Abstract
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 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 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 (κ ∼ 2.7, δ ∼ 0.8) with βN approaching the with-wall β-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-Alfvénic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvénic. Linear toroidal Alfvén 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 β 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.
Highlights
The spherical torus (ST) concept [1] has been proposed as a potential fusion reactor [2] as well as a component test facility (ST-CTF) [3]
Substantial progress has been made towards achieving the primary mission of National Spherical Torus Experiment (NSTX), which is to understand and utilize the advantages of the ST configuration by establishing attractive ST steady-state operating scenarios and configurations at high β
The scaling of Scrape-off layer (SOL) fluxes has been measured on NSTX, an extremely important issue for future ST devices, such as the proposed NHTX [53]
Summary
The spherical torus (ST) concept [1] has been proposed as a potential fusion reactor [2] as well as a component test facility (ST-CTF) [3]. As an additional mission element, NSTX exploits its unique capabilities to complement the established tokamak database and thereby supports ITER by expanding the breadth of the range of operating parameters such as lower A, very high β, high vfast/vAlfven, as well as other important plasma parameters. This broader range of experience helps clarify uncertainties in extrapolating to ITER by removing existing degeneracies in physics scalings.
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