Low Aspect Ratio FNSF Mission, Features, Readiness Issues – a discussion for feedback Martin Peng, ORNL FNST Meeting August 18-20, 2009 UCLA, Los Angeles, U.S.A.

What does CTF do? [Abdou et al.] “Scientific Exploration” to discover and investigate unexpected physical properties, and improve. “Engineering Verification” to compare designs and select winners. “Reliability Growth” to establish fusion nuclear component database for fusion DEMO (NRC required). Physical properties important to FNST can have time constants up to 106 s

D-D Tokamak Confinement Where should CTF be located in fusion neutron fluence rate and plasma burn duration? PP U.S. Demo  IFMIF (20-55 dpa/yr)  MTS (20 dpa/yr)  SNS (5 dpa/yr)  E.U. Demo 1.0 0.1 0.01 CTF-III CTF-II Fusion Neutron Fluence Rate (MW-yr/m2-yr) ITER: Burning Plasma CTF-I 105 104 103 102 106 10 107 108 Plasma Burn Duration (s) D-D Tokamak Confinement

Tungsten erosion redeposition lifetime D-D Tokamak Confinement Where should physical properties of interest to Plasma-Material Interface and Fusion Power in a full fusion nuclear environment be located in the same parameter space? Examples - updated. PP U.S. Demo  IFMIF (20-55 dpa/yr)  MTS (20 dpa/yr)  SNS (5 dpa/yr)  E.U. Demo 1.0 0.1 0.01 CTF-III CTF-II Fusion Neutron Fluence Rate (MW-yr/m2-yr) T build-up hot FS and tungsten T build-up in solid breeders ITER: Burning Plasma (TBM) T build-up in PbLi Breeder Tungsten erosion redeposition lifetime CTF-I T Recycling with wall bulk; T measurement 105 104 103 102 106 10 107 108 Fusion Burn Duration (s) D-D Tokamak Confinement

Fusion Burn Duration (s) FNSF tests physical properties of interest to Plasma-Material Interface and Fusion Power, to establish the understanding needed to begin engineering development on a FNSF-CTF PP U.S. Demo  IFMIF (20-55 dpa/yr)  MTS (20 dpa/yr)  SNS (5 dpa/yr)  E.U. Demo 1.0 0.1 0.01 CTF-III CTF-II Fusion Neutron Fluence Rate (MW-yr/m2-yr) T permeation thru hot RAFS with barriers FNSF Fusion Nuclear Science R&D using increasing burn durations & fluence rate T extraction in solid breeders ITER: Burning Plasma (TBM) T build-up in PbLi Breeder CTF-I T Recycling with wall bulk; T measurement 105 104 103 102 106 10 107 108 Fusion Burn Duration (s) D-D Tokamak Confinement

bN  0.75x no-wall limit; HH  1.25; q95  2xlimit; JavgTF  4 kA/cm2 Conservative parameters are available for R0 = 1.3m, A = 1.7 plasmas to address the FNSF mission bN  0.75x no-wall limit; HH  1.25; q95  2xlimit; JavgTF  4 kA/cm2 FNSF-CTF Performance level JET-DD JET-DT 2xJET 3xJET Tests enabled divertor FNS FNST WL (MW/m2) 0.005 0.25 1.0 2.0 Current, Ip (MA) 4.2 6.7 8.4 Field, BT (T) 2.7 2.9 3.6 Safety factor, q95 12.7 8.6 Toroidal beta, bT (%) 4.4 10.1 10.8 Normal beta, bN 2.1 3.3 3.5 Avg density, ne (1020/m3) 0.54 1.1 1.5 Avg ion Ti (keV) 7.7 7.6 10.2 11.8 Avg electron Te (keV) 4.3 5.7 7.2 BS current fraction 0.45 0.47 0.50 0.53 NBI H&CD power (MW) 26 22 44 61 Fusion power (MW) 0.4 19 76 152 NBI energy (kV) 120 235 330 Aspect ratio and plasma aggressiveness have high leverage on tradeoffs in readiness, performance, cost, and risk.

D-D Tokamak Confinement FNSF-CTF R&D can be staged: JET-level DD-only for divertor-PMI verification (requiring full remote handling); JET-level DT for FNS; 2xJET for CTF-II; 3xJET for CTF-III PP U.S. Demo  IFMIF (20-55 dpa/yr)  MTS (20 dpa/yr)  SNS (5 dpa/yr)  E.U. Demo 1.0 0.1 0.01 CTF-III CTF-II FNS R&D Fusion Neutron Fluence Rate (MW-yr/m2-yr) 3xJET-DT T permeation thru hot RAFS with barriers T extraction in solid breeders 2xJET-DT ITER: Burning Plasma (TBM) T build-up in PbLi Breeder CTF-I JET-DT T Recycling with wall bulk; T measurement 105 104 103 102 106 10 107 108 Fusion Burn Duration (s) JET-DD D-D Tokamak Confinement

Divertor testing goals at JET-level DD operation Required capabilities Peak heat fluxes  10 MW/m2 Pulse lengths progressively up to  106 s Remote handling, efficient replacement Extremely rare plasma-induced disruptions Disruption mitigation, e.g. “killer pellets” Plasma material interface diagnostics Tritium diagnostics Testing goals – measure, understand, improve, and verify Life time limiting mechanisms – divertors, plasma facing surfaces Tritium transport mechanisms – permeation, distribution at very low levels Can ITER tungsten divertor be extended up to 106 s? FNSF-CTF divertors

