FIRE Overview Dale Meade FIRE Physics Validation Review

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

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

The Next Step Option (NSO) Activity • 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 Pre. Conceptual design activity at a budget of ≈ $2 M/year with FY 04 = $0. 6 M. The Pre. Conceptual design is to be completed in FY 04.

Charge to PVR Committee • 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

ARIES Economic Studies have Defined the Plasma Requirements for an Attractive Fusion Power Plant High Power Gain Q ~ 25 - 50 nt. ET ~ 6 x 1021 m-3 ske. V Pa/Pheat = fa ≈ 90% High Power Density Pf/V~ 6 MW-3 ~10 atm Gn ≈ 4 MWm-2 Plasma Exhaust Pheat/Rx ~ 100 MW/m Helium Pumping Tritium Retention Plasma Control Fueling Current Drive RWM Stabilization Steady-State ~ 90% Bootstrap Significant advances are needed in each area. A burning plasma should address the key plasma issues for an attractive PP.

FIRE Mission: to attain, explore, understand 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

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.

FIRE Physics Objectives Burning Plasma Physics (Conventional Inductively Driven H-Mode) Q ~10 as target, higher Q not precluded fa = Pa/Pheat ~ 66% as target, up to 83% @ Q = 25 TAE/EPM stable at nominal point, access to unstable Advanced Toroidal Physics (100% Non-inductively Driven AT-Mode) Q ~ 5 as target, higher Q not precluded fbs = Ibs/Ip ~ 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 t. E Alpha ash accumulation/pumping > several t. He Plasma current profile evolution ~ 2 to 5 tskin Divertor pumping and heat removal > many tdivertor First wall heat removal > 1 tfirst-wall

Fusion Ignition Research Experiment (FIRE) • R = 2. 14 m, a = 0. 595 m • B = 10 T, (~ 6. 5 T, AT) • Ip = 7. 7 MA, (~ 5 MA, AT) • PICRF = 20 MW • PLHCD ≤ 30 MW (Upgrade) • Pfusion ~ 150 MW • Q ≈ 10, (5 - 10, AT) • Burn time ≈ 20 s (2 t. CR - Hmode) ≈ 40 s (< 5 t. CR - AT) • Tokamak Cost = $350 M (FY 02) • Total Project Cost = $1. 2 B (FY 02) 1, 400 tonne LN cooled coils Mission: to attain, explore, understand optimize magnetically-confined fusion-dominated plasmas

Characteristics of FIRE • 40% scale model of ARIES-RS plasma • Strong shaping kx = 2, dx = 0. 7, DN • All metal PFCs • Actively cooled W divertor • Be tile FW, cooled between shots • T required/pulse ~ TFTR ≤ 0. 3 g-T • LN cooled Be. Cu/OFHC TF • no inboard nt shield, allows small size • 3, 000 pulses @ full field • 30, 000 pulses @ 2/3 field • 1 shot/hr @10 T/20 s/150 MW • Site needs comparable to previous DT tokamaks (TFTR/JET).

FIRE is Aims to Address Issues Related to an Attractive PP • ARIES and SSTR/CREST studies have determined requirements for an attractive power plant. • Existing experiments, EAST, KSTAR and JT-SC would exp and high N region at low field. EAST KSTAR JT 60 -SC • ITER would expand region to N ≈ 3 and fbs ≈ 50% at moderate magnetic field. • FIRE would expand region to N≈ 4 and fbs ≈ 80% at reactor-like magnetic field. 12 Modification of JT 60 -SC Figure

FIRE Plasma Regimes Operating Modes H-Mode AT(ss) ARIES-RS/AT • Elmy H-Mode R/a 3. 6 • Improved H-Mode B (T) 10 6. 5 • Hybrid Mode Ip (MA) 7. 7 5 12. 3 -11. 3 • Two Freq ICRF ITB n/n. G 0. 85 1. 7 -0. 85 1. 2 – 1. 7 0. 9 - 1. 4 1. 8 ≤ 4. 2 4. 8 - 5. 4 25 ~77 88 - 91 3 - 5 steady H(y, 2) • Reversed Shear AT - “steady-state” (100% NI) N fbs , % 0. 7 1. 1 Burn/t. CR 2 4 8 - 6 • H-mode facilitated by dx= 0. 7, kx = 2, n/n. G= 0. 7, DN reduction of Elms. • AT mode facilitated by strong shaping, close fitting wall and RWM coils.

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 tskin) AT mode ~ 1 tskin at Snowmass, now up to 5 tskin • Diagnostics integration with FW AT diagnostics (beam seeded) • Magnet insulation needs R&D • Reduce time between shots

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 0 D and 1. 5 D 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

FIRE Confinement is a Modest Extrapolation(x 3) • Tokamaks have established a basis for scaling confinement of the diverted H-Mode. • Bt. E is the dimensionless metric for confinement time projection • nt. ET is the dimensional metric for fusion - nt. ET = B 2 t. E = B. Bt. E • ARIES-RS Power Plants require Bt. E slightly larger than FIRE due high and B. ARIES-RS (Q = 25)

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 ITER 98(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?

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. 03 1. 18 1. 25 1. 22 1. 27 Q 9 10 15 25 22 26 • ITER 98(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).

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.

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

Simulation of a Standard H-mode in FIRE - TSC • CTM ≈ GLF 23 • m = 1 sawtooth Model - Jardin et al • other effects to be added - Jardin et al FIRE, the Movie

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? 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.

Exhaust: Type II ELMs occur with strong shaping • database extended down to q 95 3. 5 • closeness to DN necessary: type II obtained in whole d-range accessible when DXp 0. 02 m (0. 35 d 0. 5) • stability analysis: edge shear stabilises lower n, squeezes eigenfunction Zohm IAEA 2002

Pure Type-II ELMy phases achieved at high bpol in the QDN configuration • Type-II ELMs may be accessible at higher Ip with higher power: to be done Type-II ELMs in “JT-60 U high-βpol” scheme Da 1. 5 MA Da 1. 35 MA Da 1. 2 MA bpol. = 1. 2 1. 3 1. 5 1. 6 2 • Te, ped and ne, ped remain high at high pol: not consistent with Type-III ELMs Te ped. (ke. V) ne pedl, (x 1019 m-2) 19. 60 Presentation to STAC • ELMs get smaller with increasing pol and frequency/irregularity increases 1. 8 1. 2 0. 9 3. 6 • Not seen with lower single-null configuration at high pol: QDN configuration may be necessary (although jedge was also different) 19. 65 Time (s) 19. 75 Jerome Pamela EFDA-CSU, 05 March 2004

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

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

q Profile is Steady-State During Flattop, t=10 - 41 s ~ 3. 2 t. CR Profile Overlaid every 2 s From 10 s to 40 s 0 10 20 30 40 , s 0 0 10 20 30 2 3 4 5 6 7 40 , s 0 0 1 40 li(3)=0. 42 10 20 30 40

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

FIRE Plasma Technology Parameters H-Mode AT(ss) ARIES-RS/AT All Metal PFCs • W divertor R/a 3. 6 • Be coated Cu tiles FW B (T) 10 6. 5 4 8 - 6 Ploss/Rx (MW/m) 17 23 94 - 66 Power Density ~ARIES Prad-div (MWm-2) < 8 5 • divertor - steady-state - Prad-FW (MWm-2) 0. 3 0. 5 <0. 5 water cooled, t ~ 2 s Pfusion (MWm-2) 5. 5 6 - 5. 3 • First wall tiles - cooled Gn (MWm-2) 2 4 - 3. 3 between pulses t ~40 s Pn(MWm-3), VV 25 50 - 40 5 2 • 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.

25 MW/m 2

Areas of Major FIRE Activities for the Near Term • Advanced Tokamak Modes (ARIES as guide) (k, d, A, SN/DN, N, fbs, ……) - 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? )

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.