Overview of Results from the National Spherical Torus Experiment (NSTX)
D.A. Gates, (E. Schuster), et al. (Collaboration Paper)
IAEA Fusion Energy Conference
Geneva, Switzerland, 13-18 October 2008
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Abstract
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The mission of 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 β 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 confirm the flow shear dependence of ion transport.
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 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 (κ ∼ 2.7, δ ∼ 0.8) with βN 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 are an important issue for all burning plasmas, including ITER.
Fast ions from NBI on NSTX are super- Alfvenic. Linear TAE thresholds
and appreciable fast-ion loss during multi-mode bursts are measured and
these results are compared to theory. RWM/RFA feedback combined with
n=3 error field control was used on NSTX to maintain plasma rotation
with β above the no-wall limit. The impact of n > 1 error fields on
stability is a important result for ITER. Other highlights are: results
of lithium coating experiments, demonstration of divertor heat load
mitigation in strongly shaped plasmas, and coupling of CHI plasmas to
OH ramp-up. These results advance the ST towards next step fusion
energy devices such as NHTX and ST-CTF.
