Liquid Lithium Limiter Experiments in CDX-U
Authors: R. Majeski, S. Jardin, R. Kaita, T. Gray, P. Marfuta, J. Spaleta, J. Timberlake, L. Zakharov, G. Antar, R. Doerner, S. Luckhardt, R. Seraydarian, V. Soukhanovskii, R. Maingi, M. Finkenthal, D. Stutman, and D. Rodgers
Date of PPPL Report: October 2004
Presented at: the 20th IAEA Fusion Energy Conference, 1-6 November 2004, Vilamoura, Portugal. The papers will be published by the IAEA as unedited proceedings in electronic format on CD-ROM and on the IAEA Physics Section web site as soon as possible after the conference. To be submitted to Nuclear Fusion.
Recent experiments in the Current Drive eXperiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, Btoroidal = 2 kG, IP = 100 kA, Te(0) ~ 100 eV, ne(0) ~ 5 x 1019 m-3) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Zeffective < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced.