A Domestic Program for Liquid Metal PFC Research in Fusion
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ORIGINAL RESEARCH
A Domestic Program for Liquid Metal PFC Research in Fusion D. Andruczyk1 • R. Maingi2 • Chuck Kessel3 • D. Curreli1 • E. Kolemen2 • J. Canik3 • B. Pint3 D. Youchison3 • S. Smolentsev4
•
Accepted: 20 September 2020 Springer Science+Business Media, LLC, part of Springer Nature 2020
Abstract While high-Z solid plasma-facing components (PFCs) are the leading candidates for reactors, it is unclear that they can survive the intense plasma material interaction (PMI). Liquid metals (LM) PFCs offer potential solutions since they are not susceptible to the same type of damage, and can be ‘‘self-healing’’. Following the Fusion Energy System Study on Liquid Metal Plasma Facing Components study that recently was completed by Kessel et al. (Fusion Sci Technnol 75:886, 2019) a domestic LM PFC design program has been initiated to develop reactor-relevant LM PFC concepts. This program seeks to evaluate LM PFC concepts for a Fusion Nuclear Science Facility (FNSF) or a Compact Pilot Plant via engineering design calculations, modeling of PMI and PFC components and laboratory experiments. The latter involves experiments in dedicated test stands and confinement devices and seeks to identify and answer open questions in LM PFC design. The new national LM PFC program is first investigating lithium as the plasma facing material for a flowing divertor PFC concept. Several flow speeds will be evaluated, ranging from * cm/s to m/s. The surface temperature will initially be held below the strongly evaporative limit in the first design; higher temperatures with strong evaporation will be considered in future concepts. Other topics of interest include: understanding of the hydrogen and helium interaction with the liquid lithium; single effect experiments on wetting, compatibility and embrittlement; and prototypical experiments for control and characterization of flowing LM. A path to plasma and future tokamak exposure of these concepts will be developed. Keywords Plasma Fusion Liquid metal Lithium PFC Divertor
Introduction At the moment solid plasma facing components (PFCs) are the leading candidates for future fusion reactors, of which tungsten is the leading solid PFC candidate for future devices. The accepted heat flux limit for tungsten can be quite high, * 5–15 MWm-2. Tungsten also has substantial resilience to physical sputtering and little to no chemical sputtering with hydrogenic species. Finally, tritium retention in tungsten is acceptably low [1, 2]. The divertor in ITER is designed with tungsten monoblock
& D. Andruczyk [email protected] 1
University of Illinois Urbana-Champaign, Urbana, IL, USA
2
Princeton Plasma Physics Laboratory, Princeton, NJ, USA
3
Oak Ridge National Laboratory, Oak Ridge, TN, USA
4
University of California Los Angeles, Los Angeles, CA, USA
tiles, along with beryllium on the first wall [3]; the designed divertor steady heat flux limit in ITER is 10 MWm-2. Despite the attractive properties of tungsten, the fusion environ
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