PMIF: Can a Linear Test Stand Provide High Plasma Heat Fluxes, Ti and Te for up to 106 s? Magnetic Mirror with RF Heating Test Target Transfer Cask Helicon PMI Test Station PMI Analysis & Preparation Stations US-J PSI July 2009

Feasibility for high heat flux in test stand, required to enable FNSF-CTF, is being answered High recycle target (Tt < 10 eV): Tu is limited to 30-35 eV, requiring q|| ~ 20-40 MW/m2 and nu ~4-8x1019/m3 Detachment (Tt < 2-3 eV): Tu is limited to 20 eV, requiring q|| ~ 10-20 MW/m2 and same nu range Can Helicon plasma at ~5 eV be heated to ~35 eV? Can Helicon reach such high nu (data so far  2-3x1019/m3)? High Recycle Detachment Target Plasma T(eV) High Recycle Detachment Upstream Plasma T(eV)

Divertor testing goals at JET-level DT operation with WL = 0.25 MW/m2 Additional testing goals – measure, understand, improve, and verify Degradation due to neutron damage – decreased thermal conductivity, material strength Enhancements to life time limiting mechanisms – divertors, plasma facing surfaces Enhancements in tritium transport mechanisms – permeation, distribution Can ITER tungsten divertor still be extended up to 106 s? Allowable stress to limit radiation induced creep FNSF-CTF divertors Note that RAFS is not expected to suffer He-embrittlement for up to 10-20 dpa [ref: Zinkle et al, TOFE 2002 review paper] Divertor and plasma facing surfaces improved to handle WL = 1-2 MW/m2 would enable the start of CTF-II and CTF-III R&D

Upper and lower breeding blankets Breeding blanket testing goals at JET-level DT operation with WL = 0.25 MW/m2 Upper and lower breeding blankets Required capabilities Peak plasma heat fluxes  1 MW/m2 Pulse lengths progressively up to  106 s Remote handling, efficient replacement Extremely rare plasma-induced disruptions Disruption mitigation, e.g. “killer pellets” Blanket tritium diagnostics Blanket performance instrumentation Testing goals – measure, understand, improve, and verify Tritium breeding and power extraction Blanket tritium transport mechanisms – permeation, distribution at power relevant levels Can ITER TBM designs be extended up to 106 s?

Mid-plane Test Blanket Modules Test Blanket Module testing goals at JET-level DT operation with WL = 0.25 MW/m2 Required capabilities Peak plasma heat fluxes  2 MW/m2 Pulse lengths progressively  106 s Remote handling, efficient replacement Extremely rare plasma-induced disruptions Disruption mitigation, e.g. “killer pellets” Blanket performance instrumentation Tritium diagnostics Testing goals – measure, understand, improve, and verify Power extraction and tritium breeding Test blanket tritium transport mechanisms – permeation, distribution at power relevant levels Can ITER TBM designs be extended up to 106 s? Mid-plane Test Blanket Modules

Single-turn toroidal field coil center post Single-turn TF center post testing goals at JET-level DT operation with WL = 0.25 MW/m2 Required capabilities JavgTF  3 kA/cm2 ITFC = 17.7 MA Lifetime  1 dpa and up to 10 dpa Remote handling, efficient replacement Testing goals – measure, understand, improve, and verify Allowable stresses of hardened and embrittled GlidCop to limit crack growth Proper combination of load bearing design (compression, locking against twisting force) Low-A power plant study estimated  3.7 kA/cm2 [ref: Song et al, FED 2005] Single-turn toroidal field coil center post GlidCop is highly resistant to He-embrittlement for > 100 dpa of fission neutron damage [ref: Cu alloy review, DOE/ER-0313/16, 1994]

Additional FNSF chamber components will require FNS R&D at 0.25 MW/m2 Positive-ion NBI (120 kV) Pulse lengths progressively up to  106 s Remote handling, efficient replacement Tritium diagnostics Testing goals – measure, understand, improve, and verify Life time limiting mechanisms – source, cryo, power handling, etc. Can ITER NBI technologies be applied and extended to very long pulses Plasma diagnostic port assemblies Can ITER diagnostic systems be extended progressively up to  106 s? FNSF-CTF

FNS R&D can be enabled using JET-level DT plasmas, providing WL = 0 FNS R&D can be enabled using JET-level DT plasmas, providing WL = 0.25 MW/m2 with Q~0.8 FNSF is the scientific R&D stage of CTF, to improve component designs before engineering and technology testing for Demo Very conservative plasmas are possible for FNSF-CTF eventually to deliver WL = 2 MW/m2, without plasma-induced disruptions for  106 s Remaining new R&D: startup, electron transport, high heat flux, TFC JET-level DD plasma adequate for divertor verification and improvements A very long pulse linear high plasma heat flux test stand being assessed JET-level DT plasma would enable full FNS R&D for PMI and Harnessing Fusion Power Divertors, plasma facing surfaces Blankets – breeding and mid-plane test modules TFC center post Conclusions applicable to normal aspect ratio designs – as a first stage of FDF US-J PSI July 2009

We will continue to work closely with the FNST stakeholders to Describe the scientific mission Develop major components of a National FNS Program Example: What are the scientific questions? – tritium transport through divertor Why are these important? – at low burn fraction, can only lose <1% for fuel self-sufficiency Basic physical models and uncertainties? – tritium implantation on plate, permeation through RAFS, barriers, GlidCop, crossing into coolant Experimental conditions needed – test stands, subsystems  test components in full fusion nuclear environment Measurements needed and techniques Improvements in understanding, or reductions in uncertainty required to determine component design for CTF-II and CTF-III testing US-J PSI July 2009