Liquid lithium wall experiments in CDX-U

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. 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 ground breaking 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 ≈1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment