Edge transport barrier (ETB) studies on the Alcator C-Mod tokamak [Phys. Plasmas 1, 1511, (1994] investigate pedestal scalings and radial transport of plasma and neutrals.Pedestal profiles show trends with plasma operational parameters such as total current I P .A ballooning-like I 2 P dependence is seen in the pressure gradient, despite calculated stability to ideal ballooning modes. A similar scaling is seen in the near scrape-off-layer for both low-confinement (L-mode) and H-mode discharges, possibly due to electromagnetic fluid drift turbulence setting transport near the separatrix. Neutral density diagnosis allows examination of D 0 fueling in H-modes, yielding profiles of effective particle diffusivity in the ETB, which vary as I P is changed. Edge neutral transport is studied using a 1D kinetic treatment. In both experiment and modeling, the C-Mod density pedestal exhibits a weakly increasing pedestal 1

Reduction of core-resonant mϭ1 magnetic fluctuations and improved confinement in the Madison Symmetric Torus ͓Dexter et al., Fusion Technol. 19, 131 ͑1991͔͒ reversed-field pinch have been routinely achieved through control of the surface poloidal electric field, but it is now known that the achieved confinement has been limited in part by edge-resonant mϭ0 magnetic fluctuations. Now, through refined poloidal electric field control, plus control of the toroidal electric field, it is possible to reduce simultaneously the mϭ0 and mϭ1 fluctuations. This has allowed confinement of high-energy runaway electrons, possibly indicative of flux-surface restoration in the usually stochastic plasma core. The electron temperature profile steepens in the outer region of the plasma, and the central electron temperature increases substantially, reaching nearly 1.3 keV at high toroidal plasma current ͑500 kA͒. At low current ͑200 kA͒, the total beta reaches 15% with an estimated energy confinement time of 10 ms, a tenfold increase over the standard value which for the first time substantially exceeds the constant-beta confinement scaling that has characterized most reversed-field-pinch plasmas.

A power-balance model, with radiation losses from impurities and neutrals, gives a unified description of the density limit (DL) of the stellarator, the L-mode tokamak, and the reversed field pinch (RFP). The model predicts a Sudo-like scaling for the stellarator, a Greenwald-like scaling, , for the RFP and the ohmic tokamak, a mixed scaling, , for the additionally heated L-mode tokamak. In a previous paper (Zanca et al 2017 Nucl. Fusion 57 056010) the model was compared with ohmic tokamak, RFP and stellarator experiments. Here, we address the issue of the DL dependence on heating power in the L-mode tokamak. Experimental data from high-density disrupted L-mode discharges performed at JET, as well as in other machines, are taken as a term of comparison. The model fits the observed maximum densities better than the pure Greenwald limit.

Abstract. High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects.