Preliminary estimates of tritium permeation and retention in the first wall of DEMO due to ion bombardment

Abstract

• Tritium losses in the DEMO first wall were investigated with T gas and ion loading. • Code comparison with TMAP and TESSIM showed good agreement in a DEMO-like scenario. • Hundreds of g of T may be retained in wall. Up to 160 mg T could permeate per day. • With bare EUROFER wall, retention below 100 g but T permeation up to several g/day. • Impact of T losses on T fuel cycle assessed with simplified particle balance model. Tritium self-sufficiency presents a critical engineering challenge for DEMO, requiring efficient breeding and extraction systems, as well as minimizing tritium losses to the surrounding systems, such as plasma-facing components, vacuum vessel, cooling system, etc. Structural and plasma-facing components will act as a tritium sink, as tritium will be accumulated in the bulk of these components due to energetic particle bombardment and may permeate out of the vacuum system. The design of the plasma-facing components will consequently directly influence the plant lifetime, operational safety and cost of any future power plant. Therefore, modeling of tritium retention and permeation in these components is required for the engineering designs of the tritium breeding and safety systems. In this work, the diffusion-transport code TESSIM-X is benchmarked against the well-established TMAP7 code and a comparison with a simplified DEMO-relevant test case is performed. The use of either code for modeling of DEMO conditions is discussed. Following this, TESSIM-X is used to provide a preliminary assessment of tritium permeation and retention in the DEMO first wall, based on the current WCLL (Water Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) breeding blanket designs.