1. Optimization of a High-  Steady-State Tokamak Burning Plasma Experiment Based on a High-  Steady-State Tokamak Power Plant D. M. Meade, C. Kessel, S. Jardin Princeton Plasma Physics Laboratory Presented at IEA Workshop on Optimization of High-  Steady-State Tokamaks http://fire.pppl.gov/iea59_optimize_workshop.html General Atomics February 14, 2005 Optimization of FIRE Based on ARIES RS/AT

2. Tokamak Based Power Plant Studies have Identified Attractive High-  Steady-State Configurations with A ≈ 4 Three decades of systematic studies in the US have surveyed the range of possibilities available for a tokamak power plant and have identified the ARIES- RS/AT (A = 4) designs as most attractive (lowest COE) possibilities. Other studies favoring highish aspect ratio for high-  steady-state include: TPX (A= 4.5), ITER-HARD (A = 4), ASSTR (A = 4) The FIRE design studies initiated in 1999 adopted the ARIERS/AT physics and plasma technology design characteristics including : Strong shaping –> Double null  x = 0.7,  x = 2, Aspect ratio ≈ 4 Reactor level B T = 6 - 10 T and plasma density = 2 - 4 x 10 20 m -3 LHCD/ICFWCD - no momentum input All metal PFCs W divertor Internal fast control coils RWM coils integrated into FW of port plugs - LN cooled Cu coils provide sufficient pulse length and small size –> low cost

3. Optimization of Cu Coil BPX (e,.g, FIRE) Optimization Depends on Goals and Constraints Realistic engineering constraints must be imposed Optimization of inductively-driven BPX with Cu Coil (e,.g, FIRE) 1991 CIT Study (LLNL Super Code-Galumbos) W. Reiersen 2000 FIRE (FIRE Sale) J. Schultz 2001 FIRE (BPSC) S. Jardin,C. Kessel, D. Meade, C. Neumeyer Optimization of High-  Steady-State Modes in FIRE 2002 SOFE Meeting FIRE AT C. Kessel 2002/2004 IAEA FIRE D. Meade et al References 1. J. Galumbos et al Fusion Tech. 13, 93, 1988 2. W. Reiersen 3. J. Schultz - 4. S. Jardin, C. Kessel et al Fusion Science and Technology 43, 161 2003 5. A High-Aspect-Ratio Design for ITER, J. C. Wesley et al Fusion Tech. 21, 1380 1992. 6. Y. Seki et al., (1991). Rep. JAERI-M 91-081, JAERI. Naka.

4. CIT Optimization Using Super Code 1989 Compact Ignition Tokamak Optimized at A = 3.5

5. Confinement (Elmy H-mode) ITER98(y,2):  E = 0.144 I 0.93 R 1.39 a 0.58 n 20 0.41 B 0.15 A i 0.19  0.78 P heat -0.69 H(y,2) Density Limit: n 20 < 0.75 n GW = 0.75 I P /  a 2 H-Mode Power Threshold: P th > (2.84/A i ) n 20 0.58 B 0.82 R a 0.81 MHD Stability:  N =  / (I P /aB) < 3.0 P AUX, Q = P FUSION /P AUX, q CYL or q MHD,, Z EFF all held fixed Engineering Constraints:1. Flux swing requirements in OH coil (V-S) 2. Coil temperature not exceed 373 o K 3. Coil stresses remain within allowables Configuration Concept: 1. OH coils not linking TF coils, or 2. OH coils linking TF coils—ST-like The Systems Code was Updated and Calibrated Based on 3-D Finite Element Stress Calculations for FIRE. Kessel, Jardin 2002

6. Optimization of Cu Coil BPX (e,.g, FIRE) Using BPSC B T = 4T 5T 6T 7T 8T 9T 10T

8. Optimization at Smaller Size and Higher Aspect Ratio as Confinement Improves Major Radius (m) 10T

9. Optimization is not Sensitive to Variation of Elongation with Aspect Ratio Major Radius (m)

10.  N = 1.8 P f /V = 5.5 MWm -3 f bs ≈ 25% A 1.5 D Simulation(TSC) is Used to Test Systems Code

11. Optimization of High-  Steady-State* Modes in a Cu Coil BPX (e,.g, FIRE) Optimization of AT Modes in a specific FIRE (Fixed A, R, B < B max, etc) Use a 0-D Systems Code to calculate a large data base (~ 50,000) of possible solutions as parameter space is scanned. Impose engineering constraints on pulse length (TF ohmic and nuclear heating, divertor target and baffle heat loads, vacuum vessel nuclear heating and first wall surface heating) to define operating space. Use J_solver and PEST to validate stability and required current profiles. Use TSC to confirm evolution of integrated discharge. * Steady-state = 100% non-inductive,  q/q < few % for several  CR,  div,  FW

