Takumi Chikada

Surface behaviour in deuterium permeation through erbium oxide coatings

Suppression of tritium permeation through structural materials is essential in order to mitigate fuel loss and radioactivity concerns. Ceramic coatings have been investigated for over three decades as tritium permeation barriers (TPBs); however, a very limited number of investigations on the mechanism of hydrogen-isotope permeation through the coatings have been reported. In this study, deuterium permeation behaviour of erbium oxide coatings fabricated by filtered arc deposition on reduced activation ferritic/martensitic steels has been investigated. The samples coated on both sides of the substrates showed remarkably lower permeability than those coated on one side, and the maximum reduction efficiency indicated a factor of 105 compared with the substrate. The different permeation behaviour between the coatings facing the high and low deuterium pressure sides has been found by the crystal structure analysis and the evaluation of the energy barriers. It is suggested that the permeation processes on the front and back surfaces are independent, and the TPB efficiency of the samples coated on both sides can be expressed by a multiplication of that of each side.

Modeling of Tritium Permeation Through Erbium Oxide Coatings

Abstract Tritium permeation through erbium oxide coatings has been modeled on the basis of experimental results. Permeation models were constructed step-by-step by the introduction of the following predominant parameters: surface coverage, grain size, and energy barrier. The surface-coverage model agreed with the imperfectly coated samples fabricated by filtered arc deposition as well as by metal-organic decomposition. The grain-boundary-diffusion model also agreed with the coatings fabricated by filtered arc deposition, though it was not applicable to the metal-organic decomposition coatings because of impurities and different layer structures. The energy-barrier model explains the contributions to the additional permeation reduction of the multilayer coatings. The discussion of permeation models provides new design concepts for the development of tritium permeation barriers.

Effect of neutron energy and fluence on deuterium retention behaviour in neutron irradiated tungsten

Deuterium (D) retention behaviours for 14 MeV neutron irradiated tungsten (W) and fission neutron irradiated W were evaluated by thermal desorption spectroscopy (TDS) to elucidate the correlation between D retention and defect formation by different energy distributions of neutrons in W at the initial stage of fusion reactor operation. These results were compared with that for Fe2+ irradiated W with various damage concentrations. Although dense vacancies and voids within the shallow region near the surface were introduced by Fe2+ irradiation, single vacancies with low concentration were distributed throughout the sample for 14 MeV neutron irradiated W. Only the dislocation loops were introduced by fission neutron irradiation at low neutron fluence. The desorption peak of D for fission neutron irradiated W was concentrated at low temperature region less than 550 K, but that for 14 MeV neutron irradiated W was extended toward the higher temperature side due to D trapping by vacancies. It can be said that the neutron energy distribution could have a large impact on irradiation defect formation and the D retention behaviour.

THERMAL INFLUENCE ON ERBIUM OXIDE COATING FOR TRITIUM PERMEATION BARRIER

Er2O3 coating for tritium permeation barrier has been fabricated on steel substrates by a filtered arc deposition method at room temperature and 973 K. Thermal expansion of the oxide layer and the substrate induced peel-off of the coating. The non-crystalline layer is thought to play a role in forming a uniform surface coating. Five cycles of permeation measurements at 773-973 K resulted in no degradation of the coating. Different permeation behaviors are seen between degassing for 12 h at 873 K and at room temperature. Low hydrogen background following degassing at 873 K helps detect the transition to deuterium permeation. The permeation flux following different degassing conditions eventually approached comparable levels.

A multi-technique analysis of deuterium trapping and near-surface precipitate growth in plasma-exposed tungsten

In this work, we examine how deuterium becomes trapped in plasma-exposed tungsten and forms near-surface platelet-shaped precipitates. How these bubbles nucleate and grow, as well as the amount of deuterium trapped within, is crucial for interpreting the experimental database. Here, we use a combined experimental/theoretical approach to provide further insight into the underlying physics. With the Tritium Plasma Experiment, we exposed a series of ITER-grade tungsten samples to high flux D plasmas (up to 1.5 × 1022 m−2 s−1) at temperatures ranging between 103 and 554 °C. Retention of deuterium trapped in the bulk, assessed through thermal desorption spectrometry, reached a maximum at 230 °C and diminished rapidly thereafter for T > 300 °C. Post-mortem examination of the surfaces revealed non-uniform growth of bubbles ranging in diameter between 1 and 10 μm over the surface with a clear correlation with grain boundaries. Electron back-scattering diffraction maps over a large area of the surface confirmed th...

