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FIRE Physics Basis

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

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FIRE Physics Basis

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

  2. FIRE Description R = 2.14 m, a = 0.595 m, x = 2.0, x = 0.7, Pfus = 150 MW AT-Mode IP = 4.5 MA BT = 6.5 T N = 4.2  = 4.7% P = 2.35  = 0.21% q(0) ≈ 4.0 q95, qmin ≈ 4.0,2.7 li(1,3) = 0.52,0.45 Te,i(0) = 15 keV Te,i = 6.8 keV n20(0) = 4.4 n(0)/n = 1.4 p(0)/p = 2.5 n/nGr = 0.85 Zeff = 2.2 fbs = 0.78 Q = 5 burn = 32 s H-mode IP = 7.7 MA BT = 10 T N = 1.80  = 2.4% P = 0.85  = 0.075% q(0) < 1.0 q95 ≈ 3.1 li(1,3) = 0.85,0.66 Te,i(0) = 15 keV Te,i = 6.7 keV n20(0) = 5.3 n(0)/n = 1.15 p(0)/p = 2.4 n/nGr = 0.72 Zeff = 1.4 fbs = 0.2 Q = 12 burn = 20 s VV baffle divertor passive plate plasma port

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

  4. FIRE Magnets • TF Coils • Limit flattop • 20 s at BT = 10 T (H-mode) • 48 s at BT = 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 PF1,2,3 PF4 CS3 CS2 PF5 CS1

  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 < fs < 1.13 and growth time is 43 < tg(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 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,)Br/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 ICRF Port Plug RWM Coil

  7. FIRE Heating and CD 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-m2 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-m2 ECCD (??MW, 170 GHz) LFS, O-mode, fundamental NTM control @ 6.5 T 20 = 0.004 A/W-m2 (at 149 GHz) =ce=170 GHz pe=ce

  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, Qmax = P+ Paux • Volumetric heating • Nuclear heating, qmax = qpeak(Z=0) • VV, Cladding, Tiles, Magnets…. • Volumetric heating • Nuclear heating, qmax = qpeak(Z=0) • Divertor • Surface heat flux • Particle heat flux, Qmax = PSOL/Adiv(part) • Radiation heat flux, Qmax = PSOL/Adiv(rad) • Volumetric heating • Nuclear heating, qmax = qpeak(divertor) VV Clad Tile plasma

  10. Power Handling Pulse length limitations VV nuclear heating (stress limit), 4875 MW-s -----> Pfus (qVVnuclear) FW Be coating temperature, 600oC -----> QFW & Pfus (qBenuclear) TF coil heating, 373oK -----> BT & Pfus (qCunuclear) PF Coil heating-AT-mode, 373oK -----> 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/m2

  11. Power Handling/Operating Space • FIRE H-mode Operating Space • N limited by NTM or ideal MHD with NTM suppression • -----> maximum Pfus • Higher radiated power in the divertor allows more operating space, mainly at higher N • -----> maximum Pfus • 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 H98 (1.03-1.1) • Highest n/nGr (0.7-1.0) • Highest n(0)/n (1.25) H98(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 Pfus Higher radiated power in the divertor allows more operating space, mainly at higher N -----> maximum Pfus 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, BT, Q increase H98(y,2) ≤ 2.0

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

  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(qmin) ≈ 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. TSC-LSC FIRE MHD Stability (3,2) surface 12.5 MW 0.65 MA n/nGr = 0.4 Q = 6.8 • 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(=Pfus/Paux) • 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 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 Bialek, Columbia Univ. Growth Rate, /s N=4.2 N

  17. Disruption Modeling Experimental database used to project for FIRE Thermal quench time ≈ 0.2 ms Ihalo/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 Thalo and halo Axisymmetric and zero-net current structures Toroidal and poloidal currents

  18. FIRE Transport and Confinement • Energy Confinement Database • E98(y,2) = 0.144 M0.19 Ip0.93 BT0.15 R1.97 0.58 n200.41 0.78 P-0.69 (m, MA, T, MW) • p*/E = 5 • Zeff = 1.2-2.2 (fBe = 1-3%, fAr = 0-0.3%) • Pedestal Database (Sugihara, 2003) • Pped(Pa) = 1.824104M1/3Ip2R-2.1a-0.573.81(1+2)-7/3(1+)3.41nped-1/3(Ptot/PLH)0.144 • ----> Tped = 5.24 ± 1.3 keV • ----> ped?? • L-H Transition • PLH(MW) = 2.84Meff-1BT0.82nL200.58Ra0.81 (2000) ----> 26 MW in flattop • PLH(MW) = 2.58Meff-1BT0.60nL200.70R0.83 a1.04 (2002) ----> 18.5-25 MW in flattop • DN has less or equal PLH compared to favored SN (Carlstrom, DIII-D; NSTX; MAST) • H-L Transition & ELM’s • Ploss > PLH although hysterisis exists in data • Type I ELM’s typically require Ploss > 1.( )PLH, expts typically > 2PLH • Type II ELM’s require strong shaping, higher density, DN ---> reduced Pdiv, H98=1 • Type III ELM’s, near Ploss ≈ PLH, or high density, reduced H98 • Active methods ----> pellets, gas puffing, impurity seeding, ergodization

  19. Pedestal Physics and ELM’s • Type I ELM trends • Reduced WELM/Wped with increasing *ped ----> inconsistent with higher Tped for high Q • Reduced WELM/Wped with increasing ||i ----> inconsistent with higher Tped for high Q • WELM/Wped correlated with Tped/Tped as nped varied, very little change in Nped/Nped • Type II ELM’s • ASDEX-U with DN and high n ----> H98 = 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 Pin JET PELM Wth Prad

  20. T(0)/T, n(0)/n, p*/E, H98, fBe, fAr POPCON Operating Space vs. Parameters H98(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

  24. 0D Advanced Tokamak Operating Space Scan ----> q95, n(0)/n, T(0)/T, n/nGr, N, fBe, fAr Constrain ----> LH = 0.16, FW = 0.2, PLH ≤ 30 MW, P ≤ 30 MW, IFW = 0.2 MA, ILH = (1-fbs)Ip, Q Screen ----> flattop(VV, TF, FW heating), Prad(div), Ppart(div), Paux< Pmax

  25. Examples of FIRE Q=5 AT Operating Points That Obtain flat/J > 3 HH < 1.75, satisfy all power constraints, Pdiv(rad) < 0.5 P(SOL)

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

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

  28. 1.5D Integrated Scenarios AT-mode Q = 5 I(bs) = 3.5 MA, I(LH) = 0.80 MA I(FW) = 0.20 MA, t(flattop)/j=3.2 n/nGr = 0.85 n(0)/<n> = 1.4 n(0) = 4.4x10^20 Wth = 34.5 MJ tE = 0.7 s H98(y,2) = 1.7 Ti(0) = 14 keV Te(0) = 16 keV Dy(total) = 19 V-s, Pa = 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

  29. Perturbation of AT-mode Current Profile 5 MW perturbation to PLH Flattop time is sufficient to examine CD control t = 12 s t = 25 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 IBS/IP (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

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