FIRE Physics Basis (detailed version) 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

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 = 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

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

Toroidal Field Coils 16 TF coils BeCu inboard legs OFHC Cu outboard legs Coil stress and heating limit TF pulse length (factor of ≥ 1.18 over  allowable) H-mode B T = 10 T and P fus = 150 MW ----> 20 s flattop AT-mode B T = 6.5 T and P fus = 150 MW ----> 48 s flattop Maximum TF ripple at R+a is 0.3% 0.3%  -particle loss H-mode 8%  -particle loss AT-mode Expect to use Fe shims for AT-mode

Poloidal Field Coils Center Stack 1, 2U&L, 3U&L PF1,2,3,4,5 U&L All CS and PF coils are CuCrZr H-mode Fiducial equilibria in discharge IM, SOD, SOH, SOB, EOB, EOH, EOD Flexibility of PF coils 0.55 ≤ li(3) ≤ 0.85 (SOB,EOB) 0.85 ≤ li(3) ≤ 1.15 (SOH,EOH)  ref -5 ≤  (Wb) ≤  ref ≤  N ≤ 3.0 Full operating space available within stress (1.3 margin) and heating allowables, except at EOB, li=0.85 where  ≤  ref -2 Rampup consumes ≈ 40 V-s Flattop consumes ≈ 3 V-s CS1 CS2 CS3 PF1,2,3PF4 PF5

Poloidal Field Coils AT-mode Fiducial equilibria in discharge IM, SOD, SOH, SOF, SOB, EOB, EOH, EOD Flexibility of PF coils 0.35 ≤ li(3) ≤ 0.65 (SOB & EOB) 2.5 ≤  N ≤ ≤  flattop (Wb) ≤ 17.5 Full operating space is available for Ip ≤ 5.0 MA PF coils can provide pulse length limitation -----> 41 s for access to op. space at Ip = 4.5 MA, and scales with Ip, li,  p, and  Inductive + non-inductive rampup consumes V-s, final state can be optimized Plasma current 100% non-inductive in flattop shape control feedback points

Poloidal Field Coils TSC simulations Free-boundary calculations with heating, CD, bootstrap current, energy and current transport, impurities, PF coils, structure and feedback systems, etc. ---> check of equilibrium coil currents ---> Volt-second consumption ---> Feedback control of vertical position, radial position, plasma current and shape

Vertical Stability Design passive structures to slow vertical instability for feedback control and provide a stability factor f s > 1.2 Passive stabilizers are 2.5 cm thick Cu, toroidally continuous on upper outboard and inboard sides 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 Passive stabilizers Cladding (ports provide poloidal cuts) Cladding (large number of poloidal cuts)

Internal Control Coils 8 OFHC Cu coils (2nd redundant coils) above and below the midplane Fast vertical position feedback control  Z RMS = 1 cm, kA-turn, V/turn > 7-14 MVA (peak) Fast radial position feedback control (antenna coupling) Analysis not completed, assuming I and V similar to vertical control Fast radial feedback is coupled with slower outer PF shape control These coils also used in startup to tailor field null

Resistive Wall Mode (RWM) Coils ICRF Port Plug RWM Coil DIII-D experience Modes are detectable at the level of 1G The C-coils can 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 projection FIRE has approximately 3-4 times the DIII-D 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 G at the plasma These fields require approximately I = f(d,Z,  )B r /  o = kA Assume we also require  wall ≈ 3 Required voltage would go as V ≈ 3  o (2d+2Z)NI/  wall ≈ 0.25 V/turn

Error Field Correction Coils Static or slow dynamic Cu coils Located outside TF and PF coils Compensating TF and PF coil/lead/etc. misalignments and other under field conditions These coils are NOT used for RWM feedback Extrapolated threshold to induce locked modes ≈ 1  T (very uncertain!) Correction coils should be capable of reducing (m=1,n=1), (2,1), and (3,1) error fields, simultaneously And provide factor of ≈ 5 reduction in net error field B r2,1 net 3 distributed coils provides poloidal mode control allowing multiple (m,n=1) suppression Recent C-Mod data shows that applied B r2,1 of 6  T removed mode-locking -----> Important since C-Mod does NOT have external rotation source No analysis performed ITER Error Coils

