[en] Well understanding of tritium (T) permeation and retention behavior in tungsten (W) requires the knowledge of hydrogen isotopes transport parameters. In the present study, the deuterium (D) permeation behavior in tungsten (W) are studied by gas driven permeation (GDP) and plasma driven permeation (PDP) methods. The D-GDP process in W agrees with the Sieverts’s law that the steady state permeation flux under different temperature are proportional to the square root of pressure. According to the Siverts’ law and Fick’s first law, the D permeability and effective diffusivity in W are obtained from the GDP results. To study the hydrogen isotopes plasma driven permeation (PDP) behavior in plasma facing materials, a linear Radio Frequency (RF) plasma device has been constructed in the radiation controlled area at Shizuoka University. In the PDP experiment, the steady state permeation flux is approximate proportional to the square root of the ion flux indicating that D-PDP process takes place in the RD regime. The D recombination coefficient on upstream surface of W is obtained by the PDP results. (author)

[en] Highlights: • In order to simulate damage distribution at operation condition, heavy ions were implanted into W. • HIDT simulation code was applied to evaluate the depth profile of D. • The total D retentions were decreased by the damage introduced near the surface region. • The hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface. -- Abstract: 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.

[en] Highlights: • Tritium release dynamics study on neutron irradiated Li₂TiO₃–Li₄SiO₄ breeder. • Tritium release behavior is influenced by the phase ratio. • Tritium recovery at low temperatures seems to be facilitated in biphasic materials. - Abstract: 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.

[en] Highlights: • To evaluate hydrogen isotope retention, W was exposed by plasma of QUEST 2017A/W. • Deuterium desorption was different from sample position of wall. • H was mainly accumulated in the deposition layer and trapped as C–H bonds. The W (tungsten) samples were placed at Top, Equator and Bottom plasma facing walls of QUEST (Q-shu University Experiment with Steady-State-Spherical Tokamak) device and exposed to 754 shots of hydrogen plasma during 2017A/W (Autumn/Winter) campaign. Thereafter, their surface morphologies and chemical states were evaluated by TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). The XPS results showed that a thick carbon layer about 3–18 nm has formed throughout the wall surface. Among them, the Bottom wall had the deposition layer with the thickness of 2 nm, which was thinner than the top wall, namely erosion-dominated. On the other hand, a thick C layer about 18 nm was deposited on the Equator wall. The additional 1 keV D₂⁺ was implanted into these samples and the D (deuterium) retention enhancement was estimated. The D₂ TDS (Thermal Desorption Spectroscopy) spectra for all the samples had two major desorption stages at 400 K and 650 K, namely the desorption of D trapped by irradiation damages and deposition layer. The erosion/deposition profile would be caused by wall position and plasma condition, like a current start-up experiment. The desorption temperature of H₂ (hydrogen) was shifted toward higher temperature side compared to that exposed to previous plasma campaign (2016 A/W), suggesting that H was mainly accumulated in the deposition layer with forming C–H bonds.

[en] 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.