Preliminary analysis on tritium permeation through the first wall of China DFLL-TBM for ITER

The Dual-Functional Lithium-Lead Test Blanket Module (DFLL-TBM) system, which is designed to demonstrate and validate the technologies of both the helium-cooled blanket and the helium/lithium-lead dual-cooled blanket, is proposed to be tested in ITER referring to the relevant several conceptual designs in China. Fast deuterium and tritium ions, escaping from the magnetic confinement and impinging on the first wall (FW), are suspected to cause both high hydrogen isotope concentration in the near surface region and high permeation into the coolant in the FW. High tritium permeation may challenge the Objective to limit the tritium release through the cooling system to 10mg-T/yr. Consequently, plasma-driven tritium permeation through the FW is an important issue of tritium safety for fusion reactors. The tritium permeation through the FW from plasma into helium coolant has been evaluated for DFLL-TBM. The influence of surface conditions on plasma facing and coolant sides, temperature gradient, beryllium layer clad on the plasma facing side, and trapping in ion- and neutron-induced defects on the tritium permeation has been considered. The results show that most of the tritium implanted in the FW recombined from the surface and reentered the plasma. The plasma-driven tritium permeation was very sensitive to the surface conditions on plasma facing side. With a higher sticking coefficient on the front side of the FW, the tritium permeation from the plasma into helium coolant was reduced largely. Compared with the bare plasma-facing side, the tritium permeation was strongly reduced with a coating layer of beryllium on the front side of the FW, so that it can be neglected. Based on tritium safety, it's necessary to clad beryllium layer on the front side of the FW and keep the surface clean to reduce the tritium permeation through the FW. (authors)