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ARIES-Advanced Tokamak Power Plant Study Physics Analysis and Issues Charles Kessel, for the ARIES Physics Team Princeton Plasma Physics Laboratory U.S.-Japan Workshop on Fusion Power Plant Design, University of Tokyo March 29-31, 2001

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ARIES-AT Physics Team UCSD –T. K. Mau –R. Miller –F. Najmabadi PPPL –S. Jardin –C. Kessel GA –V. Chan –M. Chu –R. La Haye –L. Lao –T. Petrie –G. Staebler

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Objectives for ARIES-AT Physics Extend the physics basis for ARIES-RS to increase and decrease P CD –Use 99% flux surface from free-boundary equilibrium –Use more flexible plasma pressure profile description –Increase plasma triangularity –Increase plasma elongation –Eliminate HHFW current drive

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Objectives for ARIES-AT Physics Include more detailed analysis of critical physics issues –Kink stabilization, RWM analysis with rotation and feedback control –NTM assessment –Comparison of kinetic profiles with experiments and Gyro-Landau-Fluid Model –Edge plasma/SOL/Divertor modelling –Plasma edge assumption, L-mode and H-mode

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ARIES-RS and AT Parameters

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ARIES-AT Equilibrium

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Use 99% flux surface from free- boundary equilibrium in fixed boundary equilibrium and stability analysis Old method used 95% surface and analytic plasma boundary 99% plasma surface has stronger shaping 99% plasma surface contains higher order shaping from realistic PF coils 99% plasma volume is consistent with free- boundary plasma

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Expand the plasma pressure profile description More flexible pressure profile allows better bootstrap alignment and higher ballooning limits simultaneously The HHFW current drive could then be eliminated ARIES-AT ARIES-RS

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Minimum CD power does not occur at the highest Reduction of dp/d near the plasma edge for ballooning stability reduced bootstrap current there Higher N leads to excessive bootstrap current near plasma center requiring broader density profiles These lead to larger externally driven current at highest ’s

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Increasing the plasma triangularity leads to higher Higher increases the N for ballooning modes, and increases the plasma current for fixed q 95, while higher enhances the effect N (Ip/aB)

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Increasing the plasma elongation leads to higher Higher leads to higher N for ballooning modes when is less than so long at is high enough, however continues to rise due to the strong increase in Ip

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The plasma elongation is limited by the vertical instability Vertical stability analysis showed that we could have X up to with tungsten conductors a from the plasma boundary which is inside the blanket

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ARIES-AT vertical stability and control solution Partial tungsten shells on the inboard and outboard sides, 4 cm thick, at 1100 degC. Feedback control coils are located behind the sheild. For 30 MVA limit we produce a large range of plasmas

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External kink stability is provided by a conducting shell and feedback coils For >2.3 the conducting shell must move much closer to the plasma, and higher N leads to a closer shell and higher limiting n (toroidal mode number)

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The location of the conducting wall is affected by triangularity Higher has allowed us to place the conducting shell further from the plasma, at the same location as the vertical stabilizer

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From theory rotation of the plasma with a conducting shell might stabilize the external kink mode ARIES-AT requires 9% of the Alfven speed, which is difficult to provide, and experiments indicate it may not work

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ARIES-AT uses feedback to stabilize the external kink modes Modelled after DIII-D C-coil approach Tungsten shells on outboard side slow the kink mode growth rate creating a RWM Feedback coils are 8 (or 16) window coils on the outboard side Feedback coils span 60 deg poloidally about the midplane and are outside the shield Estimated power requirements are about 10 MW Only n=1 has been addressed

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External current drive is required on axis and near the plasma edge ICRF/FW on-axis CD; 96 MHz, N || =2, P CD =5 MW, =0.032 A/W LHCD off-axis CD; 3.6 GHz, N || = , P CD =24.5 MW, = A/W; 2.5 GHz, N || =5.0, P CD =12.5 MW, =0.013 A/W

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Scaling of the CD power with temperature and Zeff is used to find the best operating point CD efficiency includes bootstrap current effect Higher Zeff reduces divertor heat load Higher temperature improves CD efficiency and leads to thermally stable operating points N || were varied for this scan

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Alternate CD sources are examined for plasma rotation and current profile control 120 keV NBI provides plasma rotation and current for >0.6, P(NBI)=44 MW for 1.15 MA HHFW at 20 ci provides current at = which is useful for control of q min and r/a(q min ), P(HHFW)=20 MW for 0.4 MA

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Neoclassical tearing modes must be stabilized to reach ideal MHD limits A combination of the LHCD current profile modification and ECCD at the 5/2 is expected to stabilize the NTM (plot below is H-mode edge), while 3/1 is in very high collisionality region

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Examination of assumed T and n profiles and those theoretically predicted by GLF23 Density is input Temperature is determined, found to be close to assumed profile GLF23 is theory based predictive model including ITG, TEM, and ETG turbulence

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Comparison of L-mode and H-mode plasma edges H-mode edge has higher T, giving more j(bootstrap), 30% lower I CD H-mode edge is unstable to 10 < n < 25 kink/ballooning/peeling modes leading to ELM’s H-mode has higher ion T at plasma separatrix which aggravates divertor heating H-mode has significant j(bootstrap) at 5/2 and 3/1 surfaces leading to NTM’s H-mode has higher edge density leading to larger radiated power fraction from plasma

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Plasma edge / SOL / Divertor solution must satisfy physics and engineering constraints P(plasma) = 388 MW SOL width cm thick 8:1 OB/IB power split Q(first wall) < 0.45 MW/m2 Q(div) < 5-6 MW/m2

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Divertor heat loads are too high with no techniques for enhanced radiation For radiated power fraction of 0.3 from plasma core (bremsstrahlung + cyclotron + line from Ar) : Q(first wall) < 0.4 MW/m2, Q(OB div) < 14 MW/m2, Q(IB div) < 3.5 MW/m2

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Radiating plasma power is required to obtain workable divertor solutions Radiating mantle, Frad=0.75, Q(fw)>0.9 MW/m2, Q(OB div)>5 MW/m2, Q(IB div)>1.2 MW/m2 --> Ar (core) Radiating divertor, Frad=0.36, Frad(div)=0.43, Q(fw) Ar (divertor)

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POPCON plot for ARIES-AT P(fus) = 1760 MW P(alpha) = 351 MW P(brem) = 55 MW P(cyc) = 18 MW P(line) = 67 MW P(aux) = 37 MW H(89p) = 2.65 H(98y) = 1.35 E = 2.0 s p*/ E = 10.0 n/nGr = 1.0 Zeff = 1.85

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PF coil locations and currents are optimized with maintenance constraints

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ARIES-AT Physics Issues and New Experimental Results Neutral particle control can allow the plasma density to exceed the Greenwald limit without confinement degradation (DIII-D, TEXTOR). Helium particle control is demonstrated with pumped divertors giving p */ (DIII-D, JT- 60) Detachment of inboard strike point plasma allows high triangularity (DIII-D). LHCD is shown to stabilize neoclassical tearing modes (COMPASS, ECCD on ASDEX-U,DIII-D). Vertical and inboard pellet launch show better penetration with lower speed (ASDEX-U, DIII-D).

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ARIES-AT Physics Issues and Experimental Results RWM n=1 kink feedback control experiments continue on DIII-D (if successful n=2 will set the -limit) Controlled impurity effects on the plasma edge, SOL, and divertor continue on many tokamaks Experiments on the behavior of Internal Transport Barriers (ITB) through control of turbulence will lead to control of T and n profiles (JET, JT-60U, DIII-D, ASDEX-U)

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A High , High f BS Configuration Have Been Developed as the Physics Basis for Advanced Tokamak Fusion Power Plants High accuracy equilibria Large ideal MHD database over profiles, shape and aspect ratio RWM stable with wall/rotation or wall/feedback control NTM stable with L-mode edge and LHCD Bootstrap current consistency using advanced bootstrap models External current drive Vertically stable and controllable with modest power (reactive) Modest core radiation with radiative SOL/divertor Accessible fueling No ripple losses 0-D consistent startup Rough kinetic profile consistency with RS /ITB experiments, examining GLF23 model consistency Several assumptions based on experimental/theoretical results

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