NSTXU Supported by Plasmawall boundary control with lithium
NSTX-U Supported by Plasma-wall boundary control with lithium divertor and associated plasma confinement improvements in NSTX Columbia U Comp. X General Atomics FIU INL Johns Hopkins U LANL LLNL Lodestar MIT Nova Photonics New York U ORNL PPPL Princeton U Purdue U SNL Think Tank, Inc. UC Davis UC Irvine UCLA UCSD U Colorado U Illinois U Maryland U Rochester U Tennessee U Washington U Wisconsin NSTX-U 短縮版 Masayuki Ono NSTX-U Project Director PPPL, Princeton University In collaboration with the NSTX-U Team 29 th プラズマ核融合学会 Symposium Plasma-wall Boundary Control Kasugashi, Japan, Nov. 29, 2012 JPS Lithium Divertor NSTX-U M. Ono Culham Sci Ctr U St. Andrews York U Chubu U Fukui U Hiroshima U Hyogo U Kyoto U Kyushu Tokai U NIFS Niigata U Tsukuba U U Tokyo JAEA Hebrew U Ioffe Inst RRC Kurchatov Inst TRINITI NFRI KAIST POSTECH Seoul National U ASIPP ENEA, Frascati CEA, Cadarache IPP, Jülich IPP, Garching ASCR, Czech Rep Nov. 29, 2012
NSTX-U Bridges Physics Gaps Toward Designing FNSF NSTX-U Fusion Nuclear Science Facility ST Pilot Plant Major Radius R 0 [m] 0. 86 0. 94 1. 3 2. 2 Aspect Ratio = R 0 / a 1. 3 1. 5 1. 6 1. 7 Plasma Current [MA] 1 2 4 10 10 20 0. 5 1 2 -3 P/R, P/S [MW/m, m 2] 10, 0. 2* 20, 0. 4* 30 60, 0. 6 1. 2 40 100, 0. 3 1 Peak Div. heat flux [MW/m 2] 10 40 (5 sec) 50 60 Toroidal Field [T] ITER ~ 10 DEMO ~ 40 - 60 • DEMO/FNSF projected to require X 10 the tungsten divertor design limit • NSTX-U aims to provide database to design FNSF in this decade • FNSF requires DEMO level divertor heat flux handling – innovative solution needed if we were to design FNSF • NSTX-U aims to test divertor solutions at FNSF/ DEMO level heat flux NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 2
NSTX Tested Various Lithium Conditioning Systems Effects of solid / liquid lithium surfaces investigated with LLD Pre-Run LLD Post-Run • Lithium conditioning systems - Li Evaporator (LITER) – 1. 3 kg evaporated during last campaign - Liquid Lithium Divertor (LLD) – Moly surface with over 300 °C temperature - Li Dropper – Fast injection into plasma - Li Pellets – Used initially NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 3
Lithium Improved H-mode Performance in NSTX Te Broadens, t. E Increases, PH Reduces, ELMs Stabilize Te broadening with lithium No lithium (129239); 260 mg lithium (129245) With Lithium Without Lithium H. W. Kugel, Po. P 2008 t. E improves with lower collisionality t. E improves with lithium Pre-discharge lithium evaporation (mg) R. Maingi, PRL (2011) NSTX-U JPS Lithium Divertor NSTX-U M. Ono S. Kaye, IAEA (2012) Nov. 29, 2012 4
Li core concentration stays well below 0. 1% for LLD temperature range of 90°C to 290°C R=135 -140 cm, t=500 -600 ms Solid Liquid • Li core concentration remained very low ≤ 0. 05%. C remains dominant impurity even after massive (hundreds of milligrams) Li evaporation • No apparent increase in Li nor C core concentration even at higher LLD surface temperature. M. Podesta, IAEA (2012) Reason for low lithium core dilution? : • Li is readily ionized ~ 6 e. V • Li is low recycling – sticks to wall • Li has high neoclassical diffusivity F. Scotti, APS (2012) NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012
Clear reduction in NSTX divertor surface temperature and heat flux with increased lithium evaporation • 2 identical shots (No ELMs) – Ip = 0. 8 MA, Pnbi ~ 4 MW – high δ, fexp ~ 20 • 2, pre-discharge lithium depositions – 150 mg: 141255 – 300 mg: 138240 • Tsurf at the outer strike point stays below 400° C for 300 mg of Li – Peaks around 800° C for 150 mg • Results in a heat flux that never peaks above 3 MW/m 2 with heavy lithium evaporation • a) • b) • Lithiated graphite • c) • d) T. Gray. IAEA 2012 NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 6
1 -D Cylindrical RLLD Model Two Point Model with Given Lithium Profile NSTX-like conditions Assumptions: R = 0. 75 m, DR = 3 cm, cylinder, deuterium ions, Z = 23 cm, At the divertor entrance Te 0 = Ti 0 = 50 e. V, ne 0 = 2 x 1013 cm-3, Two-Point Model, Constant BT Given Li profile, Li radiation = 10 -26 W/Li-electron Power and particle flux Te 0, ne 0 Divertor Entrance Li+++ Li+ Evaporated Li Divertor strike point NSTX-U JPS Lithium Divertor NSTX-U M. Ono Li radiation Z (cm) Li 0 LLD Tray M. Ono. IAEA 2012 Nov. 29, 2012
1 -D Cylindrical RLLD Two Point Model Explains NSTX LLD Observation with Modest Amount of Li Te (e. V) N-Li (cm 3) Linear lithium density profiles • 3. 00 E+13 • 6. 00 E+01 • 2. 50 E+13 • 5. 00 E+01 N-Li = 3. 9 e 17 • 2. 00 E+13 t ~ 100 ms Li ~ 30 mg /sec • 1. 50 E+13 Electron temperature profiles N-Li = 3. 24 e 16 • 4. 00 E+01 N-Li = 2. 0 e 16 • 3. 00 E+01 N-Li = 2. 0 e 17 • 2. 00 E+01 • 1. 00 E+13 N-Li = 3. 24 e 16 • 5. 00 E+12 N-Li = 3. 9 e 17 • 1. 00 E+01 • 0. 00 E+00 • 0 PII(MW/m 2) • 9. 00 E+01 • 5 • 10 • 15 • 20 • 25 Parallel Divertor Heat Flux ne • 5 • 10 • 15 • 20 • 25 Electron density profiles • 7. 00 E+14 N-Li = 3. 24 e 16 • 8. 00 E+01 • 0 (cm 3) • 6. 00 E+14 • 7. 00 E+01 N-Li = 2. 0 e 17 • 5. 00 E+14 • 6. 00 E+01 • 4. 00 E+14 • 5. 00 E+01 • 4. 00 E+01 • 3. 00 E+14 N-Li = 3. 9 e 17 • 3. 00 E+01 • 2. 00 E+14 • 2. 00 E+01 • 1. 00 E+14 • 1. 00 E+01 • 0. 00 E+00 • 0 NSTX-U • 5 • 10 • 15 Z (cm) • 20 • 25 • 0 JPS Lithium Divertor NSTX-U M. Ono • 5 • 10 • 15 Z (cm) • 20 • 25 Nov. 29, 2012
RLLD Works for Fusion Power Plant R = 6 m, DR = 10 cm, Z = 138 cm with modest amount of Li N-Li (cm 3) Li density profiles Log N-Li / m 2 -s • 4. 00 E+13 t ~ 100 ms • 3. 50 E+13 N-li = 9. 37 e 19 Li ~ 10 g /sec • 3. 00 E+13 • 2. 50 E+13 N-li = 5. 2 e 19 • 2. 00 E+13 N-li = 1 e 19 • 1. 50 E+13 • 1. 00 E+13 • 5. 00 E+12 • 0. 00 E+00 • 0 PII (MW/m 2) • 1400 • 20 • 40 • 60 • 80 • 100 • 120 • 140 • 160 Parallel Divertor Heat Flux LL Temp (°C) Te (e. V) Electron temperature profiles • 2. 50 E+02 • 1200 • 2. 00 E+02 • 1000 • 1. 50 E+02 • 800 • 600 • 1. 00 E+02 • 400 • 5. 00 E+01 • 200 • 0 • 50 • 100 • 150 • 0. 00 E+00 • 0 Z (cm) NSTX-U • 20 • 40 • 60 • 80 • 100 • 120 • 140 • 160 Z (cm) JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012
Radiative Liquid Lithium Divertor Proposed Based largely on the NSTX Liquid Lithium Divertor Research RLLD Edge Plasma Core Reacting Plasma Divertor Heat and Particles Flux B 0 First Wall / Blanket At 500°C – 700°C Li+++ Li++ 000000 Scrape Off Layer Li+ Closed RLLD Li 0 LL In LL Out Flowing LLD Tray LL In 200 – 450 °C M. Ono. IAEA 2012 NSTX-U Heat Exchanger Liquid Lithium (LL) ~ 1 l/sec Particle pumping by Li coated wall Flowing LL Particle Pumping Surfaces Li Radiative Mantle Li wall coating / condensation Li path Li Evap. / Ionization Reduced Divertor Heat and Particle Flux Divertor Strike Point LL Purification System to remove tritium, impurities, and dust JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 10
Impurity and Dust Removal by LL Circulating Loop Low melting point of Li (180°C) makes such a loop practical in reactor Divertor Heat and Particle Flux T ~ 0. 5 g /s For 1 GW-Electric Power Plant: • LL loop removes tritium ~ 0. 5 g / sec and recycle for plasma fueling. Condensed liquid lithium Pumping: LL ~ 1 l/s • Modest LL flow of ~ 1 l/sec should be sufficient to transport impurities out of the system with LL impurity level ~ 1 %. Lithium Radiative Mantle Lithium Evap. / Ionization LL 10 g / s << 1 l/s Flowing LLD Tray • Tritium (T) level ~ 0. 5 g / l before filtering or 1000 l RLLD system contains ~ 500 g of tritium or ~ one day of T burn. M. Nishikawa, Fusion Sci. & Tech. (2011) • Purification systems developed by IFMIF could provide particle removal and keep LL relatively clean ≤ 0. 1 %. H. Kondo, Fusion Eng. & Design (2011) K. Katekari, J. Energy and Power Eng. (2012) To RLLD Heat Exchanger LL Circulation Pump LL 1 l/s Imp. ~ 0. 1% • A simple filter can remove the dust generated within reactor chamber ~ 0. 1 - 10 m D. NSTX-U Fueling JPS Lithium Divertor NSTX-U M. Ono From RLLD LL ~ 1 l/s T – 0. 5 g/s Imp. ~ 1% LL T – 0. 5 g/s Cold Trap Tritium Removes Recycling T, D, H, N, C, O, and dust to ~ 0. 1% level Tritium at 500 °k Separater Deuterium & Other impurities as well as dust Nov. 29, 2012 11
Is Li PFC viable in magnetic fusion reactors? RLLD appears promising to solve outstanding Li issues First Lithium Symposium in 2010: Y. Hirooka. et al. , Nucl. Fusion (2010) Second Lithium Symposium in 2011 : M. Ono, et al. , Nucl. Fusion (2012) ü Handling high divertor heat flux, RLLD ü Removal of deuterium, tritium, and impurities from liquid lithium, Lithium-loop, FNSF ü Removal of high steady-state heat flux from divertor, RLLD ü Flowing of liquid lithium in magnetic fields, Low LL Flow Rate ü Longer term corrosion of internal components by liquid lithium, Low Temperature, FNSF, TBM R&D ü Safety of flowing liquid lithium, Low Temperature, FNSF and TBM R&D ü Compatibility of liquid lithium with a hot reactor first wall Lower temperature operation of RLLD. NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 12
Summary of NSTX-U / RLLD • Divertor PMI and liquid lithium is a high priority research for NSTX-U. • For next-step devices FNSF and DEMO, an acceptable divertor heat flux solution is needed; x 10 higher heat flux than the tungster-based divertor PFC design limit. • Lithium coating has shown to significantly improve Hmode performance and divertor heat flux in NSTX. • Radiative LL Divertor (RLLD) is proposed to solve the reactor divertor heat flux issue. • A preliminary assessment indicates a good compatibility of RLLD for fusion reactor divertor application. • NSTX upgrade is progressing well and it aims to develop Demo-FNSF relevant divertor solutions NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 13
Backup Slides NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 14
