakami, M.

Navratil, G. A.

Okabayashi, M.

Petty, C. C.

Reimerdes, H.

Scoville, J. T.

Strait, E. J.

TEAM, DIII-D

Turnbull, A, D,

Wade, M. R.

Walker, M. L.

Whyte, D. G.

UNT Digital Library

G A-A24356 CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES by E.J. Strait, J. Bialek, M.S. Chance, M.S. Chu, D.H. Edgell, J.R. Ferron, C.M. Greenfield, A.M. Garofalo, D.A. Humphreys, G.L. Jackson, R.J. Jayakumar, T.C. Jernigan, J.S. Kim, R.J. La Haye, L.L. Lao, T.C. Luce, M.A. Makowski, M. Murakami, G.A. Navratil, M. Okabayashi, C.C. Petty, H. Reimerdes, J.T. Scoville, A.D. Turnbull, M.R. Wade, M.L. Walker, D.G. Whyte and the DIN-D Team JUNE 2003 DISCLAIMER II This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. I' P G A-A24356 CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES by E.J. Strait, J. Bialek*, M.S. Chance*, M.S. Chu, D.H. Edgellt, J.R. Ferron, C.M. Greenfield, A.M. Garofalot, D.A. Humphreys, G.L. Jackson, R.J. Jayakumar , T.C. Jernigan#, J.S. Kim? R.J. La Haye, L.L. Lao, T.C. Luce, M.A. Makowski , M. Murakami#, G.A. Navratilt, M. Okabayashi*, C.C. Petty, H. Reimerdes? J.T. Scoville, A.D. Turnbull, M.R. Wade#, M.L. Walker, D.G. WhyteO and the Dlll-D Team This is a preprint of a paper to be presented at the 30th European Physical Society Conference on Controlled Fusion and Plasma Physics, SSt. Petersburg, Russia, July 7-11, 2003 and to be published in the Proceedings. *Princeton Plasma Physics Laboratory, Princeton, New Jersey. +FARTECH San Diego, California, USA kolumbia University, New York, New York, USA 'Oak Ridge National Laboratory, Oak Ridge, Tenessee, USA 'University of Wisconsin, Madison, Wisconsin, USA Lawrence Livermore National Laboratory, Livermore, California, USA Work supported by the US. Department of Energy under Contract Nos. D E-AC03-99E R54463, D E-AC02-76CH03073 and DE-FG02-89ER53297 c GENERAL ATOMICS PROJECT 30033 JUNE 2003 CONTROL OF MHJ3 IN DE-D ADVANCED TOKAMAK DISCHARGES E.J. Strait, et al. Control of MHD Stability in DIII-D Advanced Tokamak Discharges E.J. Strait', J. Bialek2, M.S. Chance2, M.S. Chul, D.H. Edgell3, J.R. Ferron', C.M. Greenfield', A.M. Gar0fal04, D.A. Humphreys', G.L. Jackson', R.J. Jaykumars, T.C. Jernigan6, J.S. Kim3, R.J. La Hayel, L.L. Laof, T.C. Lucel, M.A. Makowskis, M. Murakami6, G.A. Navratil4, M. Okabayashi2, C.C. Petty', H. Reimerdes4, J.T. Scoville1, A.D. Turnbulll, M.R. Wade6, M.L. Walkerl, D.G. Whyte7, and the DIII-D Team 1 General Atomics, P. 0. Box 85608, San Diego, California, USA 2Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA 3FAR-TECH, Inc., San Diego, California, USA 4Columbia University, New York, New York, USA SLawrence Livermore National Laboratory, Livermore, California, USA 6Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA 7 University of Wisconsin, Madison, Wisconsin USA Advanced tokamak research in DIII-D seeks to optimize the tokamak approach for fusion energy production, leading to a compact, steady state power source. High power density implies operation at high toroidal beta, f+(p)2CLO/B+, since fusion power density increases roughly as the square of the plasma pressure. Steady-state operation with low recirculating power for current drive implies operation at high poloidal beta, Pp = (p)2p~/(Bp)2, in order to maximize the fraction of self-generated bootstrap current. Together, these lead to a requirement of operation at high normalized beta, PN = fi~(aB/I), since Pp& = 25[(1+K2)/2] (P~/100)~ .  Plasmas with high normalized beta are likely to operate near one or more stability limits, so control of MHD stability in such plasmas is crucial. 1. Kink Mode Stabilization With A Resistive Wall In "advanced tokamak" scenarios, the broad current density profile associated with a large bootstrap current leads to a relatively low free-boundary kink mode beta limit, but also allows the possibility of stabilization by an ideally conducting wall. In the presence of a resistive wall, such as the DIII-D vacuum vessel, the kink mode is not completely stabilized but is converted to a slowly-growing resistive wall mode (RWM), which can be stabilized by feedback control or plasma rotation. Sustained wall stabilization of the external kink mode has been demonstrated in an advanced tokamak discharge with high normalized beta and high bootstrap current [ 11. As seen in Fig. 1, beta exceeds the calculated no-wall limit of PN - 44 for almost one second, and reaches a maximum near the estimated ideal-wall stability limit at PN - Sei. Here plasma rotation plays a major role in the stabilization, with a set of six external, non-axisymmetric coils (C-coils) providing feedback-controlled correction of n= 1 error fields that would otherwise slow the plasma rotation. Direct feedback control of the resistive wall mode has also been demonstrated in plasmas where the rotation had decayed to a value below the threshold for stabilization by rotation alone [2,3]. 