Edge Recycling and Heat Fluxes in L- and H-mode NSTX Plasmas PDF Download

Author: V. A. Soukhanovskii
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Category : National Spherical Torus Experiment (Project)
Languages : en
Pages : 4

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Author: V. A. Soukhanovskii
Publisher:
ISBN:
Category : National Spherical Torus Experiment (Project).
Languages : en
Pages : 4

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Author: M. H. Redi
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Category : National Spherical Torus Experiment (Project).
Languages : en
Pages : 9

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Author: C. E. Bush
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Category : National Spherical Torus Experiment (Project).
Languages : en
Pages : 34

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Author: V. A. Soukhanovskii
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ISBN:
Category : National Spherical Torus Experiment (Project)
Languages : en
Pages : 6

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Author: R. E. Bell
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Languages : en
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The launch of High-Harmonic Fast Waves (HHFW) routinely provides auxiliary power to NSTX plasmas, where it is used to heat electrons and pursue drive current. H-mode transitions have been observed in deuterium discharges, where only HHFW and ohmic heating, and no neutral beam injection (NBI), were applied to the plasma. The usual H-mode signatures are observed. A drop of the Da light marks the start of a stored energy increase, which can double the energy content. These H-mode plasmas also have the expected kinetic profile signatures with steep edge density and electron temperature pedestal. Similar to its NBI driven counterpart--also observed on NSTX-- the HHFW H mode have density profiles that features ''ears'' in the peripheral region. These plasmas are likely candidates for long pulse operation because of the combination of bootstrap current, associated with H-mode kinetic profiles, and active current drive, which can be generated with HHFW power.

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Languages : en
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Lithium as a plasma-facing material has many attractive features, including a reduction in the recycling of hydrogenic species and the potential for withstanding high heat and neutron fluxes in fusion reactors. Recent NSTX experiments have shown, for the first time, significant and recurring benefits of lithium coatings on plasma-facing components (PFC's) to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. They included decreases in the plasma density and inductive flux consumption, and increases in the electron temperature, ion temperature, energy confinement time, and DD neutron rate. Extended periods of MHD quiescence were also achieved, and measurements of the visible emission from the lower divertor showed a reduction in the deuterium, carbon, and oxygen line emission. Other salient results with lithium evaporation included a broadening of the electron temperature profile, and changes in edge density gradients that benefited electron Bernstein wave coupling. There was also a reduction in ELM frequency and amplitude, followed by a period of complete ELM suppression. In general, it was observed that both the best and the average confinement occurred after lithium deposition and that the increase in WMHD occurs mostly through an increase in We. In addition, a liquid lithium divertor (LLD) is being installed on NSTX this year. As the first fully-toroidal liquid metal divertor target, experiments with the LLD can provide insight into the behavior of metallic ITER PFC's should they liquefy during high-power divertor tokamak operations. The NSTX lithium coating and LLD experiments are important near-term steps in demonstrating the potential of liquid lithium as a solution to the first-wall problem for both magnetic and inertial fusion reactors.

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Languages : en
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30 MHz high-harmonic fast wave (HHFW) heating and current drive are being developed to assist fully non-inductive plasma current (I{sub p}) ramp-up in NSTX. The initial approach to achieving this goal has been to heat I{sub p} = 300 kA inductive plasmas with current drive antenna phasing in order to generate an HHFW H-mode with significant bootstrap and RF-driven current. Recent experiments, using only 1.4 MW of RF power (P{sub RF}), achieved a noninductive current fraction, f{sub NI} ≈ 0.65. Improved antenna conditioning resulted in the generation of I{sub p} = 650 kA HHFW H-mode plasmas, with f{sub NI} ≈ 0.35, when P{sub RF} ≥ 2.5 MW. These plasmas have little or no edge localized mode (ELM) activity during HHFW heating, a substantial increase in stored energy and a sustained central electron temperature of 5-6 keV. Another focus of NSTX HHFW research is to heat an H-mode generated by 90 keV neutral beam injection (NBI). Improved HHFW coupling to NBI-generated H-modes has resulted in a broad increase in electron temperature profile when HHFW heating is applied. Analysis of a closely matched pair of NBI and HHFW+NBI H-mode plasmas revealed that about half of the antenna power is deposited inside the last closed flux surface (LCFS). Of the power damped inside the LCFS about two-thirds is absorbed directly by electrons and one-third accelerates fast-ions that are mostly promptly lost from the plasma. At longer toroidal launch wavelengths, HHFW+NBI H-mode plasmas can have an RF power flow to the divertor outside the LCFS that significantly reduces RF power deposition to the core. ELMs can also reduce RF power deposition to the core and increase power deposition to the edge. Recent full wave modeling of NSTX HHFW+NBI H-mode plasmas, with the model extended to the vessel wall, predicts a coaxial standing mode between the LCFS and the wall that can have large amplitudes at longer launch wavelengths. These simulation results qualitatively agree with HHFW+NBI H-mode data that show decreasing core RF heating efficiency and increasing RF power flow to the lower divertor at longer launch wavelengths.

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Languages : en
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Lithium as a plasma-facing material has attractive features, including a reduction in the recycling of hydrogenic species and the potential for withstanding high heat and neutron fluxes in fusion reactors. Dramatic effects on plasma performance with lithium-coated plasma-facing components (PFCOs) have been demonstrated on many fusion devices, including TFTR, [1] T-11M, [2] and FT-U. [3] Using a liquid-lithium-filled tray as a limiter, the CDX-U device achieved very significant enhancement in the confinement time of ohmically heated plasmas. [4] The recent NSTX experiments reported here have demonstrated, for the first time, significant and recurring benefits of lithium PFC coatings on divertor plasma performance in both L- and H- mode regimes heated by neutral beams.

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Languages : en
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The confinement of auxiliary heated NSTX discharges is discussed. The energy confinement time in plasmas with either L-mode or H-mode edges is enhanced over the values given by the ITER97L and ITER98Pby(2) scalings, being up to 2-3 times L-mode and 1.5 times H-mode. TRANSP calculations based on the kinetic profile measurements reproduce the magnetics-based determination of stored energy and the measured neutron production rate. Power balance calculations reveal that, in a high power neutral beam heated H-mode discharge, the ion thermal transport is near neoclassical levels, and well below the electron thermal transport, which is the main loss channel. Perturbative impurity injection techniques indicate the particle diffusivity to be slightly above the neoclassical level in discharges with L-mode edge. High-harmonic fast-wave (HHFW) bulk electron heating is described and thermal transport is discussed. Thermal ion transport is found to be above neoclassical level, but thermal electron transport remains the main loss mechanism. Evidences of an electron thermal internal transport barrier obtained with HHFW heating are presented. A description of H-mode discharges obtained during HHFW heating is presented.