NSTX-U Aims to Develop Physics Understanding Needed for Designing Fusion Energy Development Facilities • Enable devices: ST-FNSF, STPilot/DEMO, ITER – Leveraging unique ST plasmas provides new understanding for tokamaks, challenges theory q Develop key physics understanding to be tested in unexplored, hotter ST plasmas q q q Study high beta plasma transport and stability at reduced collisionality, for extended pulse Prototype methods to mitigate very high heat/particle flux Move toward fully non-inductive operation NSTX-U JPS Lithium Divertor NSTX-U M. Ono Fusion Nuclear Science Facility (FNSF) ST Pilot Plant ITER NSTX-U New center-stack BT 0. 5 Ip 1 2 nd neutral beam 1 T PNBI 6 2 MA pulse 1 Nov. 29, 2012 12 MW 5 s 15
Innovative Divertor Solution Needed Expected Divertor Heat Flux ~ 10 x Tungsten Design Limit • Improved solid material to better survive heat and neutron flux – e. g. Develop fundamental understanding at microscopic level - Solid material research is needed to optimize fusion reactor PFCs and blanket issue D O Li C • Innovative divertor configuration to reduce divertor heat flux – e. g. , snowflakes, super-x - Snow flake was shown to reduce divertor heat flux by x 3 in NSTX • Liquid metal divertor? - e. g. , radiative liquid lithium divertor NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 16
TFTR Demonstrated Small Amount of Li ~ 20 mg/sec can Greatly Enhance Plasma Confinement and Fusion Reaction Effects of Li coating with DOLLOP on TFTR Improving Confinement with Reduce Edge Recycling D. K. Mansfield, NF 2001 NSTX-U JPS Lithium Divertor NSTX-U M. Ono J. D. Strachan, JNM 1992 Nov. 29, 2012 17
Simple 2 -D Diffusion Model of Li Transport Li diffuses radially then axially back to divertor wall Power and particle flux Divertor Entrance l^ Normalized 2 -D Li density • 0. 500 • 0. 450 Li cross field diffusion Li parallel diffusion in • 0. 400 • 0. 350 • 0. 300 l. II Li parallel diffusion out • 0. 250 • 0. 200 Divertor side wall • 0. 150 • 0. 100 l^ • 0. 050 LLD Tray • 0. 000 • 1 • 2 • 3 • 4 • 5 • 6 • 7 • 8 • 9 • 10 • 11 • 12 • 13 • 14 • 15 • 16 • 17 Li source l. II Normalized lithium flux on side wall Normalized Li flux on divertor side wall Normalized Li flux back onto LLT l • 13 -6 • 12 -5 • 11 -4 • 10 -3 • 9 -2 • 8 -1 ^ • 07 • 16 • 25 • 34 • 43 • 52 • 61 • 0. 000 NSTX-U • 0. 020 Lithium source (suppressed) • 0. 018 • 0. 016 • 0. 014 • 0. 012 • 0. 010 • 0. 008 • 0. 006 • 0. 004 • 0. 002 • 0. 000 • 0. 100 • 0. 200 • 0. 300 • 0. 400 • 0. 500 • 0. 600 JPS Lithium Divertor NSTX-U M. Ono l. II • 1 • 2 • 3 • 4 • 5 • 6 • 7 • 8 • 9 • 10 • 11 • 12 • 13 • 14 • 15 • 16 • 17 Nov. 29, 2012 18
Strong (~x 100) Li Radiation Level Over Coronal Eq. Expected in Divertor Due to Low Confinement Assumed radiation level in the modeling calculation for RLLD Radiation ~ N-Li/t, N-Li~ Li-inj t Radiation ~ Li-inj Divertor Heat and Particle Flux t ~ 100 ms Li paths Coronal-Equilibrium Value * D. Clayton et al. , JHU LLD Tray The Li radiation power per one atom and one electron in coronal-quilibrium ( net = infinity) and non-equilibrium regimes. S. V. Mirnov, et al. , Plasma Phys. Control. Fusion (2006) NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 19