1 General Atomics Report GA-A24356 E.J. Strait, et al. CONTROL OF h4HD IN DIII-D ADVANCED TOKAMAK DISCHARGES Recently, a new set of internal control coils (I-coils) has been installed in DIII-D, consisting of two sets of six coils, one set above and one below the midplane [4]. VALEN code calculations [SI predict that the I-coils can allow feedback-stabilized operation up to the ideal-wall limit even in the absence of plasma rotation. Preliminary results indicate that the I-coil provides feedback-driven error field correction at least as effectively as the C-coil, and stable operation above the estimated no- wall beta limit has been sustained for up to 2.5 s. Preliminary results are encouraging for the prospects of resistive wall mode control without plasma rotation. Figure 2 shows a case in which strong “magnetic braking” reduced the plasma rotation to a very low value, essentially zero in the outer half of the plasma. Feedback stabilization with the I-coil sustains the plasma at beta above the no- wall limit for about 100 ms after the outer plasma rotation reaches zero. A 20 10 0: loo0 1200 1400 1600 1800 2OOO 2200 2400 Fig. I .  (a )  An advanced tokamak discharge (65% bootstrap current and 85% total noninductive current, @4%) is stable above the ideal no-wall limit, with (b) rapid toroidal rotation made possible by &e&ack- controlled error field reduction. Time (ms) Time (ms) P Fig. 2.  (a) A discharge with a low no-wall stability limit (red curves) is stabilized by feedback after (b) the rotation at pzO.5 has decreased to Zero. A lower-beta &charge without feedback (blue curves) becomes unstable (c) as the rotation at p=0.6 decays below about 6 kHz. (d) Comparison of rotation profiles just before the onset of the resistive wall mode. comparison discharge without feedback becomes unstable when the outer plasma rotation decays below a threshold value of about 6 lcHz, even though beta is slightly smaller. 2. Neoclassical Tearing Mode Stabilization High beta plasmas are subject to the neoclassical tearing mode (NTM), where the helical perturbation to the bootstrap current caused by a magnetic island further destabilizes the island and causes it to grow. The wall-stabilized discharge shown in Fig. 1 is ultimately ended by an NTM. Stabilization of neoclassical tearing modes can be achieved through replacement of the missing bootstrap current with electron cyclotron current drive. DIII-D experiments have demonstrated suppression of the m/n=3/2 and 2/1 modes, using precise feedback control of the current drive position to within 1 cm by variation of the plasma position or toroidal field [6]. As shown in Fig. 3, once the NTM is stabilized, beta can be raised to a value higher than at the initial onset of the NTM. In this example, the feedback system is driven by minimizing the amplitude of the detected NTM and cannot follow the location when the mode amplitude vanishes. More recently, feedback control has been upgraded to provide real-time tracking of the rational surface associated with the NTM, even in the absence of an unstable mode. General Atomics Report GA-A24356 2 CONTROL OF MHD IN DIE-D ADVANCED TOKAh4AK DISCHARGES 3. Profile Control Stable, steady-state operation with high fusion performance may depend on maintaining profiles that differ from those that would naturally evolve in the plasma, so local control of the pressure, current density, and rotation profiles is an essential element of advanced tokamak plasmas. Recent DIII-D experiments have demonstrated feedback control of the local electron temperature by electron cyclotron heating, and modification of the central current density profile evolution with electron cyclotron current drive [7]. Real- time equilibrium reconstructions using motional Stark effect data are being E.J. Strait, et al. 7. 10 0. 3. 2. 1. 0. Time (ms) t Beams off Fig. 3 .  Suppression of an mln=3/2 neoclassical tearing mode. (a) When electron cyclotron wa drive is turned on, (b) the n=2 mode amplitude is reduced to zero. (c) Beta can then increase above the level where the NTM first appeared. Eventually the stabilization is lost as the increased Shaffanov shi3 moves the q=3/2 surface away ffom the ECCD resonance. implemented, and will ultimately allow real-time control of the current density profile. Ideally, real-time profile measurements should be compared to predicted stability limits in order to avoid instabilities, but reliable prediction of those limits in real time may be a significant challenge. “MHD spectroscopy,” based on the plasma’s resonant response to small-amplitude magnetic perturbations near a stability limit [8] , provides a direct, real-time measurement of plasma stability that may prove useful as a control input. 