Wave driven fast ion loss in the National Spherical Torus Experiment

Spherical tokamaks, with their relatively low toroidal field, extend fast-ion-driven instability physics to parameter ranges not normally accessed in conventional tokamaks. The low field means that both the fast-ion Larmor radius normalized to the plasma minor radius and the ratio of the fast-ion velocity to the Alfven speed are relatively large. The large Larmor radius of the ions enhances their interaction with instability modes, influencing the structure of the unstable mode spectrum. The relatively large fast-ion velocity allows for a larger population of fast ions to be in resonance with the mode, increasing the drive. It is therefore an important goal of the present proof-of-principle spherical tokamaks to evaluate the role of fast-ion-driven instabilities in fast-ion confinement. This paper presents the first observations of fast-ion losses resulting from toroidal Alfven eigenmodes and a new, fishbone-like, energetic particle mode

Observation of a High Performance Operating Regime with Small Edge-Localized Modes in the National Spherical Torus Experiment

We report observation of a high performance scenario in the National Spherical Torus Experiment with very small edge-localized modes (ELMs). These ELMs have no measurable impact on stored energy and are consistent with high bootstrap current operation with line average density approaching Greenwald scaling. The ELM perturbation is observed to typically originate near the lower divertor region, as opposed to the outer midplane for ELMs described in the literature. If extrapolable, this scenario would provide an attractive operating regime for next step fusion experiment

Effect of plasma shaping on performance in the National Spherical Torus Experiment

The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving Βt ∼40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ∼2.8 and triangularity δ∼0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S≡ q95 Ip (a Bt), which has been observed at large values of the S∼37 [MA (m·T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping has enabled operation with 1 s pulses with Ip =1 MA, and for 1.6 s for Ip =700 kA. Analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data are presented showing a reduction in peak divertor heat load due to increasing in flux expansion. © 2006 American Institute of Physics

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

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