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Co-deposited layers on surfaces of bulk tungsten divertor tiles (W lamellae) from the first campaign of JET with the ITER-Like Wall (JET-ILW, 2011–2012) were examined by means of a cross-sectional transmission electron microscope observation. The focus was on geometrical effects in impurity deposition, mainly beryllium (Be), on surfaces located in the poloidal gap separating adjacent lamellae. The study was carried out on the sides (gap surfaces) of two W lamellae from Stack C (located on the outboard part of the horizontal section of the lower divertor), from the region most exposed to the plasma (lamella C23) and in the magnetic shadow of the upstream divertor module (lamella C3). The tile manufacturing process (cold rolling) left shallow grooves, i.e. structures classified as convex (hill) and concave (valley) regions. These regions are decisive for the deposition structure. The main results were: (i) two kinds of impurity deposition features, “homogeneous” and “directional”, have been distinguished; (ii) the directional ones were characterized by nanoscale inclined vertical stripes in the deposition layer; (iii) homogeneous deposition without directional features were in the valley region. The results clearly indicate the impact of the surface finish, even in the tile gaps, on the qualitative and quantitative aspects of deposition.

Irradiation responses of binary W alloys were investigated systematically from the perspective of the binding energy of an alloying element with a W self-interstitial atom (W-SIA). Plates of W, W-0.3 at.% Cr, W-5 at.% Re, W-2.5 at.% Mo and W-5 at.% Ta alloys were irradiated at 1073 K with 6.4 MeV Fe ions to 0.26 dpa at the damage peak, where the binding energy of alloying element with W-SIA is in order of Cr > Re > Mo > Ta. The formation of vacancy-type defects (vacancies and vacancy clusters) was studied by using positron lifetime measurement. The precipitation of alloying elements was studied by using atom probe tomography (APT) and the hardness changes in the irradiated volumes were measured by the nanoindentation technique. The formation of vacancy-type defects was strongly suppressed by the addition of Cr and Re, while Ta and Mo had no noticeable suppression effect. The APT measurements showed fine Cr- and Re-rich precipitates in W-0.3 at.% Cr and W-5 at.% Re alloys, respectively, where the density of precipitates in the latter was clearly lower than that in the former. The distributions of Mo and Ta were uniform even after irradiation. Irradiation hardening was observed for all materials but that of W-5 at.% Re alloy was significantly smaller than the hardening of W, W-2.5 at.% Mo and W-5 at.% Ta alloys. These observations suggest that the irradiation hardening of W, W-2.5 at.% Mo, and W-5 at.% Ta alloys were mainly caused by vacancy-type defects. It was concluded that an alloying element with moderate binding energy with a W-SIA effectively suppresses vacancy formation without significantly enhanced precipitation and consequently mitigates irradiation hardening.

Deuterium retention was measured on neutron irradiated tungsten samples where one side of the samples had a visually clean metallic luster and the opposite side appeared to have a reacted surface film. Deuterium plasma exposure and subsequent thermal desorption from the reacted surface side produced spectra with larger total deuterium desorption at lower temperatures than from the clean surface side. For neutron irradiation, these W disk samples were installed in an irradiation capsule in such a way that one side of W sample was in contact with the surface of another W sample, and the opposite side was in contact with a SiC temperature monitor. The composition of the reacted surface was investigated using X-ray photoelectron spectroscopy and showed that SiC had interdiffused into the W samples. Neutron enhanced diffusion likely contributed to this as SiC and W are stable at temperatures exceeding the irradiation temperature. Results highlight the need to consider the surface chemistry of samples in drawing conclusions on hydrogen isotope retention of W materials and also illustrate the complexity of multi-material nuclear environments expected in fusion devices.

Proton-conducting oxides have potential applications in hydrogen sensors, hydrogen pumps, and other electrochemical devices including the tritium purification and recovery systems of nuclear fusion reactors. Although the distribution of hydrogen (H) in such oxide materials is an important aspect, its precise measurement is difficult. In the present study, the hydrogen solubility and diffusivity behavior of BaZr0.9Y0.1O2.95 (BZY), BaZr0.955Y0.03Co0.015O2.97 (BZYC), and CaZr0.9In0.1O2.95 (CZI) were studied using tritiated heavy water vapor i.e., DTO (~2 kPa, tritium (T) = 0.1%) by converting the photostimulated luminescence (PSL) values of the imaging plate (IP). The samples were exposed to DTO vapor at 673 K for 2 h or at 873 K for 1 h. The disc-shaped oxide specimens (diameter ~ 7.5 mm; thickness ~ 2.3 mm; theoretical density (TD) greater than 98%) were prepared by conventional powder metallurgy. The IP images of the specimen surfaces of all the three materials T-exposed revealed that BZY showed the most uniform T distribution with the highest tritium activity. The cross-sectional T concentration profiles of the cut specimens showed that T diffused deeper into BZY and BZYC than into CZI. The hydrogen solubility and diffusivity in the CZI specimen were lower than that in the BZY and BZYC specimens. This suggested that barium zirconates were more favorable proton conductors than calcium zirconates.

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

Poly-crystalline tungsten (W) samples were simultaneously irradiated with Helium (He) and Deuterium (D) ions using the triple-ion implantation device. He effect on D retention and transportation was studied using different combination of ion energies and He/D flux ratios in the simultaneous implantation. The experimental results show that D trapping at dislocation loops is significantly reduced in the case of 3 keV He+–3 keV D2+at He/D flux ratios over 0.6. D trapping by stronger trapping sites such as vacancies and vacancy clusters showed less dependence on the flux ratio. On the contrary, the D retention increases at each He/D flux ratio in the case of 3 keV He+–1 keV D2+compared to only D2+ implantation even the He/D flux ratio reaches a value of 1.0. TEM observations confirmed that dense dislocation loops are formed rather than He bubbles, which is responsible for the enhanced D retention in W. Keywords: Simultaneous implantation, D retention, Helium, Flux ratio, Transportation, Thermal desorption spectroscopy