Abstract Divertor tiles after Joint European Torus-ITER like wall (JET-ILW) campaigns and dust collected after JET-C and JET-ILW operation were examined by a set of complementary techniques (full combustion and radiography) to determine the total, specific and areal tritium activities, poloidal tritium distribution in the divertor and the presence of that isotope in individual dust particles. In the divertor tiles, the majority of tritium is detected in the surface region and, the areal activities in the ILW divertor are in the 0.5–12 kBq cm⁻² range. The activity in the ILW dust is associated mainly with the presence of carbon particles being a legacy from the JET-C operation. The total tritium activities show significant differences between the JET operation with ILW and the earlier phase with the carbon wall (JET-C) indicating that tritium retention has been significantly decreased in the operation with ILW.

A model of deuterium permeation through liquid (liq.) Sn supported by a nickel (Ni) substrate is proposed as follows; deuterium dissolves in liq. Sn according to the Sieverts’ law and its concentration immediately becomes uniform through the liq. Sn layer due to large diffusivity of deuterium in liq. Sn. A thin intermediate Sn-Ni alloy layer is formed at an interface of liq. Sn and the Ni substrate in a transient of deuterium permeation. Deuterium concentration and diffusivity in the alloy layer would be lower than those in liq. Sn and pure Ni, resulting in lower deuterium permeability. Consequently, the intermediate Sn-Ni alloy layer could control total deuterium permeation through the liq. Sn supported by the Ni substrate.

Abstract Tritium retention in the castellated structure of beryllium limiters used in JET with the ITER-like wall (ILW) during the first (ILW1), third (ILW3) and all three (ILW1-3) campaigns were examined and evaluated. Tritium was deposited on the surfaces inside the castellation grooves together with deuterium, beryllium, oxygen, carbon and small amounts of metallic impurities such as nickel, copper and tungsten. The tritium content after the ILW1 campaign was greater than after the ILW3 campaign. This is attributed to the steadily decreasing amount of carbon impurities in JET from campaign to campaign. The majority of tritium was retained in shallow regions in the grooves, up to 2 mm from the entrance to the gap. It was comparable on all sides of the castellation, i.e. no difference has been detected between the toroidal and poloidal gaps. Secondly, the tritium retention in the gaps was similar on all specimens independent of their position in the tokamak, while the retention on the plasma-facing surfaces clearly depended on the tile position. The tritium deposition patterns in the castellation were also compared with the deuterium distribution determined in earlier studies.

The diffusivity values of C in pure Zr and in Zr containing dissolved O (Zr(O)) were determined in a temperature range of 623–1023 K for Zr and 923–1123 K for Zr(O) from depth profiles of C obtained by glow-discharge emission spectroscopy. The C diffusivity in Zr(O) decreased as the O concentration in Zr increased and that for Zr containing 15 at % O was two orders of magnitude smaller than that for pure Zr. One-dimensional numerical calculations of Fick’s diffusion equation with a Soret effect indicated various non-uniform distributions of C in a 5-mm-thick Zr matrix under a temperature gradient of 573 to 773 K for 3 years, assuming a heat of transport of − 1.5 to + 1.5 eV. Isothermal annealing at 773 K for 10 years could result in a uniform distribution, whereas dissolution of O in the interstitials of the Zr matrix would hinder C transport through the interstitials. Graphical Abstract: As concentration of O increased in HIP Zr by 15 at%, the diffusivity of C decreased more than two orders of magnitude.For the positive Q∗c, concentration of C slightly segregated at the surface of the cooler side but had maximum peaks at a middle to a higher temperature zone, and depleted at the surface of the hotter side.For the negative Q∗c, concentration of C depleted at the surface of the cooler side and at the middle to the higher temperature zone, and highly segregated at the surface of the hotter side. [Figure not available: see fulltext.].

Effects of a gap/open space between double membranes of Ni/Ni and Pd/Ni on hydrogen permeation through the double membranes are studied. For easy detection of permeated H, T is introduced. For Ni/Ni and Pd/Ni, the influence of the gap on hydrogen permeation is not appreciable, while the permeation for Ni/Pd is significantly reduced because the gap holds H2O produced by the reaction of permeated hydrogen and the surface oxide of Ni facing toward the gap; consequently, the partial pressure of H2O in the gap becomes high and subsequent reduction of the surface oxide is prohibited. From these findings, a new double-walled tube concept for the reduction of T permeation is proposed with a combination of a rather thin front tube with its back side oxidized as a permeation barrier and a thick tube as a structure material.

Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015–2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011–2012 (ILW-1). The dominant component in ILW-1 dust was carbon (C) originating from tungsten-coated carbon fibre composite (CFC) tiles in JET-ILW divertor and/or legacy of C dust after the JET operation with carbon wall. Around 85% of the total tritium retention in ILW-1 dust was attributed to the C dust. The retention in tungsten (W) and beryllium (Be) dominated particles was 100 times smaller than the highest T retention in carbon-based particles. After ILW-3 the main component contributing to the T retention was W. The number of small W particles with T increased, in comparison to ILW-1, most probably by the exfoliation and fragmentation of W coatings on CFC tiles though T retention in individual W particles was smaller than in C particles. The detection of only very few Be-dominated dust particles found after ILW-1 and ILW-3 could imply stable Be deposits on the divertor tiles.

Tritium is required for research and development activities for the deuterium–tritium (DT) fusion reactor and fueling the DEMOnstration Power Station (DEMO). However, tritium is a very rare nuclide and must be produced artificially. Tritium production by loading Li compounds (Li rods) into burnable poison holes of a high-temperature gas-cooled reactor (HTGR) has been proposed (H. Matsuura, et al., Nucl. Eng. Des. 243 (2012) 95–101.). Al2O3 and Zr are used to prevent tritium leaks. Nuclear reaction heat caused by the nuclear reaction (e.g., 6Li(n,α)T reaction) can cause a spatial temperature profile in the Li rods and may change its tritium containment performance, because Al2O3 and Zr performance strongly depend on these temperatures. The effect of nuclear reaction heat by the 6Li(n,α)T reaction on the tritium containment performance of the Li rods was evaluated by simulation. The temperatures of the Li rods for the high-temperature engineering test reactor (HTTR) and gas turbine high-temperature reactor 300 (GTHTR300) increased by 36 K and 46 K, and the leaked tritium decreased by 32% and 37% via nuclear reaction heat, respectively.

Behavior of hydrogen trapping in defects induced by cold-working (CW) of pure aluminum (Al) and Duralumin (A2017) were examined by thermal desorption analysis (TDA). The total desorption amount of deuterium in pure Al and A2017 increased with the working rate of CW; the desorption amount for hydrogen trapped at dislocations increased though that at vacancies and their clusters was likely saturated. Strong hydrogen trapping was observed in A2017, which disappeared by CW.

The production of tritium (T) using high-temperature gas-cooled reactors (HTGRs) has been studied for a prior engineering research with T handling and initial T possession in demonstration fusion reactors. Stable containment of T in Li-loading rods during HTGR operation is a critical issue. This study investigates this for an irradiation test to examine T-containment performance in Li-loading rods and develops an analytical model of evaluating the amount of T outflow to a He coolant. The hydrogen absorption characteristics, including the deterioration of the hydrogen absorption speed after Zr has sufficiently absorbed the hydrogen, is experimentally measured assuming an HTGR setting. We present an analytical model of evaluating the T outflow from a Li rod and, on the basis of this model, estimate the total T outflow, assuming the presence of a gas-turbine high-temperature reactor of 300 MWe with a nominal capacity and a high-temperature engineering test reactor. It is demonstrated that, by loading a sufficient amount of Zr into the Li rod, the T outflow can be suppressed to less than a small percent of the total T produced during 360 days of reactor operation.

In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD.

Nondestructive analysis of tritium (T) distribution was performed by means of imaging plate technique on specimens cut from the Be limiters, W-coated carbon tiles and bulk W lamellae retrieved from the JET tokamak after the first and third experimental campaigns with the ITER-like wall. Afterwards, analyses were continued using X-ray photoelectron spectroscopy, microscopy techniques and thermal desorption spectroscopy. Co-deposits formed on the W-coated tiles in the 1st campaign showed large T retention because of high carbon content reaching up to 50 atomic %, while the carbon fraction in co-deposits after the 3rd campaign was distinctly lower. The T retention of the plasma-facing surface of the bulk W tile was smaller than that of the W-coated tiles by a factor of 20, while deposition of small amount of T was found at the side surfaces facing to the gaps in a lamella structure. The correlation of T distributions with surface morphology and the discharge conditions is discussed.

To reveal the triton transport and the tritium migration in a deuterium plasma experiment in the Large Helical Device (LHD), the distribution of the remaining tritium in divertor tiles made of graphite after the first deuterium plasma experimental campaign in 2017 was investigated. In this study, tritium contents in divertor tiles have been measured by using a full-combustion method. The asymmetric tritium retention in divertor tiles located at symmetric positions, which was found in the previous study by the surface tritium measurement using an imaging plate technique, has been confirmed by the results of the full-combustion method. The asymmetry is considered to be attributed to the asymmetric distribution of lost-points of energetic tritons in divertor. A depth profile of remaining tritium in a divertor tile estimated by using a combination of the imaging plate technique and a sputtering treatment shows that the peak of the profile locates at several micro-meters from the surface. This result suggests that the majority of the remaining tritium impinged upon the divertor tile as energetic tritons. In this study, a distribution of lost-points of energetic tritons has been calculated by using a Lorentz orbit following code (LORBIT) with taking into account divertor components. The obtained distribution has been compared with measured tritium distributions on divertor tiles. The measured and calculated distributions are similar to each other, but they are not the same. The difference between them can be attributed to plasma exposures during and after the deuterium plasma campaign.

© 2020 The Authors Oxidation kinetics of pure tungsten (W) and tungsten-rhenium (W-Re) alloys with Re contents of 1%, 5% and 15% has been examined in the oxygen gas atmosphere at 873 K. The oxidation kinetics of the W-Re alloys were well characterized by a parabolic rate law. During oxidation of the W-Re alloys, the Re oxide sublimated near the top surface of the oxide layer. The sublimation of the Re oxide may play roles as a sintering agent and/or a stress relief agent in the W oxide layer to be more resistant to oxidation than pure W. There will be an effective value of Re content in W oxide layers for suppression of oxidation depending on oxidation temperatures or atmospheres. The Re oxide in the gas phase could deposit at cooler remote area on materials surface. Furthermore, the Re oxide deposits can be easily moved or transported to different area by dissolving in a water content or a moisture of flowing gas atmosphere.

A technique to monitor the permeation behavior of tritium in metals to pure water was successfully developed. A metal membrane separated two containers: one is for tritium loading as an upstream side, and the other is for tritium permeation release as a downstream side. Tritium was loaded by gas absorption at controlled temperatures of 303 K, 323 K, and 373 K and pressures of 4 and 8 kPa at the upstream side. Pure water in the downstream side was automatically and continuously circulated to a solid scintillation counting apparatus by which the tritium concentration in the pure water was directly measured for more than 100 h. When the present technique was applied, almost diffusional permeation behavior of tritium at the nickel-water interface was demonstrated.

Hydrogen isotope retention and chemical state for the tiles exposed to plasma in the JET-ITER-like wall (ILW) during two campaigns in 2011-2012 (first campaign, ILW-1) and 2015-2016 (third campaign, ILW-3) were studied and compared by means of X-ray photoelectron spectroscopy and thermal desorption spectroscopy. In both campaigns the upper part of the inner divertor tiles was the deposition-dominated area, while erosion was observed on the outer divertor tiles. Therefore, higher deuterium retention was found on the inner divertor tiles. The major D desorption peak for the inner divertor tiles from ILW-3 was located at the temperature range of 470 degrees C to 520 degrees C, which was higher than measured after ILW-1: 370 degrees C to 430 degrees C. The XPS analyses showed the formation of a BeO layer on the ILW-3 inner divertor tiles, while after ILW-1 the layers also contained a significant amount of carbon. Deuterium retention was reduced toward the outer divertor tiles. The differences could be related to the difference in the power level in the two campaigns.

Tritium permeability through FeCrAl-oxide-dispersion-strengthened (ODS) ferritic steel containing Ce oxides (Ce-ODS steel) was measured at temperatures ranging from 373 to 623 K. Some of the Ce-ODS steel specimens were oxidized by means of an autoclave treatment at 563 K for 30 days to examine the effect of the surface oxidized layer on the tritium permeability. The tritium permeability obtained for nonoxidized specimen was consistent with that for other common ferritic steels and FeCrAl ferritic steel. For the oxidized specimen, the surface oxide layer suppressed the apparent tritium permeability. The permeability for the oxidized specimen also depended on the atmosphere of the downstream in the permeation experiment: An atmosphere containing water vapor yielded lower tritium permeability compared with a reductive one.

Surface characterization of bulk tungsten tiles (W lamellae) used during the first campaign of JET with the ITER-Like Wall (JET-ILW) was performed by means of microscopy and tritium imaging techniques. This is the first report regarding very detailed structural studies of W lamellae from the JET-ILW divertor. A special feature of the W lamellae installed in JET is the intrinsic network of micro-cracks detected on surfaces of the as-manufactured material. Analyses of different W lamellae samples on the plasma-facing surface reveal two types of surface structures caused by plasma impact: Areas with strong erosion and regions of mild plasma interaction. In regions of strong erosion, a thin modified layer (thickness of ∼20 nm) with a high density of defects including bubble-like structures has been formed. In addition, features indicating melting along edges of micro-cracks with the micro-scale plastic deformation have been identified.

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.

© 2019 Quantitative tritium inventory in dust particles from campaigns in the JET tokamak with the carbon wall (2007–2009) and the ITER-like wall (ILW 2011–2012) were determined by the liquid scintillation counter and the full combustion method. A feature of this full combustion method is that dust particles were covered by a tin (Sn) which reached 2100 K during combustion under oxygen flow. The specific tritium inventory for samples from JET with carbon and with metal walls was measured and found to be similar. However, the total tritium inventory in dust particles from the ILW experiment was significantly smaller in comparison to the carbon wall due to the lower amount of dust particles generated in the presence of metal walls.

Production of tritium using a high-temperature gas-cooled reactor (HTGR) has been studied for a prior engineering test with tritium handling and for the startup operation of a demonstration fusion reactor. For this purpose, the hydrogen absorption speed of Zr in a Li-loading rod for the reactor operation is experimentally measured, and an analysis model is presented to evaluate the tritium outflow from the Li rod in a high-temperature engineering test reactor (HTTR). On the basis of the presented model, the structure of the Li-loading rod for the demonstration test using the HTTR is proposed.

© 2018 The Authors A tritium imaging plate technique (TIPT) in combination with an electron-probe microscopic analysis (EPMA) were applied to examine tritium (T) retention characteristics in individual dust particles collected in the Joint European Torus with the ITER-like Wall (JET-ILW) after the first campaign in 2011–2012. A lot of carbon (C)-dominated dust particles were found, which would be pre-existing carbon deposits in the JET-C or released carbon particles from the remaining carbon-fiber components in the JET-ILW. Most of T was retained at the surface of and/or in the C-dominated dust particles. The retention in tungsten, beryllium and other metal-dominated dust particles is relatively lower by a factor of 10–100 in comparison with that in the C-dominated particles.

© 2018 Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified (“geological-like”) mixed-material deposition layer which mainly included Be and Ni with the thickness of ∼2 μm. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper.

Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by Li-6(n,alpha)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO2, alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod's tritium production and containment performance was presented.

Effects of the shot-peening on hydrogen permeation and retention behaviors were examined by applying tritium tracer techniques to the gas-driven hydrogen permeation experiments. Hydrogen permeability in shot-peened iron was reduced by a factor of ten in comparison with the as-received iron at lower temperatures of 298 K and 453 K. The effects were disappeared at higher temperatures than 473 K. Hydrogen was trapped at least two trapping sites induced by the shot-peening treatment. The amount of trapped hydrogen in the shot-peened surface was three times larger than that in the normal surface of the as-received iron.

To clarify factors controlling variation of chemical forms of hydrogen desorbed by permeation phenomena through the oxide dispersion strengthened ferritic (ODS) steels, permeation behaviors of hydrogen through pure alpha iron (Fe) and the ODS steel into a pure argon (Ar) gas atmosphere have been examined by hydrogen permeation experiments with a tritium tracer technique. Difference of chemical forms of hydrogen desorbed by permeation through pure Fe and the ODS steel was explained by oxygen potential at a downstream side which was determined by an amount of hydrogen supplied by permeation and a moisture content in the Ar gas atmosphere at a certain temperature. Due to very high stability of chromium oxides formed on the ODS steels, most of hydrogen desorbed as a hydrogen gas molecular form into the Ar gas atmosphere in the present experimental temperatures ranging from 303 K to 623 K.

Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011-2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign (2011-2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles were a single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. A sample from the apron of Tile I was deposition-dominated. Stratified mixed-material layers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thickness was similar to 1.5 mu m. By means of transmission electron microscopy, nano-size bubble-like structures with a size of more than 100 nm were identified in that layer. They could be related to deuterium retention in the layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition: a stratified mixed-material layer with the total thickness of 200-300 nm. The electron diffraction pattern obtained with transmission electron microscope indicated Be was included in the layer. No bubble-like structures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosion zone. This is consistent with the fact that the strike point was often located on that tile during the plasma operation. The study revealed the micro-and nano-scale modification of the inner tile surface of the JET ILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retention and dust formation. (C) 2017 Elsevier B.V. All rights reserved.

A new methodology for depth profiling of hydrogen in metals is developed applying a tritium imaging plate technique (TIPT) with cross sectional observation. Owing to its high sensitivity and wide dynamic range for tritium detection, depth distribution of hydrogen dissolved in the BCC metals such as tungsten (W) and steels are successfully obtained. The depth distributions enable us to determine reliable lattice diffusion coefficients of hydrogen in W and a ferritic/martensitic steel (F82H) within 20% errors taking into account three dimensional desorption/release from the surfaces of the sample metals. Hydrogen trapped at surface and subsurface are clearly separated from the dissolved one. In BCC metals, since the former could be much larger than the latter, observation of overall hydrogen behavior without knowing detailed depth distributions could lead to wrong estimation of diffusion coefficients and solubility.

Performance of tritium production for fusion reactors, using a high-temperature gas-cooled reactor (HTGR) is examined. From the viewpoints of tritium recovery and environmental safety, tritium outflow from Li rods should be suppressed to the same level as the liquid radioactive waste from the pressurized water reactors (PWRs) in Japan. Methods for suppressing tritium leakage from Li rods are studied. The tritium outflow is reevaluated accurately on the basis of non-equilibrium simulations and the influence of coolant temperature on tritium leakage is clarified. The approach using Zr in the Li rod to reduce the tritium pressure and the resulting suppression of tritium leakage are also investigated.The results of the non-equilibrium simulation show that the tritium outflow is approximately 40% lower than the outflow reported in a previous study. Although the electric power generation efficiency is reduced, lowering the coolant temperature to 600 K results in a reduction of the tritium outflow to similar to 1/30 compared to the outflow in the case of a coolant temperature of 800 K. The incorporation of Zr into the Li rod can suppress tritium outflow (to similar to 1/200 compared to the case without Zr) to below the outflow level in PWRs in Japan.

Release behaviors of tritium (T) from pure copper (Cu) and its alloys into pure water were examined at ambient temperature by a tritium imaging plate technique and a liquid scintillation counting technique. Two mechanisms govern the liberation of T into pure water; one is rapid release and the other is chronic release. The former is caused by diffusional release of T dissolved in bulk of Cu alloys and the latter by release of T strongly bound on/in surface oxide layers. (C) 2016 Elsevier B.V. All rights reserved.

To address the transport and trapping of hydrogen isotopes, several permeation experiments are being pursued at both Sandia National Laboratories (deuterium gas-driven permeation) and Idaho National Laboratories (tritium gas- and plasma-driven tritium permeation). These experiments are in part a collaboration between the US and Japan to study the performance of tungsten at divertor relevant temperatures (PHENIX). Here we report on the development of a high temperature (<= 1150 degrees C) gas-driven permeation cell and initial measurements of deuterium permeation in several types of tungsten: high purity tungsten foil, ITER-grade tungsten (grains oriented through the membrane), and dispersoid-strengthened ultra fine grain (UFG) tungsten being developed in the US. Experiments were performed at 500-1000 degrees C and 0.1-1.0 atm D-2 pressure. Permeation through ITER-grade tungsten was similar to earlier W experiments by Frauenfelder (1968-69) and Zaharakov (1973). Data from the UFG alloy indicates marginally higher permeability (< 10x) at lower temperatures, but the permeability converges to that of the ITER tungsten at 1000 degrees C. The permeation cell uses only ceramic and graphite materials in the hot zone to reduce the possibility for oxidation of the sample membrane. Sealing pressure is applied externally, thereby allowing for elevation of the temperature for brittle membranes above the ductile-to-brittle transition temperature. (C) 2016 Elsevier B.V. All rights reserved.

Tritium imaging plate technique (TIPT) has been applied to examine tritium (T) retention in individual particles made of titanium (Ti) with 30 and 100 mu m in diameter and tungsten (W) with 50 mu m in diameter. Distribution of T radioactivity observed by TIPT corresponded well to spatial distribution of the particles. In a limited case of uniform and high T concentration in the bulk of the individual particle, the amount of T is directly quantified from T radioactivity by a master curve method. Density and size of the particle and T concentration profiles in the bulk of the particle are important factors to change emission behavior of T beta-ray and thus accurate quantification of the amount of T in the individual particle.

The change of chemical states of niobium with oxide growth was examined in the oxide layers of Zr-2.5Nb around the first kinetic transition by the conversion electron yield - X-ray absorption near-edge structure measurements. The detailed depth profiles of niobium chemical states were obtained in both the pre- and the post-transition oxide layers of Zr-2.5Nb formed in water at 663 K for 40-280 d. The depth profiling revealed that the inner oxide layer remained protective to oxidizing species even though in the post-transition region and this excellent stability of barrierness would be attributed the suppression of hydrogen pickup.

Three tungsten samples irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to deuterium plasma (ion thence of 1 x 10(26) m(-2)) at three different temperatures (100, 200, and 500 degrees C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy was performed with a ramp rate of 10 degrees C min(-1) up to 900 degrees C, and the samples were annealed at 900 degrees 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 degrees 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 degrees C for 0.5 h. (C) 2014 Elsevier B.V. All rights reserved.

Retention behaviors of hydrogen loaded by gas absorption and plasma implantation to pure copper, pure tungsten and the F82H steels at various temperatures have been examined by the tritium imaging plate technique. Three components are distinguished in hydrogen retained near the surface region; one is an endothermic trapping component in the bulk or near the surface region, second is an exothermic trapping component induced by plasma implantation and third is a trapping component in oxide layers. The relative amount of each component in depth near the surface region of the metals can alter retention behaviors of hydrogen with respect to the loading temperatures. (C) 2014 Elsevier B.V. All rights reserved.

A tritium imaging plate technique has been applied to visualize hydrogen distribution and examine hydrogen solubility and diffusivity in a proton-conducting oxide, Y-doped BaCeO3 (BaCe0.9Y0.1O3 - alpha. Tritium charging of the BaCe0.9Y0.1O3 (-) (alpha) specimens was carried out by a gas absorption method using partially tritiated water vapor (HTO, 3 kPa, T/H similar to 10(-6)) at temperatures ranging from 673 K to 873 K for a given time. After charging, tritium distributions of the surface and cross section of the halved specimens were visualized using an imaging plate technique. From the tritium concentration and distributions of the surface and cross section, hydrogen solubility and hydrogen (tritium) diffusivity of the BaCe0.9Y0.1O3 - alpha specimens were determined. (C) 2015 Elsevier B.V. All rights reserved.

In order to understand behaviors of hydrogen uptake and permeation in pure alpha-iron (alpha Fe) during water corrosion around room temperature, hydrogen permeation experiments for a alpha Fe membrane have been conducted by means of tritium tracer techniques.Hydrogen produced by water corrosion of alpha Fe is trapped and/or blocked in/by product oxide layers to delay hydrogen uptake in alpha Fe for a moment. However, the oxide layers do not work as a sufficient barrier for hydrogen uptake. Some of hydrogen dissolved in alpha Fe could normally diffuse and permeate through the alpha Fe bulk.Assuming hydrogen dissolution at the water/Fe interface proportional to the square root of the hydrogen pressure (Sieverts' law), the partial hydrogen pressure were estimated to be 0.7, 5.0 and 9.5 kPa at 303, 323 and 348 K, respectively.

Hydrogen diffusion coefficients in a reduced activation ferritic/martensitic steel (F82H) and an oxide dispersion strengthened F82H (ODS-F82H) have been determined from depth profiles of plasma-loaded hydrogen with a tritium imaging plate technique (TIPT) in the temperature range from 298 K to 523 K. Data of hydrogen diffusion coefficients, D, in F82H are summarized as D [m2 s-1] =1.1×10-7exp(-16[kJ mol-1]/RT). The present data indicate almost no trapping effect on hydrogen diffusion due to an excess entry of energetic hydrogen by the plasma loading, which results in saturation of the trapping sites at the surface and even in the bulk. In the case of ODS-F82H, data of hydrogen diffusion coefficients are summarized as D [m2 s-1] =2.2×10-7exp(-30[kJ mol-1]/RT) indicating a remarkable trapping effect on hydrogen diffusion caused by tiny oxide particles in the bulk of F82H.

Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory at reactor coolant temperatures of 50-70 degrees C to low displacement damage of 0.025 and 0.3 dpa. After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 degrees C twice at the ion fluence of 5 x 10(25) m(-2) to reach the total ion fluence of 1 x 10(26) m(-2) in order to investigate the near-surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate the irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near-surface (<5 mu m depth) deuterium concentration increased from 0.5 at% D/W in 0.025 dpa samples to 0.8 at% D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near-surface retention via nuclear reaction analysis indicated the deuterium was trapped in bulk (at least 50 mu m depth for 0.025 dpa and 35 mu m depth for 0.3 dpa) at 500 degrees C cases even in the relatively low ion fluence of 10(26)m(-2).

In order to understand how hydrogen loaded by plasma in F82H is removed by annealing at elevated temperatures in vacuum, depth profiles of plasma-loaded hydrogen were examined by means of a tritium imaging plate technique.Owing to large hydrogen diffusion coefficients in F82H, the plasma-loaded hydrogen easily penetrates into a deeper region becoming solute hydrogen and desorbs by thermal annealing in vacuum. However the plasma-loading creates new hydrogen trapping sites having larger trapping energy than that for the intrinsic sites beyond the projected range of the loaded hydrogen. Some surface oxides also trap an appreciable amount of hydrogen which is more difficult to remove by the thermal annealing. (C) 2013 Elsevier B.V. All rights reserved.

Copper (Cu) and its alloys are candidate materials for heat sinks or cooling-tubes in a fusion reactor. Hence their tritium retention and permeation are very important safety concerns. Most data for diffusion and permeation of hydrogen in Cu so far available have been limited for rather higher temperatures and data for lower temperatures, in particular, for near room temperature (RT) are scarce. We have applied a tritium tracer technique for gaseous hydrogen permeation in pure Cu at near RT and succeeded to get reliable data for hydrogen permeation coefficients given by Phi = (2.8 +/- 0.4) x 10(-6) exp(-85 +/- 2(kJ/mol)/RT), mol m(-1) s(-1) Pa-1/2, which is reliable in very wide temperature range from 300 K to 1000 K.However, diffusion coefficients determined by the time-lag method are bending downward from the extrapolation of higher temperature data and are influenced by initial surface contamination which is removed by hydrogen loading. (C) 2013 Elsevier B.V. All rights reserved.

Hydrogen permeation and diffusion coefficients in alloys of iron (Fe) and nickel (Ni) with the Ni content of 5, 9, and 20 at.% and a crystal structure of alpha/alpha' phase have been examined around room temperature (RT) using a tritium-tracer hydrogen-permeation experiment. Hydrogen permeation coefficients around RT agree well with values extrapolated from literature data obtained at higher temperatures for the respective alloys. On the other hand, apparent hydrogen diffusion coefficients determined using the time-lag method are several orders of magnitude smaller than extrapolated from the literature data. This could be caused by surface blocking and/or barrier effects due to surface oxide and/or other impurities. Initially, hydrogen permeation is suppressed by the existence of the surface oxide. It appears that hydrogen, mostly at the upstream side or even at the downstream side, can reduce and remove the surface oxides so that normal hydrogen steady-state permeation can occur without surface blocking or barrier effects. Thus, true hydrogen diffusion coefficients for respective Fe-Ni alloys during steady-state permeation must be much larger than those estimated from the time-lag method. (C) 2013 Elsevier B. V. All rights reserved.

Depth profiles of tritium (T) loaded by gas and plasma in tungsten (W) coatings on ferritic steels have been examined by using a tritium imaging plate technique and their changes during storage and after annealing have been monitored. The depth profiles of T consisted of 4 components, (I) T trapped at impurities and defects newly introduced in the near surface region of the coating by plasma loading, (II) T trapped at the inner surfaces of the grains and dissolved in the grains resulting in a flat depth profile throughout the whole coating, (III) T dissolved and diffused into the substrate giving a decaying profile, and (IV) T trapped at the backside surface of the substrate. The results support that retention of T is mainly caused by pore diffusion of gaseous T followed by dissolution and trapping in/at each W grain, and dissolution of T into the F82H substrate to allow permeation. Release of T proceeds in an opposite way of retention but each component desorbs independently. (C) 2013 Elsevier B. V. All rights reserved.

The effects of neutron and ion irradiations on deuterium (D) retention in tungsten (W) were investigated. Specimens of pure W were irradiated with neutrons to 0.3 dpa at around 323 K and then exposed to high-flux D plasma at 473 and 773 K. The concentration of D significantly increased by neutron irradiation and reached 0.8 at% at 473 K and 0.4 at% at 773 K. Annealing tests for the specimens irradiated with 20 MeV W ions showed that the defects which play a dominant role in the trapping at high temperature were stable at least up to 973 K, while the density decreased at temperatures equal to or above 1123 K. These observations of the thermal stability of traps and the activation energy for D detrapping examined in a previous study (approximate to 1.8 eV) indicated that the defects which contribute predominantly to trapping at 773 K were small voids. The higher concentration of trapped D at 473 K was explained by additional contributions of weaker traps. The release of trapped D was clearly enhanced by the exposure to atomic hydrogen at 473 K, though higher temperatures are more effective for using this effect for tritium removal in fusion reactors.

Tungsten coating with a thickness of 0.6 mm on reduced-activation ferritic/martensitic steel (RAF/M) F82H (Fe-8Cr-2W) have been produced by Vacuum Plasma Spraying (VPS). Heat flux experiments using an electron beam and quantitative analyses about temperature profiles and thermal stress using FEM have been carried out on the VPS-W coated F82H. In addition, behavior of hydrogen penetration/permeation on the VPS-W coated F82H has been investigated by the tritium (T) tracer technique.

In order to understand the role of a plasma-sprayed tungsten (W) coating on tritium (T) permeation in a W-coated ferritic/martensitic steel (F82H), we have examined depth profiles of T in the coating and the substrate using the tritium imaging plate technique after T loading by a dc glow-discharged plasma at 453 and 573K for 2 h. Tritium loaded by plasma exposure was distributed uniformly in the whole coating, while T penetrated to the substrate by diffusion. The former is caused by T diffusion through open pores and/or along grain boundaries followed by adsorption on grain surfaces and dissolution in the grains. The main role of the W coating on T permeation is to reduce the incoming flux at the coating/substrate interface owing to pore diffusion in the coating and the effective area for T dissolution in the substrate.

Tritium retention in plasma-facing components influences the design, operation, and lifetime of fusion devices such as ITER. Most of the retention studies were carried out with the use of either hydrogen or deuterium. Tritium Plasma Experiment is a unique linear plasma device that can handle radioactive fusion fuel of tritium, toxic material of beryllium, and neutron-irradiated material. A tritium depth profiling method up to mm range was developed using a tritium imaging plate and a diamond wire saw. A series of tritium experiments (T-2/D-2 ratio: 0.2 and 0.5 %) was performed to investigate tritium depth profiling in bulk tungsten, and the results shows that tritium is migrated into bulk tungsten up to mm range.

In order to understand behavior of tritium (T) on surface and in bulk of metals exposed to T plasma, both surface activities and depth profiles of T were periodically observed by a tritium imaging plate technique during storage in air at room temperature (RT) for over 1 year. In the T depth profiles, T localized within a depth of sub mm from the surface was clearly distinguished from T in the bulk. The former was attributed to strong trapping by some defects produced by the plasma exposure and remained quite longer during the storage, while the latter was released from the surfaces by diffusion. T surface activity measured on the plasma-exposed surface changed in a complicated way with time due to removal of T by isotopic replacement with H in ubiquitous H(2)O and T supply from the bulk in the course of the diffusional release.

Applying a tritium tracer technique, we have investigated hydrogen plasma driven permeation (PDP) through tungsten (W) near room temperature. The technique was confirmed to give reliable data on diffusion and permeation coefficients of pure W for gas driven permeation (GDP), and then it was applied to observe PDP in W near room temperature. It was found that PDP in earlier phase was controlled by diffusion giving reliable diffusion coefficients. Taking literature data at higher temperatures and present ones near room temperature determined from PDP into account, we have proposed new diffusion coefficients D(Upper) (limit) = (3.8 +/- 0.4)x10(-7) exp ((-39.8 +/- 1.5) (kEmol)/RT), m(2)s(-1). (1) The activation energy for permeation determined by PDP was similar to that by GDP. The extrapolation of the present data to higher temperature agreed well with Frauenfelder's data, suggesting the activation energy of around 65 kJ/mol for permeation is quite reasonable. However prolonged measurements resulted in significant reduction of PDP. The cause of the reduction was attributed to the increase of reemission owing to surface cleaning and/or roughening by incidence of energetic hydrogen.

In order to obtain heat of transport, Q*, of tritium in V-4Cr-4Ti alloy (NIFS-HEAT-2), thermomigration experiments have been carried out at a temperature ranging from 333 to 471 K. Tritium homogeneously distributed in the specimen bar was forced to migrate by an applied temperature gradient. The resulting tritium profile was visualized by an imaging plate technique and Q* was determined from the profile according to a thermomigration theory. The obtained value of Q* was about +20 kJ/mol for the initial hydrogen concentration of 0.008 at.%, and no appreciable temperature dependence was observed. In order to examine the effect of thermomigration on tritium retentions in pure V and the alloy used as cooling tubes or first walls of the blanket, a simple model calculation was made under designed temperature gradients in fusion reactors. The result showed that tritium retention could be enhanced about 10-20% compared with the case without thermomigration. (C) 2010 Elsevier B.V. All rights reserved.

In a fusion reactor, tritium retention and permeation in structure materials are very important safety concerns. Most data for diffusion and permeation of hydrogen in metals so far available have been limited for rather higher temperatures and, in particular, no data are available for high-Z metals near room temperature (RT). We have tried to observe gaseous hydrogen permeation through metals near RT applying a tritium tracer technique, which is a very powerful tool to detect quite small amount of hydrogen (tritium) by a liquid scintillation counting (LSC) method. After confirming the reliability of the method for the determination of diffusion and permeation coefficients in pure Ni, it was applied to hydrogen permeation in W near RT, and diffusion and permeation coefficients of hydrogen in W were determined,D = (3.42 +/- 0.68) x 10(-9) exp((-37.8 +/- 1.2)(kJ/mol)/RT), m(2) s(-1),andphi = (1.21 +/- 0.24) x 10(-5) exp((-57.8 +/- 0.9)(kJ/mol)/RT), mol m(-3) s(-1) Pa-1/2. (C) 2010 Elsevier B.V. All rights reserved.

Tritium tracer techniques are applied to observe behavior of hydrogen (tritium (T)) in near surface regions of Mo and W loaded by gaseous absorption (GAS) and a glow discharge (GDC). GDC produces blisters on both Mo and W surfaces and Tritium Auto-RadioGraph (TARG) showed the thickness of blister skins is larger than the escaping depth of T beta-electrons, around 1 mu m. For GAS specimens, T evolution is likely controlled by diffusion giving diffusion coefficients of,D-Mo = 1.5 x 10(-7) exp (-41 kJ/mol/RT) m(2) s(-1)D-w = 4.3 x 10(-9) exp (-38 kJ/mol/RT)m(2) s(-1)at 273-323 K. GDC specimens show much smaller diffusion coefficients with higher activation energies and T release continues very long, suggesting T release from blisters. (C) 2011 Elsevier B.V. All rights reserved.

Hydrogen including a trace amount of tritium was loaded on the edge surface of an F82H rod. After the loading, the rod was held at 298 or 323 K to allow hydrogen diffuse in and release out. Tritium tracer techniques have been applied to determine hydrogen depth profiles and hydrogen release rates by using an tritium imaging plate technique and a liquid scintillation counting technique, respectively. The depth profiles were composed of a surface localized component within 200 mu m of the surface and a diffused component extending over 1 mm in depth. The apparent hydrogen diffusion coefficients obtained from the depth profile of the diffused component are near the extrapolated value of the literature data determined at higher temperatures. The surface localized component, which is attributed to trapping at surface oxides and/or defects, was released very slowly to give apparent diffusion coefficients much smaller than those determined from the diffused component. (C) 2010 Elsevier B.V. All rights reserved.

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials. Published by Elsevier B.V.

Tungsten is a primary candidate of plasma facing material in ITER and beyond, owing to its good thermal property and low erosion. But hydrogen solubility and diffusivity near ITER operation temperatures (below 500 K) have scarcely studied. Mainly because its low hydrogen solubility and diffusivity at lower temperatures make the detection of hydrogen quite difficult. We have tried to observe hydrogen plasma driven permeation (PDP) through nickel and tungsten near room temperatures applying a tritium tracer technique, which is extremely sensible to detect tritium diluted in hydrogen. The apparent diffusion coefficients for POP were determined by permeation lag times at first time, and those for nickel and tungsten were similar or a few times larger than those for gas driven permeation (GDP). The permeation rates for POP in nickel and tungsten were larger than those for GDP normalized to the same gas pressure about 20 and 5 times larger, respectively. (C) 2010 Elsevier B.V. All rights reserved.

We have applied a tritium imaging plate technique to measure the tritium distribution profile on surface and in bulk of various metal materials after exposure to a deuterium-tritium plasma in a linear plasma experimental apparatus. The experimental tritium concentration profiles in mm range are interpreted according to a simple hydrogen diffusion model in each metal. We found that a significant amount of tritium is localized in near surface regions and is clearly distinguishable from tritium diffused in the bulk. The amount of surface tritium is not likely correlated to bulk properties (diffusivity and solubility), but is related to trapping in surface defects or metal impurities such as oxide and carbide. (C) 2010 Elsevier B.V. All rights reserved.

In order to understand the role of a plasma-sprayed tungsten (W) coating on tritium (T) permeation in a W-coated ferritic/martensitic steel (F82H), we have examined depth profiles of T in the coating and the substrate using the tritium imaging plate technique after T loading by a dc glow-discharged plasma at 453 and 573 K for 2 h. Tritium loaded by plasma exposure was distributed uniformly in the whole coating, while T penetrated to the substrate by diffusion. The former is caused by T diffusion through open pores and/or along grain boundaries followed by adsorption on grain surfaces and dissolution in the grains. The main role of the W coating on T permeation is to reduce the incoming flux at the coating/substrate interface owing to pore diffusion in the coating and the effective area for T dissolution in the substrate. © 2011 The Royal Swedish Academy of Sciences.

Gamma cells using p-type Si substrates with various resistivities were fabricated with a vacuum evaporation method. The energy conversion efficiency from gamma ray to electric power successfully reached about 2% for the gamma cell with a resistivity of 50∼100 Ω·cm.

Tritium release behavior and surface tritium behavior were separately examined for typical fcc and bcc metals by using tritium tracer techniques. Pure copper (Cu), iron (Fe), nickel (Ni) and molybdenum (Mo) were loaded with hydrogen including a trace amount of tritium and then immersed into water at around room temperatures. Then, the tritium release rate into the water was examined by a liquid scintillation counting technique and the surface tritium concentration by a tritium imaging plate technique.The tritium release from the metals is attributed to the release of dissolved tritium by diffusion from the normal interstitial sites, and the first order desorption of trapped one with detrapping energies of 64, 72 and 25 kJ mol(-1) for Cu, Fe and Mo. respectively. Overall release behavior is varied depending on the ratio of dissolved and trapped amounts of tritium. (C) 2010 Elsevier B.V. All rights reserved.

The tritium imaging plate (TIP) technique has been applied to visualize penetration profiles of hydrogen (tritium) loaded in pure tungsten (W) by a dc glow discharge at a temperature ranging from 473 to 673 K. The penetration profile consists of two components, i.e. a highly localized one in the near-surface region (sub-mm in depth), and another, deep penetrating one (several mm in depth). An apparent hydrogen diffusion coefficient is determined from the latter to be D(2) = (3 +/- 2 x 10(-7)) x exp(-0.39 +/- 0.03 inverted right perpendiculareVinverted left perpendicular/kT), which agrees well with the extrapolation of Frauenfelder's data obtained at elevated temperatures. The near-surface localized one is attributed to hydrogen trapping with a trapping energy of 0.84 +/- 0.04 eV.

Under appropriate conditions, exposing tungsten to a high flux D plasma creates near-surface blisters and other changes in surface morphology. We have characterized the sizes of blisters formed at different temperatures (147 degrees C <= T(surface) <= 704 degrees C) and performed a surface analysis to elucidate factors that influence blister formation. Tungsten targets that were exposed to low energy (70 eV) D ions at a flux of 1.1 x 10(22) m(-2) s(-1) in the tritium plasma experiment (TPE) were considered. We used AES to analyze the surface for evidence of implanted impurities. Blister diameters and heights were quantified using SEM imagery and vertical scanning interferometry. Given the likelihood of D precipitation in blisters, we expect that the data obtained here could be incorporated into a computational model to better simulate the diffusion and desorption of D in W. With this in mind, we present an analysis of thermal desorption profiles showing the release of D from the surface.

in order to study blistering mechanisms of Molybdenum (Mo), hydrogen distributions at and around blisters formed on Mo surfaces are examined by Tritium (T) radio-luminography or autoradiography (TARG). TARG shows that large amount of hydrogen (T) is accumulated at and near grain boundaries and some blisters are covered with Ag precipitates representing T under the blister skins. Two independent types of blistering mechanisms seem to occur on Mo surface simultaneously. One is typical blistering due to bubble coalescence accompanying plastic deformation of the blister skins and only very thin blister skins allow T detection by TARG. Another is exfoliation or cracking of a grain caused by mechanical fracturing of the grain boundaries and/or defect clusters due to brittle nature of Mo, remaining tritium on the fractured surface. (C) 2009 Elsevier B.V. All rights reserved.

We have revisited hydrogen behavior in Fe-Ni alloys with the tritium evolution technique. Applying a 3D analytical solution of Fick's diffusion equation to the tritium evolution curve from disk shaped samples, hydrogen retention and apparent diffusion coefficients in Fe-Ni alloys with Ni content from 6 to 50 at.% are determined around RT. In gamma phase region, diffusion coefficients were not appreciably changed and increased with lattice parameter. In the alpha' phase region, two diffusion components were distinguished: For fast diffusion components, diffusion coefficients have quite good agreement with the literature value and decreased with increase of lattice parameter. The slower diffusion components are likely attributable to retained gamma phase precipitates or Ni(3)Fe dispersed in the alpha' matrix and must be very important to understand hydrogen embrittlement and tritium safety handling. (C) 2008 Elsevier B.V. All rights reserved.

Hydrogen distribution around blisters on aluminum (AI) and molybdenum (Mo) was examined by tritium radio-luminography, i.e. tritium autoradiography (TARG) and an imaging plate technique. Tritium accumulated in the blisters on Al surface was successfully visualized at the first time. The tritium density in the blisters was found to increase with their radius to the power of 2.3. This supports the blister mechanism of bubble coalescence but the blister shape was flattened along the surface with increasing their size. For Mo. tritium distribution was not well correlated with blisters, and the bubbles coalescence was not clearly observed, too. But the erosion or exfoliation of thick layers with wider area than blisters were observed and hydrogen was released by the exfoliation of the thick surface layers, remaining not tritium on the exfoliated surface. Such exfoliation is very likely caused by mechanical stress given by accumulated hydrogen at trapping site such as grain boundaries, intrinsic defect, or self trapping. (C) 2009 Elsevier B.V. All rights reserved.

The tritium imaging plate (TIP) technique has been applied to visualize penetration profiles of hydrogen (tritium) loaded in pure tungsten (W) by a dc glow discharge at a temperature ranging from 473 to 673 K. The penetration profile consists of two components, i.e. a highly localized one in the near-surface region (sub-mm in depth), and another, deep penetrating one (several mm in depth). An apparent hydrogen diffusion coefficient is determined from the latter to be D2 = (3±2 × 10-7) × exp(-0.39±0.03 [eV]/kT), which agrees well with the extrapolation of Frauenfelder's data obtained at elevated temperatures. The near-surface localized one is attributed to hydrogen trapping with a trapping energy of 0.84±0.04 eV. © 2009 The Royal Swedish Academy of Sciences.

Under appropriate conditions, exposing tungsten to a high flux D plasma creates near-surface blisters and other changes in surface morphology. We have characterized the sizes of blisters formed at different temperatures (147°C≤Tsurface≤704°C) and performed a surface analysis to elucidate factors that influence blister formation. Tungsten targets that were exposed to low energy (70 eV) D ions at a flux of 1.1×1022 m-2 s-1 in the tritium plasma experiment (TPE) were considered. We used AES to analyze the surface for evidence of implanted impurities. Blister diameters and heights were quantified using SEM imagery and vertical scanning interferometry. Given the likelihood of D precipitation in blisters, we expect that the data obtained here could be incorporated into a computational model to better simulate the diffusion and desorption of D in W. With this in mind, we present an analysis of thermal desorption profiles showing the release of D from the surface. © 2009 The Royal Swedish Academy of Sciences.

Hydrogen release behaviors from the 8Cr2W stainless steel (RAF/M) around RT are examined by using tritium tracer techniques, and trapping effects of bulk and surface are discussed In the overall release, three different release stages are clearly distinguished giving three different diffusion coefficients and release amounts which indicate the existence of different kinds of trapping. In addition, the appreciable amount of hydrogen (tritium) is trapped on the surface and/or surface oxides of RAF/M, but they are hardly released and show no influence on the overall hydrogen release behavior.At very low hydrogen concentration, almost all hydrogen atoms are trapped at the deepest trapping site, probably M23C6, and the sites are easily saturated. With increasing the hydrogen concentration, the shallower trapping sites are occupied Remaining hydrogen atoms seem to be in normal (interstitial) sites, whose amount increases with the square root of the hydrogen loading pressure, but they are still influenced by trapping with lattice imperfections and/or grain boundaries.

