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epr_2012_majeski_red.pdf2013-02-15 09:31:25Dick Majeski

Liquid metal walls for fusion reactors and the Lithium Tokamak eXperiment (LTX)

Author: Dick Majeski
Requested Type: Consider for Invited
Submitted: 2012-12-07 12:24:55

Co-authors: T. Abrams, D. Boyle, E. Granstedt, J. Hare, C. M. Jacobson, R. Kaita, T. Kozub, B. LeBlanc, D. P. Lundberg, M. Lucia, E. Merino, J. Schmitt, D. Stotler, PPPL, T. M. Biewer, J. M. Canik, T. K. Gray, R. Maingi, A. G. McLean, ORNL, S. Kubota, W. A. Peebles,

Contact Info:
Princeton Plasma Physics Lab
P.O. Box 451
Princeton, NJ   08543

Abstract Text:
Liquid metal walls offer the potential for eliminating several of the most vexing issues for solid (e.g. tungsten) divertor options. A liquid metal system is not subject to neutron damage. The solid substrate structure is subject to neutron damage, but not to plasma-material interactions, therefore the choice and fabrication of the substrate does not require low sputtering – or even refractory - materials. In the case of a fast flowing, or self-cooled, liquid metal wall, the substrate need not exhibit high thermal conductivity, which allows consideration of ferritic steels or silicon carbide composites, both of which are known to have high (fission) neutron tolerance, but poor heat removal characteristics. Therefore, liquid metal walls may well present the only solution for plasma-facing components in compact, high fusion power density systems. A liquid metal system would almost certainly incorporate provisions for replacement of eroded material, and for redistribution of redeposited fluid. As a result, erosion/redeposition of plasma-facing component (PFC) material is far less of an issue than for solid PFCs. In D-T, a liquid metal system would either sequester tritium for removal with the fluid, in the case of lithium, or not be subject to appreciable long-term retention of tritium, in the case of tin. With either liquid metal, the in-vessel tritium inventory is mitigated as a result. This presentation will elaborate on the rationale for developing liquid metal PFCs, and discuss the development path. The costs of deploying a full-wall liquid metal PFC in a large confinement device (of which only two, DIII-D and NSTX-U, are scheduled to be in operation in coming years) are very high. Hence experimentation with large surface area liquid metal systems seems well-suited to smaller devices. One such experiment is the Lithium Tokamak eXperiment (LTX). LTX is a low aspect ratio tokamak with R=0.4 m, a=0.26 m, and kappa=1.5. The toroidal field is 2.1 kG, plasma current less than 100 kA, and discharge duration less than 50 msec. LTX is fitted with a conformal 1 cm thick heated copper liner or shell, which covers 85% of the total plasma surface area (5 square meters). The plasma-facing surface of the shell is clad with stainless steel, and is conformal to the last closed flux surface (a close-fitting wall). The shell can be heated to maximum temperature of 350 C, and coated with lithium. In addition, a system to fill the lower segments of the shell with several hundred grams of liquid lithium has been installed and employed. The role envisioned for LTX in the development of liquid metal PFCs will be discussed, as well as the most recent results from lithium wall experiments. Results with both solid and liquid walls will be presented. LTX is a collaborative effort between PPPL and ORNL, with additional participation by UCLA, LLNL, and Johns Hopkins University. This work is supported by USDoE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

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University of Texas

Workshop on Exploratory Topics in Plasma and Fusion Research (EPR2013)
February 12-15, 2013
Fort Worth, Texas

EPR 2013