Tritium removal from the LHD first-wall by the hydrogen plasma discharge

Abstract During the 19th experimental campaign in the Large Helical Device (LHD) in 2017, deuterium (D) plasma experiments were performed for the first time in the LHD. Tritons were generated by the D-D reaction in the core plasma during the plasma discharges. To remove the tritium in plasma facing surfaces for safety in maintenance, hydrogen (H) plasma experiments were performed for one month after the four months of D plasma experiments. In this study, the removal effect of the tritium by H plasma discharges was investigated. Six stainless steel (SUS316L) samples were introduced into a position of the first-wall of the LHD in the last month of the D experiments, which were exposed to the D plasma for one month and then, three samples were continued to exposure to the following H plasma for another one month. The experimental results showed that one-half of the tritium in the surface region of the SUS sample was removed, which was measured by the TIPT (Tritium imaging plate technique). Further, hydrothermal treatments were performed and the results showed that the tritium retention in the bulk of the SUS materials was also reduced. Meanwhile, the carbon content in the near-surface region was also reduced after H plasma experiments, which reduces the possibility of the formation of the C H chemical bond.