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

• Published in: Nuclear materials and energy 2019-01, Vol.18 (C), p.56-61
• Main Authors: Doerner, R.P., Tynan, G.R., Schmid, K.
• Format: Article
• Language: English
• Subjects: Co-deposition; Nuclear Science & Technology; Plasma-material interactions; Self-sufficiency; Tritium
• ISSN: 2352-1791; 2352-1791
• DOI: 10.1016/j.nme.2018.12.006

•The impact of co-deposition of fuel with eroded wall material is estimated and its impact on tritium self-sufficiency calculated.•Tungsten appears to satisfy the conditions required for self-sufficiency.•Beryllium, carbon, silicon and boron do not appear to satisfy the self-sufficiency requirements.•Purification time of flowing liquid surfaces appears to be the figure of merit in evaluating their compatibility with self-sufficiency. 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.