ANALYSIS OF TRITIUM/DEUTERIUM RETENTION AND PERMEATION IN FW/DIVERTOR INCLUDING GEOMETRIC AND TEMPERATURE OPERATING FEATURES

Available data and mathematical formulations concerning tritium transport in the FW/Divertor with tungsten and beryllium as plasma facing materials were implemented in the commercial code COMSOL Multiphysics. The goal is to develop a CAD-based multiphysics modeling capability so that FW/Divertor temperature and geometric features can be readily taken into consideration while tritium permeation to the primary coolant in a prototypical PFC can be more realistically addressed. This development began with the simulation of ion implantation experiments, validated against existing laboratory experimental results. Analysis shows that with ITER FW where Be is used as the plasma facing material, the low operating temperature, erosion, and the dwell time greatly hinder tritium bulk diffusion, permeation, and inventory accumulation. However, under DEMO high-temperature operating conditions, tritium can quickly diffuse through tungsten to structural material and reach a steady state inventory after a relatively short time. Additionally, its permeation to the coolant can be reduced when the Soret effect is considered. The findings and challenges of developing a 3-D predictive capability for tritium transport in a FW/Divertor PFC are discussed