1. FIRE Physics Basis C. Kessel for the FIRE Team Princeton Plasma Physics Laboratory FIRE Physics Validation Review March 30-31, 2004 Germantown, MD AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc FIRE Collaboration http://fire.pppl.gov

2. FIRE Description H-mode I P = 7.7 MA B T = 10 T  N = 1.80  = 2.4%  P = 0.85    = 0.075% q(0) < 1.0 q 95 ≈ 3.1 li(1,3) = 0.85,0.66 T e,i (0) = 15 keV  T e,i  = 6.7 keV n 20 (0) = 5.3 n(0)/  n  = 1.15 p(0)/  p  = 2.4 n/n Gr = 0.72 Z eff = 1.4 f bs = 0.2 Q = 12  burn = 20 s R = 2.14 m, a = 0.595 m,  x = 2.0,  x = 0.7, P fus = 150 MW AT-Mode I P = 4.5 MA B T = 6.5 T  N = 4.2  = 4.7%  P = 2.35    = 0.21% q(0) ≈ 4.0 q 95, q min ≈ 4.0,2.7 li(1,3) = 0.52,0.45 T e,i (0) = 15 keV  T e,i  = 6.8 keV n 20 (0) = 4.4 n(0)/  n  = 1.4 p(0)/  p  = 2.5 n/n Gr = 0.85 Z eff = 2.2 f bs = 0.78 Q = 5  burn = 32 s port plasma divertor baffle passive plate VV

3. FIRE Magnet Layout TF Coil CS1 CS2 CS3 PF1,2,3 PF4 PF5 Error field correction coils Fast vertical and radial position control coil RWM feedback coil Fe shims

4. FIRE Magnets CS1 CS2 CS3 PF1,2,3PF4 PF5 TF Coils Limit flattop 20 s at B T = 10 T (H-mode) 48 s at B T = 6.5 T (AT-mode) TF ripple (max) = 0.3% 0.3%  loss H-mode 8%  loss AT-mode (Fe shims) PF Coils Provide H-mode operation 0.55 ≤ li(3) ≤ 0.85 (SOB,EOB) 0.85 ≤ li(3) ≤ 1.15 (SOH,EOH)  ref -5 ≤  (Wb) ≤  ref +5 1.5 ≤  N ≤ 3.0  ramp = 40 V-s,  flat = 3 V-s Provide AT-mode operation 0.35 ≤ li(3) ≤ 0.65 (SOB & EOB) 2.5 ≤  N ≤ 5.0 7.5 ≤  flattop (Wb) ≤ 17.5 Ip ≤ 5.0 MA  ramp = 20 V-s

5. FIRE Magnets Vertical stability Cu passive plates, 2.5 cm thick For most unstable plasmas (full elongation and low pressure), over the range 0.7 < li(3) < 1.1, the stability factor is 1.3 < f s < 1.13 and growth time is 43 <  g (ms) < 19 Internal Control Coils Fast vertical position control Fast radial position control (antenna) Startup assist Error Correction Coils Static to slow response Correct PF and TF coil, lead, etc. misalignments ITER Error Coils

6. FIRE Magnets ICRF Port Plug RWM Coil Resistive Wall Mode Coils DIII-D Modes are detectable at the level of 1G C-coils produce about 50 times this field The necessary frequency depends on the wall time for the n=1 mode (which is 5 ms in DIII-D) and they have  wall ≈ 3 FIRE FIRE has approximately 3-4 times the plasma current, so we might be able to measure down to 3-4 G If we try to guarantee at least 20 times this value from the feedback coils, we must produce 60-80 G at the plasma These fields require approximately I = f(d,Z,  )B r /  o = 5-6.5 kA Assume we also require  wall ≈ 3 Required voltage would go as V ≈ 3  o (2d+2Z)NI/  wall ≈ 0.25 V/turn

7. FIRE Heating and CD  =  ce =170 GHz  pe =  ce ICRF (20+ MW, 70-115 MHz) Ion heating @ 10 T He3 minority and 2T at 100 MHz Ion heating @ 6.5 T H minority and 2D at 100 MHz Electron heating/CD @ 6.5 T 70-75 MHz,  20 = 0.14-0.21 A/W-m 2 LHCD (30MW, 5 GHz) n || ≈ 1.8-2.5,  n || = 0.3 NTM control @ 10 T Bulk CD/NTM @ 6.5 T  20 = 0.16 A/W-m 2 ECCD (??MW, 170 GHz) LFS, O-mode, fundamental NTM control @ 6.5 T  20 = 0.004 A/W-m 2 (at 149 GHz)

8. ICRF Heating and CD Want to reduce power required to drive on- axis current 2 strap antenna and port geometry provides only 40% of ICRF power in good CD part of the spectrum 4 strap antenna can provide 60% of power in good CD part of spectrum Expanding antenna cross-section and going to 4 straps reaches 80% in good CD part of spectrum

9. Power Handling First wall Surface heat flux Plasma radiation, Q max = P  + P aux Volumetric heating Nuclear heating, q max = q peak (Z=0) VV, Cladding, Tiles, Magnets…. Volumetric heating Nuclear heating, q max = q peak (Z=0) Divertor Surface heat flux Particle heat flux, Q max = P SOL /A div (part) Radiation heat flux, Q max = P SOL /A div (rad) Volumetric heating Nuclear heating, q max = q peak (divertor) plasma VVClad Tile

10. Power Handling Pulse length limitations VV nuclear heating (stress limit), 4875 MW-s -----> P fus (q VV nuclear ) FW Be coating temperature, 600 o C - ----> Q FW & P fus (q Be nuclear ) TF coil heating, 373 o K -----> B T & P fus (q Cu nuclear ) PF Coil heating-AT-mode, 373 o K --- --> Ip, li,  p, and  (not limiting) Component limitations Particle power to outboard divertor < 28 MW Radiated power on (inner&outer) divertor/baffle < 6-8 MW/m 2

11. Power Handling/Operating Space FIRE H-mode Operating Space  N limited by NTM or ideal MHD with NTM suppression -----> maximum P fus Higher radiated power in the divertor allows more operating space, mainly at higher  N -----> maximum P fus Majority of operating space limited by TF coil flattop ----->  flattop ≤ 20 s High Q (≈15-30) operation obtained with Low impurity content (1-2% Be) Highest H 98 (1.03-1.1) Highest n/n Gr (0.7-1.0) Highest n(0)/  n  (1.25) H 98(y,2) ≤ 1.1

12. Power Handling/Operating Space FIRE AT-mode Operating Space  N is limited by ideal MHD w/wo RWM feedback -----> maximum P fus Higher radiated power in the divertor allows more operating space, mainly at higher  N -----> maximum P fus Majority of operating space limited by VV nuclear heating ----->  flattop ≤ 20-50 s Design solutions to improve VV nuclear heating limit, could reach PF coil limit, function of Ip Number of current diffusion times accessible is reduced as  N, B T, Q increase H 98(y,2) ≤ 2.0

13. FIRE Particle Handling V HFS = 125 m/s Parks, 2003 Cryopumping in slanted ports Midplane pumping for pumpdown & bakeout HFS (v max = 125 m/s, determined by ORNL), LFS, VL Parks HFS modeling, deposition to axis WHIST analysis indicates n(0)/  n  ≤ 1.25

14. MHD Stability H-mode Sawtooth ---> unstable, weak impact on burn, coupling to global modes? NTM’s ---> unstable or stable?, LHCD  ’ stabilization, reduce  N if near threshold, experiments with little or no NTM impact (DIII-D, JET, ASDEX-U) Ideal MHD ---> over range of profiles  N (n=1 or ∞) ≈ 3 AT-mode NTM’s ---> unstable or stable? q(  ) > 2 everywhere, r/a(q min ) ≈ 0.8, ECCD/OKCD, LHCD multiple spectra Ideal MHD ---> no wall/feedback,  N (n=1) ≈ 2.5-2.8 ---> with wall/feedback, VALEN analysis indicates 80-90% of with- wall  N -limit (5-6), however, n=2,3 have lower  N -limits? Other MHD issues Ballooning/peeling modes, unstable with H-mode edge Alfven and energetic particle modes, H-mode stable (unless higher  N ), AT not analized No external rotation source

15. FIRE MHD Stability Neoclassical Tearing Modes H-mode Threshold for NTM’s is uncertain Sawteeth and ELM’s are expected to be present and can drive NTM’s Typical operating point is at low  N and  P Can lower  N further if near threshold Lower Hybrid CD at the rational surfaces Compass-D demonstrated LH stabilization Analysis by Pletzer and Perkins showed stabilization was feasible (PEST3) Lowers Q(=P fus /P aux ) EC methods require high frequencies at FIRE field and densities ----> 280 GHz DIII-D (Luce)  N ≤ 3, NTM weak impact ASDEX-U, JET (Gunter) frequently interrupted NTM confinement degradation JET (3,2) surface 12.5 MW 0.65 MA n/n Gr = 0.4 Q = 6.8 TSC-LSC normal FIR-NTM Weak NTM, FIR-NTM

16. MHD Stability RWM Stabilization AT-mode RWM stabilization with feedback coils, VALEN analysis indicates 80-90% of ideal with wall limit for n=1 Coils in every other port, very close to plasma n = 1 stable with wall/feedback to  N ’s around 5.0-6.0 n = 2 and 3 appear to have lower  N limits in presence of wall, possibly blocking access to n = 1 limits H-mode edge stability will depend on pedestal parameters; width, height, and location Growth Rate, /s NN  N =4.2 Bialek, Columbia Univ.

17. Disruption Modeling Experimental database used to project for FIRE Thermal quench time ≈ 0.2 ms I halo /Ip  TPF ≤ 0.5 dIp/dt rates for current quench ≤ 3 MA/ms (worst), and 1 MA/ms (typical) TSC used to provide plasma evolution Hyper-resistivity for rapid j redistribution T halo and  halo Axisymmetric and zero-net current structures Toroidal and poloidal currents

18. FIRE Transport and Confinement Energy Confinement Database  E 98(y,2) = 0.144 M 0.19 Ip 0.93 B T 0.15 R 1.97  0.58 n 20 0.41  0.78 P -0.69 (m, MA, T, MW)  p * /  E = 5 Z eff = 1.2-2.2 (f Be = 1-3%, f Ar = 0-0.3%) Pedestal Database (Sugihara, 2003) P ped (Pa) = 1.824  10 4 M 1/3 Ip 2 R -2.1 a -0.57  3.81 (1+  2 ) -7/3 (1+  ) 3.41 n ped -1/3 (P tot /P LH ) 0.144 ----> T ped = 5.24 ± 1.3 keV ---->  ped ?? L-H Transition P LH (MW) = 2.84M eff -1 B T 0.82 n L20 0.58 Ra 0.81 (2000) ----> 26 MW in flattop P LH (MW) = 2.58M eff -1 B T 0.60 n L20 0.70 R 0.83 a 1.04 (2002) ----> 18.5-25 MW in flattop DN has less or equal P LH compared to favored SN (Carlstrom, DIII-D; NSTX; MAST) H-L Transition & ELM’s P loss > P LH although hysterisis exists in data Type I ELM’s typically require P loss > 1.( )  P LH, expts typically > 2  P LH Type II ELM’s require strong shaping, higher density, DN ---> reduced P div, H 98 =1 Type III ELM’s, near P loss ≈ P LH, or high density, reduced H 98 Active methods ----> pellets, gas puffing, impurity seeding, ergodization

19. Pedestal Physics and ELM’s Type I ELM trends Reduced  W ELM /W ped with increasing  * ped ----> inconsistent with higher T ped for high Q Reduced  W ELM /W ped with increasing  || i ----> inconsistent with higher T ped for high Q  W ELM /W ped correlated with  T ped /T ped as n ped varied, very little change in  N ped /N ped Type II ELM’s ASDEX-U with DN and high n ----> H 98 = 1- 1.2 and reduction in divertor heat flux by 3  JET with high  and high n ----> mixed Type I+II, no reduction in confinement and 3  reduction in ELM power loss P in W th P rad P ELM JET

20. POPCON Operating Space vs. Parameters T(0)/  T , n(0)/  n ,  p * /  E, H 98, f Be, f Ar H 98(y,2) must be ≥ 1.1 for robust operating space

21. 1.5D Integrated Simulations H-mode Tokamak Simulation Code (TSC) Free-boundary Energy and current transport Density profiles assumed GLF23 & MMM core energy transport Assumed pedestal height/location ICRF heating, data from SPRUCE Bootstrap current, Sauter single ion Porcelli sawtooth model Coronal equilibrium radiation Impurities with electron density profile PF coils and conducting structures Feedback systems on position, shape, current Use stored energy control Snowmass E2 simulations for FIRE Corsica, GTWHIST, Baldur, XPTOR

22. 1.5D Integrated Simulations H-mode FIRE H-mode, GLF23

23. 1.5D Integrated Simulations H-mode FIREQPaux(MW)Tped(keV) TSC GLF 10.313.54.5 10.07.53.8 10.0 4.1 10.012.54.4 10.015.04.7 10.020.05.4 Baldur MMM 4.530.02.5* 7.010.02.5* XPTOR/12 GLF 5.020.03.0 10.020.04.0 15.020.05.0 Corsica GLF 4.012.52.5 6.012.54.0 10.012.55.0

24. 0D Advanced Tokamak Operating Space Scan ----> q 95, n(0)/  n , T(0)/  T , n/n Gr,  N, f Be, f Ar Constrain ---->  LH = 0.16,  FW = 0.2, P LH ≤ 30 MW, P ≤ 30 MW, I FW = 0.2 MA, I LH = (1-f bs )Ip, Q Screen ---->  flattop (VV, TF, FW heating), P rad (div), P part (div), P aux < P max

25. Examples of FIRE Q=5 AT Operating Points That Obtain  flat /  J > 3 nn nnTT TT BTBT q 95 IpH f Gr f BS P cd PP z eff f Be f Ar t/  0.52.601.58.176.54.25 1.710.80.8027.527.82.081%.3%3.58 0.52.932.07.286.54.25 1.570.90.8030.931.41.771%.2%3.95 0.753.101.57.836.53.754.821.460.90.8033.136.51.892%.2%3.07 0.752.911.07.716.54.004.521.620.90.8524.728.61.771%.2%3.52 0.753.231.57.006.54.004.521.541.00.8527.532.02.081%.3%4.40 0.752.441.58.906.54.25 1.740.80.9116.028.02.202%.3%3.65 1.003.491.07.356.53.505.161.361.00.8332.638.61.771%.2%3.00 1.003.261.07.606.53.754.821.541.00.8923.930.12.013%.2%4.00 1.002.441.59.596.54.004.521.650.80.9513.631.52.323%.3%3.29 HH < 1.75, satisfy all power constraints, Pdiv(rad) < 0.5 P(SOL)

26. 1.5D Integrated Simulations AT-mode Ip=4.5 MA Bt=6.5 T  N =4.1 t(flat)/  j=3.2 I(LH)=0.80 P(LH)=25 MW f BS =0.77 Z eff =2.3 q(0) =4.0 q(min) = 2.75 q(95) = 4.0 li = 0.42,  = 4.7%,  P = 2.35

27. 1.5D Integrated Scenarios AT-mode t = 12-41 s

28. 1.5D Integrated Scenarios AT-mode n/nGr = 0.85 n(0)/ = 1.4 n(0) = 4.4x10^20 Wth = 34.5 MJ  E = 0.7 s H98(y,2) = 1.7 Ti(0) = 14 keV Te(0) = 16 keV  (total) = 19 V-s, P  = 30 MW P(LH) = 25 MW P(ICRF/FW) = 7 MW (up to 20 MW ICRF used in rampup) P(rad) = 15 MW Zeff = 2.3 Q = 5 I(bs) = 3.5 MA, I(LH) = 0.80 MA I(FW) = 0.20 MA, t(flattop)/  j=3.2

