The effects of rhenium (Re) addition on deuterium (D) retention in neutron-irradiated tungsten (W) were investigated. Pure W and W-5Re (5 at.%) alloy samples were irradiated with neutrons at High ...Flux Isotope Reactor using MFE-RB-19 J capsule. The sample temperature and the damage level were 864 K and 0.35 dpa for pure W and 792 K and 0.26 dpa for W-5Re alloy. A portion of the samples was exposed to D plasma at Tritium Plasma Experiment at Idaho National Laboratory at 823 K to a fluence of 5 × 1025m−2. Vacancy-type defects in neutron-irradiated samples were examined using positron annihilation spectroscopy (PAS); D retention after plasma exposure was evaluated by thermal desorption spectrometry (TDS).
TDS measurements revealed that D retention in the neutron-irradiated W-5Re alloy was similar to that in the unirradiated W sample, whereas a significant increase in D retention was observed in neutron-irradiated W. Thus, Re addition significantly suppressed the increase in D retention after neutron irradiation. This effect was attributed to the suppression of vacancy-type defect formation, as confirmed by PAS.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
DT+ ion irradiation with energy of 0.5 and 1.0keV 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 1×1017He/cm2, and at 1×1018He/cm2 it became smaller compared to 1×1017He/cm2. 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.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
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.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
Remaining tritium in the vacuum vessel after the first deuterium plasma experimental campaign conducted over four months was investigated in the large helical device (LHD) for the first time in ...stellarator/heliotron devices by using the tritium imaging plate technique. In-vessel components such as divertor tiles and first wall panels, and long-term material probes retrieved from the vacuum vessel were analyzed. The in-vessel component in which tritium remained most densely is the baffle part of divertor tiles made of graphite retrieved from the inboard-side divertor. Asymmetric tritium retention is observed on divertor tiles located at magnetically symmetric positions, and can be attributed to the toroidal field direction dependence of the asymmetric loss of energetic tritons generated by deuterium-deuterium nuclear fusion reactions. On the first wall, tritium remained in a deposited layer, which mainly consists of carbon.
The deuterium retention/desorption behavior of F82H has been studied to clarify the effect of radiation damage with respect to the impurity layer. To evaluate the effects of oxygen impurities, ...non-oxidized and thermally oxidized samples were prepared with the introduction of radiation damage through He+ irradiation. The amount of deuterium retained in the thermally oxidized samples was less than that of the non-oxidized samples. The amount of deuterium retained in the oxidized sample increased after He+ irradiation. The retention/desorption behaviors were discussed in the surface chemical state.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
Lactobacillus brevis KB290 (KB290), a plant-derived probiotic lactic acid bacterium, reportedly improves gut health and stimulates immune function. Here we extensively investigated the geno-, acute, ...subacute, and subchronic toxicity of KB290 and its bacterial translocation potential. KB290 was non-mutagenic in the bacterial reverse mutation assay by the preincubation method. In the single oral dose toxicity test, KB290 at ⩾10
9 cfu/ml was nontoxic at maximum capacity (20
ml/kg). When 10
8, 10
9, or 10
10 cfu/kg was administered daily to rats by gavage for 2
weeks (subacute assay), we observed no clear treatment-related effect and no evidence of bacterial translocation from the gastrointestinal tract. When it was administered for 13
weeks (subchronic assay), we again observed no clear treatment-related effect and no significant toxicological effect. Based on those results, we consider 10
10 cfu/kg per day, the highest dose tested, to be the no observed adverse effect level (NOAEL). These results suggest that KB290 is safe for human consumption.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
•D retention in W irradiated by high-dose neutrons at high temp. were investigated.•Neutron irradiation temperature affected D retentions; 894 > 1074 ≈ 1379 K.•TDS and D implantation simulation were ...performed to estimate retention properties.•Detrapping energy, depth and D/W were estimated.
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.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPCLJ, UPUK, ZAGLJ, ZRSKP
•The purpose is to evaluate the effects of induced damage and surface modification on the surface of F82H for deuterium retention/desorption behavior.•2 types of the samples, the non-oxidized and the ...oxidized F82H, were examined.•The amount of retained deuterium of the non-oxidized F82H was larger than that of the oxidized F82H.•The desorption behavior was quite difference between the non-oxidized and the oxidized F82H.•The oxide layer of F82H greatly influenced the deuterium retention/desorption behavior.
The effects of induced damage on hydrogen isotope retention in F82H with or without thermal oxidation were investigated using thermal desorption spectroscopy. To induce damage and modify the surface, glow discharge pre-irradiated Ar+ ions was examined. In non-oxidized samples, the amount of desorbed deuterium increased with Ar+ ion fluence. Oxygen depletion in the surface layer of non-oxidized samples from the Ar+ ion irradiation, which resulted in bulk diffusion of deuterium, is responsible for the increase in deuterium retention. A comparison between non-oxidized and oxidized samples clearly indicated that the surface oxide layer greatly influenced deuterium retention/desorption behaviors of F82H.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
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 1keV 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.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
In order to examine the effect of neon glow discharge on hydrogen or helium removal, neon glow discharge was conducted for the stainless steel after the exposure to hydrogen or helium glow discharge, ...and then the amount of desorbed hydrogen or helium and retained neon were evaluated. Large hydrogen desorption was observed at the initial period of the neon discharge following the hydrogen discharge. The removal ratio of retained hydrogen by the neon discharge with 2
h was 1.3 times larger than that by the argon discharge, and a half of that by the helium discharge. In the case of the neon discharge following the helium discharge, the removal ratio of retained helium was 4 times larger than that by the argon discharge. The amount of retained neon was an order of magnitude smaller than that of helium retained in the stainless steel.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
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