4 th General Scientific Assembly of Asia Plasma and Fusion Association (APFA) Hangzhou, China, October , 2003 AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc, FIRE Collaboration Exploring Advanced Burning Plasma Regimes at Reactor Power Densities Dale Meade for the FIRE Collaboration

Topics to be Discussed Vision for Magnetic Fusion Power Plant Conventional Mode Operation in FIRE Advanced Mode Operation in FIRE O-D Systems analysis 1.5-D Tokamak Code Simulation RWM Stabilization Concept Issues Needing R&D Concluding Remarks

High Power Density ~ 6 MW -3 ~10 atm 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.

W7-AS

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 Development Considerations for FIRE Address key physics issues for an advanced reactor burning plasma scenarios similar to ARIES controlled burn of high power density plasma with Q >5, f BS ≈ 80% Focus technology on areas coupled to the plasma high power density plasmas plasma facing components plasma control technologies Limit scope/size of the device size comparable to today’s largest tokamaks to reduce cost only integrate items that are strongly coupled: plasma-PFCs These are some of the biggest challenges for fusion, success in these areas would lead to an attractive Demo.

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 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

0-D Power/Particle Balance Identifies Operating Space for FIRE - AT Heating/CD Powers –ICRF/FW, ≤ 30 MW –LHCD, ≤30 MW Using CD efficiencies –  (FW)=0.20 A/W-m 2 –  (LH)=0.16 A/W-m 2 P(FW) and P(LH) determined at r/a=0 and r/a=0.75 I(FW)=0.2 MA I(LH)=I p (1-f bs ) Scanning B t, q 95, n(0)/, T(0)/, n/n Gr,  N, f Be, f Ar Q= Constraints: –  flattop /  CR determined by VV nuclear heat (4875 MW-s) or TF coil (20s at 10T, 50s at 6.5T) –P(LH) and P(FW) ≤ max installed powers –P(LH) + P(FW) ≤ P aux –Q(first wall) < 1.0 MWm -2 with peaking of 2.0 –P(SOL) - P div (rad) < 28 MW –Q div (rad) < 8 MWm -2 Generate large database and then screen for viable points

ARIES-like AT Plasmas with Q ≈ 10 in FIRE Q = accessible  N = accessible f bs = accessible  flat /  CR = accessible If we can access….. H98(y,2) = P div (rad) = P(SOL) Z eff = n/n Gr = n(0)/ =

“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

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.

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 Areas of Major FIRE Activities for the Near Term The goal is to develop advanced operating modes that can be used as the design basis for FIRE.

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. 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. The U.S. Administration has shown an interest in fusion and has approved joining the ITER negotiations. Congress has also shown interest with Authorization bills that support ITER if it goes ahead, and support FIRE if ITER does not go ahead. This is consistent with the consensus in the U.S. fusion community. Concluding Remarks

The Terra Cotta Warriors have returned to fight for fusion.