Progress towards Steady State at Low Aspect Ratio on the National Spherical Torus Experiment (NSTX)

Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S ≡ q95 (Iρ⁄ αΒτ)) of ∼41 (MA m−1 Τ−1) simultaneous with a record plasma elongation of κ ≡ β ⁄ α ∼ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance

Noninductive Current Generation in NSTX using Coaxial Helicity Injection

Coaxial Helicity Injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration that any CHI discharges previously produced in a Spheromak or a Spherical Torus (ST)

Beta-limiting MHD Instabilities in Improved-performance NSTX Spherical Torus Plasmas

Global magnetohydrodynamic stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized beta and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable beta. As a result of these improvements, peak beta values have reached (not simultaneously) {beta}{sub t} = 35%, {beta}{sub N} = 6.4, <{beta}{sub N}> = 4.5, {beta}{sub N}/l{sub i} = 10, and {beta}{sub P} = 1.4. High {beta}{sub P} operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 second with sustained periods of {beta}{sub N} {approx} 6 above the ideal no-wall limit and near the with-wall limit. Details of the {beta} limit scalings and {beta}-limiting instabilities in various operating regimes are described

Physics Results from the National Spherical Torus Experiment

The National Spherical Torus Experiment (NSTX) produces plasmas with aspect ratio A {triple_bond} R/a = 0.85m/0.68m {approx} 1.25, at plasma currents up to 1.5 MA with vacuum toroidal magnetic field up to 0.6 T on axis. The plasmas are heated by up to 6 MW of High-Harmonic Fast Waves (HHFW) at a frequency 30 MHz and by 7 MW of deuterium Neutral Beam Injection (NBI) at an energy up to 100 keV. Since January 2004, NSTX has been operating, routinely at toroidal fields up to 0.45 T, with a new central conductor bundle in the toroidal field coil

Characterization of the plasma current quench during disruptions in the National Spherical Torus Experiment

A detailed analysis of the plasma current quench in the National Spherical Torus Experiment [M.Ono, et al Nuclear Fusion 40, 557 (2000)] is presented. The fastest current quenches are fit better by a linear waveform than an exponential one. Area-normalized current quench times down to .4 msec/m2 have been observed, compared to the minimum of 1.7 msec/m2 recommendation based on conventional aspect ratio tokamaks; as noted in previous ITPA studies, the difference can be explained by the reduced self-inductance at low aspect ratio and high-elongation. The maximum instantaneous dIp/dt is often many times larger than the mean quench rate, and the plasma current before the disruption is often substantially less than the flat-top value. The poloidal field time-derivative during the disruption, which is directly responsible for driving eddy currents, has been recorded at various locations around the vessel. The Ip quench rate, plasma motion, and magnetic geometry all play important roles in determining the rate of poloidal field change

Accounting of the Power Balance for Neutral-beam-heated H-Mode Plasmas in NSTX

A survey of the dependence of power balance on input power, shape, and plasma current was conducted for neutral-beam-heated plasmas in the National Spherical Torus Experiment (NSTX). Measurements of heat to the divertor strike plates and divertor and core radiation were taken over a wide range of plasma conditions. The different conditions were obtained by inducing a L-mode to H-mode transition, changing the divertor configuration [lower single null (LSN) vs. double-null (DND)] and conducting a NBI power scan in H-mode. 60-70% of the net input power is accounted for in the LSN discharges with 20% of power lost as fast ions, 30-45% incident on the divertor plates, up to 10% radiated in the core, and about 12% radiated in the divertor. In contrast, the power accountability in DND is 85-90%. A comparison of DND and LSN data show that the remaining power in the LSN is likely to be directed to the upper diverto

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

Introduction Edge characterization experiments have been conducted in NSTX to provide an initial survey of the edge particle and heat fluxes and their scaling with input power and electron density. The experiments also provided a database of conditions for the analyses of the NSTX global particle sources, core fueling, and divertor operating regimes

A Simple Apparatus for the Injection of Lithium Aerosol into the Scrape-Off Layer of Fusion Research Devices

A simple device has been developed to deposit elemental lithium onto plasma facing components in the National Spherical Torus Experiment. Deposition is accomplished by dropping lithium powder into the plasma column. Once introduced, lithium particles quickly become entrained in scrape-off layer flow as an evaporating aerosol. Particles are delivered through a small central aperture in a computer-controlled resonating piezoelectric disk on which the powder is supported. The device has been used to deposit lithium both during discharges as well as prior to plasma breakdown. Clear improvements to plasma performance have been demonstrated. The use of this apparatus provides flexibility in the amount and timing of lithium deposition and, therefore, may benefit future fusion research devices