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Progress towards steady state on the National Spherical Torus Experiment (NSTX)

This presentation highlights the progress made towards achieving steady state operation on the NSTX, including advancements in MHD stability, transport, non-inductive startup, wave physics, and boundary physics.

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Progress towards steady state on the National Spherical Torus Experiment (NSTX)

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  1. Office of Science Supported by Progress towards steady state on the National Spherical Torus Experiment (NSTX) College W&M Columbia U Comp-X General Atomics INEL Johns Hopkins 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 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 U Tokyo JAERI Ioffe Inst RRC Kurchatov Inst TRINITI KBSI KAIST ENEA, Frascati CEA, Cadarache IPP, Jülich IPP, Garching ASCR, Czech Rep U Quebec Presented by David A. Gates* for the NSTX Group *Princeton Plasma Physics Laboratory, Princeton, NJ at MIT PSFC March 9, 2007 Cambridge, MA 1

  2. Outline • Introduction • Progress towards steady state operation • MHD stability • Transport • Non-inductive startup • Wave physics • Boundary physics 2

  3. R/a~4, k=2, qa=4 R/a~1.3, k=2, qa=12 The Spherical Torus (ST) is a Low Aspect Ratio Tokamak • ST maximizes the field line length in the good curvature (stable) region - strong toroidicity - Benefit to stability and confinement • Strong natural shaping -Enhances stability and bootstrap current 3

  4. NSTX Addresses Key Issues for Fusion Energy Development, ITER Physics and Plasma Science • Enable attractive CTF for success of DEMO: addresses unique ST development issues, synthesis of ST and tokamak results • Support USBPO and ITPA activities: can address physics in ITER physics regimes • Complement and extend toroidal confinement physics: operating regimes unique to ST, overlap with that of conventional aspect ratio tokamaks 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% 4

  5. STs Can Lead to Attractive Fusion Systems • Component Test Facility (CTF) will be needed after ITER to carry out integrated DEMO power testing and development • ST enables highly compact CTF with full remote maintenance and high duty factor, and it provides potentially attractive reactor configuration • Issue: Plasma formation and current ramp-up Peng et al, PPCF 47,B263(2005) 5

  6. AT/ST requires high bootstrap fraction, fbs, simultaneous with high t • High fbs and high t competing requirements (at fixed shape and N) • Progress for ST and advanced tokamak given by the “sustainable t” sus fbst ~ SN2, where S = q95(Ip/(aBt))] is the “shaping factor” • If Nmax = CTroyon, then only shape improves sus • sus increases linearly with increasing S [S = q95(Ip/(aBt))] • Component Test Facility requires t ~ 20% and fbs ~ 50%, sus = 10% • ARIES-ST requires t ~ 40% and fbs ~ 90% • NSTX has optimized shaping with new PF coils for high triangularity and elongation • NSTX has achieved record values of elongation and shape factor • Leads directly to record values of the sus for the ST • For NSTX 100% non-inductive operation with N ~ 7 only with strong shaping 6

  7. Continuous shaping enhancements have enabled progress towards steady state • Shaping factor has nearly doubled since 2001 - record  ~ 3 • Major upgrades include • Power supply modifications • Control system upgrades • PF Coil modifications • Real time equilibrium reconstruction for shape control  = 1.8,  = 0.6, S ~ 22  = 2.0,  = 0.8, S ~ 23  = 2.3, = 0.6, S ~ 27  = 2.75, = 0.8, S ~ 37  = 3, = 0.8, S ~ 41 2004 2006 2002-3 2005 2001 7

  8. Reduction in control system latency increases elongation (2004) Plasma elongation () plotted versus normalized internal inductance (li) • Vertical instability determines upper bound on boundary elongation  = b/a • Control system latency (time between an event and the mitigating control action) major factor in vertical controllability • Faster control system leads to better plasmas Each point in the plot represents one EFIT equilibrium in the NSTX database 8

  9. Upgrade of PF1A coil has enabled simultaneous achievement of high  and  Old PF1A Coil • Triangularity [=(Rx-R0)/a] increases edge q for given Ip, Bt by increasing length of the plasma boundary on high field side • Location of X-point (Rx,Zx) determined on NSTX by PF1A coil • Coil was modified to enable simultaneous achievement of high  and  Modified PF1A Coil 9

  10. Implementation of rtEFIT improves shape control reproducibility Boundary overlay time window • rtEFIT used for boundary control • Define locations where boundary should be, control on difference between flux at the control points and flux at the x-point • Collaboration with General Atomics - very successful • Has lead to marked improvements in reliability of shape control ~3cm includes MHD perturbations Control points Gates, PoP 2006 10

  11. Shape control modifications substantial increase in NSTX operating space Pulsed average 1/2p*tvs S (All years) Pulse averaged t versus pulse length Data are sorted by year and by S • Strong correlation between improved shaping and the achieved plasma performance • Improvements in 2006 are due to improved reliability at high shaping (rtEFIT) 11

  12. High Performance Can Be Sustained For Several Current Redistribution Times at High Non-Inductive Current Fraction • p and NBI current drive provide up to 65% of plasma current  • Relative to earlier results,High bN  H89P now sustained 2  longer TRANSP non-inductive current fractions 116313G12 D. Gates, PoP 13, 056122 (2006) 12

  13. MHD-Induced Redistribution of NBI Current Drive Contributes to NSTX “Hybrid”-Like Scenario as Proposed for ITER • Fast ion transport converts peaked JNBI to flat or hollow profile • Redistribution of NBICD makes predictions consistent with MSE qmin>1 for entire discharge, increases during late n=1 activity n=1 mode onset n=1 mode onset • High anomalous fast ion transport needed to explain neutron rate discrepancy during n=1 J. Menard, PRL 97, 095002 (2006) 13

  14. Integrated Modeling Points to Importance of Shaping, Reduced ne, and Increased Te/tE for Higher fNI and High bN • n20(0)=0.85, • =2.2 • H98=1.1 • N = 5.6 • q(0) = 1.15 • n20(0)=0.36, • =2.2 • H98=1.1 • N = 5.6 • q(0) = 1 @ 0.8 s • n20(0)=0.75, • =2.55 • H98=1.35 • bN = 6.6 • q(0) = 1.4 n(0)=0.75e20 14

  15. Fully Non-Inductive Scenario at Higher bN Requires Higher Confinement, Higher q, Strong Plasma Shaping Target Experiment (116313) • Higher k for higher q, bP, fBS • High d for improved kink stability k= 2.3, dX-L = 0.75 dRSEP = -1cm k= 2.6, dX-L = 0.85 dRSEP = -2mm • Need 60% higher T, 25% lower ne • Lithium? • higher q0 qmin 2.4 (higher with-wall limit bN< 7.2) 15

  16. Copper passive conductor plates SS Vacuum Vessel internal sensors VALEN Model of NSTX 6 ex-vessel midplane control coils Dynamic Error Field Correction (DEFC) Extended Pulse • Six EF/RWM Coils powered by “SPA” supplies with up to 6 kA-turn currents at 1 kHz • Supported by unique RWM sensors (24 BR and 24 BZ internal coils + over 150 other magnetic sensors) with 51 ch. Plasma rotation sustained after correction of intrinsic error fields J. Menard et al., IAEA 2006 16

  17. RWM Actively Stabilized at Low, ITER-Relevant Rotation • First demonstration in an ST • Plasma rotation reduced by non-resonant n=3 magnetic braking • No-wall b-limit computed by DCON • Optimize RWM control • Fully understand stabilization • physics Sabbagh et al., PRL 97 (3006) 04500 17

  18. Observed Rotation Follows Neoclassical Toroidal Viscosity (NTV)Theory Magnetic braking due to applied n=3 field • First quantitative agreement with NTV theory • Due to plasma flow through non-axisymmetric field • Trapped particle, 3-D field spectrum important • Computed using experimental equilibria • Viable physics for simulations of rotation dynamics in future devices (ITER, CTF) Zhu et al., PRL 96 (2006) 225002 18

  19. Both Internal and External Modes Can be b-Limiting in NSTX Resistive Wall Modes can limit bT at low-q (Sabbagh et al., NF, 44 [2004] 560) Discharge (in black) collapses as rotation flattens and decreases Non-linear M3D results consistent with experiment Maintaining high rotation is a key to stabilizing both modes bT (%) bT  31% bT  23% ff(0) (kHz) Mode Bq (Gauss) q0 (w/o MSE) 19

  20. NSTX Addresses Transport & Turbulence Issues Critical to Both Basic Toroidal Confinement and Future Devices • NSTX offers a novel view into plasma T&T properties • NSTX operates in a unique part of dimensionless parameter space: R/a, bT, (r*,n*) • Dominant electron heating with NBI: relevant to a-heating in ITER • Excellent laboratory in which to study electron transport: electron transport anomalous, ions close to neoclassical • Large range of bT spanning e-s to e-m turbulence regimes • Strong rotational shear that can influence transport • Localized electron-scale turbulence measurable (re ~ 0.1 mm) 20

  21. Dedicated H-mode Confinement Scaling Experiments Have Isolated the BT and Ip Dependences Scans carried out at constant density, injected power (4 MW) 0.50 s 0.50 s 21

  22. Dedicated H-mode Confinement Scaling ExperimentsHave Revealed Some Surprises Strong dependence of tE on BT Weaker dependence on Ip H98y,2 ~ 0.9 → 1.1 → 1.4 H98y,2 ~ 1.4 → 1.3 → 1.1 4 MW 4 MW tE,98y,2 ~ BT0.15 tE,98y,2 ~ Ip0.93 NSTX tE exhibits strong scaling at fixed q tE~Ip1.3-1.5 at fixed q tE,98y,2~Ip1.1 at fixed q 22

  23. Variation of Electron Transport Primarily Responsible for BT Scaling Broadening of Te & reduction in ce outside r/a=0.5 with increasing BT Ions near neoclassical Neoclassical 23

  24. Ion Transport Primarily Governs Ip Scaling- Ions Near Neoclassical Level - GTC-Neo neoclassical: includes finite banana width effects (non-local) ci,GTC-NEO (r/a=0.5-0.8) 24

  25. Microwave scattering system measures reduced fluctuations (n/n) in both upper ITG/TEM and ETG ranges during H-mode ~ ELMs Turbulence Measurements + Gyrokinetic Calculations Have Helped Identify Possible Sources of Transport Ion and electron transport change going from L- to H-modes Electron transport reduced, but remains anomalous Ion transport during H-phase is neoclassical - Localized measurement (axis to edge) - Excellent radial resolution (6 cm) 25

  26. Strongly Reversed Magnetic Shear L-mode Plasmas Achieve Higher Te and Reduced Transport Linear GS2 calculations indicate reduced region of mtearing instability for RS plasma GS2 calcs also indicate ETG stabilized by RS F. Levinton, APS 2006 26

  27. CHI Is Being Used for Solenoid-Free Startup Transient CHI: Axisymmetric reconnection leads to the formation of closed flux surfaces 27

  28. 160 kA of Closed Flux Current Produced in NSTX by Transient CHI Ip decays after ICHI→0 High Ip flux closure EFIT reconstruction • 2006 discharges operated at high RF and injector flux • Magnetic sensors and flux loops used in reconstruction • Plan to optimize high current CHI discharges, couple to OH, HHFW R. Raman et al., PRL 97 (2006) 175002 28

  29. Measured EBW Coupling Efficiency k||=3 m-1 PRF = 2 MW 30% 0.02 7 m-1 20% BRF (au) 0.01 10% 14 m-1 0% 0.1 0.2 Time (sec) EBW/HHFW Coupling Studies Being Carried Out B-X-O mode coupling understood in L-mode at fce; coupling in H-mode low HHFW htg efficiency improved at high BT and k|| L-mode 29

  30. NSTX Accesses ITER-Relevant Fast-Ion Phase-Space Regime • ITER will operate in new, small r* regime for fast ion transport • k^r ≈ 1 means "short" wavelength Alfvén modes • Fast ion transport expected from interaction of many modes • NSTX can access multi-mode regime via high bfast / btotal and vfast / vAlfven NSTX observes that multi-mode TAE bursts induce larger fast-ion losses than single-mode bursts: 1% neutron rate decrease: 5% neutron rate decrease: E. Fredrickson, Phys. Plasmas 13, 056109 (2006) 30

  31. Alfvén Cascades (RSAE) Observed at Low be on NSTX (also on MAST) Frequency chirp indicates evolution of qmin Use for q-reconstruction, MSE verification 31

  32. Angelfish Identified as Form of Hole-Clumps, Consistent with Theory • Mode satisfies Doppler-shifted resonance condition for TRANSP calculated fast ion distribution • Growth rate estimates from theory is 0.04; from observation is 0.053 • Engineering of fast-ion phase space can suppress deleterious instabilities • Berk et al., (PoP, ‘99) 32

  33. Peak Heat Flux Can Be Reduced By Plasma Shaping • Flux expansion decreases peak heat flux despite reduced major radius • Compare single-null & double-null configurations with triangularity  ≈ 0.4 at X-point and high triangularity =0.8 double-null plasmas • Measure heat flux with IR thermography of carbon divertor tiles • Peak heat flux decreases as 1 : 0.5 : 0.2 • ELM character changes: Type I  Mixed  Type V 33

  34. Peak Heat Flux Can Be Reduced With No Loss of Confinement 34

  35. Lithium Evaporator (LITER) Produced Particle Pumping and Improved Energy Confinement in H-mode Plasmas Wtot 20% higher in post-Li H98y,2=1.1→1.3 post-Li 35

  36. Edge Imaging Has Been Key to Studying Edge Turbulence Phenomena (Blobs, ELMs) ELM dynamics and rotation have been measured Excellent “blob” measurements have allowed connection to theory 36

  37. Summary • NSTX normalized performance approaching ST-CTF level • Only ST in world with advanced mode stabilization tools and diagnostics • Unique tools for understanding core and edge transport and turbulence • Uniquely able to mimic ITER fast-ion instability drive with full diagnostics • Improved understanding of HHFW and EBW coupling/heating efficiency • Demonstrated 160kA closed-flux plasma formation in NSTX using CHI • Developing understanding and unique tools for heat flux and particle control 37

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