AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc FIRE Collaboration FIRE Exploring the Plasma Physics of an Advanced Fusion Reactor Dale Meade Forum on the Future of Fusion Fusion Power Associates November 19-21,2003

Topics to be Discussed Vision for Magnetic Fusion Power Plant Critical Issues for Magnetic Fusion A Big Question for Fusion - FIRE Goals Status of FIRE Conventional Mode Operation Advanced Mode Operation Future FIRE/US Burning Plasma Activities Issues Needing R&D Concluding Remarks

High Power Density P f /V~ 6 MW -3 ~10 atm  n ≈ 4 MWm -2 High Power Gain Q ~ n  E T ~ 6x10 21 m -3 skeV P  /P heat = f  ≈ 90% Steady-State ~ 90% Bootstrap ARIES Economic Studies have Defined the Plasma Requirements for an Attractive Fusion Power Plant Plasma Exhaust P heat /R x ~ 100MW/m Helium Pumping Tritium Retention Plasma Control Fueling Current Drive RWM Stabilization Significant advances are needed in each area. In addition, the plasma phenomena are non-linearly coupled.

Confining Field How does a fusion-dominated plasma self-organize in the laboratory?

Reactor studies ARIES and SSTR/CREST have determined requirements for a reactor. 12 ITER would expand region  to  N ≈ 3 and f bs ≈ 50% at moderate magnetic field. FIRE would expand region to  N ≈ 4 and f bs ≈ 80% at reactor-like magnetic field. Attractive Reactor Regime is a Big Step From Today Modification of JT60-SC Figure Existing experiments, EAST, KSTAR and JT-SC would exp- and high  N region at low field. KSTAR JT60-SC EAST

Fusion Ignition Research Experiment (FIRE) R = 2.14 m, a = m B = 10 T, (~ 6.5 T, AT) I p = 7.7 MA, (~ 5 MA, AT) P ICRF = 20 MW P LHCD ≤ 30 MW (Upgrade) P fusion ~ 150 MW Q ≈ 10, (5 - 10, AT) Burn time ≈ 20s (2  CR - Hmode) ≈ 40s (< 5  CR - AT) Tokamak Cost = $350M (FY02) Total Project Cost = $1.2B (FY02) 1,400 tonne LN cooled coils Mission: to attain, explore, understand and optimize magnetically-confined fusion-dominated plasmas

Characteristics of FIRE 40% scale model of ARIES-RS plasma All metal PFCs Actively cooled W divertor Be tile FW, cooled between shots T required/pulse ~ TFTR ≤ 0.3g-T LN cooled BeCu/OFHC TF no neutron shield, allows small size 3,000 full field (H-Mode) 30,000 2/3 field (AT-mode) X3 repetition rate since SNMS Site needs comparable to previous DT tokamaks (TFTR/JET).

FIRE Plasma Systems are Similar to ARIES-AT  x = 2.0,  x = 0.7 Double null divertor Very low ripple 0.3% (0.02%) NTM stability: LH current profile modification (  ’) at 10T 180 GHz, B o = 6.6T No ext plasma rotation source Vertical and kink passive stability: tungsten structures in blanket, feedback coils behind shield n=1 RWM feedback control with coils - close coupled 80 (90%) bootstrap current 30 MW LHCD and 5 MW (25 MW capable) ICRF/FW for external current drive/heating Tungsten divertors allow high heat flux Plasma edge and divertor solution: balancing of radiating mantle and radiating divertor, with Ar impurity n/n Greenwald ≈ 0.9, (ARIES-AT) H(y,2) = 1.4 (ARIES-AT) High field side pellet launch allows fueling to core, and  P * /  E = 5 (10) allows sufficiently low dilution

FIRE Could Access Confinement Similar to ARIES-RS Tokamaks have established a basis for scaling confinement of the diverted H-Mode. B  E is the dimensionless metric for confinement time projection n  E T is the dimensional metric for fusion - n  E T =  B 2  E =  B. B  E ARIES-RS Power Plants require B  E slightly larger than FIRE due high  and B. ARIES-RS (Q = 25)

FIRE Plasma Regimes Operating Modes Elmy H-Mode Improved H-Mode Reversed Shear AT - OH assisted - “steady-state” (100% NI) H-ModeAT(ss)ARIES-RS/AT R/a B (T) I p (MA) n/n G H(y,2) –  N 1.8≤ f bs,% 25 ~ Burn/  CR steady H-mode facilitated by  x = 0.7,  x = 2, n/n G = 0.7, DN reduction of Elms. AT mode facilitated by strong shaping, close fitting wall and RWM coils.

No He Pumping Needs He pumping technology

“Steady-State” High-  Advanced Tokamak Discharge on FIRE time,(current redistributions)

q Profile is Steady-State During Flattop, t= s ~ 3.2  CR , s i (3)= Profile Overlaid every 2 s From 10s to 40s

Improved FIRE AT-Mode (  N > 5 without wall) 100% Non-Inductive “Steady State”static analysis L-mode edge Ip = 4.5 MA B T = 6.5 T  N = 4.5  = 5.4%  p = 2.33 li(1) = 0.54 li(3) = 0.41 q min = 2.61 p(0)/  p  = 2.18 n(0)/  n  = 1.39  N (n=1) = 6.2  N (n=2) = 5.2  N (n=3) = 5.0 C. Kessel, Active MHD Control Workshop, Nov 2003

Improved FIRE AT-Mode (  N > 4 without wall) 100% Non Inductive TSC-LSC Dynamic Simulations TSC-LSC equilibrium Ip=4.5 MA Bt=6.5 T q(0)=3.5, q min =2.8  N =4.2,  =4.9%,  p=2.3 li(1)=0.55, li(3)=0.42 p(0)/  p  =2.45 n(0)/  n  =1.4 Stable n=  Stable n=1,2,3 with no wall √V/Vo L-mode edge C. Kessel, Active MHD Control Workshop, Nov 2003

FIRE Plasma Technology Parameters All Metal PFCs W divertor Be coated Cu tiles FW Power Density ~ARIES divertor - steady-state - water cooled,  ~ 2s First wall tiles - cooled between pulses  ~40s H-ModeAT(ss)ARIES-RS/AT R/a B (T) P loss /R x (MW/m) P rad-div (MWm -2 ) 5 < 8 5 P rad-FW (MWm -2 ) <0.5 P fusion (MWm -2 )  n (MWm -2 ) P n (MWm -3 ), VV The FIRE divertor would be a significant step toward an ARIES-like DEMO divertor. FIRE AT performance is presently limited by the first wall power ( , n) handling.

25 MW/m 2

R&D Needed for Advanced Tokamak Burning Plasma Scaling of energy and particle confinement needed for projections of performance and ash accumulation. Benchmark codes using systematic scans versus density, triangularity, etc. Determine effect of high triangularity and double null on confinement,  -limits, Elms, and disruptions. Continue RWM experiments to test theory and determine hardware requirements. Determine feasibility of RWM coils in a burning plasma environment. Improve understanding of off-axis LHCD and ECCD including effects of particle trapping, reverse CD lobe on edge bootstrap current and Ohkawa CD. Development of a self-consistent edge-plasma-divertor model for W divertor targets, and incorporation of this model into core transport model.

Advanced Tokamak Modes (ARIES as guide) (  A, SN/DN,  N, f bs, ……) - RWM Stabilization - What is required and what is feasible? - Integrated Divertor and AT - Plasma Control (fast position control, heating, current-drive, fueling) High Power Density Plasma Facing Components Development - High heat flux, low tritium retention Diagnostic Development and Integration Integrated Simulation of Burning Plasmas Physics Validation Review (Jan 2004) Areas of Major FIRE Activities for the Near Term

FIRE would be able to access quasi-stationary burning plasma conditions. In addition, a reactor-relevant“steady-state” advanced tokamak operating mode with power densities approaching ARIES-AT could be explored on FIRE. Progress continues to be made in this area. There are a number of high leverage physics R&D items to be worked on for operation in the conventional mode and the advanced mode for FIRE and ITER. There needs to be an increased emphasis on physics R&D for advanced modes and high power density PFCs that lead to an attractive fusion reactor. We need to take advantage of the interest and opportunity afforded by the ITER initiative. We also need to broaden interest in fusion, burning plasma physics and the associated scientific program. This is consistent with the consensus in the U.S. fusion community, FESAC, NRC and Energy Bill. Concluding Remarks

Uncle Sam Asks the Deep Question about Fusion Will the US Fusion Program self-organize to produce fusion?