Liquid Lithium (LL) as Divertor PFC Material Handling heat & paticle flux while improving plasma performance! • Low melting temperature (180°C) makes LL in natural state in reactor environment • LL is resilient against high heat flux - It can vaporize, ionize but it can be collected, renewed and recycled • LL could protect solid surface from high heat flux! - Cooling due to vaporization and ionization Radiation could provide potentially very high cooling in the divertor region due to low confinement. T. D. Rognlien and M. E. Rensink, Physics of Plasmas 9, 2120 (2002). NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 20
7. Compatibility with liquid lithium with a hot reactor first wall? ü Closed RLLD configuration permits operation at lower T < 450 °C. Edge Plasma Lower RLLD Operating Temperature: • Prevents excessive Li vaporization pressure. Core Reacting Plasma • Provides natural collection (pumping) surfaces for entire reactor chamber. 000000 Scrape Off Layer • Permits use of iron based material for substrates and structural material. Closed RLLD • Reduces Li corrosive issues. • Provides safer LL utilization. NSTX-U First Wall / Blanket At 500°C – 700°C LL In JPS Lithium Divertor NSTX-U M. Ono LL Out Flowing LLD Tray LL In 200 – 450 °C Nov. 29, 2012 21
PPPL Liquid Metal R&D for Future PFCs For NSTX-U and Future Fusion Facilities Design studies focusing on thin, capillary-restrained liquid metal layers – Combined flow-reservoir system in “soaker hose” concept – Building from high-heat flux cooling schemes developed for solid PFCs – Optimizing for size and coolant type (Helium vs. supercritical-CO 2) LL Loop Test Stand Laboratory work establishing basic technical needs for PFC R&D – Construction ongoing of LL loop at PPPL – Tests of LI flow in PFC concepts in the next year – Coolant loop for integrated testing proposed M. Jaworski et al. , PPPL NSTX-U JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 22
Surface Analysis Facilities to Elucidate Plasma-Surface Interactions In Collaboration with B. Koel et al. , Princeton University • The Surface Science and Technology Laboratory (SSTL) with three surface analysis systems and an ultrahigh vacuum deposition chamber. • The Surface Imaging and Microanalysis Laboratory (SIML) with a Thermo VG Scientific Microlab 310 -F High Performance Field Emission Auger and Multi-technique Surface Microanalysis Instrument. • Recently solid lithium and Li coated TZM were examined using X-ray photoelectron spectroscopy (XPS), temperature programmed desorption (TPD), and Auger electron spectroscopy (AES) in ultrahigh vacuum conditions and after exposure to trace gases. - Determined that lithiated PFC surfaces in tokamaks will be oxidized in about 100 s depending on the tokamak vacuum conditions. (C. H. Skinner et al. , PSI_20 submitted to J. Nucl. Mater. ) NSTX-U • X-ray photo-electron spectroscopy • Lithium coated TZM being examined by TPD and AES. JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 23
Rapid Progress is Being Made on NSTX Upgrade First Plasma Anticipated in Summer 2014 Old center stack. NEW Center Stack TF OD = 20 cm q TF OD = 40 cm TF quadrant assembled NSTX-U q 2 nd neutral beam moved into place JPS Lithium Divertor NSTX-U M. Ono Nov. 29, 2012 24
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