4. Disruption Mitigation In the event that avoidance of instabilities by profile control and suppression of instabilities by direct feedback control both fail, a disruption may follow. Disruption mitigation by injection of a high-pressure impurity gas jet leads to a radiative thermal quench and rapid current decay, reducing runaway electrons, thermal loads, and electromagnetic forces on plasma facing components [9]. In recent DIII-D experiments, real-time detection of off-normal conditions has been successfully used to trigger such a controlled termination. In Fig. 4, the vertical position control is deliberately switched off to create a disruption. The control system detects the excursion in vertical position and triggers a high-pressure neon jet. More sensitive detection and earlier triggering increase the radiated power and reduce the halo currents and associated electromagnetic forces. Similar detection schemes have been developed for density limit and locked-mode disruptions. 5. Conclusions Experiments in DIII-D have demonstrated many of the elements needed for control of MHD stability in advanced tokamak plasmas. Stabilization of the external kink by a resistive wall with rapid plasma rotation is an effective technique, made possible by feedback-con- trolled reduction of error fields. The feasibility of stabilizing resistive wall modes and neo- classical tearing modes by direct suppression of the instabilities has been demonstrated. The tools for real-time profile control are at hand, including profile measurements and means 3 General Atomics Report GA-A24356 E.J. Strait, et al. CONTROL OF MHD IN DID-D ADVANCED TOKAMAK DISCHARGES of localized heating and current drive, while MHD spectroscopy provides a direct measurement of proximity to a stability limit. Gas jet injection can terminate the discharge safely if a stability limit is exceeded. A number of scientific and technical issues remain for implementing these techniques in a burning plasma. A better understanding of the physics of plasma rotation and rotational stabilization is needed for extrapolation to a burning plasma, which is likely to have little or no torque from neutral beam heating. Models of RWM feedback stabilization in a rotating plasma are being developed [lo], and validation is needed. Nonmagnetic methods such as reflectometry may allow resistive wall mode detection at a threshold of 1 part in lo3 or lo4 of the ambient magnetic field in long pulses where the accuracy of inductive sensors is a challenge. In general, active control of profiles and of the instabilities themselves will require flexible systems of actuators and complete 0 -1 0 -20 2 1 0 1 0 2 0 uilibrium Position 4 Plasma Current (MA) 20 40 60 Time After Loss of Vertical Stabilization (ms) Fig. 4.  Real-time triggering of a high-pressure gas jet during a vertical displacement event increases the radiative dissipation and reduces the halo currents during the disruption. measurements in real time of the internal state of the plasma. An integrated system of real- time measurement and control remains to be demonstrated. Finally, although the gas jet mitigation technique appears to scale favorably to the burning-plasma regime, this must be shown through comparative experiments on existing tokamaks. Work supported by U.S. Department of Energy under Contract Nos. DE-AC03- 99ER54463, DE-AC02-76CH03073, DE-FG03-99ER8279 1, DE-FG02-89ER53297, W-7405-ENG-48, DE-AC05-000R22725, and DE-FG02-89ER53296. A.M. Garofalo, et al., Phys. Plasmas 9, 1997 (2002). M. Okabayashi, et al., Phys Plasmas 8, 2071 (2001). E.J. Strait, et al., Nucl. Fusion 43, 430 (2003). G.L. Jackson, et al., “Initial Results from the New Internal Magnetic Field Coils for Resistive Wall Mode Stabilization in the DUI-D Tokamak,“ this conference. J. Bialek, et al., Phys. Plasmas 8, 2170 (2001). R.J. La Haye, et al., Phys. Plasmas 9, 2051 (2002); “Complete Suppresion of the m/n=2/1 Neoclassical Tearing Mode Using Radially Localized Electron Cyclotron Current Drive on DIII-D and the Requirements of ITER,” this conference. C.M. Greenfield, et al., “Progress Toward Fully Noninductive High Beta Discharges in DIII-D,” this conference. H. Reimerdes, et al., “Resistive Wall Modes and Plasma Rotation in DUI-D,” this conference. D.G. Whyte, et al., Phys. Rev. Lett. 89, 055001 (2002). M.S. Chance, et al., “Modeling of the Feedback Stabilization of the Resistive Wall Mode in Tokamaks.” this conference. General Atomics Report GA-A24356 4 

CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES

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CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES

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