Observations of the relationship between ionospheric central polar cap and dayside throat convection velocities, and solar wind/IMF driving

Convection observations from the Southern Hemisphere Super Dual Auroral Radar Network are presented and examined for their relationship to solar wind and interplanetary magnetic field (IMF) conditions, restricted to periods of steady IMF. Analysis is concentrated on two specific regions, the central polar cap and the dayside throat region. An example time series is discussed in detail with specific examples of apparent direct control of the convection velocity by the solar wind driver. Closer examination, however, shows that there is variability in the flows that cannot be explained by the driving. Scatterplots and histograms of observations from all periods in the year 2013 that met the selection criteria are given and their dependence on solar wind driving is examined. It is found that on average the flow velocity depends on the square root of the rate of flux entry to the polar cap. It is also found that there is a large level of variability that is not strongly related to the solar wind driving

Investigations of Low and Moderate Harmonic Fast Wave Physics on CDX-U

Third harmonic hydrogen cyclotron fast wave heating studies are planned in the near term on CDX-U to investigate the potential for bulk ion heating. In preparation for these studies, the available radio-frequency power in CDX-U has been increased to 0.5 MW. The operating frequency of the CDX-U radio-frequency transmitter was lowered to operate in the range of 8-10 MHz, providing access to the ion harmonic range 2* {approx} 4* in hydrogen. A similar regime is accessible for the 30 MHz radio-frequency system on the National Spherical Torus Experiment (NSTX), at 0.6 Tesla in hydrogen. Preliminary computational studies over the plasma regimes of interest for NSTX and CDX-U indicate the possibility of strong localized absorption on bulk ion species

Performance projections for the lithium tokamak experiment (LTX)

Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized

Enhanced energy confinement and performance in a low-recycling tokamak,

ABSTRACT Extensive lithium wall coatings and liquid lithium plasma-limiting surfaces reduce recycling, with dramatic improvements in ohmic plasma discharges in the Current Drive eXperimentUpgrade (CDX-U). Global energy confinement times increase by up to 6×. These results exceed confinement scalings such as ITER98P(y,1) by 2-3×, and represent the largest increase in confinement ever observed for an ohmic tokamak plasma

Plasma Performance Improvements with Liquid Lithium Limiters in CDX-U

The use of flowing liquid lithium as a first wall for a reactor has potentially attractive physics and engineering features. The Current Drive experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has begun experiments with a fully toroidal liquid lithium limiter. CDX-U is a compact [R = 34 cm, a = 22 cm, Btoroidal = 2 kG, IP =100 kA, T(subscript)e(0) {approx} 100 eV, n(subscript)e(0) {approx} 5 x 10{sup 19} m-3] short-pulse (<25 msec) spherical tokamak with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, can be filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium prior to the experiment. The total area of the tray is approximately 2000 cm{sup 2}. The tokamak edge plasma, when operated in contact with the lithium-filled tray, shows evidence of reduced impurities and recycling. The reduction in re cycling and impurities is largest when the lithium is liquefied by heating to 250 degrees Celsius. Discharges which are limited by the liquid lithium tray show evidence of performance enhancement. Radiated power is reduced and there is spectroscopic evidence for increases in the core electron temperature. Furthermore, the use of a liquid lithium limiter reduces the need for conditioning discharges prior to high current operation. The future development path for liquid lithium limiter systems in CDX-U is also discussed

Results from the CDX-U Lithium Wall and NSTX Lithium Pallet Injection and Evaporation Experiments

CDX-U has been operated with the vacuum vessel wall and limiter surfaces nearly completely coated with lithium, producing dramatic improvements to plasma performance. Discharges achieved global energy confinement times up to 6 ms, exceeding previous CDX-U results by a factor of 5, and ITER98P(y,1) scaling by 2-3. Lithium wall coatings up to 1000 {angstrom} thick were applied between discharges by electron-beam-induced evaporation of a lithium-filled limiter and vapor deposition from a resistively heated oven. The e-beam power was modest (1.6 kW) but it produced up to 60 MW/m2 power density in a 0.3 cm{sup 2} spot; the duration was up to 300 s. Convective transport of heat away from the beam spot was so effective that the entire lithium inventory (140 g) was heated to evaporation (400-500 C) and there was no observable hot spot on the lithium surface within the beam footprint. These results are promising for use of lithium plasma-facing components in reactor scale devices. Lithium coating has also been applied to NSTX carbon plasma-facing surfaces, to control the density rise during long-duration H-modes for non-inductive current sustainment. First, lithium pellets were injected into sequences of Ohmically heated helium plasmas in both center stack limiter (CSL) and lower single-null divertor (LSND) configurations to deposit a total of 25-30 mg of lithium on the respective plasma contact areas. In both cases, the first subsequent L mode, deuterium discharge with NBI showed a reduction in the volume-average density by a factor {approx}3 compared to similar discharges before the lithium coating. Recently, a lithium evaporator was installed aimed toward the graphite tiles of the lower center stack and divertor. Twelve depositions, ranging from about 10 mg to 5 g of lithium, were performed. The effects on LSND L-mode, double-null divertor (DND) H-mode, and DND reversed-shear plasmas were variable but, immediately after coating, there were decreases in the density and significant increases in electron and ion temperature, neutron rate, confinement time, and edge flow velocity, and reductions in H-mode ELM frequency. For several days of operation after lithium coating, the ratio of oxygen to carbon emission was lower than with boronization

Progress towards high-performance, steady-state spherical torus

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structue fro that in more moderate aspect ratio devices. The ST experiments are being conducted in various US research facilities. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks