1. AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT, ORNL, PPPL, SNL, SRS, UCLA, UCSD, UIIC, UWisc FIRE Collaboration http://fire.pppl.gov FIRE Overview Dale Meade FIRE Physics Validation Review DOE Germantown, MD March 30, 2004

2. Topics to be Discussed NSO/FIRE Mission, Objectives Critical Issues for Burning plasma Experiment Status of FIRE and Progress since Snowmass Characteristics of FIRE Conventional Mode Operation Advanced Mode Operation Power Handling Approach Summary

3. The purpose of the Next Step Options activity is to investigate and assess various opportunities for advancing the scientific understanding of fusion energy, with emphasis on plasma behavior at high energy gain and for long duration. The Next Step Options (NSO) study has been organized as a national integrated physics/engineering design activity within the Virtual Laboratory for Technology (VLT). The NSO program’s objective is to develop design options and strategies for burning plasmas in the restructured fusion sciences program, considering the international context. Examples of specific tasks to be pursued include investigation of a modular program pathway, with initial emphasis on the burning plasma module. The initial effort has been focused on a design concept called the Fusion Ignition Research Experiment (FIRE) that includes both burning plasma physics and advanced toroidal physics mission objectives. NSO-PAC: 15 members, Chaired by Tony Taylor, 5 meetings, last report attached FIRE effort evolved form the US ITER Design Home Team,and involved >15 institutions and >50 individuals. FIRE has been engaged in a PreConceptual design activity at a budget of ≈ $2M/year with FY04 = $0.6M. The PreConceptual design is to be completed in FY04. The Next Step Option (NSO) Activity

5. Are the mission and objectives identified by FIRE appropriate to answer the critical burning plasma issues in a major next step experiment? Is the proposed physical device sufficiently capable and flexible to answer the critical burning plasma issues proposed? What areas are deficient and what remedies are recommended? What areas need supporting R&D from the base program (experimental, theory and modeling)? Agenda Charge to PVR Committee

6. High Power Density P f /V~ 6 MW -3 ~10 atm  n ≈ 4 MWm -2 High Power Gain Q ~ 25 - 50 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. A burning plasma should address the key plasma issues for an attractive PP.

7. FIRE Mission FIRE Mission: to attain, explore, understand and optimize fusion-dominated plasmas. NSO-PAC The first part is to create and control a burning plasma. President’s Science Advisor to NRC BPAC Nov 1992 Create and understand a controlled, self-heated, burning starfire on earth. FESAC Key Overarching Theme

8. Confining Field The overarching issue for a burning plasma is whether a self-heated plasma with a self-generated confining magnetic field can be created and controlled.

9. 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> 10  E Alpha ash accumulation/pumping> several  He Plasma current profile evolution~ 2 to 5  skin Divertor pumping and heat removal> many  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 83% @ Q = 25 TAE/EPMstable at nominal point, access to unstable

11. 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) 1,400 tonne LN cooled coils Mission: to attain, explore, understand and optimize magnetically-confined fusion-dominated plasmas

12. Characteristics of FIRE 40% scale model of ARIES-RS plasma Strong shaping  x = 2,  x = 0.7, DN 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 inboard nt shield, allows small size 3,000 pulses @ full field 30,000 pulses @ 2/3 field 1 shot/hr @10T/20s/150 MW Site needs comparable to previous DT tokamaks (TFTR/JET).

14. ARIES and SSTR/CREST studies have determined requirements for an attractive power plant. 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. FIRE is Aims to Address Issues Related to an Attractive PP 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

17. FIRE Plasma Regimes Operating Modes Elmy H-Mode Improved H-Mode Hybrid Mode Two Freq ICRF ITB Reversed Shear AT - “steady-state” (100% NI) H-ModeAT(ss)ARIES-RS/AT R/a 3.6 3.6 4 B (T) 10 6.5 8 - 6 I p (MA) 7.7 5 12.3-11.3 n/n G 0.70.851.7-0.85 H(y,2) 1.11.2 – 1.70.9 - 1.4  N 1.8≤ 4.24.8 - 5.4 f bs,% 25 ~7788 - 91 Burn/  CR 2 3 - 5steady 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.

19. Snowmass Assessment of FIRE H-mode confinement - OK (but uncertain) Stabilization of NTMs in H-Mode needs more study Elm/disruption power handling - both ITER/FIRE Plasma pulse length H-mode same as ITER(2  skin ) AT mode ~ 1  skin at Snowmass, now up to 5  skin Diagnostics integration with FW AT diagnostics (beam seeded) Magnet insulation needs R&D Reduce time between shots

20. Snowmass on Confinement There is confidence that ITER and FIRE will achieve burning plasma performance in H–mode based on an extensive experimental database. Based on 0D and 1.5D modeling, all three devices have baseline scenarios which appear capable of reaching Q = 5 – 15 with the advocates’ assumptions. ITER and FIRE scenarios are based on standard ELMing H–mode and are reasonable extrapolations from the existing database. More accurate prediction of fusion performance of the three devices is not currently possible due to known uncertainties in the transport models. An ongoing effort within the base fusion science program is underway to improve the projections through increased understanding of transport. Executive Summary

21. FIRE Confinement is a Modest Extrapolation(x3) 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)

23. Since Snowmass Confinement Projections have Improved New two term scaling published by ITPA leads to Q > 10 for FIRE DEMO shot for FIRE (JET 52009)- q 95 = 2.9, H(y,2) =1.2,  N = 2.1, n/n GW = 0.7, small sawteeth, no NTMs High triangularity and modest n/n GW in existing experiments continues to lead to increased confinement relative to ITER98(y,2). C-Mod experiments show slightly improved (10%) confinement for DN relative to SN plasmas. Recent experiments with DIII-D Hybrid mode project to Q ≈ 10 to 20 However need to strengthen data base for non-rotating plasmas both beam heated and ICRF only is confinement for all metal PFCs different from carbon PFCs?

24. Recent ITPA Results on Confinement CDBM (Cordey et al) extended H-mode scaling to a two term (core and pedestal model). IAEA FEC 2002 and Nucl. Fusion 43 No 8 (August 2003) 670-674 H(y, 2) 1.0 1.03 1.18 1.25 1.22 1.27 Q 9 10 15 25 22 26 ITER98(y,2) is pessimistic relative to  scans in DIII-D and JET. A new scaling is being evolving from ITPA CDBM March 8-11 meeting that will reduce adverse  scaling (similar to electrostatic gyro-Bohm model). Increased pedestal pressure dependence on triangularity (Sugihara-2003).

25. FIRE-Like Discharge in JET without NTMs n/n GW ≈ 0.75 q 95 ≈ 2.9 H(y,2) ≈ 1.2  N ≈ 2.1 A good shot to test models.

26. No He Pumping Creation and control of a Burning Plasma with strong self-heating

27. FIRE, the Movie Simulation of a Standard H-mode in FIRE - TSC CTM ≈ GLF23 m = 1 sawtooth Model - Jardin et al other effects to be added - Jardin et al

28. An integrated burning plasma simulation capability would be of great benefit to: Understand burning plasma phenomena based on existing exp’ts Refine the physics and engineering design of a BP experiment Provide a real time control algorithm for self-driven burning plasma, and to optimize experimental operation Analyze experimental results and help transfer knowledge to other magnetic configurations. An Integrated BP Simulation Capability is Needed Burning plasmas are complex, non-linear and strongly-coupled systems. highly self driven (83% self-heated, 90% self-driven current) plasmas are needed for power plant scenarios. Does a burning plasma naturally evolve to a self-driven state?

29. database extended down to q 95  3.5 closeness to DN necessary: type II obtained in whole  -range accessible when  Xp  0.02 m (0.35    0.5) stability analysis: edge shear stabilises lower n, squeezes eigenfunction Exhaust: Type II ELMs occur with strong shaping Zohm IAEA 2002

30. T e ped. (keV) n e ped l, (x10 19 m -2 ) 1.5 MA 1.35 MA 1.2 MA DD DD DD DD DD Pure Type-II ELMy phases achieved at high  pol in the QDN configuration 1.2 1.3 1.5 1.6 1.8 Type-II ELMs in “JT-60U high-β pol ” scheme ELMs get smaller with increasing  pol and frequency/irregularity increases T e,ped and n e,ped remain high at high  pol : not consistent with Type-III ELMs Type-II ELMs may be accessible at higher I p with higher power: to be done Not seen with lower single-null configuration at high  pol : QDN configuration may be necessary (although j edge was also different) 1.2 0.9 3.6 2 19.6019.6519.75 Time (s)  pol. = Presentation to STAC Jerome Pamela EFDA-CSU, 05 March 2004

32. Alpha Particle Driven Instabilities in FIRE HINST - locally unstable NOVA-globally stable Gorelenkov et al, Nuc Fus 43(2003) 594 Nominal H-Mode plasma case Need to analyze alpha driven modes in FIRE and ITER AT modes

33. “Steady-State” High-  Advanced Tokamak Discharge on FIRE 0 1 2 3 4 time,(current redistributions)

34. q Profile is Steady-State During Flattop, t=10 - 41s ~ 3.2  CR 0 10 20 30 40, s 0 10 20 30 40 i (3)=0.42 0 1 2 3 4 5 6 7 Profile Overlaid every 2 s From 10s to 40s

35. Application to ITER is also being studied as part of ITPA.(IAEA paper)

37. 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 3.6 3.6 4 B (T) 10 6.5 8 - 6 P loss /R x (MW/m) 17 23 94 - 66 P rad-div (MWm -2 ) 5 < 8 5 P rad-FW (MWm -2 ) 0.3 0.5 <0.5 P fusion (MWm -2 ) 5.5 5.5 6 - 5.3  n (MWm -2 ) 2 2 4 - 3.3 P n (MWm -3 ), VV 25 2550 - 40 The FIRE divertor would be a significant step toward an ARIES-like DEMO divertor. FIRE AT pulse length is presently limited by nuclear heating of the vacuum vessel.

38. 25 MW/m 2

41. Advanced Tokamak Modes (ARIES as guide) (  A, SN/DN,  N, f bs, ……) - RWM Stabilization - What is required and what is feasible? - Integration of detached divertor and Advanced Tokamak - Plasma Control (fast position control, heating, current-drive, fueling) High Power Density Handling - Divertor R&D: High heat flux, low tritium retention - First Wall and Vacuum Vessel for high neutron wall loading Diagnostic Development and Integration with First Wall/fusion environment Integrated Simulation of Burning Plasmas - exploration of fusion-dominated plasmas (self organized?) Areas of Major FIRE Activities for the Near Term

42. Summary of Introduction The FIRE mission and design is aimed toward “creating and controlling burning plasmas” first in conventional and then advanced modes. FIRE is aimed to address nearly all the BP issues associated with both H-mode and ARIES-like AT modes. Progress has been made toward addressing FIRE issues raised at Snowmass, more examples in subsequent talks. Continuing progress in tokamak research, and coordination by the ITPA has strengthened the physics basis for ITER and FIRE. Several areas of physics and technology R&D important for burning plasmas will be identified in subsequent talks.