Tritium (hydrogen) accumulation and release processes at MnS precipitates and surrounding alpha Fe matrix area at room temperature (RT) were studied by means of tritium autoradiography (TARG) using a pseudo-binary alloy of Fe-MnS. Hydrogen accumulation at the MnS precipitates at RT was clearly observed but only at a limited occasion. The process involves diffusion, solution and trapping in a complex way including a temperature effect. TARG is proved to be a very good technique to obtain hydrogen area profiles in a near surface region, whereas it is only a snap shot at a particular time and temperature. It could lead us to a totally different interpretation of the accumulation process without detailed dependencies of hydrogen diffusivity and solubility in inclusion species and alpha Fe on temperatures. (C) 2007 Elsevier B.V. All rights reserved.

Tritium diffusion behavior in a V-4Cr-4Ti (NIFS-Heat-2) alloy has been examined with a tritium tracer technique. Firstly, a small amount of tritium (T) was implanted into the specimen surface, and then the specimen was diffusionannealed at temperatures ranging from 373 K to 573 K. The diffusion depth profile of T in the specimen was measured with a tritium imaging plate (IP) technique to determine the diffusion coefficient. The obtained diffusion coefficient of tritium in V-4Cr-4Ti is expressed as D-t (cm(2)/s) = (7.5 +/- 0.2) x 10(-4) exp(-0.13(eV)/kT), which is lower than that in pure vanadium, and is comparable with literature values of protium in a V-4Ti alloy taking the isotope mass effect into consideration. (c) 2007 Elsevier B.V. All rights reserved.

Tritium diffusion in a vanadium alloy (V-4Cr-4Ti) has been investigated at temperatures ranging from 230 K to 573 K. Tritium was loaded into the surface layers of the alloy specimen with an ac-glow discharge. Before and after diffusion annealing of the specimen, tritium diffusion profiles were measured by means of an imaging plate (IP) technique. Tritium diffusion coefficients (D-T), which were evaluated by fitting a numerical solution of the diffusion geometry employed here to the obtained diffusion profiles, were a little smaller than those for pure V with the activation energy of 0.13 +/- 0.01 eV. Below 320 K, in addition, the Arrhenius plot of DT bent downwards showing a larger activation energy of 0.19 +/- 0.01 eV, probably owing to the trapping effect of both of Cr and Ti. The effect of alloying elements on tritium diffusion and the influence of tritium release from the surface were discussed. (c) 2007 Elsevier B.V. All rights reserved.

Two characteristic hydrogen migration processes in the tempered martensite steel are distinguished by means of tritium autoradiography: (i) long range transport in the ferrite matrix, and (ii) local hydrogen accumulation in the cementite precipitates and/or at boundaries between the cementite precipitates and the surrounding ferrite matrix. (c) 2005 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Hydrogen distributions around non-metallic inclusions in steels are successfully characterized with high-resolution tritium autoradiography. The autoradiographs show that hydrogen accumulation characteristics around the inclusions depend on types of the inclusions. In the case of MnS, hydrogen was inhomogeneously distributed in the ferrite matrix surrounding the MnS inclusion, probably because hydrogen is trapped in defects formed around MnS. The inhomogeneous distribution of hydrogen may be originated from the asymmetric stress field produced by a contraction of the MnS phase in the heat treatment, i.e. the inhomogeneous volumetric change of MnS owing to its larger thermal expansion than that of the ferrite phase. In the case of Al2O3, hydrogen was intensely localized at boundary layers of the ferrite matrix surrounding the Al2O3 inclusion. This could be attributed to hydrogen trapping at defects introduced by a residual stress in the boundary layers of the ferrite matrix due to larger contraction of the ferrite phase than that of the Al2O3 phase on cooling. Similarly hydrogen was accumulated in the surrounding ferrite matrix but more widely distributed around Cr carbide probably because difference in the thermal expansion between the Cr carbide and ferrite phases is. less than that between the Al2O3 and ferrite phases.

The impact properties of an oxidized and a hydrogenated Zircaloy have been Studied with an instrumented Charpy machine by using a strip Charpy V-notch specimen (1 mm thick by 4 mm wide). Fracture processes Such as crack initiation and propagation were examined using load-displacement Curves obtained in this Study. In the case of the hydrogenated specimen containing preferentially oriented hydrides, an appreciable decrease in the absorbed energy was observed in the crack propagation rather than in the crack initiation. From results of fractographs of the specimen, it was suggested that the reduction of the crack propagation energy of hydrogenated specimen could be attributed to the change of the stress state in the Zircaloy matrix, which was caused by the fracture of hydride in the inner part of specimen. In the case of the specimen oxidized at 973 K for 60 min, on which an oxide layer (4 pin in thickness) and oxygen incursion layer (4 mum) were formed, the surface layers affected the crack initiation process. The growing oxygen incursion layer, in particular, resulted in the constraint of plastic deformation of the Zircaloy matrix not only in the crack initiation but also in the crack propagation as its thickness increased.

Tungsten coating with a thickness of 0.6 mm on reduced-activation ferritic/martensitic steel (RAF/M) F82H (Fe-8Cr-2W) have been produced by Vacuum Plasma Spraying (VPS). Heat flux experiments using an electron beam and quantitative analyses about temperature profiles and thermal stress using FEM have been carried out on the VPS-W coated F82H. In addition, behavior of hydrogen penetration/permeation on the VPS-W coated F82H has been investigated by the tritium (T) tracer technique.

プラズマ対向材料中のトリチウム挙動に及ぼす中性子照射の影響を明らかにするため,候補材であるタングステンをオークリッジ国立研究所の研究炉High Flux Isotope Reactor(HFIR)で中性子照射した上で,アイダホ国立研究所の線型プラズマ装置Tritium Plasma Experiment(TPE)にて同位体である重水素の高フラックスプラズマにばく露し,捕獲重水素濃度と昇温脱離挙動を調べた.照射欠陥の捕獲効果により水素同位体滞留量が著しく増大すると共に,加熱処理による除去が困難となるため,同位体交換法等の新たなトリチウム除去技術の開発が必要であることが示された.

DTイオンやHeイオン,および壁材料イオンが直接照射される第一壁では,トリチウムの蓄積や拡散,あるいは壁材料の損耗や堆積などのトリチウム・物質移行現象が起こり,ブランケット寿命や炉内トリチウム挙動に大きな影響を及ぼす.ここでは,これらのトリチウム・物質移行現象解明のため,高密度プラズマ照射装置(PISCES-B(UCSD),TPE(INL))を用い,タングステン等の壁材料にD,He,T,Beを含むプラズマやパルスレーザーを照射して,表面状態変化,水素同位体吸蔵・拡散特性,および損耗特性を調べた結果を報告する.

プラズマ対向材料中のトリチウム挙動に及ぼす中性子照射の影響を明らかにするため,候補材であるタングステンをオークリッジ国立研究所の研究炉High Flux Isotope Reactor(HFIR)で中性子照射した上で,アイダホ国立研究所の線型プラズマ装置Tritium Plasma Experiment(TPE)にて同位体である重水素の高フラックスプラズマにばく露し,捕獲重水素濃度と昇温脱離挙動を調べた.照射欠陥の捕獲効果により水素同位体滞留量が著しく増大すると共に,加熱処理による除去が困難となるため,同位体交換法等の新たなトリチウム除去技術の開発が必要であることが示された.