12. 0-D Operating Space Analysis for FIRE AT Heating/CD Powers –ICRF/FW, 30 MW –LHCD, 30 MW Using CD efficiencies –  (FW)=0.20 A/W-m2 –  (LH)=0.16 A/W-m2 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 G,  N, f Be, f Ar Q=5 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 MW/m 2 with peaking of 2.0 –P(SOL)-Pdiv(rad) < 28 MW –Q div (rad) < 8 MW/m 2

13. FIRE’s Q = 5 AT Operating Space A data base of ~ 50,000 operating points is calculated with 0-D code Engineering constraints are imposed to generate the operational boundaries shown below Potential operating points are examined in more detail-PEST, TSC, etc 300 100 0 200 400 1.0 1.2 1.4 1.6 1.8 2.0 4.0 5.0 3.0 2.0 4.5 3.5 2.5 0.00.2 0.4 0.6 0.8 1.0 H-Mode

14. FIRE AT Mode Operating Range is Limited by Nuclear Heating of Vac Vessel & First Wall Not by Cu Coils Q = 5 Nominal operating point Q = 5 P f = 150 MW, P f /V p = 5.5 MWm -3 (ARIES) ≈ steady-state 4 to 5  CR Physics basis improving (ITPA) required confinement H factor and  N attained transiently C-Mod LHCD experiments will be very important First Wall is the main limit Improve cooling revisit FW design Opportunity for additional improvement (optimization).

15. “Steady-State” High-  Advanced Tokamak Discharge on FIRE P f /V = 5.5 MWm -3  n ≈ 2 MWm -2 B = 6.5T  N = 4.1,  t = 5% f bs = 77% 100% non-inductive Q ≈ 5 H98 = 1.7 n/n GW = 0.85 Flat top Duration = 48  E = 10  He = 4  cr FT/P7-23

16. Cool 1st Wall ARIES AT (  N ≈ 5.4, f bs ≈ 90%) 12 OFHC TF (≤ 7 T) Additional Opportunities to Optimize FIRE for the Study of ARIES AT Physics and Plasma Technologies

17. Concluding Remarks FIRE is very close to the optimum aspect ratio and size for an inductively-driven H-Mode burning plasma experiment using LN-cooled coils. The present FIRE configuration is also capable of producing AT plasmas with characteristics approaching those of ARIES-RS with pulse lengths sufficient to study High-  Steady-state burning plasmas with fusion power densities of ≈ 5 MW m -3. The present FIRE AT regimes are limited by the first wall and vacuum vessel and not the TF coil. - improve FW and Vac Vessel cooling ––> 6  CR - change TF conductor to OFHC (B t ≤ 7T) ––> ≈ 12  CR A bottoms up optimization of a FIRE for AT operation only has not been done, need more information on confinement scaling in AT mode.

18. FIRE-AT Approaches the Parameters Envisioned for ARIES-Power Plant Plasmas Physics Items FIRE-ATARIES-RS/AT xx 2.0 xx 0.7 ConfigurationDN  N  t ~ 4, 5%~ 5, 9% % Non-inductive100 % bootstrap7790 Equilibration %~100steady Plasma rotationVery low RWM Coils (rel. to First Wall) Integrated With FW Inside TF Outside VV On axis CDICFW Off axis CDLH Technology Items FIRE-ATARIES- RS/AT B(T)6.56 - 8 I p (MA)511 Core Power Density (MWm -3 ) 55 - 6 FW -  N (MW/m -2 ) 24 FW - P rad (MWm -2 )15100 First WallBeMo Div Target (MWm -2 )5 - 20 Divertor TargetWW Pulse Length(s,  cr ) 40, 5months Cooling Divertor, First Wall Steady inertial Steady steady