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Functional ceramic coatings have been investigated for several decades to mitigate tritium permeation through structural materials in a fusion reactor blanket. For the establishment of a liquid blanket system, the coatings require not only tritium permeation reduction but also the tolerance to irradiation damage induced by neutrons and corrosion by liquid tritium breeders. In this study, deuterium permeation measurements under exposure to liquid lithium–lead were performed for a heavy-ion-irradiated zirconium oxide coating sample to elucidate the synergy of irradiation and corrosion on hydrogen isotope permeation. The irradiated coating decreased the deuterium permeation flux under lithium–lead exposure by more than two orders of magnitude in comparison with the unirradiated one. Each activation energy of permeation and diffusion of the coating sample was larger than that of the unirradiated one at around 450 °C and decreased to the level of the unirradiated one at 600 °C. The results indicated that the voids formed by ion irradiation aggregated in grain boundaries of the coating, resulting in an increase in the energy barrier of diffusion at low temperatures and then a decrease due to the defect recovery at high temperatures. Besides, a corrosion product layer formed on the coating during lithium–lead exposure increased the energy barrier of solution. The synergy of irradiation and corrosion on permeation was developed at lower temperatures before the defect recovery.

Ceramic coatings have been developed for nearly half a century to mitigate tritium leakage through steels and corrosion of the components in fusion reactor blanket systems. In recent years, irradiation effects of ceramic coatings on their characteristics and functions have been elucidated through heavy-ion irradiation tests. However, most of analytical methods used in these studies were destructive and time-consuming. In this study, yttrium oxide coatings were fabricated by radio-frequency magnetron sputtering, and the electrical impedance measurements were conducted for the coatings to develop a convenient approach for understanding of irradiation effects. The conductivity obtained from Nyquist plots showed lower values for the undamaged sample than for the damaged ones below 350 °C, while the relationship of the conductivity was reversed at 350 °C and higher temperatures. The ac conductivity of the undamaged sample showed a frequency dependence from 200 °C to 550 °C; however, that of the damaged ones did not show at 300 °C and lower temperatures and indicated a tendency similar to undamaged one above 300 °C. These tendencies of the conductivity at low temperatures were attributed to the partial recovery of the irradiation damage, while the remaining irradiation defects and/or irradiation-induced grain growth affected the electrical properties at higher temperatures. The electrical impedance spectroscopy methods have a potential to detect the influence of irradiation on the coating.

Deuterium permeation through a fusion-relevant ferritic steel F82H with and without erbium oxide coatings under γ-ray irradiation has been investigated in a temperature range from 300 to 700 °C. The deuterium permeation flux through F82H sample increased by γ-ray irradiation at lower temperatures below 450 °C. The irradiation effect increased with dose rate, and the percentage of the permeation flux gain might be several percent under the dose rate of a few Gy s‒1. Temperature of the F82H sample surface rose by about 0.5 °C depending on the dose rate, and so the γ-ray irradiation effect is mainly attributed to γ-heating. On the other hand, at higher temperature above 500 °C, no appreciable change of the deuterium permeation was observed. Similarly, the deuterium permeation flux through erbium oxide coated samples increased under γ-ray irradiation at lower temperatures (350‒450 °C), but no appreciable change of permeation flux through coatings was observed at higher temperatures (600‒700 °C). The coating surface temperature increased at lower sample temperatures by γ-heating. Keywords: Tritium, Permeation, γ-ray, Irradiation, F82H, Erbium oxide

For the establishment of liquid tritium breeding concepts, static lithium-lead corrosion tests for single-layer erbium oxide coatings and erbium oxide-metal multi-layer coatings were carried out, followed by deuterium permeation measurements. Grain boundary corrosion of erbium oxide coatings was confirmed by the static immersion tests at 550 and 600 °C. An erbium oxide-iron two-layer coating sustained its structure after lithium-lead immersion at 600 °C for up to 3000 h. The results of gas-driven deuterium permeation measurements for multi-layer coatings immersed at 550 and 600 °C indicated that crystallization and grain growth of erbium oxide would sufficiently occur during the immersion at 600 °C, and then the sample showed lower permeabilities in the first measurements at lower temperature. On the other hand, permeation reduction factors of the sample immersed at 550 °C were estimated to be 100‒200 in the temperature range of 400‒550 °C. A corrosion layer formed on the coating surface might work as an additional diffusion barrier after the permeation test at 600 °C. Keywords: Lithium-lead, Tritium permeation barrier, Corrosion, Erbium oxide

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

Yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steels by metal organic decomposition with a dip-coating technique, and their deuterium permeation behaviors were investigated. The microstructure of the coatings varied with heat-treatment temperature: amorphous at 670ºC (amorphous coating) and crystallized at 700ºC (crystallized coating). Deuterium permeation flux of the amorphous coating was lower than the uncoated steel by a factor of 5 at 500ºC, while that of the crystallized coating was lower by a factor of around 100 at 400‒550ºC. The permeation fluxes of both coatings were drastically decreased during the measurements at higher temperatures by a factor of up to 790 for the amorphous coating and 1000 for the crystallized one, indicating a microstructure modification occurred by an effect of test temperature with hydrogen flux. Temperature dependence of deuterium diffusivity in the coatings suggests that the decrease of the permeation flux has been derived from a decrease of the diffusivity. Characteristic permeation behaviors were observed with different annealing conditions; however, they can be interpreted using the permeation mechanism clarified in the previous erbium oxide coating studies.