1. FIRE Activities with Emphasis on Technology Needs VLT PAC Meeting MIT September 4, 2003 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 Dale Meade for the FIRE Collaboration

2. Topics to be Discussed Vision for Magnetic Fusion Power Plant Conventional Mode Operation in FIRE Advanced Mode Operation in FIRE Engineering Status Issues Needing R&D Physics Validation Review (PVR) Concluding Remarks

3. House Appropriations FY 04 (pending conference) “The additional $10,800,000 includes $4,000,000 for burning plasma experiments, including support for ITER and for the domestic FIRE project, $5,200,000 for fusion technology, and $1,600,000 for advanced design and analysis work.” Senate Appropriations FY 04 (pending conference) “The Department's proposed fiscal year 2004 budget proposes to cut severely long-term activities in fusion technology and advanced design that will have significant impact on the ultimate attractiveness of fusion power. The Committee recommends that, within available funds, the Department should make adjustments to redress the imbalance resulting from these cuts.” House Energy Bill (HR-6) (pending conference) “If at any time during the negotiations on ITER, the Secretary determines that construction and operation of ITER is unlikely or infeasible, the Secretary shall send to Congress, as part of the budget request for the following year, a plan for implementing the domestic burning plasma experiment known as FIRE, including costs and schedules for such a plan.” Senate Energy Bill (S-14) (108 th version, 107 th version is pending conference) “In the event that ITER fails to go forward within a reasonable period of time, the Secretary shall send to Congress a plan, including costs and schedules, for implementing the domestic burning plasma experiment known as the Fusion Ignition Research Experiment.” Background for FIRE Planning

5. High Power Density ~ 6 MW -3 ~10 atm High Power Gain Q ~ 25 - 50 n  E T ~ 6x10 21 m -3 skeV P  /P heat = f  ≈ 90% Steady-State ~ 90% Bootstrap ARIES Studies have Defined Requirements for an Attractive Fusion Power Plant and Critical Issues. Plasma Exhaust P heat /R x ~ 100MW/m Helium Pumping Tritium Retention Plasma Control Fueling Current Drive RWM Stabilization Significant advances (> 10) are needed in each area. In addition, the plasma phenomena and technology are non-linearly coupled.

6. Reactor studies ARIES and SSTR/CREST have determined requirements for a reactor. 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. Existing experiments, KSTAR and JT-SC would expand high  N region at low field. The Attractive Reactor Regime is a Big Step From Today

7. 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, AT) Burn time ≈ 20s (2  CR -Hmode) ≈ 40s (< 5  CR -AT) Tokamak Cost = $350M (FY02) Total Project Cost = $1.2B (FY02) 1,400 tonne Mission: to attain, explore, understand and optimize magnetically-confined fusion-dominated plasmas

8. Characteristics of FIRE 40% scale of ARIES plasma xsection All metal PFCs Actively cooled W divertor Be tile FW, cooled between shots T inventory ~ TFTR LN cooled BeCu/OFHC TF no neutron shield, small a 3,000 full pulses 30,000 2/3 pulses X3 repetition rate since SNMS Site needs comparable to previous DT tokamaks.

9. 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 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 f equilib (J(r)) 86 95 - 99 100 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.

10. 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 ) 0.2 0.3 0.2 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 performance is presently limited by the first wall power ( , n) handling.

11. No He Pumping Needs He pumping technology

12. 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 (5,2) @ 10T ECCD @ 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, 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

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

14. 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

16. A “Design” of a RWM coil and Diagnostics Integrated with the First Wall Shielded Port Assembly is Needed.

17. Response to Snowmass and NSP-PAC Critiques readjusted radial build of the center stack and the configuration to increase major radius to 2.14 in response to NSO/Snowmass tripled pulse repetition rate by cooling both sides of TF inner leg SBIR on looks promising for increasing lifetime shots vacuum vessel/divertor support stresses now OK for disruptions in the 2m machine, need to scale up to 2.14m machine development of “steady-state” ARIES-like mode (  N ~ 4, f bs ~ 80%) > doubled the pulse length of AT mode up to 5  CR Work in Progress scaling disruptions and disruption stresses to the 2.14m machine modeling of edge and divertor plasma power handling for 2.14 m and to extend power handling capability for AT modes feasibility study of resistive wall mode coil integrated with first wall study of a generic diagnostic integrated with shield and first wall evaluation of new proposal for FWCD launcher, and LHCD launcher. extending TSC simulations to latest versions of GLF23 transport model participating in ITPA, oral talk at EPS and APFA 2003 in China. FY 2003 Progress on FIRE

18. TF - robust, not the main limit to pulse length (vac vessel and first wall) Insulation 3x10 10 Rads similar to ARIES TF coils (10 11 allowable) SBIR underway on insulation. PF – need to understand the fatigue issues for AT modes possible design mod to increase shots potential for ultra fast control system using RWM coils. PFC and Divertor – a major issue at ITPA Divertor – W rod targets, W dome – actively cooled disruption load calculations underway redesign of mounting configuration underway (remote handling) plasma edge/divertor modeling underway (UEDGE) First Wall – Be coated Cu tiles (similar to ITER) Close to thermal limit for disruption mitigation using gas jets Disruption heat loads on first wall may be problem Can APEX results for high power density be used/tested on FIRE? FIRE Engineering Status and Tasks

19. Blankets/Shielding Need neutron streaming calculations for diagnostic integration 3-D neutron calculations when configuration has been finalized. Internal shielding – low activation material? Blanket test modules - ~ 2 MWm -2 for 40 s, useful? Vacuum Vessel –disruption loads and nuclear heating (19MWm -3 to 0.04 MWm -3 ) Integration of RWM coil and port plug Integration of diagnostics with vacuum vessel. Improved first wall cooling, partial Remote Handling - Ongoing, good connection to ITER/ORNL experience Pumping and fueling Feasibility of moveable port plug for base pumping HFS Pellet injection is a high leverage item for efficient tritium use, density profile peaking and burn control. FIRE Engineering Status and Tasks (2)

20. Tritium Handling and Retention (tritium consumption/shot similar to TFTR ~0.25gT) ARIES needs < 0.04% retention for one yr, ITER needs <0.2% for one yr, FIRE needs < 2% for one yr. DT experiments had 20 – 40 % Can W rod system provide 0.04% retention? Is this a reactor solution? Interaction between Be FW, W div and boronization Power Supply - Construction Project R&D ICRF (20 MW, 70 – 120 MHz for heating and on axis current drive) Sources available Launcher with moveable tuning in vac under investigation (ORNL) Extended 4 strap antenna Dog-leg with two piece port plug (ORNL) Full analysis for thermal and disruption loads on launcher (future) ECF (170 GHz for feedback stabilization of NTM in AT) Use ITER source development Collaborating with MIT(Decker/Bers) on OKCD calculations Must have sensitive instability sensors since power/cost is an issue FIRE Engineering Status and Tasks (3)

21. LHCD ( 20 MW, ~ 5 GHz for off-axis current drive) Some source development may be needed Launcher is a major R&D item (future) NBI ( for Diagnostics) R&D needed for intense pulsed beam Safety Major issues have been addressed TFTR levels of tritium reduce risks, more sites available Design Integration Major issues are OK, but lots of detailed work to be done. Diagnostics, RWM coils, neutron streaming, remote handling Materials TF insulation for ~ 10 11 Rads CuCrZr fatigue properties Low activation structural - should they be investigated? FIRE Engineering Status and Tasks (4)

22. 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 - High heat flux, low tritium retention Diagnostic Development and Integration Integrated Simulation of Burning Plasmas Areas of Major FIRE Activities for FY 2004

23. Plasma Facing Components (Divertor and First Wall materials) - high power density, long pulse capability - elm erosion and disruption survivability - low tritium retention Vacuum Vessel (blanket modules and shielding port plugs) - blanket module test assemblies (nuclear heating, low activation) - disruptions - integrate with closely coupled control/stabilization coils and diagnostics Plasma Heating, Current Drive and Fueling - development/design of ICRF, LHCD systems for BP scenarios - interface with fusion environment ( esp. launchers) Diagnostics Development and Design Integration - new diagnostics for J(r), E(r), fluctuations, alpha particles - integration with fusion environment( e.g., radiation induced conductivity) Development of Advanced Operating Modes and Plasma Control systems Areas of Synergy and Possible Joint Work (FIRE, ITER)

24. Fusion Community assessment of FIRE’s capability to accomplish program Respond to previous reviews of FIRE - Engineering Review June 2001 - Snowmass Technical assessment - NSO-PAC Recommendations Incorporate recent tokamak results (e.g., from ITPA meetings, etc) Since resources are limited, - Propose to carry this out in late November/early December 2003. - Focus on key issues FIRE Physics Validation Review (PVR)

25. FIRE is able to access quasi-stationary burning plasma conditions. In addition, an interesting “steady-state” advanced tokamak operating mode with power densities approaching ARIES-AT appears to be feasible on FIRE. There are a number of R&D items to be worked on for operation in the conventional mode and the advanced mode. Items of greatest interest for FIRE, and perhaps the U.S. are: high power density all-metal PFCs, associated modeling, and technologies to facilitate high performance AT operation like RWM coils, RFCD launchers and diagnostic integration with the vacuum vessel. 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. FIRE needs to be in a position to move ahead if ITER fails to go forward in a “reasonable period of time.” Concluding Remarks