1 / 31

ARIES- A dvanced T okamak Power Plant Study Physics Analysis and Issues

This study examines the physics of the ARIES-AT tokamak power plant, with objectives to increase plasma elongation and triangularity, eliminate HHFW current drive, and address critical physics issues such as kink stabilization, RWM analysis, NTM assessment, and edge plasma modeling.

rkoons
Download Presentation

ARIES- A dvanced T okamak Power Plant Study Physics Analysis and Issues

An Image/Link below is provided (as is) to download presentation Download Policy: Content on the Website is provided to you AS IS for your information and personal use and may not be sold / licensed / shared on other websites without getting consent from its author. Content is provided to you AS IS for your information and personal use only. Download presentation by click this link. While downloading, if for some reason you are not able to download a presentation, the publisher may have deleted the file from their server. During download, if you can't get a presentation, the file might be deleted by the publisher.

E N D

Presentation Transcript

  1. ARIES-Advanced TokamakPower Plant StudyPhysics 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

  2. 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 ARIES-AT Physics Team

  3. Objectives for ARIES-AT Physics • Extend the physics basis for ARIES-RS to increase b and decrease PCD • 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

  4. 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

  5. ARIES-RS and AT Parameters

  6. ARIES-AT Equilibrium

  7. 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

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

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

  10. Increasing the plasma triangularity leads to higher b Higher d increases the bN for ballooning modes, and increases the plasma current for fixed q95, while higher k enhances the effect b = bN (Ip/aB)

  11. Increasing the plasma elongation leads to higher b Higher k leads to higher bN for ballooning modes when k is less than 2.2-2.5 so long at d is high enough, however b continues to rise due to the strong increase in Ip

  12. The plasma elongation is limited by the vertical instability Vertical stability analysis showed that we could have kX up to 2.2-2.3 with tungsten conductors 0.3-0.35a from the plasma boundary which is inside the blanket

  13. 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

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

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

  16. 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

  17. 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 ARIES-AT uses feedback to stabilize the external kink modes

  18. External current drive is required on axis and near the plasma edge ICRF/FW on-axis CD; 96 MHz, N||=2, PCD=5 MW, h=0.032 A/W LHCD off-axis CD; 3.6 GHz , N||=1.65-3.5, PCD=24.5 MW, h=0.024-0.053 A/W; 2.5 GHz, N||=5.0, PCD=12.5 MW, h=0.013 A/W

  19. 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

  20. Alternate CD sources are examined for plasma rotation and current profile control 120 keV NBI provides plasma rotation and current for r>0.6, P(NBI)=44 MW for 1.15 MA HHFW at 20wci provides current at r=0.7-0.9 which is useful for control of qmin and r/a(qmin), P(HHFW)=20 MW for 0.4 MA

  21. Neoclassical tearing modes must be stabilized to reach ideal MHD b 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

  22. 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

  23. Comparison of L-mode and H-mode plasma edges H-mode edge has higher T, giving more j(bootstrap), 30% lower ICD 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

  24. Plasma edge / SOL / Divertor solution must satisfy physics and engineering constraints • P(plasma) = 388 MW • SOL width 0.8-2.1 cm thick • 8:1 OB/IB power split • Q(first wall) < 0.45 MW/m2 • Q(div) < 5-6 MW/m2

  25. 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 0.0018 Ar) : Q(first wall) < 0.4 MW/m2, Q(OB div) < 14 MW/m2, Q(IB div) < 3.5 MW/m2

  26. 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 --> 0.0035 Ar (core) Radiating divertor, Frad=0.36, Frad(div)=0.43, Q(fw)<0.45 MW/m2, Q(OB div)<6 MW/m2, Q(IB div)<1.5 MW/m2--> 0.0026 Ar (divertor)

  27. 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 tE = 2.0 s tp*/tE = 10.0 n/nGr = 1.0 Zeff = 1.85

  28. PF coil locations and currents are optimized with maintenance constraints

  29. 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 tp*/tE = 3-15 (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).

  30. 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 b-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)

  31. 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 A High b, High fBS Configuration Have Been Developed as the PhysicsBasis for Advanced Tokamak Fusion PowerPlants

More Related