R. Woolley

TL;DR: The National Spherical Torus eXperiment (NSTX) is a MA-class spherical tokamak (ST) facility in the US actively developing the physics basis for an ST-based FNSF as discussed by the authors.

TL;DR: Menard et al. as mentioned in this paper developed a blanket system capable of tritium breeding ratio TBR ≈ 1, a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, and a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard field coils outside the vacuum chamber and superconducting to reduce power consumption.

TL;DR: Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory as discussed by the authors, and the results showed significant reduction in both recycling and in impurity levels.

TL;DR: In this paper, the magnetohydrodynamics (MHD) features associated with a free surface flow in a fusion-relevant magnetic field environment, and determine what LM-free surface flow option is most suitable for lithium divertor particle pumping and surface heat removal applications in a near-term experimental plasma device, such as NSTX.

TL;DR: In this article, a next-step spherical torus (NSST) device is proposed to provide a sufficient physical basis for the design of a compact component test facility (CTF).