Low recycling and high power density handling physics in the Current Drive Experiment-Upgrade with lithium plasma-facing components

Abstract

The Current Drive Experiment-Upgrade [T. Munsat, P. C. Efthimion, B. Jones, R. Kaita, R. Majeski, D. Stutman, and G. Taylor, Phys. Plasmas 9, 480 (2002)] spherical tokamak research program has focused on lithium as a large area plasma-facing component (PFC). The energy confinement times showed a sixfold or more improvement over discharges without lithium PFCs. This was an increase of up to a factor of 3 over ITER98P(y,1) scaling [ITER Physics Basis Editors, Nucl. Fusion 39, 2137 (1999)], and reflects the largest enhancement in confinement ever seen in Ohmic plasmas. Recycling coefficients of 0.3 or below were achieved, and they are the lowest to date in magnetically confined plasmas. The effectiveness of liquid lithium in redistributing heat loads at extremely high power densities was demonstrated with an electron beam, which was used to generate lithium coatings. When directed to a lithium reservoir, evaporation occurred only after the entire volume of lithium was raised to the evaporation temperature. The ability to dissipate a beam power density of about 60MW∕m2 could have significant consequences for PFCs in burning plasma devices.

Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

“…Liquid lithium has been shown to effectively self-generate flows which can dissipate very high heat loads -up to 60 MW/m 2 for 300 seconds in tests with a small electron beam system in the CDX-U experiments. 30 Experiments with electron beam heating of liquid lithium targets at the University of Illinois have produced similar results, along with evidence that thermoelectric effects -the self-generation of flows through thermoelectric currents and the resultant J × B forces -are probably responsible. 38 CPS systems have demonstrated similar capabilities -25 MW/m 2 steady state, and greater than 50 MW/m 2 transiently, with an estimated maximum heat flux in excess of 100 MW/m 2 , although at high surface temperatures and correspondingly high evaporation rates.…”

“…The first experiment to employ a large-area liquid lithium limiter as a tokamak PFC was conducted on the CDX-U device, also at the Princeton Plasma Physics Laboratory, from 2001 through 2005. 30 The CDX-U design employed an electrically heated toroidal tray, which was filled with liquid lithium. The tray provided an exposed area of liquid lithium of 2000 cm 2 as the lower plasma limiting surface, and was typically operated at a temperature of 300 -400 °C during plasma discharges.…”

Performance projections for the lithium tokamak experiment (LTX)

“…Liquid lithium limiter experiments in CDX-U demonstrated a significant, more than five-fold enhancement of confinement, when the lithium surface was in contact with the plasma edge [1,5,6]. The evolution of a typical low recycling discharge has previously been described [3,7].…”

Simulations of NSTX with a Liquid Lithium Divertor Module

“…4(a)] drops roughly a factor of 40 over this range of recycling coefficients; the peak D α emission rate decreases by a similar factor (60). In contrast, the liqiuid lithium tray experiments on CDX-U yielded D α emission rates only about a factor of three lower than that obtained with a bare, stainless steel tray [12]. This disparity underscores the practical difficulty in preparing and maintaining a lithium surface capable of approaching the theoretical minimum recycling level.…”