Evaluation of hydrogen retention behavior for damaged tungsten exposed to hydrogen plasma at QUEST with high temperature wall

The undamaged W (tungsten) and 0.3 dpa (displacement per atom) damaged W samples by 6 MeV Fe2+ (Fe ion) irradiation were installed in the first wall of QUEST (Q-shu University Experiment with Steady-State-Spherical Tokamak) device and exposed to H (hydrogen) plasma in 2018A/W (Autumn / Winter) or 2019S/S (Spring / Summer) campaign to evaluate the impact of damages and impurities on hydrogen isotope retention. The surface morphology and chemical states of constituent atoms were observed by TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). It was found that thick Al deposit were found for the samples in 2019S/S campaign, which would come from the insulating plate during the CHI (Coaxial Helicity Injection) discharge. The additional 1 keV D2+ was implanted into both of these samples and D (deuterium) retention enhancement was evaluated by TDS (Thermal Desorption Spectroscopy). The downward ion toroidal drift changed the impurity deposition and damage profiles. In 2018A/W and 2019S/S, the D retentions for undamaged W samples had position independent, indicating that plasma would be well-controlled in this configuration. In case of Fe2+ damaged samples, irradiation damages clearly changed the D retention characteristics.