Surface Treatment of a Lithium Limiter for Spherical Torus Plasma Experiments
Authors: R. Kaita, R. Majeski, R. Doerner, G. Antar, J. Timberlake, J. Spaleta, D. Hoffman, B. Jones, T. Munsat, H. Kugel, G. Taylor, D. Stutman, V. Soukhanovskii, R. Maingi, S. Molesa, P. Efthimion, Menard, M. Finkenthal, and S. Luckhardt
Date of PPPL Report: March 2001
Submitted to: Fusion Engineering and Design
The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. As part of investigations to determine the feasibility of this approach, plasma interaction questions in a toroidal plasma geometry are being addressed in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The first experiments involved a toroidally local lithium limiter (L3). Measurements of pumpout rates indicated that deuterium pumping was greater for the L3 compared to conventional boron carbide limiters. The difference in the pumpout rates between the two limiter types decreased with plasma exposure, but argon glow discharge cleaning was able to restore the pumping effectiveness of the L3. At no point, however, was the extremely low recycling regime reported in previous lithium experiments achieved. This may be due to the much larger lithium surfaces that were exposed to the plasma in the earlier work. The possibility will be studied in the next set of CDX-U experiments, which are to be conducted with a large area, fully toroidal lithium limiter.