1. AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc NSO Collaboration http://fire.pppl.gov Implications of Recent Physics Advances for the Design and Operation of Burning Plasma Experiments such as ITER and FIRE Dale Meade Princeton Plasma Physics Laboratory Symposium on Fusion Engineering Knoxville, TN September 27, 2005

2. High Power Density P f /V~ 6 MWm -3 p ~10 atm  n ≈ 4 MWm -2 High Gain Q ~ 25 - 50 n  E T ~ 6x10 21 m -3 skeV P  /P heat = f  ≈ 90% Low rotation Steady-State ~ 90% Bootstrap ARIES Studies have Defined the Plasma Requirements for an Attractive Fusion Power Plant and hence DEMO 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. A Broad International Approach is needed to provide the R&D Needed for DEMO

3. Advanced Toroidal Physics (100% Non-inductively Driven AT-Mode) Q~ 5 as target, higher Q not precluded f bs = I bs /I p ~ 80% as target, ARIES-RS/AT≈90%  N ~ 4.0, n = 1 wall stabilized, RWM feedback Quasi-Stationary Burn Duration (use plasma time scales) Pressure profile evolution and burn control> 20 - 40  E Alpha ash accumulation/pumping> 4 - 10  He Plasma current profile evolution~ 2 to 5  skin Divertor pumping and heat removal> 10 - 20  divertor First wall heat removal> 1  first-wall FIRE Physics Objectives Burning Plasma Physics (Conventional Inductively Driven H-Mode) Machine parameters optimized for H-Mode Q~10 as target, higher Q not precluded f  = P  /P heat ~ 66% as target, up to 83% @ Q = 25 TAE/EPMstable at nominal point, access to unstable

4. FIRE is Based on ARIES-RS Vision for DEMO 40% scale model of ARIES-RS plasma ARIES-like all metal PFCs Actively cooled W divertor Be tile FW, cooled between shots Close fitting conducting structure ARIES-level toroidal field LN cooled BeCu/OFHC TF ARIES-like current drive technology FWCD and LHCD (no NBI/ECCD) No momentum input Site needs comparable to previous DT tokamaks (TFTR/JET). T required/pulse ~ TFTR ≤ 0.3g-T

5. Fusion Ignition Research Experiment (FIRE) R = 2.14 m, a = 0.595 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) at new site. 1,400 tonne LN cooled coils Mission: to attain, explore, understand and optimize magnetically-confined fusion-dominated plasmas

6. Extended H-Mode and AT operating ranges Benefits of FIRE high triangularity, DN and moderate n/n G Extended H-Mode Performance based ITPA scaling with reduced  degradation, and ITPA Two Term (pedestal and core) scaling (Q > 20). Hybrid modes (AUG, DIII-D,JET) are excellent match to FIRE n/n G, and projects Q > 20. Slightly peaked density profiles (n(0)/ = 1.25)enhance performance. Elms mitigated by high triangularity, disruptions in new ITPA physics basis will be tempered somewhat. Continued Progress on Existing Tokamaks Improves FIRE and ITER Design Basis since FESAC and NRC Reviews

7. New ITPA  E Scaling Opens Ignition Regime for FIRE Systematic scans of  E vs  on DIII-D and JET show little degradation with  in contrast to the ITER 98(y, 2) scaling which has  E ~  -0.66 A new confinement scaling relation developed by ITPA has reduced adverse scaling with  see eq. 10 in IAEA-CN-116/IT/P3-32. Cordey et al. A route to ignition is now available for the wall stabilized regime  N ≤ 3. Stable side Unstable side Q = 30Q = 15 Old Scaling New Scaling

8. FIRE Conventional H-Mode Operating Range Expanded Nominal operating point Q =10 P f = 150 MW, 5.5 MWm -3  n = 2 MWm -2 Power handling improved P f ~ 300 MW, 10 MWm -3 Physics basis improved (ITPA) DN enhances  E  N DN reduces Elms Hybrid mode has Q ~ 20 Engineering Design Improved Pulse repetition rate tripled divertor & baffle integrated Advanced DEMO power density of 5 - 10 MWm -3 could be produced.

9. H-Mode and hybrid mode do not provide the basis for an attractive steady-state DEMO. AT operation based on reversed (negative) shear with high bootstrap fraction (>80%), high  N ~ 4 approaches regime needed for DEMO. Fortunately, both FIRE and ITER can operate in the AT regime. Advanced Tokamak Operation of a Burning Plasma is needed to Provide Physics Design Basis of DEMO

10. “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 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

11. FIRE AT Mode Limited by First Wall not TF Coil 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

13. Cool 1st Wall ARIES AT (  N ≈ 5.4, f bs ≈ 90%) 12 OFHC TF (≤ 7 T) Opportunities to Optimize FIRE for the Study of ARIES AT Physics and Plasma Technologies 050100 time (sec)

14. First Results from ITER-AT Studies Using TSC,TRANSP and NOVA-K Goal is Steady-State,  N ≈ 3.5, f bs > 60% f bs, Q > 5 using NINB, ICFW and LHCD First case has  N ≈ 2.5, f bs ≈ 44%, ≈ 97% non-inductive and Q ≈ 5. Optimize CD mix & startup to flatten q profile C. Kessel et al: Simulations of ITER Hybrid Operating Mode-SOFE 1A-108

15. RWM Coil Concept for ITER Baseline RWM coils located outside TF coils Applying FIRE-Like RWM Feedback Coils to ITER Increases  limit for n = 1 from  N = 2.5 to ~4 Integration and Engineering feasibility of internal RWM coils is under study. VALEN Analysis Columbia University No-wall limit FIRE-like RWM coils would have large stabilizing effect on n=1 G. Navratil, J. Bialek Columbia University FIRE-like RWM coils would be located inside the vacuum vessel behind shield module but inside the vacuum vessel on the removable port plugs. 7 RWM Coils in every other port except NBI,

16. ITER with FIRE would provide a strong basis for Adv. DEMO FIREARIES-RS ITER FIREARIES-RS Fusion Gain10(H), 5(AT) 25 (AT) Fusion Power (MW)500 - 3501502170 Power Density(MWm -3 )0.65.66.2 Wall Loading  n (MWm -2 ) 0.624 Pulse Duration (s) (  CR, % equilibrated) 500 - 3000 2 -10, 86 - >99.9% 20 - 35 2 - 5, 86 - >99% 20,000,000 steady Mass of Fusion Core (tonnes)23,0001,40013,000

17. Concluding Remarks Both ITER and FIRE could operate in the AT regime, but to more fully exploit AT operation some modifications would be needed. Areas of additional R&D for FIRE and ITER include: high power density all metal PFCs and actively cooled first wall internal feedback coils to allow higher beta (power-density) plasmas Optimization of FIRE design based on AT operation. A Broadened Approach with FIRE as a supporting burning plasma experiment would reduce ITER Technical risk and help fully exploit ITER’s ultimate capability to provide the basis for an advanced DEMO.

18. Status and Plans for FIRE DOE Physics Validation Review of FIRE passed. March 30-31, 2004 FIRE Pre-Conceptual Activities are completed. September 30, 2004 Ready to begin FIRE Conceptual Design Activities. Now Assessing feasibility of using AT as design basis for FIRE AT progress (DIII-D, JT-60U, JET, C-Mod) is nearly there required values of  N, H(y,2), f bs, q, n/n G achieved but not simult Assessing impact of using AT as design basis for FIRE B ≤ 7T would allow use of OFHC TF reducing coil construction costs and coil power while allowing increased pulse lengths. Cost savings from coils and power supplies would be used to offset costs of active cooling of first wall and current drive system. Optimize FIRE parameters using AT as the design basis.(Future)

19. The proposed RWM Coils would be in the Front Assembly of Every other Mid-Plane Port Plug Assembly except for the Four NBI ports. RWM coils located behind shielding module on Port Plug