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• Tritium existence in the LHD first-wall was confirmed. • More than one-half of tritium can be removed by H plasma discharges. • Tritium amounts in the bulk region of the SUS materials were reduced. 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. [ABSTRACT FROM AUTHOR]

Tungsten (W) is one of the most promising candidates for plasma facing materials in the fusion reactor. Helium (He) in W can influence the retention of hydrogen isotopes. In the present study, W targets were simultaneously irradiated by He+-H 2 + or He+-D 2 + ion beams with the energies of 1 keV or 3 keV, at fixed temperatures in a range of room temperature (R.T.) to 1073 K. Mechanisms of He and hydrogen interaction in W were discussed, especially from the point of He retention, which was characterized by the high-temperature thermal desorption spectroscopy (TDS) and glow-discharge optical emission spectroscopy (GD-OES) measurements. It is found that He desorption shifts to a lower temperature range for the W simultaneously irradiated by 3 keV He+-1 keV H 2 + at 573 K, under He+ fluence up to 1 × 1022 He+m−2. Transmission electron microscope (TEM) observation and annealing treatment at the temperature of 873–1073 K show that the increased He uptake is caused by the formation of dislocation. Enhanced retention amounts for the hydrogen isotopes were also confirmed. Amounts of the dislocation loops introduced by the H 2 +-only irradiation can be reduced by annealing treatment at 873 K, while that introduced by He+ irradiation are quite stable, which grows larger at elevated temperatures. With an increase of H 2 + energy, Helium uptakes at both weak trapping sites and bubbles are increased, while the amounts of hydrogen retention are decreased. It suggests that hydrogen ion has a significant influence on the He trapping sites at the irradiation temperature up to 573 K, while the hydrogen retention is determined by the distribution of He bubbles and dislocation loops. • Helium was mainly trapped by dislocation loops in tungsten irradiated by 3 keV He+- 1 keV H 2 + at 573 K. • Helium concentration in tungsten under 3 keV He+- 1 keV H 2 + irradiation was decreased at 873 K. • Increased hydrogen isotope uptake in tungsten was found when helium was mainly trapped at dislocation loops. [ABSTRACT FROM AUTHOR]

Co-deposition of wall material and ions from plasmas occurs on material surfaces in fusion devices. In this study, we focus on tungsten (W) deposition layer in helium (He) plasmas (He-W co-deposition) formed at surface temperatures of 473–773 K. Usually, He irradiation at the surface temperature below 1000 K does not lead to significant morphology changes; we will show from transmission electron microscope observations that the roughness of the co-deposition layer has a significant temperature dependence even when the temperature is less than 800 K. We conducted deuterium plasma irradiation to the He-W co-deposition layer. It was found that the deuterium retention on the He-W co-deposition layer increased significantly with increasing the formation temperature of the co-deposition layer. [ABSTRACT FROM AUTHOR]

• Hydrogen isotope exchange behavior on C-W mixed material was studied by TDS. • The temperature dependence showed that hydrogen isotope exchange was the most effective at 523 K. • It can be said that hydrogen isotope replacement is efficient to remove retained tritium on the surface region of mixture layer. Hydrogen isotope exchange behavior on carbon (C)- tungsten (W) mixed material layer on W was studied as a function of gas exposure time and temperature. It was found that hydrogen isotope exchange enhanced the reduction of residual hydrogen effectively compared by only heating. The temperature dependence showed that hydrogen isotope exchange was the most effective at 523 K and above 573 K, thermal annealing reduced the total hydrogen isotope retention. It can be said that hydrogen isotope replacement is efficient to remove retained tritium on the surface region of mixture layer. [ABSTRACT FROM AUTHOR]