Overview of results from the National Spherical Torus Experiment (NSTX)
D.A. Gates, ..., E. Schuster, et al. (Collaboration Paper)
Nuclear Fusion 49 (2009) 104016 (14pp)
Abstract
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The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate
to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component
test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high beta operation. To better
understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale
fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the
flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium
wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved
coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface
wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation.
Plasmas have been maintained on NSTX at very low internal inductance li ~ 0.4 with strong shaping (kappa ~ 2.7, delta ~ 0.8)
with betaN approaching the with-wall beta-limit for several energy confinement times. By operating at lower collisionality in this
regime, NSTX has achieved record non-inductive current drive fraction fNI ~ 71%. Instabilities driven by super-Alfvenic ions will
be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvenic. Linear toroidal
Alfven eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared
with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field
amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with beta above the
no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width
scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection
plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.