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0.8 MeV and 6 MeV iron (Fe) ions were implanted into tungsten (W) to produce the irradiation damages with the various damage distributions. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed to evaluate the D retention behavior on damage distribution in W. The experimental results showed that the total D retentions were decreased by increasing the damage concentration introduced near the surface region by 0.8 MeV Fe ion implantation. The retention of D trapped by vacancy clusters and voids, which would be the stable trapping sites with higher trapping energies, were reduced, suggesting that the recombination of D atom into D2 on the W surface was enhanced due to D accumulation near the surface region. It can be said that the hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface.

Deuterium (D) retention behaviors for the 3 keV Helium (He+) implanted damaged-Tungsten (W) and 10 keV Carbon (C+) - 3 keV He+ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He+ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He+ fluence. The D retention in the deeper region was reduced by He+ implantation with higher He+ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C+ - He+ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C+ - He+ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

Six single crystal tungsten specimens were neutron irradiated to a dose of 0.1 displacements per atom (dpa) at three different irradiation temperatures (633 K, 963 K, and 1073 K) at the High Flux Isotope Reactor in Oak Ridge National Laboratory under the US-Japan PHENIX project. A pair of neutron-irradiated tungsten specimens was exposed to deuterium (D) plasma to D ion fluence of 5.0 × 1025 m−2 at three different exposure temperatures (673 K, 873 K, and 973 K) at the Tritium Plasma Experiment in Idaho National Laboratory. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (Tritium Migration Analysis Program, TMAP) were used to understand D behavior in neutron-irradiated tungsten. A broad D desorption spectrum from the plasma-exposure temperature up to 1173 K was observed. Total D retention up to 1.9 × 1021 m−2 and near-surface D concentrations up to 1.7 × 10−3 D/W were experimentally measured from the 0.1 dpa neutron-irradiated single crystal tungsten. Trap density up to 2.0 × 10−3 Trap/W and detrapping energy ranging from 1.80 to 2.60 eV were obtained from the TMAP modeling.

Targets of Reduced Activation Ferritic Martensitic (RAFM) steel F82H were exposed to low-energy (200 eV), high flux (about 1022 D/m2s) deuterium (D) plasma at 623–773 K to various D fluences in the range from 1 × 1025 to 2.5 × 1026 D/m2. The surface morphology of the plasma-exposed targets was examined with a field-emission scanning electron microscope. Cross-sectional observations of nano-structures formed on the F82H target surfaces were performed using a transmission electron microscope equipped with an energy dispersive X-ray spectrometer. It has been shown that nano-sized fiber-like layers are formed on the target surfaces under D plasma exposure. Micro-sized surface morphology pattern depends on the D fluence. As the D fluence increases, clusters of the fiber-like layers begin to be formed and organized into ordered structure.

Tritium-containing polycrystalline tungsten, which was pre-irradiated with helium prior to tritium exposure, was isothermally heated under a vacuum or deuterium gas in order to investigate the effect of the deuterium on tritium removal from the tungsten. It is found that in the case of 3 h of baking at temperatures varying from 423 K and 573 K, tritium retention after baking in deuterium gas became smaller than that in a vacuum. At the baking temperature of 573 K, tritium removal was hastened in the case of baking in deuterium atmosphere. The results indicate that the baking in deuterium atmosphere is effective to remove tritium from tungsten. The acceleration of diffusion of tritium in bulk and recombination of tritium with deuterium at the surface might be possible mechanisms.

Eight samples of F82H ferritic/martensitic steel were irradiated at 300 K with 20 MeV W ions to the damage level of 0.54 displacements per atom at the damage peak. Three samples were afterwards annealed in vacuum at 423 K for 72 h and then at 373 K for 106 h. Three other samples were annealed in H2 atmosphere at 100 kPa at the same annealing temperatures and durations. All samples were exposed at room temperature to D2O vapor at the partial pressures in the range from 3 to 6 kPa for 365 to 1181 days. After termination of the D2O vapor exposure, the surfaces of the samples were partly covered by small (≤1 mm in diameter) drops of D2O water. Thus, the damaged samples were exposed to a mixture of D2O vapor and water. Trapping of deuterium at the ion-induced defects in the damage zone was examined by the D(3He, p)4He nuclear reaction. It has been found that the W-ion-induced defects generated in the F82H samples are decorated by deuterium diffusing from the surface.

