AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc NSO Collaboration FIRE - A Test Bed for Advanced DEMO (ARIES) Physics and Plasma Technology Dale Meade Princeton Plasma Physics Laboratory 7th International Symposium on Fusion Nuclear Technology Tokyo, Japan May 24, 2005

High Power Density P f /V~ 6 MWm -3 ~10 atm  n ≈ 4 MWm -2 High Gain Q ~ n  E T ~ 6x10 21 m -3 skeV P  /P heat = f  ≈ 90% Low rotation 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. A Broad International Approach is needed to provide the R&D Needed for DEMO

Can a fusion fire be controlled and sustained? A fusion plasma will be a complex non-linearly coupled system that can not be simulated by a large number non-burning experiments: one addressing confinement, another addressing MHD, etc as separate issues New Issues for a self-heated Burning Plasma (Q > 10) include: Global energy balance Local burn profile control Alpha ash removal Bootstrap Current Profile control (f bs /I p > 70%) The difficulty of controlling and sustaining a fusion fire should not be underestimated. - a second burning plasma experiment would greatly reduce the risk of burning plasma problems in ITER.

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>  E Alpha ash accumulation/pumping>  He Plasma current profile evolution~ 2 to 5  skin Divertor pumping and heat removal>  divertor First wall heat removal> 1  first-wall FIRE Physics Objectives Burning Plasma Physics (Conventional Inductively Driven H-Mode) Q~10 as target, higher Q not precluded f  = P  /P heat ~ 66% as target, up to Q = 25 TAE/EPMstable at nominal point, access to unstable

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) at new site. 1,400 tonne LN cooled coils Mission: to attain, explore, understand and optimize magnetically-confined fusion-dominated plasmas

FIRE is Based on ARIES-RS Vision 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

FIRE has Passed DOE Physics Validation Review The DOE FIRE Physics Validation Review (PVR) was held March in Germantown. The Committee included: S. Prager, (Chair) Univ of Wisc, Earl Marmar, MIT, N. Sauthoff PPPL, F. Najmabadi, UCSD, Jerry Navratil, Columbia (unable to attend), John Menard PPPL, R. Boivin GA, P. Mioduszewski ORNL, Michael Bell, PPPL, S. Parker Univ of Co, C. Petty GA, P. Bonoli MIT, B. Breizman Texas, PVR Committee Consensus Report: The FIRE team is on track for completing the pre-conceptual design within FY 04. FIRE would then be ready to launch the conceptual design. The product of the FIRE work, and their contributions to and leadership within the overall burning plasma effort, is stellar. Is the proposed physical device sufficiently capable and flexible to answer the critical burning plasma science issues proposed above? The 2002 Snowmass study also provided a strong affirmative answer to this question. Since the Snowmass meeting the evolution of the FIRE design has only strengthened ability of FIRE to contribute to burning plasma science.

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. Significant Progress on Existing Tokamaks Improves FIRE (and ITER) Design Basis since FESAC and NRC Reviews

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 ~  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 if high  N regime can be stabilized. Stable side Unstable side

FIRE Conventional H-Mode Operating Range Expanded Nominal operating point Q =10 P f = 150 MW, 5.5 MWm -3 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 MWm -3 could be produced.

“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

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

25 MW/m 2

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 time (sec)

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 RWM Coils in every third port, no shield module 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. Baseline RWM Coils

The proposed RWM Coils would be in the Front Assembly of Every 3rd Mid-Plane Port Plug Assembly RWM coils located behind shielding module on Port Plug

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) Power Density(MWm -3 ) Wall Loading  n (MWm -2 ) Pulse Duration (s) (  CR, % equilibrated) , 86 - >99.9% , 86 - >99% 20,000,000 steady Mass of Fusion Core (tonnes)23,0001,40013,000

Concluding Remarks 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 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 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 help provide the basis for advanced DEMO.