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Management of tritium inventory remains one of the grand challenges in the development of fusion energy, and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER’s plasma-facing materials – Be, C, and W – and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this paper together with recommendations for ITER. Basic parameters of diffusivity, solubility, and trapping in Be, C, and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping, but long-term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be- and C-containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak, and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.
Skinner et al. (Sat,) studied this question.