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

The National Spherical Torus Experiment (NSTX) Research Program and Progress Towards High Beta, Long Pulse Operating Scenarios

A major research goal of the National Spherical Torus Experiment is establishing long-pulse, high-beta, high-confinement operation and its physics basis. This research has been enabled by facility capabilities developed over the last two years, including neutral-beam (up to 7 MW) and high-harmonic fast-wave heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with <beta {sub T}> up to 35%. Normalized beta values often exceed the no wall limit, and studies suggest that passive wall mode stabilization is enabling this for broad pressure profiles characteristic of H-mode plasmas. The viability of long, high bootstrap-current fraction operations has been established for ELMing H-mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fueling are likely contributing to a reduction in H-mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary-heated plasmas examined thus far. High-harmonic fast-wave (HHFW) effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is by comparing of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. A peak heat flux of 10 MW/m superscript ''2'' has been measured in the H-mode, with large asymmetries in the power deposition being observed between the inner and outer strike points. Noninductive plasma start-up studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current-drive techniques have begun