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Tritium is a radioisotope of hydrogen emitting low energy β-rays in disintegration to

Neutron-irradiated tungsten (W) samples were exposed to helium (He)–seeded deuterium (D) plasmas using a linear plasma device called Tritium Plasma Experiment in order to investigate the synergetic...

In inertial confinement fusion (ICF), a fuel target containing deuterium and tritium is used. In recent ICF experiments on the Gekko XII LFEX facility at the Institute of Laser Engineering at Osaka...

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) > 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., 19 pages, 12 figures

Nondestructive measurement of tritium (T) content in solid materials is important for safe and cost-effective disposal of contaminated wastes, and beta-ray induced X-ray spectrometry (BIXS) has bee...

The deuterium (D) permeation behavior for 1 displacement per atom Fe2+ damaged tungsten (W) was studied by the gas-driven permeation method and compared with undamaged W. The results of thermal des...

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.

Tungsten (W) samples were damaged by neutron and 6.4 MeV Fe-ion irradiation above 1000 K simulating the divertor operation temperature. Deuterium (D) retention properties were examined by decorating the damaged W with D and subsequent thermal desorption spectroscopy (TDS) measurements. Vacancy clusters were the major D trapping site in the W irradiated with Fe-ion at 873 K, although D retention by vacancy clusters decreased in the W irradiated with Fe-ion at 1173 K due to dynamic annealing. The D de-trapping activation energy from vacancy clusters was found to be 1.85 eV. D retention in neutron damage W was larger than that damaged by Fe-ion due to the uniform distribution of irradiation defects. The D desorption behaviors from neutron damaged W was simulated well by assuming the D de-trapping activation energy to be 1.52 eV.

Negatively charged nitrogen-vacancy (NV) centers in diamond have emerged as promising candidates for a wide range of quantum applications, especially quantum sensing of magnetic field. Implementation of nanostructure into diamond is powerful for efficient photon collection of NV centers and chip-scale miniaturization of the device, which is crucial for sensitive and practical diamond magnetometers. However, fabrication of the diamond nanostructure involves technical limitations and can degrade the spin coherence of the NV centers. In this study, we demonstrate the hybrid integration of a silicon nitride grating structure on a single-crystal diamond by utilizing transfer printing. This approach allows the implementation of the nanostructure in diamond using a simple pick-and-place assembly, facilitating diamond-based quantum applications without any complicated diamond nanofabrication. We observed the intensity enhancement in the collected NV emissions both theoretically and experimentally using the integrated grating structure. By applying the increased photon intensity, we demonstrate the improved magnetic sensitivity of the fabricated device. The proposed hybrid integration approach will offer a promising route toward a compact and sensitive diamond NV-based magnetometer.

We investigated deuterium (D) retention in three W samples irradiated with MeV neutrons at high damage level of 0.39 ~ 0.74 displacements per atom (dpa) at high temperatures, 894 K, 1074 K and 1379 K. The W specimens were exposed to high-flux (~1 × 1022 m−2 s−1) and high-fluence (~5 × 1025 m−2) D plasma at 873 K in the Tritium Plasma Experiment. Broad desorption peaks extended from 900 K to 1200 K were observed for the neutron-irradiated W by thermal desorption spectroscopy (TDS). The retention in neutron-irradiated specimens was much larger than for an un-irradiated specimen. The highest D retention was obtained for a specimen irradiated at 894 K. With increasing neutron irradiation temperature, the retention was reduced about by half at 1074 K and further increase of the temperature (1379 K) resulted in comparable retention. In addition, one-dimensional diffusion calculations (D desorption in TDS and D depth distribution in plasma exposure) were performed to derive retention parameters (the detrapping energy, the depth occupied by D atoms and D/W ratio) from experimental D retention properties of neutron-irradiated W. By TDS simulation calculation, simple dependences of the peak temperature, height and width of TDS peaks on the retention parameters were obtained with total retention in the orders of 1019 ~ 1022 m−2. The calculation of the depth distribution of trapped D atoms made a relationship between the D/W ratio and the depth occupied by D atoms after plasma exposure at relevant conditions. By comparing the relationship (the D/W and the depth) with that obtained from the experimental results, we estimate each retention parameters for the specimens irradiated by high-dose neutrons at the high temperatures. And, we discussed the neutron-irradiation temperature dependence of D retentions.

W samples, previously irradiated with 20 MeV W ions at room temperature to a displacement-damage level of 0.23 displacements per atom (dpa) at the peak of displacements, were exposed to a deuterium-tritium (D2 ― DT ― T2) gas mixture with a tritium content of 6% at a total pressure of 1.2 kPa and temperatures of 773 K and 973 K for 3 h. The concentration of tritium retained in the displacement-damage zone of these W samples was determined by a method combining chemical etching and subsequent analysis of tritium in the etching solution using liquid scintillation counting (CE-LSC). (Some figures may appear in colour only in the online journal)

Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.

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 (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

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+.

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.

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.

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.