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

Tritium beta-ray induced X-ray spectra in various gas mediums were simulated by Monte Carlo simulation using Geant4 tool kit. The simulated beta-ray induced X-ray spectrum (s-BIX spectrum) was composed of the bremsstrahlung component and characteristics X-rays from constituent elements. The total number of photons in s-BIX spectrum decreased with increasing pressure of medium except argon. In argon medium, the characteristics X-ray of argon was generated by beta particles from tritium decay, and the contribution of Ar-Kα and -Kβ compensated the reduction of bremsstrahlung generated by solid matter with increasing argon pressure. At 0.001 atm of medium pressure, the total counts in s-BIX spectrum was independent from gas medium. Therefore, the gas medium dependence in BIXS at low pressure (less than 0.001 atm) was not serious issue.

Surface topography of and deuterium (D) retention in reduced activation ferritic-martensitic Eurofer’97 and ferritic oxide dispersion strengthening ODS-16Cr steels have been studied after exposure at 600 K to low-energy (70 and 200 eV), high-flux (∼1022 D/m2s) pure D and D-10%He plasmas with D fluence of 2 × 1025 D/m2. The methods used were scanning electron microscopy, energy-scanning D(3He,p)4He nuclear reaction, and thermal desorption spectroscopy. As a result of the plasma exposures, nano-sized structures are formed on the steel surfaces. After exposure to pure D plasmas, a significant fraction of D is accumulated in the bulk, at depths larger than 8 μm. After exposures to D-He plasmas, D is retained mainly in the near-surface layers. In spite of the fact that the He fluence was lower than the D fluence, the He retention in the steels is one order of magnitude higher than the D retention.

The purpose of this study is to elucidate helium (He) retention behavior in tungsten (W) under simultaneous He and hydrogen (H) irradiation. Polycrystalline-W was irradiated by He+ and H2+ simultaneously with the energy of 1.0 keV and 3.0 keV. He+ fluences were (0.5, 1.0, 10) × 1021 He+ m−2 and H2+ fluence was 1.0 × 1022 H+ m−2,respectively. After irradiation, He desorption behavior was investigated by high temperature thermal desorption spectroscopy (HT-TDS) in the temperature range of R.T.-1773 K. Micro-structure changes of W after irradiation were observed by TEM. It was found that simultaneous irradiation with different H2+ energy significantly changed He retention behavior. 1.0 keV H2+ suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H2+ facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H2+.

Double strand breaks (DSBs) of DNA molecules in tritiated water was examined under sterilized and non-sterilized conditions using a single molecule observation method. The genome DNA of bacteriophage T4 GT7 was immersed in sterilized tritiated water (5.2 MBq/cm3) and non-sterilized tritiated water (4.2 MBq/cm3) for 1, 7 and 14 day(s). Then the length of DNA molecules was measured using a fluorescence microscope after intercalation of fluorescent dye. The dose rate was 1.4–1.7 × 10−2 Gy/h and the dose level was 0.41–5.8 Gy. The rate of DSBs induced by β-rays from tritium was successfully evaluated under the sterilized conditions and the value comparable with the DSB rate under γ-ray irradiation (Noda et al., Scientific Reports 7 (2017) 8557) was obtained. The length of DNA molecules in non-sterilized tritiated water was clearly shorter than that in the sterilized tritiated water. This observation suggested that the effects of tritium was far weaker than that of microorganisms (e.g. bacteria) and impurities in water even at the tritium concentration as high as 5.2 MBq/cm3.

To study the effect of the content of chromium (Cr) in the tungsten (W) matrix on the vacancy formation and retention of hydrogen isotopes, the samples of the W-0.3 at.% Cr alloy were irradiated with 6.4 MeV Fe ions in the temperature range of 523–1273 K to a damage level of 0.26 displacement per atom (dpa). These displacement-damaged samples were exposed to D2 gas at a temperature of 673 K and a pressure of 100 kPa to decorate ion-induced defects with deuterium. The addition of 0.3 at.% Cr into the W matrix resulted in a significant decrease in the retention of deuterium compared to pure W after irradiation especially at high temperature (≥773 K). Positron lifetime in W-0.3 at.% Cr alloy irradiated at 1073 K was almost similar to that for non-irradiated one. These facts indicate the suppression of the formation of vacancy-type defects (monovacancies and vacancy clusters) by 0.3 at.% Cr addition, which leads to the significant reduction in deuterium retention in W-0.3 at.% Cr alloy.

Deuterium trapping at irradiation-induced defects in tungsten, a candidate material for plasma facing components in fusion reactors, was revealed by positron annihilation spectroscopy. Pure tungsten was electron-irradiated (8.5 MeV at ∼373 K and to a dose of ∼1 × 10−3 dpa) or neutron-irradiated (at 573 K to a dose of ∼0.3 dpa), followed by post-irradiation annealing at 573 K for 100 h in deuterium gas of ∼0.1 MPa. In both cases of electron- or neutron-irradiation, vacancy clusters were found by positron lifetime measurements. In addition, positron annihilation with deuterium electrons was demonstrated by coincidence Doppler broadening measurements, directly indicating deuterium trapping at vacancy-type defects. This is expected to cause significant increase in deuterium retention in irradiated-tungsten.

In this study, the cracking behavior and microstructural, mechanical and thermal characteristics of tungsten–rhenium (W–Re) binary alloys fabricated by laser powder bed fusion (L-PBF) were investigated. Four bulk specimens were prepared by L-PBF: pure W, W–1%Re, W–3%Re and W–10%Re (percentages indicate the mass percent of Re). High-density bulk specimens (relative density > 98.0%) were obtained for pure W and W–Re alloys under the same laser irradiation conditions. The columnar grains elongated along the building direction were gradually refined as the Re content increased. The most remarkable grain refinement was observed for the W–10%Re alloy. Hardness under a high-temperature environment increased with increasing Re content; the micro-Vickers hardnesses of pure W and W–10%Re at 400 °C were 179 ± 4 HV0.1/30 and 281 ± 5 HV0.1/30, respectively. Observations with a scanning electron microscope revealed that the 10 mass% Re addition resulted in a shorter and narrower crack morphology in comparison with pure W and consequently reduced crack area by 59%. Furthermore, the anisotropy of the thermal diffusivity was mitigated in the high Re content specimens, suggesting that, at high Re content, thermal diffusivity is affected less by cracks than by the effect of Re atoms on heat carrier transfer via isotropic scattering.

The number of DNA double strand breaks (DSBs) of genome-sized DNA was quantitatively evaluated under several types of radiation sources. The characteristics of radiation damages by different linear energy transfer (LET) and the effect of radiation protection by tea catechins were also studied as a function of its concentration. After β-rays and γ-rays irradiation to samples, the length of genome size DNA molecules (bacteriophage T4 GT7 DNA; 166 kbp) was measured by single molecule observation method using fluorescence microscope, which can estimate DSBs quantitatively. It was found that the number of DSBs was increased with increasing LET due to high radical density. By addition of EGCg ((-)-epigallocatechin gallate), the number of DSBs was reduced with a small concentration of 1 µM.

Four sets of targets were used in this study: (1) Fe targets surrounded with 304 type stainless steel composed of mid-Z elements: Fe, Cr, Ni, and Mn (designated as Fe[304SS] targets), (2) Fe targets surrounded with high-Z tungsten (designated as Fe[W] targets), (3) Cu targets surrounded with mid-Z copper (designated as Cu[Cu] targets), and (4) Cu targets surrounded with high-Z tungsten (designated as Cu[W] targets). The targets were exposed to low-energy (140 and 200 eV), high-flux (about 1022 D/m2s) deuterium (D) plasmas at various temperatures in the range from 355 to 740 K. The surface morphology of the Fe and Cu targets is found to be dependent strongly on atomic number of re-deposited species and on the exposure temperature. For the Fe[W] and Cu[W] targets, due to formation of the W-enriched nano-sized structures on the target surfaces, the sputtering erosion yield is lower than that for the Fe[304SS] and Cu[Cu] targets, respectively. For the Fe[304SS], Fe[W], and Cu[W] targets, the sputtering erosion yield is increased distinctly as the exposure temperature rises from 355 to 740 K.

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.

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.

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.

Dependence of irradiation temperature on tritium retention has been studied using stainless steel type 316L (SS316L) as a model sample. A mixture of D2+ and DT+ ions was used and two kinds of ion energy, 0.5 and 2.5 keV, were applied. In case of irradiation tests by 0.5 keV, tritium retention decreased with increasing temperature up to 523 K, while above this temperature it contrarily showed an increase tendency. Such a concave change was not observed for irradiation tests at 2.5 keV. The retention was almost same until 400 K, but above this temperature it decreased gradually. It was seen from the analyses by X-ray photoelectron spectroscopy that most of surface was initially covered with the carbon and oxygen species at room temperature. Among metallic elements, constituents such as Fe and Ni were metallic states more than 60 % at room temperature, while metallic chromium atoms were little observed. Both fractions of the metallic chromium and iron atoms in the major base metals of SS316 L increased with an increase in temperature, but metallic nickel atoms relatively decreased. It was suggested, therefore, that real surface states of the irradiation materials play an important role for behavior of tritium retention.

The estimation of tritium permeation rate through the plasma facing wall into coolant is required to discuss tritium balance in a D-T fusion plant, to design tritium recovery system and to perform safety assessments. In this work, tritium permeation rates in the blanket first wall and the divertor were estimated by numerical analysis for simplified multi-layer structures with considering the temperature distribution in recent JA DEMO condition. The permeation rate in the blanket first wall, which was a double layer consisting of tungsten and F82H, was estimated to be 0.69 g/day. The permeation rate in the divertor, which was a triple layer consisting of tungsten, copper and copper alloy or F82H, was estimated to be 0.013 g/day. When the permeation rate in tritium breeding region in the blanket can be reduced by three orders of magnitude due to a permeation barrier, total tritium permeation rate in the blanket and the divertor was estimated to be 0.71 g/day.

To investigate systematically the effects of Re and other elements on deuterium (D) retention, W and binary W-alloys (Re, Mo and Ta) were irradiated with 6.4 MeV Fe ions at high temperatures (1073–1273 K) and then exposed to D2 gas at 673 K. Depth profiles of D were measured by nuclear reaction analysis (NRA), and D retention was determined by thermal desorption spectrometry (TDS) and NRA. The addition of 5 at.% Re into W reduced the content of D trapped at radiation-induced defects created by irradiation with the Fe ions at 1273 K to the peak damage level of 5 displacements per atom (dpa). At Re fractions of 1, 3 and 5 at.%, comparable effects on D retention were observed after irradiation to 0.5 dpa at 1073 K. The addition of Mo and Ta to W resulted in no visible effects in D retention.

