Preliminary assessment of tritium permeation and retention in the European Water Cooled Lithium Lead Test Blanket Module with TESSIM-X

Abstract

• T inventory in the ITER EU Water-Cooled Lithium-Lead Test Blanket Module investigated. • T from vacuum vessel (gas, neutrals) and solute T in PbLi included in modeling. • Effect of neutron damaging on trap concentrations included in models. • T retention in the range of 0.75 g is expected over plasma operation period. • Comparison with the TMAP7 code performed. Tritium self-sufficiency is a key requirement for future fusion power plants. Therefore, it will be necessary to minimize tritium losses to the surrounding systems, such as the tritium breeding modules, plasma-facing components, cooling system, etc. These components and systems will act also as a tritium sink, as tritium will be retained in the metal walls in traps produced during manufacturing or induced by neutron irradiation. The design of the tritium breeding systems and plasma-facing components will therefore have a direct impact on the performance, operational safety and cost of any future power plant. Consequently, accurate modeling of tritium transport and retention in these systems is needed for future engineering designs of the tritium breeding systems and supporting safety requirements. In this work, tritium permeation and retention was modeled with the diffusion-transport code TESSIM-X for the current European WCLL (Water Cooled Lithium Lead) Test Blanket Module (TBM) for the ITER TBM Programme, with the inclusion of neutron-induced traps. Trap saturation is considered for dpa values of 0.25 and above, leading to an increase in trap concentrations relative to undamaged material of roughly a factor 3. The code was also benchmarked against the well-established TMAP7 code.