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NSTX-U Status and Plan

NSTX-U. Supported by. NSTX-U Status and Plan. Culham Sci Ctr U St. Andrews York U Chubu U Fukui U Hiroshima U Hyogo U Kyoto U Kyushu 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

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NSTX-U Status and Plan

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  1. NSTX-U Supported by NSTX-U Status and Plan Culham Sci Ctr U St. Andrews York U Chubu U Fukui U Hiroshima U Hyogo U Kyoto U Kyushu 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 Columbia U CompX 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 Masayuki Ono NSTX-U Project Director PPPL, Princeton University In collaboration with the NSTX-U Team The First A3 Foresight Workshop on Spherical Torus (ST)  Jan. 14-16, 2013 SNU, Seoul, Korea

  2. Talk Outline • NSTX-U Mission • NSTX Experimental Overview • NSTX-U Construction Status • NSTX-U Experimental Plan • Summary

  3. NSTX-U Mission Elements Fusion applications of low-A spherical tokamak (ST) • Develop plasma-material-interface (PMI) solutions for next-steps • Exploit high divertor heat flux from lower-A/smaller major radius • Fusion Nuclear Science/Component Test Facility (FNSF/CTF) • Exploit high neutron wall loading for material and component development • Utilize modular configuration of ST for improved accessibility, maintenance • Extend toroidal confinement physics predictive capability • Access strong shaping, high b, vfast / vAlfvén, and rotation, to test physics models for ITER and next-steps (see NSTX, MAST, other ST presentations) • Long-term: reduced-mass/waste low-A superconducting Demo

  4. NSTX Upgrade will access next factor of two increase in performance to bridge gaps to next-step STs Low-A Power Plants ARIES-ST (A=1.6) JUST (A=1.8) VECTOR (A=2.3) • * Includes 4MW of high-harmonic fast-wave (HHFW) heating power Key issues to resolve for next-step STs Confinement scaling (electron transport) Non-inductive ramp-up and sustainment Divertor solutions for mitigating high heat flux Radiation-tolerant magnets (for Cu TF STs)

  5. NSTX Upgrade will address critical plasma confinement and sustainment questions by exploiting 2 new capabilities n / n e Greenwald Previous center-stack New center-stack ST-FNSF constant q, b, r* 2x higher BT and IP increases T, reduces n* toward ST-FNSF to better understand confinement Provides 5x longer pulses for profile equilibration, NBI ramp-up NSTX Upgrade ? ITER-like scaling Normalized e-collisionality ne*  ne /Te2 TF OD = 40cm TF OD = 20cm IP=0.95MA, H98y2=1.2, bN=5, bT = 10% BT = 1T, PNBI = 10MW, PRF = 4MW 0.95 0.72 2x higher CD efficiency from larger tangency radius RTAN 100% non-inductive CD with q(r) profile controllable by: tangency radius, density, position RTAN [cm] __________________ 50, 60, 70, 130 60, 70,120,130 70,110,120,130 New 2nd NBI Present NBI J. Menard, et al., Nucl. Fusion 52 (2012) 083015

  6. A schematic of the new center-stack and the TF joint area New TF-Flex-Bus Designed and Tested to Full Cycles TF cooling lines TF flex-bus TF Coil CHI bus PF Coil 1c PF Coil 1b CS Casing PF Coil 1a OH Coil

  7. The NSTX-U Inner TF Bundle Manufacturing Stages New Zn-Cl-Free Soldering Technique Developed

  8. NSTX-U Support Structural Upgrades 4x Electromagnetic Forces

  9. Relocation of the 2nd NBI beam line box from the TFTR test cell into the NSTX-U Test Cell.

  10. 2nd NBI alignment performed and confirmed

  11. Beam-line Component Refurbishment Ion Dump Calorimeter upgrade Bending Magnet • 11

  12. JK cap tack welded to the vacuum vessel after completing alignments, and full welding is now underway (Jan. 3, 2013)

  13. NBI Duct and Torus Vacuum Pumping System (TVPS) components being procured and fabricated Rectangular bellows Exit spool piece 40” Torus Isolation (Gate) Valve received Spool section & supports TVPS valves, hardware, TMPs, and shields Circular bellows

  14. NSTX In-Vessel View and CHI Gap Protection Enhancement Expect x 10 Higher Heat Load Into the CHI Gap CHI Gap Center Stack Secondary Passive Plates PF 1C PF 1C NBI Armor HHFW Antenna CHI Gap Primary Passive Plates CHI Gap In-board Divertor Out-board Divertor

  15. Non-inductive ramp-up from ~0.4MA to ~1MA projected to be possible with new centerstack (CS) + more tangential 2nd NBI • New CS provides higher TF (improves stability), 3-5s needed for J(r) equilibration • More tangential injection provides 3-4x higher CD at low IP: • 2x higher absorption (4080%) at low IP = 0.4MA • 1.5-2x higher current drive efficiency TSC simulation of non-inductive ramp-up from IP = 0.1MA, Te=0.5keV target at BT=1T More tangential 2nd NBI Present NBI

  16. NSTX-U CHI Start-up Configurations X 2 Higher CHI Driven Currents Expected

  17. NSTX-U ECH/EBW System for Non-Inductive Start-Up and Sustainment 28 GHz – 1MW Gyrotron by U. of Tsukuba A schematic of the NSTX-U ECH/EBW launcher

  18. Stability control improvements significantly reduce unstable RWMs at low li and high bN; improved stability at high bN/li Unstable RWM Stable / controlled RWM Resonant Field Amplification (RFA) vs. bN/li • Disruption probability reduced by a factor of 3 on controlled experiments • Reached 2 times computed n = 1 no-wall limit of bN/li = 6.7 • Lower probability of unstable RWMs at high bN/li unstable mode • Mode stability directly measured in experiments using MHD spectroscopy • Stability decreases up to bN/li = 10 • Stability increasesat higher bN/li • Presently analysis indicates consistency with kinetic resonance stabilization S.A. Sabbagh J. Berkery IAEA

  19. Disruptivity studies and warning analysis of NSTX database are being conducted for disruption avoidance in NSTX-U Disruptivity Warning Algorithms bN q* li All discharges since 2006 • Physics results • Low disruptivity at relatively high bN ~ 6; bN / bNno-wall(n=1) ~ 1.3-1.5 • Consistent with specific disruption control experiments, RFA analysis • Strong disruptivity increase for q* < 2.5 • Strong disruptivity increase for very low rotation • Results • ~ 98% disruptions flagged with at least 10ms warning, ~ 6% false positives • False positive count dominated by near-disruptive events S. Gerhardt IAEA • Disruption warning algorithm shows high probability of success • Based on combinations of single threshold based tests

  20. NSTX “Snowflake” Divertor Configuration resulted in significant divertor heat flux reduction and impurity screening Higher flux expansion (increased div wetted area) Higher divertor volume (increased div. losses) • Maintained stable “snowflake” configuration for 100-600 ms with three PF coils • Maintained H-mode confinement with core carbon reduction by 50 % • NSTX-U control coils will enable improved and up-down symmetric snowflake configurations V. Soukhanovskii, NF 2009

  21. Lithium Improved H-mode Performance in NSTX Te Broadens, tE Increases, PH Reduces, ELMs Stabilize Te broadening with lithium No lithium (129239);260mg lithium (129245) With Lithium Without Lithium H. W. Kugel, PoP 2008 tE improves with lower collisionality tE improves with lithium Pre-discharge lithium evaporation (mg) S. Kaye, IAEA (2012) R. Maingi, PRL (2011)

  22. 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 • 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. Liquid Solid M. Podesta, IAEA (2012) Reason for low lithium core dilution?: • Li is readily ionized ~ 6 eV • Li is low recycling – sticks to wall • Li has high neoclassical diffusivity F. Scotti, APS (2012)

  23. Clear reduction in NSTX divertor surface temperature and heat flux with increased lithium evaporation • a) • b) • Lithiated graphite • c) • d) T. Gray. IAEA 2012 • 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/m2 with heavy lithium evaporation

  24. Radiative Liquid Lithium Divertor Proposed Based largely on the NSTX Liquid Lithium Divertor Research Divertor Heat and Particles Flux Edge Plasma B0 000000000000 Liquid Lithium (LL) ~ 1 l/sec RLLD Core Reacting Plasma First Wall / Blanket At 500°C – 700°C Particle pumping by Li coated wall Flowing LL Particle Pumping Surfaces Li Radiative Mantle Li wall coating / condensation Scrape Off Layer Li+++ Li path Li++ Closed RLLD Li Evap. / Ionization Reduced Divertor Heat and Particle Flux Flowing LLD Tray 200 – 450 °C Li+ Heat Exchanger LL In LL In LL Out Divertor Strike Point Li0 LL Purification System to remove tritium, impurities, and dust M. Ono. IAEA 2012

  25. 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-CO2) 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 PPPL Liquid Metal R&D for Future PFCs For NSTX-U and Future Fusion Facilities Divertor Heat and Particle Flux Lithium Radiative Mantle Liquid Lithium Divertor Tray (LLDT) 200°C – 400°C Valves EM Pumps Impurities M. Jaworski et al., PPPL

  26. Draft NSTX-U Research Plan Being Formulated

  27. Draft NSTX-U Research Facility Plan Being Formulated Upgrade Outage 1.5  2 MA, 1s  5s Advanced PFCs, 5s  10-20s 0.3-0.5 MA CHI 0.5-1 MA CHI Start-up and ramp-up New center-stack Extend NBI duration or implement 2-4 MW off-axis EBW H&CD 0.2-0.4 MA plasma gun up to 1 MA plasma gun ECH/EBW 1MW 2 MW Boundary physics Diagnostics for high-Z wall studies Divertor cryo-pump Divertor Thomson U.S. FNSF conceptual design including aspect ratio and divertor optimization Materials and PFCs All High-Z PFCs Hot High-Z FW PFCs U or L Mo divertor U + L Mo divertor Li granule injector Flowing Li divertor or limiter module Full toroidal flowing Li divertor Upward LiTER Lithium MGI disruption mitigation tests Enhanced RFA/RWM sensors NCC coils NCC SPA upgrade MHD Transport & turbulence DBS, PCI or other intermediate-k High kq dB polarimetry Waves and Energetic Particles HHFW straps for EHO, *AE Dedicated EHO or *AE antenna HHFW feedthru & limiter upgrade 2nd NBI Scenarios and control Snowflake control Rotation control qmin control Control integration

  28. Summary • NSTX-U Aims to Develop Physics Understanding Needed for Designing Fusion Energy Development Facilities (ST-FNSF, ITER, DEMO, etc.) • Develop key toroidal plasma physics understanding to be tested in unexplored, hotter ST plasmas • Upgrade Project has made good progress in overcoming key design challenges • Project on schedule and budget, ~45-50% complete • Aiming for project completion in summer 2014 • Detailed NSTX-U Research Plan is being developed

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