Return To: Session FTP1 - Technology Developments
Prev Page: (FTP1/20) Development of 100 GHz Band High Power
Next Page: (FTP1/22) Operating Windows of Pebble Divertor

(FTP1/21) Liquid Lithium Experiments in CDX-U

R. Majeski¹⁾, R. Doerner²⁾, R. Kaita¹⁾, G. Antar²⁾, J. Timberlake¹⁾, J. Spaleta¹⁾, D. Hoffman¹⁾, B. Jones¹⁾, T. Munsat¹⁾, H. Kugel¹⁾, G. Taylor¹⁾, D. Stutman³⁾, V. Soukhanovskii³⁾, R. Maingi⁴⁾, S. Molesa⁵⁾, M. Ulrickson⁶⁾, P. Efthimion¹⁾, J. Menard¹⁾, M. Finkenthal³⁾, S. Luckhardt²⁾, D. G. Whyte²⁾, R. Causey⁶⁾, D. Buchaneauer^6)
 
1) Princeton Plasma Physics Laboratory, Princeton, NJ USA
²⁾ University of California at San Diego, La Jolla, CA USA
³⁾ Johns Hopkins University, Baltimore, MD USA
⁴⁾ Oak Ridge National Laboratory, Oak Ridge, TN USA
⁵⁾ Hope College, Holland, MI USA
⁶⁾ Sandia National Laboratory, Alberquerque, NM USA

Abstract.  Abstract. The initial results of experiments involving the use of liquid lithium as a plasma facing component in the Current Drive Experiment - Upgrade (CDX-U) are reported. Studies of the interaction of a steady-state plasma with liquid lithium in the Plasma Interaction with Surface and Components Experimental Simulator (PISCES-B) are also summarized. In CDX-U a solid or liquid lithium covered rail limiter was introduced as the primary limiting surface for spherical torus discharges. Deuterium recycling was observed to be reduced, but so far not eliminated, for glow discharge-cleaned lithium surfaces. Some lithium influx was observed during tokamak operation. The PISCES-B results indicate that the rates of plasma erosion of lithium can exceed predictions by an order of magnitude at elevated temperatures. Plans to extend the CDX-U experiments to large area liquid lithium toroidal belt limiters are also described.

Read the full paper in PDF format.