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AbstractThe current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokkasho Fusion Institute is reported. In order to advance the Japanese DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach activity and ITER-related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D; design of a fusion neutron source and development of an accelerator, ITER Test Blanket System; tritium handling technology; and information technology infrastructures, including a supercomputer system and a remote experimentation system, are carried out for a Fusion DEMO reactor.

AbstractTungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.

AbstractThe effects of tritium on the corrosion behavior of chromium were investigated in the present study, since it was suggested in the previous studies that the elution of chromium, which is one of the main constituent elements of passive layer of SUS 304 stainless steel, during passivation through further oxidation induced by oxidative radiolysis products would be the key reaction for the enhancement of the corrosion of SUS304 stainless steel induced by tritium. As the experimental results of the dependence of both dissolved oxygen and tritium concentration on the anodic behavior of chromium, it was found that the self-passivation of chromium induced by dissolved oxygen was inhibited in the tritiated solution, as found in the previous studies for SUS304 stainless steel. Therefore, it was considered that the elution of chromium by highly oxidative radiolysis products would induce a passivation inhibitory effect onto SUS304 stainless steel in a tritiated solution, resulting in an enhancement of the corrosion.

AbstractThe effects of displacement damage, plasma exposure and heat loads on T retention in reduced-activation ferritic/martensitic (RAFM) steels were investigated by exposing the steels to DT gas at 473 K. Despite enormous change in surface morphology, T retention in the heat-loaded specimen was comparable with that in the unloaded specimen. The exposure to plasma resulted in a drastic increase in T retention at the surface and/or sub surface. However, the T trapped at the surface/subsurface was easily removed by maintaining the specimens in air at ~300 K. Formation of radiation-induced defects led to a significant increase in T retention, and T trapped in the defects was not removed at ~300 K. These observations suggest that displacement damages have the largest effects on T retention at ~473 K.

AbstractTritium release from thermal neutron-irradiated Li4SiO4is initiated with the annihilation of E’-centers by recovering O-with diffusion of O-. Electron Spin Resonance (ESR) shows that differences in the formation of irradiation damage between 14 MeV and thermal neutrons in Li4SiO4 result in different tritium release behaviors. The kinetics for the annihilation of irradiation defects has been determined. The contribution of elastic collisions by 14 MeV neutrons was much higher than that of thermal neutrons. Isothermal annealing experiments show that annihilation of irradiation defects consisted of two processes: namely, the fast and slow annihilation processes. Their activation energies were determined to be 0.13 and 0.39 eV, respectively. Comparing the experimental results for the thermal and 14 MeV neutronirradiated Li4SiO4shows that the activation energies of the slow annihilation process were significantly different. These results relate to the density of irradiation defects, which in turn depend on the contribution of the recoil particles produced by nuclear reactions to form irradiation damaged sites.

AbstractHe+implantation effects on the Retention behavior of hydrogen isotopes implanted into 35% oxygen-contained boron film was studied by means of SEM, AFM, XPS and TDS. It was found that the D retention for only D2+implanted film was the highest and it decreased for pre-He+implanted film and post-He+implanted film. From the SEM and AFM images, the surface morphology of the oxygen-contained boron film was partly cracked, indicating that B2O3was formed in the film. From the TDS and XPS results, the defective structure and the formation of B-D-B bond, B-D bond and B-O-D bond were observed by He+and D2+implantation, respectively. It was suggested that oxygen was trapped as B-O bond. The reaction with implanted D2+was preceded in different mechanism.

Hydrogen retention in graphite tiles exposed to hydrogen discharges at the JT-60 open divertor has been investigated by means of thermal desorption spectroscopy (TDS). Most of the plasma facing area was covered with re-deposited layers of maximum thickness of about 70?µm appearing at the inner divertor region. Major parts of retained hydrogen were thermally desorbed as hydrogen molecules with a peak temperature of around 970?K. Almost all the hydrogen atoms were retained homogeneously in the re-deposited layers with an averaged hydrogen concentration of ~0.03 in H/C, which is much smaller than the saturated hydrogen concentration (H/C= 0.4-1.0). Since the saturated hydrogen concentration in carbon materials decreases with increasing temperature, the re-deposited carbon layers are very likely subjected to higher temperatures during the discharges, which are supported by the higher release temperature of hydrogen in TDS. This result suggests that hydrogen retention can be significantly reduced with higher wall temperatures.

The annihilation behaviors of radiation defects in neutron-irradiated LiAlO2were investigated by means of Electron Spin Resonance (ESR). It was found that the annihilation of radiation defects consisted of two processes, the fast and the slow processes. The activation energies of them were determined to be 0.14 ± 0.01 eV and 0.58 ± 0.01 eV, respectively. The F+-center was found to act as a trapping site of tritium by comparing its annihilation behavior with that of tritium release. Taking the results obtained in the present and the previous works in consideration, it can be said that the annihilation process of oxygen vacancies is of very important because tritium release from the bulk of a breeder starts just after the slow annealing process becomes dominant. Therefore, to understand the slow annihilation process of radiation defects is an important key to clarify the mechanism of tritium release from ceramic breeder materials.

Retention characteristics of deuterium and hydrogen retained in graphite tiles placed in the divertor region of JT-60U were investigated by thermal desorption spectroscopy (TDS). The deuterium retained in the near surface of all graphite tiles was mostly replaced by hydrogen due to exposure to hydrogen plasma at the final stage operations, resulting in main deuterium retention in the deeper region. The dominant species desorbed from the divertor tiles were H2, HD, D2and CH4. The smallest retention of hydrogen isotopes (H+D) was observed in the outer divertor tile which was eroded with maximum of 20 μm depth. The amount of H+D retained in the inner divertor tiles covered by the re-deposited layers increased with the thickness of the re-deposited layers. Hydrogen isotopes concentration ((H+D)/C) in the re-deposited layers was ~0.02, which was much smaller than those observed in JET and other devices.