Analysis of Plasma-Driven Tritium Permeation Through the First Wall of DFLL-TBM in ITER

Abstract

Tritium permeation through the first wall (FW) from the plasma into helium coolant is evaluated for a dual-functional lithium-lead test blanket module (DFLL-TBM). The effect of the surface conditions on the plasma facing and coolant sides, both temperature gradient and beryllium layer clad on the plasma facing side, as well as trapping in defects on the tritium permeation is considered. The results show that most of the tritium implanted in FW re-entered the plasma. The plasma-driven tritium permeation is very sensitive to the surface conditions on the plasma facing side. With a higher sticking coefficient on the plasma-facing side, the tritium permeation into helium coolant is significantly reduced. The tritium permeation is strongly reduced with a beryllium layer clad on the front side of FW. The plasma driven tritium permeation will not seriously impact the tritium safety of DFLL-TBM. Based on tritium safety, it is reasonable to clothe the beryllium layer on FW and keep the surface clean to reduce the plasma driven tritium permeation.