29. Perturbation of AT-mode Current Profile 5 MW perturbation to P LH Flattop time is sufficient to examine CD control t = 12 s t = 25 s t = 41 s

30. Conclusions The FIRE device design provides sufficient/flexible/relevant operating space to examine burning plasma physics –Sufficient to provide burning conditions (Q ≥ 10 inductive and Q ≥ 5 AT, does not preclude ignition) –Flexible to accommodate uncertainty and explore various physics regimes –Relevant to power plant plasma physics and engineering design The subsystems on FIRE, within their operating limits, are suitable to examine burning plasma physics ----> subject to R&D in some cases –Auxiliary heating/CD –Particle fueling and pumping –Divertor/baffle and FW PFC’s –Magnets –Diagnostics

31. Conclusions Burning plasma conditions can be accessed and studied in both standard H-mode and Advanced Tokamak modes. The range of AT performance has been expanded significantly since Snowmass –FIRE can reach 1-5  j, and examine current profile control –Design improvement to FW tiles could extend flattop times further –FIRE can reach 80-90% of ideal with wall limit, with RWM feedback –FIRE can reach high I BS /I P (77% in 1.5D simulation) –Identified that radiative mantle/divertor solutions significantly expand operating space –FIRE will pursue Fe shims for AT operation The physics basis for FIRE’s operation is based on current experimental and theoretical results, and projections based on these continue to provide confidence that FIRE will achieve the required burning plasma performance

32. Issues/Further Work Magnets –Ripple reduction, design Fe shims for AT mode –Continue equilibrium analysis –Complete plasma breakdown and early startup –Complete internal control coil analysis –RWM coil design/integration into port plugs, time dependent analysis –Error field control coil design Heating and CD –Continue ICRF antenna design, disruption loads, neutron/surface heating –Engineering of 4 strap expanded antenna option –More detailed design of LH launcher, disruption loads, neutron/surface heating –Complete 2D FP/expanded LH calculations for FIRE specific cases –Continue examination of EC/OKCD for NTM suppression in AT mode –Pursue dynamic simulations/PEST3 analysis of LH NTM stabilization for both H- mode and AT-mode

33. Issues/Further Work Power Handling –Pulse length limitations from VV nuclear heating, design improvements –FW tile design, material choices, impacts on magnetics –Continue divertor analysis, UEDGE and neutrals analysis for integrated heat load, pumping,and core He concentration solutions –Continue examination of ITPA ELM results and projections, encourage DN strong triangularity experiments –DN up-down imbalance, implications for divertor design (lots of work on DII-D) –Disruption mitigation strategies, experiments Particle Handling –Continue pellet and gas fueling analysis in high density regime of FIRE –Neutrals analysis for pumping –Be behavior as FW material and intrinsic impurity –Impurity injection, core behavior, and controllability –Particle control techniques: puff and pump, density feedback control, auxiliary heating to pump out core, etc. –Wall behavior, no inner divertor pumping, what are impacts?

34. Issues/Further Work MHD Stability –LH stabilization of NTM’s, analysis and experiments (JET, JT-60U and C-Mod) –Examine plasmas that appear not to be affected by NTM’s (current profile) –Early (before they are saturated) stabilization of NTM’s with EC/OKCD –Continue to develop RWM feedback scheme in absense of rotation –Identify impact of n=2,3 modes in wall/feedback stabilized plasmas –Examine impact of no external rotation source on transport, resistive and ideal modes –Alfven eigenmodes/energetic particle modes, onset and accessibility in FIRE Plasma Transport and Confinement –Continue core turbulence development for H-mode, ITPA –Establish AT mode transport features, ITB onset, ITPA –Pedestal physics and projections, and ELM regimes, ITPA –Impact of DN and strong shaping on operating regimes, Type II ELMs –Improvements to global energy confinement scaling, single device trends –Expand integrated modeling of burning plasmas