Bulletin of the American Physical Society
50th Annual Meeting of the Division of Plasma Physics
Volume 53, Number 14
Monday–Friday, November 17–21, 2008; Dallas, Texas
Session CO3: NSTX and Spherical Torus |
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Chair: Stephen Knowlton, Auburn University Room: Reunion A |
Monday, November 17, 2008 2:00PM - 2:12PM |
CO3.00001: Overview of recent highlights from MAST N.J. Conway The MAST programme in support of the physics basis for ITER, DEMO and future devices based on spherical tokamaks is being facilitated by a number of recent enhancements. Twelve internal coils are now in place (6 each upper and lower, for n=3), which will be used in ELM mitigation studies at currents of up to 2 kA, and for TAE excitation and damping experiments. Magnetic perturbations from external (n=2) coils have already been used to double the ELM frequency in low collisionality H-modes. A second long-pulse PINI neutral heating beam has been installed (total beam power 5 MW). The Thomson scattering diagnostic has been upgraded with 4 more Nd:YAG lasers (8 in total, 240 Hz aggregate rate) and the number of spatial channels is being further increased to 120. A multi-chord MSE diagnostic has been constructed and installed (35 spatial channels, 2.5 cm resolution), and will be a key tool in the study of non-inductive current-drive. The present MAST campaign is focussed on exploiting our new capabilities, particularly in the areas of confinement {\&} transport, stability, ELMs, current-drive, non-inductive start-up and exhaust physics. This presentation will provide an overview of our recent results and future plans. [Preview Abstract] |
Monday, November 17, 2008 2:12PM - 2:24PM |
CO3.00002: An Overview of NSTX 2008 Results M.G. Bell Lithium, applied between discharges to the entire lower divertor, has suppressed ELMs in H-mode plasmas and produced a reduction of the density, a broadening of the temperature profiles, and an improvement in energy confinement by up to 25{\%} in NBI-heated H-mode plasmas. With lithium, the efficiency of high-harmonic fast-wave heating also improved and electron heating occurred with a directed wave spectrum peaked at k$_{\vert \vert }$ $\approx $ 3m$^{-1}$. Partial detachment of the divertor and flux expansion have mitigated the divertor heat flux. Deleterious MHD instabilities with low mode numbers were avoided by using non-axisymmetric coils to apply both n = 3 error field correction to maintain plasma rotation, and n = 1 feedback control~to suppress resonant error field amplification and to stabilize the resistive-wall mode. The transport of toroidal momentum and effects of plasma rotation on confinement have been investigated. Coaxial helicity injection was used to generate initial plasma currents which were ramped by induction to 0.7MA and achieved H-mode with NBI. Plasmas with substantial beam-driven and bootstrap current have been maintained for up 1.8s with a very low loop voltage through field error correction and increased elongation. [Preview Abstract] |
Monday, November 17, 2008 2:24PM - 2:36PM |
CO3.00003: Advanced Scenario Development on NSTX D.A. Gates, S.P. Gerhardt, H. Kugel, J.E. Menard, S.A. Sabbagh The NSTX plasma operational space has been expanded to simultaneously include the extreme shaping regime ($\kappa \sim 3$, $\delta \sim 0.8$) and $\beta_N \sim 6$, an increase from initial attempts in 2007 which achieved $\beta_N \sim 4$. The observed non-inductive current drive fraction is calculated to be $f_{BS} \sim 65 \%$. Many such discharges now extend beyond the maximum toroidal field flat-top as determined by thermal limitations, an important achievement for a device such as NSTX with severely limited transformer flux. These plasmas are demonstration discharges for proposed future spherical torus devices, such as NHTX and ST-CTF which plan to operate in this strongly shaped regime. These plasmas utilized several additional advanced control techniques. In particular, the non-axisymmetric coil set on NSTX was operated with simultaneous n=1 RFA/RWM suppression and preprogrammed n=3 error field correction. The non-axisymmetric correction fields were observed to maintain plasma rotation, thereby suppressing deleterious MHD mode activity. Lithium evaporation was used to improve plasma confinement, and may have contributed to control of the plasma density, which was observed to be lower than in similar discharges without lithium. [Preview Abstract] |
Monday, November 17, 2008 2:36PM - 2:48PM |
CO3.00004: Study of a correlation between shear Alfven activity and electron transport in NSTX D. Stutman, L. Delgado, K. Tritz, M. Finkenthal, N. Gorelenkov, E. Fredrickson, S. Kaye, E. Mazzucato A rapid increase in central (r/a$\le $0.4) electron heat transport with increasing beam power is observed in NSTX beam heated H-modes. The large gap between electron heat and ion particle diffusivity suggests electron loss along stochastic field lines. The stability analysis indicates that central thermal gradients are too weak to drive any known micro-instability. Since the only constituent having substantial gradients inside r/a$\le $0.4 are the non-thermal beam ions, we advance that it is this component that drives electron transport, through the persistent shear Alfven Eigenmode (*AE) activity they induce. Using simple beam stepping experiments a potential correlation is found between central electron transport and global Alfven (GAE) activity. Plasmas having rapid transport show also intense, broadband GAE activity, while plasma with low transport are essentially GAE free. In addition, the non-linear increase in electron transport with beam power seems correlated with a threshold in the GAE mode superposition. The initial theoretical assessment of a possible GAE/electron transport connection suggests indeed that multiple modes may induce stochastic transport of trapped electrons. [Preview Abstract] |
Monday, November 17, 2008 2:48PM - 3:00PM |
CO3.00005: Effects of Lithium-Coated Plasma-Facing Components on NSTX Discharges R. Kaita, H. Kugel Experiments on National Spherical Torus Experiment (NSTX) have demonstrated the efficacy of lithium coatings on plasma-facing components (PFC's) in providing significant and recurring improvements to plasma performance. The benefits of such coatings have been observed in both L- and H- mode divertor plasmas, with auxiliary heating from neutral beams and high-harmonic radiofrequency heating. Important upgrades in 2008 were the addition of a second LIThium EvaporatoR (LITER) to increase the PFC coating rate, and shutters that eliminated the need for the LITER's to cool to stop evaporation prior to a discharge. As in previous lithium coating experiments, there were decreases in the plasma density and ELM activity, and increases in the electron temperature, ion temperature, energy confinement time, and DD neutron rate. A notable enhancement over operation with one LITER was the larger percentage increase in pulse duration over plasmas prior to lithium evaporation. The more efficient flux consumption this suggests could be from changes in conductivity, due to broader temperature profiles from the lower edge recycling lithium-coated PFC's are expected to provide. [Preview Abstract] |
Monday, November 17, 2008 3:00PM - 3:12PM |
CO3.00006: Enhancement of Edge Stability with Lithium Wall Coatings in NSTX R. Maingi, R.E. Bell, B.P. LeBlanc, R. Kaita, S.M. Kaye, H.W. Kugel, D.K. Mansfield, T.H. Osborne ELM reduction or elimination while maintaining high confinement is essential for ITER, which has been designed for H-mode operation. Large ELMs are thought to be triggered by exceeding either edge current density and/or pressure gradient limits (peeling, ballooning modes). Stability calculations show that spherical tori should have access to higher pressure gradients and pedestal heights than higher R/a tokamaks, owing to access to second stability regimes[...1]. An ELM-free regime was recently observed in the NSTX following the application of lithium onto the graphite plasma facing components[......2]. ELMs were eliminated in phases[.....3], with the resulting pressure gradients and pedestal widths increasing substantially. Calculations with TRANSP have shown that the edge bootstrap current increased substantially, consistent with second stability access. These ELM-free discharges have a substantial improvement in energy confinement, up to the global $\beta _{N} \quad \sim $ 5.5 limit. * Supported by US DOE DE-FG02-04ER54520, DE-AC-76CH03073, and DE-FC02-04ER54698. [.1] P. B. Snyder, et. al., \textit{Plasma Phys. Contr. Fusion} \textbf{46} (2004) A131. [2] H. W. Kugel, et. al., \textit{Phys. Plasma} \textbf{15} (2008) {\#}056118. [3] D. M. Mansfield, et. al., \textit{J. Nucl. Materials} (2009) submitted. [Preview Abstract] |
Monday, November 17, 2008 3:12PM - 3:24PM |
CO3.00007: ELM Destabilization by Magnetic Perturbations at NSTX J. Canik, R. Maingi, T. Evans, T. Osborne, S.P. Gerhardt, J.-K. Park, S. Sabbagh, Z. Unterberg The destabilization of edge-localized modes (ELMs) by the application of magnetic perturbations using external coils has been observed on the National Spherical Torus Experiment. The perturbation is applied using a set of midplane coils external to the vacuum vessel, in an n=3 configuration. When the external field is applied during an otherwise ELM-free period of a discharge, ELMs begin within 50 ms, with an apparent threshold perturbation level necessary for the triggering to occur. Although calculations predict an ergodization of the edge magnetic field due to the perturbation, no strong changes in the pedestal temperature and density profile are observed. However, the toroidal rotation is reduced due to braking by the applied field. The effect is dependent on plasma shape, with triggered ELMs being larger and less frequent at lower elongation. This magnetic triggering has been used as an ELM-pacing technique to reduce impurity accumulation in the high-confinement, ELM-free H-modes that occur with lithium evaporation. [Preview Abstract] |
Monday, November 17, 2008 3:24PM - 3:36PM |
CO3.00008: Edge Localised Mode mitigation by Resonant Magnetic Perturbations on MAST Eric Nardon, Andrew Kirk ELM mitigation is essential for both ITER and a possible spherical tokamak Component Test Facility. Externally applied Resonant Magnetic Perturbations (RMPs) have given promising results at DIII-D and JET and are now considered seriously for ITER. The physics of ELM mitigation by RMPs remains, however, not fully understood. This motivated the implementation on MAST of a set of 12 in-vessel coils dedicated to ELM mitigation during the last shutdown. The coils were designed so as to be able, in the vacuum approximation (i.e. neglecting the plasma response) to ergodise the magnetic field in the region $\sim $0.8$<\psi $pol,N$<$1 ($\sim $4 times broader than the pedestal). In DIII-D, this was found to be sufficient to lead to complete ELM suppression. After having described the coils, the first experimental results will be presented, focusing not only on the impact of the RMPs on the ELM size and frequency, but also on the plasma density, temperature and rotation profiles, edge radial electric field, divertor strike zones and edge filaments, looking in particular for signs of the magnetic field ergodisation. [Preview Abstract] |
Monday, November 17, 2008 3:36PM - 3:48PM |
CO3.00009: Global Mode Stability and Active Control in NSTX S.A. Sabbagh, J.W. Berkery, J.M. Bialek, R.E. Bell, D.A. Gates, S. Gerhardt, B. LeBlanc, J. Manickam, J.E. Menard, R. Betti, B. Hu Active feedback is being used in various NSTX experiments to control mode-induced disruptions. Feedback on sensors measuring toroidal mode number n = 1 was used to control resistive wall modes (RWM) that onset at intermediate levels of plasma rotation and normalized beta near or above the ideal no-wall limit. The RWM can convert to an internal kink that either quickly damps, or leads to tearing modes that saturate or damp. In contrast, plasmas have been passively stabilized with zero rotation at the q = 2 surface, challenging the idea that RWM stability at low rotation ensures stability at higher rotation. Kinetic modification to ideal stability theory can show reduced RWM stability at intermediate plasma rotation speed using experimentally reconstructed equilibria. Rotation profiles are varied by n $>$ 1 field correction or magnetic braking. Lithium evaporation also produced unusually broad rotation profiles. Non-resonant magnetic braking was demonstrated with an n = 2 field configuration and found to be stronger in plasmas where lithium evaporation was used. [Preview Abstract] |
Monday, November 17, 2008 3:48PM - 4:00PM |
CO3.00010: Recent Results from High Harmonic Fast Wave Experiments on NSTX G. Taylor, R.E. Bell, J.C. Hosea, B.P. LeBlanc, C.K. Phillips, E.J. Valeo, J.R. Wilson, L.A. Berry, E.F. Jaeger, P.M. Ryan, J.B. Wilgen, P.T. Bonoli, J.C. Wright, R.W. Harvey, H. Yuh 30 MHz high harmonic fast wave (HHFW) heating and current drive experiments in NSTX at an axial toroidal field of 0.55 T show significantly improved core power deposition and heating efficiency at lower launched toroidal wavenumbers (k$_{\vert \vert })$ compared to operation at or below 0.45 T. In addition, lithium wall conditioning has been effectively used to reduce the edge density resulting in the first observation of HHFW core heating at k$_{\vert \vert }$ = 3 m$^{-1}$ in a deuterium plasma. Record core electron temperatures of 5 keV were reached with 3.1 MW HHFW power, and for the first time core HHFW electron heating of NBI-driven deuterium H-mode plasmas was obtained. Motional Stark effect measurements of the current driven in 0.55 T L-mode helium plasmas are consistent with predictions from AORSA and TORIC full-wave simulations. These improved HHFW heating results are attributed to moving the onset density for perpendicular fast wave propagation in the plasma further from the wall [1]. [1] J.C. Hosea, \textit{et al}., Phys. Plasmas \textbf{15}, 056104 (2008) [Preview Abstract] |
Monday, November 17, 2008 4:00PM - 4:12PM |
CO3.00011: Divertor Heat Flux Mitigation in High-Performance H-mode Plasmas in NSTX V.A. Soukhanovskii, R. Maingi, C.E. Bush, R. Raman, R. Maqueda, D.A. Gates, J.E. Menard, S.F. Paul, A.L. Roquemore, R.E. Bell, R. Kaita, H.W. Kugel, B.P. LeBlanc Divertor heat flux mitigation scenarios based on the radiative divertor and high poloidal magnetic flux expansion divertor geometry are studied in highly-shaped 1.0 - 1.2 MA, 6 MW NBI-heated H-mode discharges in NSTX. Radiative divertor performance was optimized by varying the additional divertor D$_2$ injection rate and therefore, the divertor radiated power (due to intrinsic carbon radiation) and ion momentum sink. Significant steady-state divertor peak heat flux reduction, from 8-12 MW/m$^2$ to 2- 4 MW/m$^2$ was obtained in a partially detached divertor regime with minimal core confinement degradation. In a separate experiment, the dependence of high flux expansion divertor parameters, including heat and particle fluxes, recombination rate, neutral pressure, and radiated power, on flux/area expansion factors was systematically measured by varying the X-point height and outer strike point radius. Implications of the divertor geometry for scrape-off layer power and momentum balance will be discussed using estimates from analytic 1D transport and impurity radiation modeling. Supported by the U.S. DOE under Contracts DE-AC52- 07NA27344, DE-AC02-76CH03073, DE-AC05-00OR22725, and W-7405-ENG-36. [Preview Abstract] |
Monday, November 17, 2008 4:12PM - 4:24PM |
CO3.00012: Asymmetric edge biasing for SOL control in NSTX Stewart Zweben, Ricardo Maqueda, Lane Roquemore, Charles Bush, Robert Kaita, Henry Kugel, Robert Marsala, Yevgeny Raitses, Ronald Cohen, Ryutov Dmitri In theory, the SOL strike location at a divertor plate can be actively controlled using local convective cells created by electrostatic biasing. An experiment was done on NSTX to test this idea using a small set of electrodes in the far-SOL near the outer midplane. Significant changes were observed in the local density profile near these electrodes when they were positively (but not negatively) biased with respect to the vessel ground. This was expected from the theory, as was the observation that sign of these changes reversed with the direction of the local radial ExB drift. Biasing one electrode with respect to another drew a significantly lower current, as expected, but produced a smaller density change. The SOL turbulence motion was also viewed using the GPI diagnostic; however, little convective cell motion was detected $\sim $1 m along B from these electrodes. A planned upgrade to allow the biasing of small electrodes at the divertor plate of NSTX will be described. This work was supported by USDOE Contract DE-AC02-76CHO3073. [Preview Abstract] |
Monday, November 17, 2008 4:24PM - 4:36PM |
CO3.00013: Demonstration of coupling CHI started discharges to induction in NSTX R. Raman, B.A. Nelson, T.R. Jarboe, D. Mueller, M.G. Bell, L. Roquemore, B. LeBlanc, H.W. Kugel, V. Soukhanovskii Experiments in NSTX have now unambiguously demonstrated the coupling of toroidal plasmas produced by the technique of Coaxial Helicity Injection (CHI) to inductive sustainment and ramp-up of the toroidal plasma current. In these discharges, the central Ohmic transformer with zero pre-charge was used to apply an inductive loop voltage to the decaying CHI started discharges. This resulted in an initial slower decay of the plasma current followed by a ramp up as the electron temperature increased. The coupled discharges have ramped up to $>$700kA. In addition, discharges that used 4MW of neutral beam heating, transitioned into an H-mode demonstrating compatibility of this startup method with conventional high-performance H-mode operation. The electron temperature in the coupled discharges reached over 800eV and the resulting plasma had low inductance, which is preferred for long-pulse high performance discharges. These exciting new results from NSTX in combination with the world record 160kA non-inductively generated startup currents in a ST or Tokamak previously obtained in NSTX demonstrate that CHI is a viable solenoid-free plasma startup method for future STs and Tokamaks. \textit{This work supported by U.S. DOE Contracts {\#} DE-AC02-76CH03073and }\textit{DE-FG02-99ER54519 AM08} [Preview Abstract] |
Monday, November 17, 2008 4:36PM - 4:48PM |
CO3.00014: Point-source DC helicity injection on the Pegasus toroidal experiment D.J. Battaglia, M.W. Bongard, B.A. Kujak-Ford, E.T. Hinson, B.T. Lewicki, A.J. Redd, A.C. Sontag Point-source plasma guns are used to form tokamak-like plasmas via DC helicity injection on the Pegasus Toroidal Experiment. This has been demonstrated using plasma guns mounted in the lower divertor and near the outboard midplane. The plasma guns generate high-current density ($\sim$ 0.6 kA/cm$^{2}$) plasma filaments with minimal impurity production. Under appropriate conditions, the discrete helical filaments can transition into a tokamak-like magnetic topology. Magnetic measurements suggest the tokamak-like plasma expands into the vacuum region during the current ramp, consistent with radial force equilibrium. A significant increase in the line-integrated density indicates improved particle confinement, and the line-averaged density can approach the Greenwald density limit ($>$ 1 x 10$^{19}$ m$^{-3}$). After gun shutoff, the tokamak-like plasma persists for several milliseconds. The maximum plasma current that can be created by the plasma guns is described by radial force balance, helicity balance and the requirements for Taylor relaxation. Using these limits, a simple model for point-source DC helicity injection with arbitrary plasma gun geometry is presented. [Preview Abstract] |
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