A new spectroscopic diagnostic on the National Spherical Torus Experiment (NSTX) [J. Spitzer, M. Ono, et al., Fusion Technology 30, 1337] measures the velocity distribution of ions in the plasma edge simultaneously along both poloidal and toroidal views. An anisotropic ion temperature is measured during high power High Harmonic Fast Wave (HHFW) rf heating in helium plasmas, with the poloidal ion temperature roughly twice the toroidal ion temperature. Moreover, the measured spectral distribution suggests that two populations of ions are present and have temperatures of typically 500 eV and 50 eV with rotation velocities of -50 km/s and -10 km/s, respectively (predominantly perpendicular to the local magnetic field). This bi-modal distribution is observed in both the toroidal and poloidal views (for both He + and C 2+ ions), and is well correlated with the period of rf power application to the plasma. The temperature of the hot component is observed to increase with the applied rf power, which was scanned between 0 and 4.3 MW. The 30 MHz HHFW launched by the NSTX antenna is expected and observed to heat core electrons, but plasma ions do not resonate with the launched wave, which is typically at > 10 th harmonic of the ion cyclotron frequency in the region of observation. A likely ion heating mechanism is parametric decay of the launched HHFW into an Ion Bernstein Wave (IBW). The presence of the IBW in NSTX plasmas during HHFW application has been directly confirmed with probe measurements. IBW heating occurs in the perpendicular ion distribution, consistent with the toroidal and poloidal observations. Calculations of IBW propagation indicate that multiple waves could be created in the parametric decay process, and that most of the IBW power would be absorbed in the outer 10 to 20 cm of the plasma, predominantly on fully stripped ions. These predictions are in qualitative agreement with the observations, and must be accounted for when calculating the energy budget of the plasma.

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-Research on the Spherical Torus (or Spherical Tokamak) is being pursuedto explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MAclass 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 US, 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), non-inductive 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 b T of up to 35% with the near unity central b T have been obtained. NSTX will be exploring advanced regimes where b T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta-poloidal -3 -regime, where discharges with a high non-inductive fraction (~60% bootstrap current + NBI current drive) were sustained over the resistive skin time. Research on radiofrequency based heating and current drive utilizing HHFW and EBW is also pursued on NSTX, Pegasus, and CDX-U. For non-inductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500 kA. In parallel, startup using RF 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 is 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.

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