ICRF Heating and CD Frequency range MHz 2 strap antennas 4 ports, total power 20 MW H-mode B T = 10 T, minority He3 and 2T at 100 MHz Frequency range allows heating at a/2 on HFS and LFS AT-mode B T = 6.5 T, ion heating at minority H and 2D at 100 MHz Frequency range allows ion heating at a/2 on HFS and LFS Electron heating/CD at MHz CD efficiency  20 = A/W-m 2 SPRUCE analysis n He3 /n e =2% P ICRF =11.5 MW  =100 MHz T He3 (0)=10.2 keV P He3 =60% P T =10% P D =2% P e =26%

ICRF Heating and CD B T = 10 TB T = 6.5 T Vacuum Toroidal Field Resonances Be

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

ICRF Heating and CD AORSA full wave analysis continues including fast alpha and Be impurity effects 75 MHz 70 MHz P e =0.44 P T =0.15 P Be =0.30 P e =0.65 P T =0.32 P Be =0.0  20 =0.14  20 =0.17

Lower Hybrid Current Drive Frequency 5 GHz Spectrum n || ≈ ,  n || = 0.3 Power of 30 MW, in 2 ports Upgrade to baseline design H-mode Used for NTM control for B T = 10 T Used for non-inductive CD for hybrid discharges AT-mode Used for bulk CD for B T = 6.5 T CD efficiency  20 ≈ 0.16 A/W-m 2 at 6.5 T (30-50% higher from 2D FP calcs.) Used for NTM control f > 2  f LH RF power flux is 53 MW/m 2 Need 0.57 m 2 per waveguide for 30 MW Each waveguide is 5.7 cm(tall)  0.65 cm Have ≈ 1500 waveguides

Lower Hybrid Current Drive Trapped electron effects reduce CD efficiency Reverse power/current reduces forward CD Less than 1.0 MW is absorbed by alphas Recent modeling with CQL and ACCOME/LH19 improves CD efficiency, 30-50% increase, but right now…….. B T = 8.5T ----> 0.25 A/W-m2 B T = 6.5T ----> 0.16 A/W-m2 Benchmarks with ACCOME, CURRAY and LSC 3.7 GHz, 750 kW, 1000s sources available ITER estimate for 5 GHz, 1.0 MW, CW sources was 1.15 euro/watt ACCOME TSC-LSC

Electron Cyclotron  =  ce =170 GHz  pe =  ce Rays are bent as they approach  =  pe Rays are launched with toroidal directionality for CD Frequency of 170 GHz to utilize ITER R&D LFS, O-mode, fundamental FIRE has high density and high field Cutoff of EC when  =  pe AT-mode Lower B T = 6.5 T LFS deposition implies trapping reduction of CD, however, Ohkawa effect provides more CD than standard EC Current required, scaled as Ip  N 2 from DIII-D and ASDEX-U expts for (3,2) mode ----> drive 200 kA to suppress from saturated state ----> requires 100 MW!

Electron Cyclotron Bt=6.5 T Bt=7.5 T Bt=8.5 T RoRo+a fce=182 fce=142 fce=210 fce=164 fce=190fce= GHz 200 GHz q min (3,1) J. Decker, MIT 145≤  ≤155 GHz -30 o ≤  L ≤-10 o midplane launch 10 kA of current for 5 MW of injected power r/a(q min ) ≈ 0.8 r/a(3,1) ≈ Does (3,1) require less current than (3,2)? Local *,  *, Re m effects so close to plasma edge? 170 GHz may be adequate, but 200 GHz is better fit for FIRE parameters

Neutral Beam Injection (Difficult) R tan = 0 m R tan = 0.75 m R tan = 1.7 m R tan = 0 m 16 TF Coils Need 1 MeV to get 50% of power inside a/2 R tan = 0.75 m and higher Must go to 12 TF coils, pinwheel ports, and Fe inserts Need > 1 MeV to get 50% of power inside a/2 Plasma rotation for R tan = 1.7 m Assumed 120 keV & 8 MW ----> deposited r/a > 0.65 Dominated by j  B rotation giving v/v Alfven ≤ 0.5%

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 VVCladTile

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

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 ( ) Highest n/n Gr ( ) Highest n(0)/  n  (1.25) H 98(y,2) ≤ 1.1

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 ≤ 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

Particle Fueling/Pumping Gas Fueling SystemPellet Fueling SystemRemarks Design fueling rate200 torr-l/s for 20 s Pumping cap. 200 torr-l/s Operational fuel rate torr-l/s torr-l/sIsotopic fueling Fuel isotopeD(95-99%), T,H(5-1%)T(40-99%), D(60-1%)D-rich edge, T-rich core Impurity fuel rate25 torr-l/sPrefer gas Impurity speciesHe3, Ne, Ar, others Rapid shutdownMassive gas puffKiller pellet or liquid jetDisruption/VDE Pellet sizes3-4 mm Require ≈ 1-2  tritons/s for FIRE H-mode ---> g T injected per shot (20 s) ---> 5% of injected tritium consumed HFS launch, limited to 125 m/s (test actually performed at ORNL to find pellet speed limit) LFS and VL can reach much higher velocities VL is at major radius, therefore not expected to provide improvement over LFS

Particle Fueling/Pumping 16 cryocondensation/diffusion pumps, 8 above and 8 below midplane, every other port Backed by turbo/drag pumps H 2 O pumped on 1 m long 30 o K entrance duct H and impurities pumped by cryocondensation, liquid He He pumped by turbo/drag pump located outside bio-shield, viscous drag compression (200 l/s conductance) Cooling requirement for 16 cryopumps at 200 torr-l/s and nuclear heating (0.03 W/cm 3 ) is 48 W, and liquid He flow rate is 64 l/hour for all 16 pumps Regeneration is done into the turbo/drag pumps Pumpdown and vessel bakeout utilize midplane pump, to provide minimum of 2000 l/s to reach or less base pressure

Particle Fueling/Pumping V = 125 m/s Parks, 2003 WHIST simulation of FIRE H- mode discharge (Houlberg) Assume uniform pellet deposition Obtains some density peaking with sufficient pumping

MHD Stability Sawtooth H-mode Unstable to internal kink, r/a(q=1) ≈ 0.35 m -----> coupling to other global modes? Porcelli sawtooth model (  W MHD +  W KO +  W fast ), incorporated into TSC indicates effect on fusion performance is weak Pedestal/bootstrap broadens j profile Rapid reheat of sawtooth volume  -particles providing stabilization Complete stabilization would require RFCD since FIRE does not have high energy minority species The q=1 surface can be removed from the plasma by 1.2 MA off-axis CD Reduction of Ip to 6.0 MA

MHD Stability Neoclassical Tearing Modes H-mode Stable or unstable? 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 TSC-LSC (3,2) surface 12.5 MW 0.65 MA n/n Gr = 0.4 Q = 6.8

MHD Stability FIR-NTMs ususal NTMs S. Günter et al., PRL ASDEX-U Current profile modification

FIRE MHD Stability Despite (3,2) NTM excellent confinement: H 98y =1.4,  N = 3.3 (LHCD ctr-CD in start-up phase) JET Current profile modification

MHD Stability Ideal MHD Stability H-mode n=1 external kink and n=∞ ballooning modes H-mode Stable without a wall/feedback Under various profile conditions  N ≤ 3 ballooning unstable in pedestal region depending on pedestal width and magnitude Intermediate n peeling/ballooning modes H- mode Unstable, primary candidate for ELM’s Type I ELM’s are divertor lifetime limiting, must access Type II, III P loss /P LH ≈ in flattop, not > 2 like many present experiments FIRE has high triangularity (  x = 0.7) in Double Null and high density Active methods to reduce  W ELM include pellets, impurities, ergodization,… Self-consistent ohmic/bootstrap equilibria

MHD Stability Neoclassical Tearing Modes AT-mode Unstable or Stable? q(  ) > 2 everywhere, so rational surfaces are (3,1), (5,2), (7,3), (7,2)… r/a(q min ) ≈ 0.8 r/a(3,1) ≈ Local *,  *, Re m effects so close to plasma edge? L-mode or H-mode conditions Examining EC stabilization at 170 GHz LFS absorption, Ohkawa CD dominates Scaling from (3,2) expts indicates high power ----> early detection required LH using two spectra, one for bulk CD and other for NTM suppression

MHD Stability Ideal MHD Stability AT-mode n= 1, 2, and 3…external kink and n = ∞ ballooning modes n = 1 stable without a wall/feedback for  N < n = 2 and 3 have higher limits without a wall/feedback Ballooning stable up to  N < 6.0, unstable in pedestal region of H-mode edge plasmas. RWM stabilization with feedback coils, VALEN analysis indicates 80-90% of ideal with wall limit for n=1 n = 1 stable with wall/feedback to  N ’s around n = 2 and 3 appear to have lower  N limits in presence of wall, possibly blocking access to n = 1 limits Intermediate n peeling/ballooning modes –Unstable under H-mode edge conditions Growth Rate, /s NN  N =4.2 Bialek, Columbia Univ.

MHD Stability JSOLVER Equilibria TSC-LSC Simulation Equilibria

MHD Stability Other MHD Issues H-mode and AT-mode Alfven eigenmodes and energetic particle modes Snowmass assessment indicated stable for H-mode, although access to shorter pulse high P fus plasmas should destabilize AT-mode not analyzed Error fields from coil misalignments, etc. ----> install Cu window coils outside TF coil, stationary to slow response FIRE does not have an external source of rotation Transport, sheared rotation Resistive instabilities, sheared to bulk rotation RWM, bulk rotation Plasma self-rotation (C-Mod), is it sufficient for some stabilization

Disruption Modeling Comments frequency10-30%30% plasma development, 10% for repetitive operation number300 (900)300 at maximum W mag and W th thermal energy35 MJ themal quench duration0.2 ms ( )single or multi-step quench fraction of W th to divertor80-100% (or ver low for mitigation)conduction to targets, 2-1 asymmetry fraction of W th to first wall/baffle< 30% (or nearly 100% for mitigation)radiation inside to outside divertor split5 to 1need data on DN poloidal localization3x typical SOL width(1x to 10x)this may not be reliable magnetic energy25 MJ current quench6 ms (2-600 ms) Ip decay rate1 MA/ms typical, 3 MA/ms maximum fraction of W mag to FW by radiation80-100%peaking factor of 2 fraction of W mag to FW by conduction0-20% VDE frequency10% of disruptionsuncertain due to up-down symmetry halo current fraction, I halo /I p 0.4 (0-0.5) halo current toroidal peaking2 (1.2-4) (I halo /Ip)  TPF 0.5 typical maximumexperimental data boundary is 0.75 runaway electron current50% of IpHighly uncertain, high density in FIRE makes this much weaker localization of runaway deposition< 1 m 2 PFC and wall alignment

Disruption Modeling Experimental database used to project for FIRE Thermal quench time I halo /Ip  TPF dIp/dt rates for current quench

Disruption Modeling TSC simulation of disruption Critical structures modeled; VV’s (SS), passive plates (Cu), cladding (Cu), divertor (Cu), baffle (Cu), midplane port regions (SS) Zero-net current constraint on divertor, baffle, midplane port regions Provide poloidal current paths for halo/structure currents

Disruption Modeling Thermal quench,  t = 0.2 ms Ip drop, -2.9 MA/ms Diamagnetic flux Rapidly drop pressure over 0.2 ms Use hyper-resistivity to broaden current Plasma temperature drops to eV, current is shared with halo region depending on T halo (2-7 eV) and halo width Ip drops at rate determined largely by T halo Plasma shrinks rapidly, then plasma is converted to a circuit

Disruption Modeling TSC simulation produces for Engr. analysis Toroidal structure currents, fields and forces Poloidal structure current, fields, and forces Plasma toroidal currents on a grid Halo/poloidal plasma currents at structure interfaces Global plasma and PF coil data

Disruption Mitigation Utilize fueling technology to mitigate electromagnetic effects of disruptions Massive gas puff into DIII-D ----> peak halo currents reduced by 50% by He and D puffing, and toroidal asymmetry reduced Ne, Ar, and CH 4 pellets into DIII-D - ---> peak halo currents reduced by 50% with Ne and Ar pellets, and toroidal asymmetry reduced from 3 to 1.1 Cryogenic liquid jet being developed Low Z impurity pellets (LiD) if runaway electrons not an issue Snowmass assessment indicated large radiated power to FW could cause Be melting DIII-D Massive Gas Puff System

FIRE Transport and Confinement Energy Confinement Database  E 98(y,2) = M 0.19 Ip 0.93 B T 0.15 R 1.97  0.58 n  0.78 P (m, MA, T, MW)  p * /  E = 5 Z eff = (f Be = 1-3%, f Ar = 0-0.3%) Pedestal Database (Sugihara, 2003) P ped (Pa) =  10 4 M 1/3 Ip 2 R -2.1 a  3.81 (1+  2 ) -7/3 (1+  ) 3.41 n ped -1/3 (P tot /P LH ) > T ped = 5.24 ± 1.3 keV ---->  ped ?? L-H Transition P LH (MW) = 2.84M eff -1 B T 0.82 n L Ra 0.81 (2000) ----> 26 MW in flattop P LH (MW) = 2.58M eff -1 B T 0.60 n L R 0.83 a 1.04 (2002) ----> 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

Pedestal Physics and ELM’s Pedestal physics Intermediate n peeling/ballooning modes ----> ballooning destabilized by high p’ and low j ----> peeling modes destabilized by high j and low p’ Stronger shaping raises p ped Stability analysis distinguishes n ped and T ped through  * ped (n ped /T ped 2 ) ---> j BS Higher n ped leads to mode envelope narrows and lowers j BS ---> smaller  W ELM weak shapingstrong shaping ELITE projections for FIRE

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 = 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

Pedestal Physics and ELM’s Active methods for ELM mitigation JET argon seeding in Type I, f rad > 0.65, H 98 ≈ 1, n/n Gr > 0.7, Q div reduced by 2  Type III, f rad > 0.7, H 98 ≈ , n/n Gr > 0.7 Pellets that trigger ELMs, avoiding large infrequent Type I ELMs Ergodization of plasma edge region, use coils to produce high (m,n) field that perturb only ELM region JET

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

1.5D Integrated Simulations 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

1.5D Integrated Simulations FIRE H-mode, GLF23

1.5D Integrated Simulations FIREQPaux(MW)Tped(keV) TSC GLF Baldur MMM * * XPTOR/12 GLF Corsica GLF

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

Observations from 0D Analysis for Burning Plasma AT In order to provide reasonable fusion gain Q≥5, can’t operate at low density to maximize CD efficiency Density profile peaking is beneficial (pellets or ITB), since broad densities increase required H 98 and P CD Access to high density relative to Greenwald density, in combination with high bootstrap current fraction gives the lowest required H 98 H 98 ≥1.4 are required to access  flattop /  curr diff > 3, however, the ELMy H- mode scaling law is known to have a  degradation that is not observed on individual experiments Radiative core/divertor solutions are a critical area for the viability of burning AT experiments due to high P  +P CD, suggesting impurity control techniques Access to higher radiated power fractions in divertor enlarge operating space significantly Access to higher  flattop /  j decreases at higher  N, higher B T, and higher Q, since  flattop set by VV nuclear heating

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/  %.3% %.2% %.2% %.2% %.3% %.3% %.2% %.2% %.3%3.29 HH < 1.75, satisfy all power constraints, Pdiv(rad) < 0.5 P(SOL)

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

1.5D Integrated Scenarios AT-mode t = s

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

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

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

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

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

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?

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 on 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