This paper reviews recent results pertaining to hydrogen isotope behavior in neutron and heavy ion damaged W. Accumulation of damage in W creates stable trapping sites for hydrogen isotopes, thereby changing the observed desorption behavior. In particular, the desorption temperature shifts higher as the defect concentration increases. In addition, the distribution of defects throughout the sample also changes the shape of TDS spectrum. Even if low energy traps were distributed in the bulk region, the D diffusion toward the surface requires additional time for trapping/detrapping during surface-to-bulk transport, contributing to a shift of desorption peaks toward higher temperatures. It can be said that both of distribution of damage (e.g. hydrogen isotope trapping sites) and their stabilities would have a large impact on desorption. In addition, transmutation effects should be also considered for an actual fusion environment. Experimental results show that production of Re by nuclear reaction of W with neutrons reduces the density of trapping sites, though no remarkable retention enhancement is observed.

Hydrogen addition to a sweep gas of a solid breeder blanket module has been proposed to enhance tritium recovery from the surface of the breeders. However, the influence of hydrogen addition on the bred tritium behavior is not understood completely. Tritium behavior in the simplified blanket module of Li 2 TiO 3 pebbles was numerically calculated considering diffusion in the grain bulk, surface reactions on the grain surface and permeation through the cooling pipe. Although a partial pressure of T 2 increases with increasing a partial pressure of H 2 in the sweep gas, it was estimated that tritium permeation rate to the cooling water decreases. Additionally, the release duration of water vapor generated by the reaction of the pebbles and hydrogen is shortened with increasing a partial pressure of H 2 . Tritium inventory in the grain bulk and the grain surface occupies 99.6 % of total tritium inventory in the blanket module.

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.

After introduction of Japanese history or recent topics on tritium (T)-relating research and T-handling capacity in facilities or universities, present activities on T engineering research in Japan are summarized in short in terms of T accountability on safety. The term of safety includes wide processes from T production, assay, storing, confinement, transfer through safety handling finally to shipment of its waste. In order to achieve reliable operation of fusion reactors, several unit processes included in the T cycle of fusion reactors are investigated. Especially, the following recent advances are focused: T retention in plasma facing materials, emergency detritiation system including fire case, T leak through metal tube walls, oxide coating and water detritiation. Strict control, storing and accurate measurement are especially demanded for T accountability depending on various molecular species. Since kg-order T of vaporable radioisotope (RI) will be handled in a fuel cycle or breeding system of a fusion reactor, the accuracy of

In this study, 1 keV DT + ion irradiation was performed on tungsten pre-irradiated with helium. The thermal desorption behavior and the reduction of tritium retention during vacuum preservation at room temperature, as well as isochronal annealing were investigated using an IP technique, taking advantage of the fact that tritium detection is nondestructive and is highly sensitive. At a pre-irradiated helium fluence of 1 × 10 17 He/cm 2 , retained tritium tended to be desorbed at higher temperatures when compared to no helium irradiation case. Tritium retention during preservation in vacuum and during isochronal annealing became smaller with increasing helium fluence up to 1 × 10 17 He/cm 2 . At a helium fluence of 1 × 10 18 He/cm 2 , the reduction of tritium retention was found to be greater compared to 1 × 10 17 He/cm 2 . The results indicate that helium irradiation has a significant influence not only on the thermal desorption temperature of tritium but on longtime tritium reduction at room and elevated temperatures.

An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

In this study, the effects of fabrication conditions on the microstructure, pore characteristics and gas retention of pure tungsten specimens prepared by laser powder bed fusion (LPBF) were investigated. The LPBF specimens contained micro- and nano- sized pores as internal defects. By optimizing the laser irradiation conditions, the formation of micropores was suppressed. The densest LPBF specimen was obtained when the input energy density was adjusted to be 411 J/mm3, and the relative density of the specimen measured by utilizing Archimedes' principle reached 98.58 ± 0.25% (19.03 ± 0.05 g/cm3). Heat treatment at a high temperature (1900 °C) was effective to reduce the number density of nanopores. The compositions and amounts of internal gas in the specimens were examined using a quadrupole mass spectrometer. The results indicated that the specimens contained the argon (Ar), which was used as the shielding gas. The Ar retention was correlated with the number density of nanopores in the specimens and not with the density of micropores. The Ar retention decreased to almost half that in the as-fabricated specimen after the heat treatment at 1900 °C. These observations suggest that Ar gas was trapped in nanopores in the LPBF material. It was demonstrated that a part of the trapped Ar dissolved and diffused in grains during the heat treatment and was released to the outside.

Single crystalline W (tungsten) samples irradiated at 633, 963 and 1073 K by neutrons to a damage level of 0.1 dpa were exposed to a high-flux D (deuterium) plasma at 673, 873 and 973 K, respectively, in TPE (Tritium Plasma Experiment) at INL (Idaho National Laboratory). Deuterium desorption was analyzed by TDS (Thermal Desorption Spectroscopy), and D depth profiles were determined by NRA (Nuclear Reaction Analysis) at SNL (Sandia National Laboratories). HIDT (Hydrogen Isotope Diffusion and Trapping) simulation code was applied to evaluate D behavior for neutron-damaged W at higher temperature. The D retention at depths up to 3 μm for the neutron-damaged sample at 673 K was two orders of magnitude larger than that for undamaged tungsten, and its D desorption spectrum had a single broad stage at around 900 K. As the neutron irradiation/plasma exposure temperature increased, D retention was largely reduced, and the desorption temperature was shifted to higher temperatures above 1100 K. The D depth profiles by NRA also showed D migration toward bulk by higher temperature irradiation, compared to undamaged W. The HIDT simulation indicated that the major binding energy of D was changed from 1.43 eV to 2.07 eV at higher neutron irradiation and plasma exposure temperatures, suggesting that some vacancies and small vacancy clusters would aggregate to form larger voids, or depopulation of weak traps at high D plasma exposure temperatures. It can be said that more stable trapping sites played dominant roles in the D retention at higher neutron irradiation and plasma exposure temperature. The binding energy by HIDT simulation was almost consistent with the reported value by TMAP, but the consideration of not only total D retention measured by TDS but also D depth profile by NRA led to the more accurate D behavior in neutron-damaged W.

Tritium permeation and retention are serious problems in D-T fusion reactors from the viewpoint of fuel efficiency and radiological safety. Functional ceramic coatings have been intensively studied for the development of tritium permeation barriers for several decades, while reports about tritium retention in the ceramic coatings are scarce. Moreover, irradiation may affect tritium retention in fusion materials, which is important to precisely evaluate tritium inventory in the reactor. In this study, the gamma-ray irradiation effect on deuterium retention in reduced activation ferritic/martensitic steel and three kinds of ceramic coatings were investigated through deuterium exposure, gamma-ray irradiation using cobalt-60 gamma-ray sources and deuterium depth profile measurements. The amount of deuterium retention in yttrium oxide, silicon carbide, and zirconium oxide coatings decreased after the irradiation in the dose rate of 2.43 Gy s−1, while no clear change in the retention was observed at the lower dose rate. From these results, the gamma-irradiation effect on deuterium retention would have a threshold dose rate. Diffusion and desorption of deuterium would be accelerated by excitation of deuterium via energy transfer from electrons generated by Compton scattering.

In the present study, successive exposure to tritium gas and deuterium gas was performed at elevated temperatures for tungsten samples with irradiation defects created by helium ion irradiation to investigate the effects of hydrogen isotope exchange on tritium removal. It was found that, under the present heating conditions, heating treatment in deuterium gas was more effective for tritium removal than heating in a vacuum. In addition, heating in deuterium gas was effective to remove tritium retained in a deep region of sample comparing to the case of heating in a vacuum. The enhanced tritium release was explained by enhanced tritium detrapping due to trap occupation by deuterium and/or the increase in probability of surface recombination under deuterium gas exposure.

To study the formation mechanism of stable α-Al2O3 scales, the oxidation behavior of Fe–Al(Cr) diffusion coating on China low activation martensitic steel has been investigated under the oxygen partial pressure ranging from 1 to 20,000 Pa at 1253 K. A single, continuous Al2O3 scale with the maximum thickness of about 2000 nm was formed on the Fe–Al(Cr) diffusion layer. The phase transformation of alumina scales on the surface of Fe–Al(Cr) layer was studied at different oxidation times ranging from 3 to 180 min. With the increase in oxygen partial pressure, the phase transformation time of α-Al2O3 is decreased. The metastable γ-Al2O3 and transition α-(Al0.948Cr0.052)2O3 phases were formed in the earlier oxidation process and finally transformed to the stable α-Al2O3 phase, which were detected by grazing incidence angle X-ray diffraction and confirmed by transmission electron microscopy. This implies that Cr shows the third element effect and serves as a template for the nucleation of the stable α-Al2O3.

The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900 °C, and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium-free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined.

Three tungsten samples irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to deuterium plasma (ion fluence of 1 × 1026 m−2) at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy was performed with a ramp rate of 10 °C min−1 up to 900 °C, and the samples were annealed at 900 °C for 0.5 h. These procedures were repeated three times to uncover defect-annealing effects on deuterium retention. The results show that deuterium retention decreases approximately 70% for at 500 °C after each annealing, and radiation damages were not annealed out completely even after the 3rd annealing. TMAP modeling revealed the trap concentration decreases approximately 80% after each annealing at 900 °C for 0.5 h.

DT + ion irradiation with energy of 0.5 and 1.0 keV was performed on helium pre-irradiated tungsten and the amount of retained tritium and the long-term release of retained tritium in vacuum was investigated using an IP technique and BIXS. Tritium retention and long-term tritium release were significantly influenced by helium pre-irradiation. The amount of retained tritium increased until it reached 1 × 10 17 He/cm 2 , and at 1 × 10 18 He/cm 2 it became smaller compared to 1 × 10 17 He/cm 2 . The amount of retained tritium in tungsten without helium pre-irradiation largely decreased after several weeks preservation in vacuum, and the long-term release rate during vacuum preservation was retarded by helium pre-irradiation. The results indicate that the long-term tritium release and the helium irradiation effect on it should be taken into account for more precise estimation of tritium retention in the long-term use of tungsten in fusion devices.

W samples were irradiated at 300 K with 6 MeV Fe ion with damage concentrations in the range from 0.0003 to 1.0 displacements per atom (dpa) and then implanted at 300 K with 500 eV D ions to a fluence of 5 × 10 21 D/m 2 . Deuterium retention in the damaged samples was examined in situ by thermal desorption spectrometry (TDS). Simulation of the TDS spectra was performed using the Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code to reveal the binding energies for deuterium captured by the ion-induced defects. It has been shown that the deuterium TDS spectra consist of two or three peaks (depending on the damage concentration) at about 400, 600 and 800 K, and can be simulated by the HIDT simulation code with the use of hydrogen-trap binding energies of 0.65, 1.25, and 1.55 eV.

Fatigue crack propagation (FCP) tests were performed in NaCl solution under controlled cathodic potentials to achieve the hydrogen charged condition where anodic dissolution does not occur so much to understand the effect of hydrogen on FCP behavior of wrought magnesium alloy AZ61. FCP rates were accelerated under the hydrogen charged conditions compared with dry air. A grazing incidence X-ray diffraction (GIXRD) and a thermal desorption spectrometry (TDS) analysis revealed that FCP rates have no relation to hydrogen compounds formed near the crack surface. This indicates the acceleration could be mainly attributed to hydrogen diffusion and hydrogen embrittlement is dominant in the FCP behavior of AZ61.

Surface topography and deuterium retention in polycrystalline hot-rolled W and W–5%Re have been examined after exposure to a low-energy (76 eV), high flux (around 1022 D/m2 s) deuterium plasma to an ion fluence of 1026 D/m2 at various temperatures. The methods used were confocal laser scanning microscopy and the D(3He, p)4He nuclear reaction at 3He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 348 to 673 K, small blisters of size in the range from about 1 to about 15 μm, depending on the exposure temperature, are formed on the W and W–5%Re surfaces. In the W–5%Re, the deuterium retention demonstrates its maximum at exposure temperature of 463 K, while in the W this maximum is shifted to 523 K. A difference in the temperature dependence of the D retention for the W and W–5%Re is explained, as a rough approximation, by temperature dependences of the ductility of these materials.

To evaluate hydrogen isotope retention enhancement in W by plasma exposure, the stress relieved tungsten samples were placed at three or four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (Higher heat load area) and ER (erosion dominated area) during 2011 (15th) or 2012 (16th) plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼6700 shots of hydrogen plasma in a 2011 plasma experiment campaign and ∼5000 shots in a 2012 plasma campaign. Thereafter, additional 1.0 keV deuterium ion implantation was performed to evaluate the change of hydrogen isotope retention capacity by plasma exposure. It was found that more than 50 times of hydrogen retention enhancement for DP sample was derived compared to that for pure W. In especially, the carbon-dominated mixed-material layer would control the hydrogen isotope retention for all the area except for erosion-dominated area, indicating that the chemical structure for carbon-dominant mixed-material layer would govern the H and D retention enhancement for most area by long-term plasma exposure. Therefore, the surface area for carbon material would be one of key issues for the determination of hydrogen isotope retention in first wall, even if all tungsten first walls will be used.

DT + ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both the amount of retained tritium and the reduction of retained tritium after preservation in vacuum were investigated using an IP technique and BIXS. In addition, the relationship between the retention properties of tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. The amount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1 keV of DT + irradiation, graphite showed no reduction of the amount of retained tritium after six months preservation while that of carbon film decreased by approximately 20% after 40 days preservation. It was suggested that this difference might be associated with differences in the microstructure between graphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately half after 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decrease in tritium retention during preservation became slower. Tungsten's properties of releasing tritium while preserved in vacuum would be a useful tool for the reduction/removal of retained tritium.

Stable structures of hydrogen atoms trapped in a divacancy in tungsten and their binding energies are presented on the basis of first-principle calculations. The hydrogen atoms are favorable sitting in the vicinity of octahedral interstitial sites (O-sites) next to the divacancy. Besides, hydrogen atoms preferentially occupy O-sites located in the center of the divacancy. As hydrogen atoms increases, O-sites located in the periphery of the divacancy are also occupied by the hydrogen atoms. The divacancy in tungsten is energetically unstable, compared with two isolated monovacancies. However, the divacancy is extremely stabilized by the hydrogen atom trapping. The binding energy of the divacancy depends on the sort of the hydrogen isotope.

Using a tritium imaging plate technique, kinetics of tritium dissolution and its solubility in several oxides were examined. Mirror-polished single crystals of alumina, spinel and zirconia were used as specimens, which were exposed to 133 Pa of a tritium(T)–deuterium(D) gas mixture (T/(T + D) ∼ 0.17) at temperatures ranging from 673 to 973 K for 1–5 h. The T surface activity on the specimens increased with increasing temperature and exposure time, it almost saturated at 873 K and reached 2 × 10 5 Bq/cm 2 (1 × 10 14 T/cm 2 ), and no clear difference appeared among the types of specimens. The T activity in the oxide bulk also increased with temperature, in which there was a trend for the oxides: spinel ≧ zirconia ≧ alumina. In the T dissolution process for all oxides, the concentration gradient due to its diffusion was not observed even for short exposure times: the T density was almost uniform over the specimens in transition states and increased with exposure time up to the saturated value. These experimental results suggested that the rate-controlling process of T dissolution in the temperature region should be not its diffusion in the oxides but dissociation of hydrogen molecules (T–D mixture in this case) into atoms, its adsorption on the surface and/or T penetration from the surface into the bulk.

The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C β-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ∼3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.