Overview of results from the National Spherical Torus Experiment (NSTX)

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 l i ∼ 0.4 with strong shaping (κ ∼ 2.7, δ ∼ 0.8) with β N 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 f NI ∼ 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.

Spectral effects on fast wave core heating and current drive

“…After extensive wall conditioning and avoidance of MHD activity that in general can degrade the overall confinement and stability and time evolution of a discharge, HHFW heating in deuterium plasmas at the higher magnetic field of 0.55 T was as successful as that in helium plasmas [3], as seen in figure 6. Indeed, central electron temperatures of 5 keV have now been achieved in both He and D plasmas at B T = 0.55 T with the application of 3.1 MW HHFW at -150º antenna phasing [19]. The rf induced change in total stored energy still dropped precipitously below -60˚ phasing for this initial phase scan in deuterium.…”

“…Shown in figure 2 are ray paths in the toroidal (a,c) and poloidal (b,d) cross sections, respectively, obtained with the GENRAY code [5,6] for waves launched with parallel wave numbers, k // , over the range accessible with the 12-strap antenna for typical OH target conditions [n e0 = 3.3×10 19 m -3 , T eo = 1.1 keV] in (a) and (b), and for the rf-heated plasma conditions [slightly hollow density profile, n e0 =2.3 × 10 19 m -3 , peak ne = 2.7×10 19 at ρ ~ 0.3, T eo = 2.8 keV] in (c) and (d). The solid dot at the end of each ray path represents the point at which at least 80% of the power launched on that ray has been absorbed.…”

Ebw Physics of Ece in NSTX

“…Relevant publications: [11,12,[15][16][17][18][19][20][21][22] By means of the coupled ray-tracing and Fokker-Planck simulations, using AMR with LUKE, we have thoroughly investigated electron Bernstein wave heating and current drive prospects for spherical tokamaks, particularly for four typical present and planned plasmas: NSTX Land H-mode, MAST Upgrade and NHTX [11,12]. Figure 5 - Figure 8 show the current drive efficiency for the NSTX L-and H-mode, for MAST Upgrade and for NHTX.…”

On the Limits of Compactness of Tokamak-Based Neutron Sources

“…В качестве обоснования приводятся экспериментальные данные, полученные на токамаке NSTX. Действительно, в [16] сообщается о поддержании вытянутости k  2,7 при значении  N ~ 5,5 в течение времени 0,5 с (около двух скиновых времен) и аспектном отношении А ~ 1,4. Дополни-тельным основанием возможностей достижения большой вытянутости плазмы в этих токамаках служат эксперименты на токамаке TCV, где ста-бильно получают вытянутость k  2,5-2,7 [17] даже при аспектном отношении А ~ 3,5.…”