Preliminary assessment of the tritium inventory and permeation in the plasma facing components of ITER

This paper discusses preliminary quantitative predictions for the tritium inventory in- and permeation through the first-wall and divertor PFCs of ITER. The primary plasma facing material under consideration is beryllium, with possible use of tungsten or carbon fibre composites (CFCs) on high-heat-flux surfaces. We use state-of-the-art tritium transport models, in conjunction with design parameters, and loading conditions anticipated for the first-wall, baffle, limiter and divertor. The analysis includes the synergistic effects of erosion on tritium implantation and trapping, which are expected to play a key role, particularly in the divertor regions where the interaction of the plasma with the surfaces will be most severe. The influence of several key parameters that strongly affect tritium build-up and release is assessed. Finally, we discuss the uncertainties in materials properties under ITER operating conditions and the R&D needed to resolve these uncertainties.