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The W (tungsten) samples were placed at top, equator and bottom walls of QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) device and exposed 1238 shots of hydrogen plasma during 2016A/W (Autumn/Winter) campaign with normal wall temperature of 473 K (maximum temperature of 523 K). Thereafter, the surface morphology was evaluated by color measurement, TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). Thick deposition layers were formed on the samples placed at the equator and bottom walls. On the other hand, thin mixed material layer was deposited on the top wall, where large H (hydrogen) retention was observed, which would be caused by dynamic plasma wall interaction (erosion and deposition) with higher H flux. Low H retention was confirmed for bottom wall, where higher wall temperature without He discharge would contribute. The additional 1 keV D2+ was implanted into these samples and deuterium retention enhancement was estimated. It was clearly found that the irradiation damages would induce more deuterium trapping than the formation of C–D bond.

0.8 MeV and 6 MeV iron (Fe) ions were implanted into tungsten (W) to produce the irradiation damages with the various damage distributions. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed to evaluate the D retention behavior on damage distribution in W. The experimental results showed that the total D retentions were decreased by increasing the damage concentration introduced near the surface region by 0.8 MeV Fe ion implantation. The retention of D trapped by vacancy clusters and voids, which would be the stable trapping sites with higher trapping energies, were reduced, suggesting that the recombination of D atom into D2 on the W surface was enhanced due to D accumulation near the surface region. It can be said that the hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface.

Biphasic Li2TiO3–Li4SiO4 tritium breeders with different phase ratios were prepared by the solution combustion synthesis method. The as-prepared biphasic Li2TiO3–Li4SiO4 and single-phase materials were irradiated by the thermal neutron with the flux of 2.75 × 1013 n cm−2 s−1 at Kyoto University Research Reactor (KUR). Tritium is generated based on the following reaction: 6Li+1n→3T+4He+4.8 MeV. The results of tritium release experiment show that tritium release starts at the lower temperature range for the biphasic materials. The tritium release temperature and the shape of tritium-TDS (Thermal desorption spectroscopy) spectra are dependent on the phase ratio of Li2TiO3 to Li4SiO4. In the case of Li2TiO3/Li4SiO4 = 2.0, tritium release starts at 400 K, and the tritium migration is mainly controlled by the diffusion process. With the increase of Li4SiO4 content in the biphasic materials, tritium release is influenced by the de-trapping process.

Deuterium (D) retention behaviors for the 3 keV Helium (He+) implanted damaged-Tungsten (W) and 10 keV Carbon (C+) - 3 keV He+ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He+ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He+ fluence. The D retention in the deeper region was reduced by He+ implantation with higher He+ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C+ - He+ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C+ - He+ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

To evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He+ was introduced into W bulk by 201 – 1000 keV He+, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W. Keywords: He in damaged W, Hydrogen isotope retention behavior, Damaged W, Fusion

The purpose of this study is to elucidate helium (He) retention behavior in tungsten (W) under simultaneous He and hydrogen (H) irradiation. Polycrystalline-W was irradiated by He+ and H2+ simultaneously with the energy of 1.0 keV and 3.0 keV. He+ fluences were (0.5, 1.0, 10) × 1021 He+ m−2 and H2+ fluence was 1.0 × 1022 H+ m−2,respectively. After irradiation, He desorption behavior was investigated by high temperature thermal desorption spectroscopy (HT-TDS) in the temperature range of R.T.-1773 K. Micro-structure changes of W after irradiation were observed by TEM. It was found that simultaneous irradiation with different H2+ energy significantly changed He retention behavior. 1.0 keV H2+ suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H2+ facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H2+.

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.

Establishment of effective tritium removal method was one of important issues for the development of fusion reactor from the view of fuel recycle and safe operation. The deuterium (D) removal efficiency in tungsten (W) by energetic hydrogen (H) ions under room temperature and baking under 623 K were studied by thermal desorption spectroscopy (TDS). Iron (Fe) damaged W with various damage level by 6 MeV Fe2+ was adopted to simulate neutron irradiation damages. To understand the D removal behavior, the desorption of D2 was measured in-situ by a quadrupole mass spectrometer (QMS) during H2+ implantation and baking. The in-situ results showed that the desorption of D2 started after H2+ implantation and became slowly with the increment of H2+ implantation time. After H2+ implantation, part of D trapped by dislocation loops, vacancy clusters and voids could be removed by hydrogen isotope exchange. However, the removal efficiency by hydrogen isotope exchange decrease obviously as the presence of irradiation damages. The D trapped by dislocation loops and vacancy clusters can be removed by baking with high efficiency. It is worth to note that the D trapped by voids cannot be removed by baking leading to the lower D removal efficiency for W with high damage level.

This paper reviews recent achievement of hydrogen isotope behavior for damaged tungsten. To demonstrate neutron irradiation, the irradiation damages were introduced into W by energetic Fe2+ irradiation and D retention behavior was examined by thermal desorption spectroscopy (TDS). The D trapping behavior was evaluated using Hydrogen Isotope Diffusion and Trapping (HIDT) code. It was found that D trapping states consisted of two-four stages with the trapping energy of 0.60 eV, 0.85 eV, 1.15-1.25 eV and 1.55 eV depending on the damage concentration and distribution. Based on these experimental results, the hydrogen isotope retention behavior in actual fusion condition was demonstrated. It was found that most of hydrogen isotope was retained in tungsten wall even if the wall temperature was kept at operation temperature.