Liquid Lithium Wall Experiments in CDX-U
R. Kaita, R. Majeski, S. Luckhardt, R. Doerner, M. Finkenthal, H. Ji, H. Kugel, D. Mansfield, D. Stutman, R. Woolley, S. Zakharov, and S. Zweben
Date of PPPL Report: October 1999
Presented at: The 18th IEEE/NPSS (Institute for Electrical and Electronics Engineers/Nuclear and Plasma Science Society) Symposium on Fusion Engineering (SOFE) which was held in Albuquerque, NM, on October 25-29, 1999.
The concept of a flowing lithium first wall for a fuison reactor may lead to a significant advance in reactor design, since it could virutally eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new groundbreaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only approximately 1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment.