Implications of PMI and wall material choice on fusion reactor tritium self-sufficiency

Journal Article · 11 December 2018 · Nuclear Materials and Energy

DOI: https://doi.org/10.1016/j.nme.2018.12.006 · OSTI ID:1494198

Doerner, R. P. ^[1]; Tynan, G. R. ^[2]; Schmid, K. ^[3]

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1. Univ. of California, San Diego, CA (United States)
2. Univ. of California, San Diego, CA (United States); Southwestern Inst. of Physics, Chengdu, Sichuan (China)
3. Max-Planck Inst. für Plasmaphysik, Garching (Germany)

Tritium self-sufficiency is a critical issue for the production of nuclear fusion energy. Here we quantify the impact of co-deposition of eroded wall material and fuel on the tritium particle balance in a hypothetical reactor system. The expected ITER plasma parameters and geometry are used to estimate the amount of eroded material from a full tungsten, beryllium or carbon device. Measured D concentrations in co-deposits are extrapolated to the wall temperature expected in future reactors and used along with these eroded flux estimates to determine the net loss probability of tritium from the device due to co-deposition. The use of liquid divertor surfaces is also considered with the amount of tritium residing in the recirculating liquid estimated. The general conclusion, from a tritium self-sufficiency viewpoint, is that one should avoid low-Z materials that readily form hydrogen bonds, in favor of high-Z non-hydride forming materials.

• Research Organization: Univ. of California, San Diego, CA (United States)
• Sponsoring Organization: USDOE Office of Science (SC)
• Grant/Contract Number: FG02-07ER54912
• OSTI ID: 1494198
• Alternate ID(s): OSTI ID: 1609930
• Journal Information: Nuclear Materials and Energy, Vol. 18, Issue C; ISSN 2352-1791
• Publisher: ElsevierCopyright Statement
• Country of Publication: United States
• Language: English