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

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) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} 500 kA. In parallel, start-up using radio-frequency current drive and only external poloidal field coils are being developed on NSTX. 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. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor