NSTX Research Results and Plans

Office of Science Supported by NSTX Research Results and Plans College W&M Colorado Sch Mines Columbia U Comp-X FIU General Atomics INL Johns Hopkins U Lehigh U LANL LLNL Lodestar MIT Nova Photonics New York U Old Dominion U ORNL PPPL PSI Princeton U SNL Think Tank, Inc. UC Davis UC Irvine UCLA UCSD U Colorado U Maryland U Rochester U Washington U Wisconsin R. J. Hawryluk for NSTX Team IEA LT/PD ExCo Meeting June 3-4, 2008 General Atomics Culham Sci Ctr York U Chubu U Fukui U Hiroshima U Hyogo U Kyoto U Kyushu U Kyushu Tokai U NIFS Niigata U U Tokyo JAEA Ioffe Inst RRC Kurchatov Inst TRINITI KBSI KAIST POSTECH ASIPP ENEA, Frascati CEA, Cadarache IPP, Jülich IPP, Garching IPP AS CR

NSTX participation in International Tokamak Physics Activity benefits both ST and tokamak/ITER research Actively involved in 17 joint experiments – contribute/participate in 24 total Macroscopic stability • MDC-2 Joint experiments on resistive wall mode physics • MDC-3 Joint experiments on neoclassical tearing modes including error field effects • MDC-12 Non-resonant magnetic braking • MDC-13: NTM stability at low rotation Transport and Turbulence • CDB-2 Confinement scaling in ELMy H-modes: b degradation • CDB-6 Improving the condition of global ELMy H-mode and pedestal databases: Low A • CDB-9 Density profiles at low collisionality • TP-6.3 NBI-driven momentum transport study • TP-9 H-mode aspect ratio comparison Wave Particle Interactions • MDC-11 Fast ion losses and redistribution from localized Alfvén Eigenmodes Boundary Physics • PEP-6 Pedestal structure and ELM stability in DN • PEP-9 NSTX/MAST/DIII-D pedestal similarity • PEP-16 C-MOD/NSTX/MAST small ELM regime comparison • DSOL-15 Inter-machine comparison of blob characteristics • DSOL-17 Cross-machine comparison of pulse-by-pulse deposition Advanced Scenarios and Control • SSO-2.2 MHD in hybrid scenarios and effects on q-profile • MDC-14: Vertical Stability Physics and Performance Limits in Tokamaks with Highly Elongated Plasmas

NSTX contributes broadly to fundamental toroidal science in support of future STs and ITER • Macroscopic Stability • Transport and Turbulence • Waves and Energetic Particles • Boundary Physics • Scenario Integration and Control

2 Passive plates Blanket modules Port Control Coils 1 Control Coils Z(m) 0 ITER plasma boundary -1 ITER vessel 0 1 2 -2 R(m) RWM control performance using ITER-like mid-plane coil geometry is being validated with VALEN code NSTX / ITER RWM control • VALEN prediction of optimal 270 phase difference between mode BP and applied BR consistent with experiment MDC-2

25 40 20 30 15 20 10 10 5 0 0 0.9 0.9 1.0 1.0 1.1 1.1 1.2 1.2 1.3 1.3 1.4 1.4 1.5 1.5 1.6 1.6 NSTX is obtaining data to systematically test toroidal rotation damping theories for STs and ITER Rotation evolution during n = 2 braking Rotation evolution during n = 3 braking • NSTX: Radial width of flow-damping region decreases for n=2  n=3 wf(kHz) Narrower braking region DR  5cm Braking region DR  10cm MDC-12 127488 124010 R(m) R(m) • 3D field and n* dependence of flow damping critical issue for ITER RMP • Next steps: • Analyze non-resonant NTV profile, examine resonant effects, lower n* • Joint XP proposed on MAST (didn’t see strong n = 2 braking, while JET has)

NSTX has unique diagnostics and plasma regimes for studying electron turbulence and critical gradient physics High-k scattering diagnostic (Dr=3 cm) k-range of fluctuations in ETG/high-k TEM range Expt’l R/LTe HHFW heating Jenko et al., critical R/LTe for ETG (fit to GS2 numerical results) Linear GS2 analysis High fluctuation level when R/LTe is greater than critical value for ETG

NSTX is also developing a deeper understanding of angular momentum confinement – important for next-steps • Analysis of rotation evolution after magnetic braking turn-off enables separation of diffusion and pinch terms in momentum transport • Inferred cf is much higher with finite vf-pinch in fit to momentum flux Measured pinch velocity consistent with predictions of momentum transfer from turbulence from gyro-kinetic analysis. TP-6.3

NSTX accesses broad range of fast ion parameters, and a broad range of fast particle modes • Cartoon at right illustrates NSTX operational space, as well as projected operational regimes for ITER (a’s only), ST-CTF, ARIES-ST • Also shown are parameters where typical fast particle modes (FPMs) have been studied. • Conventional beam heated tokamaks typically operate with Vfast/VAlfven < 1. • CTF in avalanche regime motivates studies of fast ion redistribution • ITER with NBI also unstable to AE • Higher* of NSTX compensated by higher beam beta Cartoon is over-simplification and there are other dependences, such as q profile

NSTX finds AE avalanches can induce fast-ion redistribution and/or loss - potentially important for ITER and CTF • As power is raised, first see AE • then chirping AE • then avalanches, multi-mode transport • Avalanches are strong bursts of multiple AE modes (2 ≤ n ≤ 6) overlapping in space and frequency • Avalanchescorrelate with neutron drops indicating fast ion redistribution and/or loss 15% drop in neutron rate MDC-11

127312 900 A 127320 950 A 127319 1000 A 315 1300 A 127317 1100 A Type I ELMs can be destabilized in ELM-free discharges  need improved understanding of RMP ELM control for ITER • n=3 RMP current below ELM destabilization threshold • n=3 RMP current just above threshold (~950A/turn) • ELM frequency increases with increasing RMP current

Improved Lithium wall conditioning can mitigate ELMs, and double tE • Reproducible ELM elimination from Li • Plasma density reduced • Pulse-length extended • Power must be reduced to avoid b limit • Confinement time doubled (up to 80ms) • Dual Lithium evaporators (LITERs) now provide complete toroidal coverage of lower divertor • Improved performance vs. 1 LITER • High-performance operation with NO between-shot He glow No Li (black)With Li (red)

Combination of optimized n=3 EF correction, n=1 RFA/RWM feedback, and Li conditioning enhances NSTX performance 116313 – no Li or mode control 129125 – with Li and mode control Late n=1 rotating modes avoided Record pulse-length = 1.8s Flux consumption reduced by Li conditioning and sustained high b bN 5 sustained for 3-4 tCR EF/RWM control helps sustain rotation, high b MDC-2, 3 Transition to regime with larger, more frequent ELMs

Summary: NSTX is addressing key science issues for next-step STs and ITER • RWM control and magnetic braking theories being tested • Energy & momentum transport assessed vs. IP, BT, rotation • Alfvén Eigenmode avalanches observed to transport fast-ions • RMPs can destabilize ELMs, while Li can suppress ELMs • High performance sustained using improved MHD control + Li

NSTX has identified 4 near-term goals to address understanding and performance gaps to next-step STs Largest gaps are in collisionality and normalized density: Next-step STs are projected to have: n* 10-100x lower, ne/nGW 2-4x lower 1. Increase and understand beam-driven current at lower ne, n* 2. Increase and understand H-mode confinement at low n* 3. Demonstrate & understand non-inductive start-up & ramp-up 4. Sustain bN and understand MHD near and above no-wall limit Prioritized Goals for FY09-10:

Backup Slides

NSTX addresses key issues for plasma science, fusion energy development, and ITER Physics • Determine the physics principles of Spherical Torus (ST) confinement • Complement and extend conventional aspect ratio tokamaks • Explore attractive configurations for: • Plasma-Material Interface Testing • Component Test Facility (CTF) • Demonstration Power Plant • Support preparation of burning plasma research in ITER • Participate in the ITPA and USBPO Major Radius R0 0.85 m Aspect Ratio A 1.3 Elongation k 2.8 Triangularity d 0.8 Plasma Current Ip 1.5 MA Toroidal Field BT 0.55 T Pulse Length 1.5 s NB Heating (100 keV) 7 MW bT,tot up to 40%

PEGASUS NSTX NHTX/CTF Normal Tokamak ITER Confinement Scaling vs. b, Electron Transport NSTX accesses uniquely wide range of plasma parameters Wide range of T up to ~ 40 % Confinement scaling with wide range of T up to ~ 40 % Boundary physics with ITER-level heat flux Unique Energetic Particle Physics Multi-mode physics, *AE avalanches, full set of diagnostics: including MSE for j(r)

Ions: tE Ip Neoclassical (r/a=0.5-0.8) Electrons: tE BT High-k data 3.5 kG 5.5 kG NSTX is developing a deeper understanding of ion and electron energy transport for STs and for ITER CDB-2 • Electron & ion confinement scale differently, and different than at higher A: • Ion tE IP , electron tE BT • Ion tE IP consistent with neoclassical ion transport • High-k scattering data indicates ce correlated w/ high-k density fluctuations • Correlation holds both spatially and versus BT • Consistent with ETG at large r/a (i.e. in Te gradient region)

Vertical control experiments for ITER find maximum recoverable displacement DZMAX/a < 10% Upward shape • DZMAX/a < 10% similar to that measured at higher A ~ 3 • But, observe up/down asymmetry in recoverable DZMAX/a • 10% for upward motion and 23% for downward motion • May be result of asymmetry in conducting structure Termination point Downward shape Termination point MDC-14

Liquid Lithium Divertor (LLD) Near-term upgrades support 3 highest-priority goals • Liquid lithium divertor for pumping, and to investigate other benefits of Li: • Improved confinement • Reduction/elimination of ELMs • Compatibility of LLD with high flux expansion • Longer-term: steady-state high-heat-flux handling 2. BES to complement existing high-k scattering diagnostic • Measure full wavenumber spectrum of turbulence • Determine modes responsible for anomalous transport of energy & momentum 3. Upgrade HHFW for higher PRF+ ELM resilience • Determine if HHFW can ramp-up IP in H-mode (BS+RF overdrive) • Determine if HHFW can heat high-bN advanced H-mode scenarios • HHFW/ICRF also important for NHTX/CTF/ITER

NSTX Research Results and Plans