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To evaluate the damage depth profile effect on hydrogen isotope retention in tungsten (W), combination usage of 0.8 MeV and 6.0 MeV Fe ions were implanted into W with the damage concentrations between 0.03 and 0.1 dpa. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed with the flux of 1.0 × 1018 D+ m−2 s−1 up to the fluence of 1.0 × 1022 D+ m−2, and the D retention behavior was evaluated by thermal desorption spectroscopy (TDS). The experimental results indicated that 6.0 MeV Fe ion irradiation would introduce vacancies and voids into bulk that were clearly controlled by the damage concentration, and the voids would become the most stable D trapping sites. It was found that D de-trapping from irradiation defects at lower temperature would be enhanced by the accumulation of defect near the surface due to 0.8 MeV Fe ion irradiation.

Influence of mixed material layer formation on hydrogen isotope retention in W exposed to 2014 Large Helical Device (LHD) experiment campaign was evaluated by Thermal Desorption Spectroscopy (TDS), X-ray Photoelectron Spectroscopy (XPS), Scanning Electron Microscope (SEM) and Transmission Electron Microscope (TEM). It was found that a lot of hydrogen isotopes were trapped by the carbon-dominated mixed-material layer deposited on the plasma facing materials. Most of He was also trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K, respectively. However, the hydrogen retention behavior for erosion dominated area was clearly different from those for deposition dominated area and typical Plasma Wall Interaction (PWI) area, where He bubbles were introduced near the sample surface, leading to the introduction of various types of trapping sites in W.

Deuterium (D) retention behavior for the sequential 6 MeV iron (Fe) and 10 keV carbon (C) implanted tungsten (W) were evaluated by thermal desorption spectroscopy (TDS) and β-ray-induced X-ray spectroscopy (BIXS) to understand the synergetic effect of defect formation and C existence on D retention behavior for W under various damage distribution profiles. The experimental results indicated that retention of D trapped by dislocation loops was controlled by 10 keV C + implantation. The D retention was reduced in the sequential Fe 2+ − C + implanted W with higher C + fluence in comparison to that with lower C + fluence due to the formation of C-W layer which suppressed D diffusion toward the bulk and dense defects at the surface which reduce effective D diffusion coefficient. On the other hand, the amount of D trapped by the defects in the deeper region than C + implantation region (50 nm) was increased due to the formation of dense defects by 6 MeV Fe 2+ implantation within the depth of 1.5 μm.

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 concentr...

In this study, we measured deuterium (D) gas-driven permeation through tungsten (W) foils that had been pre-damaged by helium ions (He + ). The goal of this work was to determine how ion-induced damage affects hydrogen isotope permeation. At 873 K, the D permeability for W irradiated by 3.0 keV He + was approximately one order of magnitude lower than that for un-damaged W. This difference diminished with increasing temperature. Even after heating to 1173 K, the permeability returned to less than half of the value measured for un-damaged W. We propose that this is due to nucleation of He bubbles near the surface which potentially serve as a barrier to diffusion deeper into the bulk. Exposure at higher temperatures shows that the D permeability and diffusion coefficients return to levels observed for undamaged material. It is possible that these effects are linked to annealing of defects introduced by ion damage, and whether the defects are stabilized by the presence of trapped He.

All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s−1 with Ar flow rate of 13.3 sccm.Using this HI-TDS system, D and T desorption behaviors for implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trap...

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 670 oC (amorphous coating) and crystallized at 700 oC (crystallized coating). Deuterium permeation flux of the amorphous coating was lower than the uncoated steel by a factor of 5 at 500 oC, while that of the crystallized coating was lower by a factor of around 100 at 400‒550 oC. 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.

Usage of CuO and water bubbler is one of the conventional and convenient methods for tritium recovery. In present work, influence of CuO particle size and diameter of reaction tubing on the tritium recovery was evaluated. Reaction rate constant of tritium with CuO particle has been calculated by the combination of experimental results and a simulation code. Then, these results were applied for exploring the dependence of reaction tubing length on tritium conversion ratio. The results showed that the surface area of CuO has a great influence on the oxidation rate constant. The frequency factor of the reaction would be approximately doubled by reducing the CuO particle size from 1.0 mm to 0.2 mm. Cross section of reaction tubing mainly affected on the duration of tritium at the temperature below 600 K. Reaction tubing with length of 1 m at temperature of 600 K would be suitable for keeping the tritium conversion ratio above 99.9%. The length of reaction tubing can be reduced by using the smaller CuO particle or increasing the CuO temperature.

Effects of annealing after/under iron (Fe) ion irradiation on deuterium (D) retention behavior in tungsten (W) were studied. The D2 TDS spectra as a function of heating temperature for 0.1 dpa damaged W showed that the D retention was clearly decreased as the annealing temperature was increased. In particular, the desorption of D trapped by voids was largely reduced by annealing at 1173 K. The TEM observation indicated that the size of dislocation loops was clearly grown, and its density was decreased by the annealing above 573 K. After annealing at 1173 K, almost all the dislocation loops were recovered. The results of positron annihilation spectroscopy suggested that the density of vacancy-type defects such as voids, was decreased as the annealing temperature was increased, while its size was increased, indicating that the D retention was reduced by the recovery of the voids. Furthermore, it was found that the desorption temperature of D trapped by the voids for damaged W above 0.3 dpa was shifted toward higher temperature side. These results lead to a conclusion that the D retention behavior is controlled by defect density. The D retention in the samples annealed during irradiation was less than that annealed after irradiation. This result shows that defects would be quickly annihilated before stabilization by annealing during irradiation.

Pure tungsten (W) samples were respectively installed on the two positions of typical plasma wall interaction area (PI) and erosion dominated position (ER) on the first wall in the Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan. After the experiment campaign in 2014, these samples were picked up and thermal desorption spectroscopy (TDS) were applied to evaluate their desorption behavior of He which was implanted by He discharge. It was found that a carbon-dominated mixed-material layer with impurities, such as Fe, Cr, Mo, O and N, etc., was deposited layer by layer during the plasma exposure on the PI sample. On the other hand, W-C mixed layer was formed on the ER sample. The results showed that a large amount of He was trapped in the samples on both PI and ER, and the total He retention for ER is about twice as large as that of PI. He was trapped in various types of trapping sites in the ER sample and their desorption peaks were located at temperatures of about 425 K, 755 K, 1130 K and 1630 K. For PI sample, most of He was trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K. The additional 3.0 keV helium ion (He+) implantation was performed for several samples to investigate the He retention characteristics in these samples and it was found that no additional desorption stage was found. These results suggested that the He discharge history and its deposition on the W plasma facing wall would affect He desorption behavior of W.

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.

Temperature dependence on deuterium (D) retention for He+ implanted tungsten (W) was studied by thermal desorption spectroscopy (TDS) to evaluate the tritium retention behavior in W. The activation energies were evaluated using Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code and found to be 0.55 eV, 0.65 eV, 0.80 eV and 1.00 eV. The heating scenarios clearly control the D retention behavior and, dense and large He bubbles could work as a D diffusion barrier toward the bulk, leading to D retention enhancement at lower temperature of less than 430 K, even if the damage was introduced by He+ implantation. By comparing the D retention for W, W with carbon deposit and tungsten carbide (WC), the dense carbon layer on the surface enhances the dynamic re-emission of D as hydrocarbons, and induces the reduction of D retention. However, by He+ implantation, the D retention was increased for all the samples.