Estimation of Tritium Permeation Rate to Cooling Water in Fusion DEMO Condition

Abstract The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle & Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium two-component system. In the representative recent DEMO condition, the following tritium permeation\ rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day.

SiC coating as hydrogen permeation reduction and oxidation resistance for nuclear fuel cladding

Silicon carbide (SiC) coating is one of the countermeasures for the prevention of oxidation and hydrogen embrittlement of fuel claddings because SiC has high resistance of oxidation and hydrogen permeation. Hydrogen permeation and oxidation experiments for the cladding materials with SiC coatings were conducted in unirradiated conditions. The sputtering method was employed to make SiC coatings. In the hydrogen permeation experiment, 316 type of stainless steel (SS316) was used as a base material of the coating. SS316 with SiC coatings showed hydrogen permeation reduction by one order of magnitude. In the oxidation experiments, Zircaloy 4 (Zry-4) and SS316 were used as base materials of the coatings. The weight gain of the Zry-4 specimens with a SiC coating decreased by about one-fifth compared to the uncoated ones at 750 °C and 1200°C. This phenomenon was observed for SS316 at 750 °C as well. The peel-off of the coating was observed in several experiments, and it is considered that the peel-off was caused by the difference of the thermal expansions between coatings and base materials. Thicker coatings showed better oxidation resistance, but thinner coatings showed more tolerance of peel-off.

Effect of Heating Temperature on Deuterium Retention Behavior for Helium/Carbon Implanted Tungsten

Abstract The effect of heating temperature on deuterium (D) retention behavior for helium (He+) / carbon (C+) implanted tungsten (W) was studied. It was found that D retention behavior for He+ implanted W was not limited by the size of the He bubbles. The microstructure observation showed that the large helium bubbles were formed near the surface for He+ implanted W at 1173 K, suggesting that the D retention was reduced by the growth of the helium bubbles. In addition, to evaluate the effect of implantation ion species at high temperature, D retention behavior for He+ implanted W at 1173 K was compared with that for C+ implanted W at 673 K. It is concluded that the D retention depends on ion species, which makes different kinds of damages like He bubbles for He+ implantation and vacancy-ion complex (voids) for C+ implantation.

Dynamics for HT and HTO Recovery Through Water Bubbler and CuO Catalyst

Dynamics of tritium recovery using CuO catalyst and water bubbler was studied as a function of gas flow rate and CuO temperature. The rate constant of tritiated water formation by CuO catalyst at the temperature above 500 K was determined to be k [s-1] = 5.4×105 exp (-0.65 eV/kBT). For the flow rate less than 50 sccm, it was found that the reaction rate will be controlled by the desorption rate of HTO on the surface of CuO. These results were applied for the design of tritium removal system at radiation-controlled area. It was concluded that the reactor tubing with 1.0 meter length at 600 K will be suitable to reduce the tritium concentration less than 1/1000 and the longer reactor tubing will be required if the operation temperature will be lower than 600 K.

Impact of Annealing on Deuterium Retention Behavior in Damaged W

Abstract The annealing effects on deuterium (D) retention for 0.1–1.0 dpa iron (Fe) ion damaged W were studied as a function of annealing duration. The D2 spectra for Fe damaged W with lower defect concentration showed that D trapped by vacancy clusters was clearly decreased as increasing annealing duration due to the recovery of vacancy clusters. On the other hand, at higher defect concentration, the desorption peak of D trapped by voids was shifted toward higher temperature side, which would be caused by aggregation of vacancies and vacancy clusters. It can be said that the recovery and aggregation behavior of defects are controlled by defect concentration. By disappearing of desorption of D trapped by vacancy clusters after annealing for longer duration, the desorption of D trapped by vacancies was increased, which could be explained by following two possibilities. One is that the retention of hydrogen isotope trapped by monovacancy was increased. The other is that number of vacancies during annihilation process of vacancy cluster were formed by annealing.