室温～473Kの低温度領域において、低放射化フェライト・マルテンサイト鋼F82Hにトリチウムを含んだ水素を注入し、トリチウムイメージングプレート法によりその注入された水素の深さ分布を測定した。得られた深さ分布にフィックの拡散方程式の解析解をフィッティングすることにより水素拡散係数を決定した。

酸化膜内の添加元素の化学状態や結晶構造、応力分布等に注目して、ジルコニウム合金酸化膜の成長にともなう酸化膜内の特性変化を調べ、酸化膜成長による水素吸収特性の変化について考察した。

予めヘリウムを照射したタングステンの表面にプラズマからトリチウムを注入し、その注入されたトリチウムの表面分布および内部深さ分布をイメージングプレート法により調べた。これらの結果をもとに、ヘリウム注入領域および内部深さ方向へのトリチウムの進入／滞留機構を議論する。

プラズマから多孔質W被覆層に573 K近傍の一定温度でトリチウム（T）を注入し、その直後にイメージングプレート（IP）法により被覆層中のT深さ分布を調べたところ、被覆層中にほぼ均一にTが分布していたことから、空隙（粒界）拡散によって進入したTがW粒子または空隙の表面に捕獲されている可能性を前学会にて報告した。今回は、この被覆層中のTがどのように放出されるのかを明らかにするために、被覆層を233 K、大気中で保持した後、再び同様に被覆層中のT深さ分布を調べた。3ヶ月間経過後、被覆層中のTはほぼ均一に分布した状態で、その濃度が1/3にまで低下していた。この結果は、W被覆層からの放出が捕獲されたTの脱捕獲によって律速されていることを示唆している。

トリチウム（T）プラズマに曝されたタングステン（W）被覆フェライト鋼へのTの進入過程における多孔質W被覆層の役割を明らかにするため、DCグロー放電プラズマによりTをW被覆層表面に注入した後、注入面に垂直な断面を切り出し、イメージングプレート法により被覆層およびフェライト鋼基板中のT深さ分布を調べた。注入面近傍の高いT濃度を除き、被覆層中にはTが均一に分布していた。被覆層中に比べ、界面ではT濃度が半分程度まで低下していた。フェライト鋼基板中ではTが拡散により進入している様子が観察された。これらの結果から、多孔質W被覆層はTの空隙（粒界）拡散によりプラズマからのT入射フラックスを低減し、フェライト鋼基板へのTの溶解の有効表面積を低下させる役割を果たしていると考えられる。

低放射化フェライト/マルテンサイト鋼（F82H）表面にプラズマ溶射法によりタングステン（W）被覆層を形成させた試料を用い、DCグロー放電注入法によりW被覆層表面にトリチウムを入射させた。この後、トリチウム入射面に垂直方向の断面を切り出し、イメージングプレート法によりトリチウムの入射方向からの進入深さ分布を調べた。これらの結果をもとに、W被覆層内部、界面およびF82H内部それぞれへのトリチウムの進入/滞留のメカニズムを議論する。

本研究グループは、高温トリチウムガスに曝された金属材料について、表面に捕獲されたトリチウムと内部に溶解したトリチウムとを分離しそれぞれの挙動を研究してきた。今回は、トリチウムプラズマにさらされた金属材料について、照射後の表面トリチウム濃度および内部で濃度プロファイルの時間推移を測定し、材料表面からのトリチウム脱離速度および内部のトリチウム拡散係数の温度依存性をも導出した。

室温付近の金属の水素（トリチウム）透過データを拡充することは、プラズマの当たらないリモートエリアの壁材料や冷却水配管材料のトリチウム透過挙動を理解する上で重要である。本研究では、DCグロー放電により室温近傍において金属膜にトリチウムを負荷し、膜を透過してきたトリチウムを液体シンチレーションカクテルにより捕集し、カクテル中のトリチウム量を定量することにより、金属膜の水素透過係数、拡散係数および溶解度を求めた。

FCC金属およびBCC金属表面にトリチウムを含んだ水素によるブリスターを形成させ、トリチウムオートラジオグラフィによりブリスター近傍の水素集積状態を可視化・定量化した。 ブリスター形成時の温度、ブリスタ形成後の時間等を様々に変化させ、ブリスター近傍の水素分布がどのように変化するかを調べ、ブリスタ形成機構を議論する。

シリコンおよびCdTe製半導体放射線電池に、線量率を200～5000Gy/hの間で変化させながら室温にてガンマ線を照射した．得られた起電力は線量率のほぼ自乗に比例して増加した．また、高線量において照射損傷によるものと思われる起電力低下が見られた．

トリチウムトレーサー技術を用いて、金属表面と内部のトリチウム挙動を同時に、かつ個別に調べ、両者の関連性を調べた。実験は、ガス吸収法によりトリチウムを負荷したFCC金属およびBCC金属について、金属からのトリチウム放出速度、金属の表面および切断面のトリチウム分布およびその時間変化を調べるもので、得られた結果を、水素の拡散・捕獲モデルに基づいて解析し、両者の関連性を明らかにする。

ジルカロイ‐４板材（4x8x0.9 mm3）の表面に酸素を50ミクロン程度の深さまで溶解させ、その後400ppm水素を溶解、厚さ方向に温度勾配（高温側設定温度400℃、低温側150℃）をかけ24時間程度熱拡散させた。試料断面の水素化物の分布を観察した結果、酸素を溶解していないジルカロイ‐４板材では、水素化物が低温側に集積し、一部は表面にブリスターを形成しているのに対し、酸素を溶解した試料では酸素濃度が高い領域には水素化物の集積は見られず、温度勾配下でも酸素溶解層は水素溶解の障壁になることが分かった。

Zry-4およびZry-2の実被覆管材料を試料とし、被覆管軸方向および径方向それぞれについてトリチウムの拡散係数測定を行った．実験では、試料表面にトリチウムを室温でガス放電法により注入し、真空中で200～400℃、15分～２時間程度拡散焼鈍し、トリチウムの拡散距離が0.5mm程度となるようにした。トリチウムの分布をイメージングプレートで測定し、拡散係数を決定した．得られたトリチウムの拡散係数値は、合金の種類、拡散方向にかかわらずほぼ同じ値を示し、これらに依存しないことがわかった．また、過去に報告されているZr金属中の値とも大きな差は見られなかった．

長さ25mmの板状のバナジウム合金（NIFS-HEAT 2)試料に溶解したトリチウムの電界拡散実験を１００℃で行った。定常濃度分布に達した後、試料中のトリチウム分布をイメージングプレートを用いて測定した。濃度分布と印可した電界の強度値から、トリチウムの有効電荷Z*の値を0.2と決定した．得られたZ*値は、純バナジウム中の値1.2より小さかった。これは合金成分によるトリチウムのトラップの影響が大きいものと考えられる．

トリチウム透過実験を行い、室温付近の低温度領域における様々な金属膜の水素透過挙動を調べた。また実験後の金属膜表面および断面のトリチウムプロファイルをイメージングプレート法により詳細に調べ、水素透過・拡散挙動と表面トリチウム偏析挙動との関係を検証した。

アイダホ国立研究所（INL）のトリチウムプラズマ実験装置（TPE）において、ステンレス鋼、モリブデンおよびタングステンへのトリチウムプラズマ照射実験を行った。これら材料表面および断面のトリチウムプロファイルをイメージングプレート法により計測し、表面および内部のトリチウム蓄積挙動を明らかにした。

金属表面にトリチウムを含んだ水素を注入し、形成された水素誘起ブリスタに集積する水素（トリチウム）を、オートラジオグラフ法を用いて可視化しブリスタ形成機構について考察する。

トリチウムルミノグラフィーにより、鋼表面から注入されたトリチウムの表面分布と深さ分布を同時に測定し、トリチウム（水素）の表面偏析が内部への溶解・拡散へ及ぼす効果を調べた。 陰極電解法により導入されたトリチウムの表面分布は、極めて不均一であったが、これらは極表面の酸化膜近傍に捕獲されたものであり、鋼中のトリチウム拡散・透過に与える影響は小さいことが示唆された。

バナジウム合金（V-4Cr-4Ti)中に溶解したトリチウムの分布をイメージングプレートを用いて測定し、拡散係数Dと熱拡散における輸送熱Q*を決定した。DとQ*の値は純バナジウムのものより大きく、合金成分によるトラップ効果が顕著であった。さらにその結果をもとに合金成分によるトラップ効果を評価し、主要なトラップサイトはTiによるものでその束縛エネルギーはおよそ0.08eVであることを明らかにした。さらに実験で得た拡散係数および輸送熱の値を用いて、核融合炉ブランケット材中で想定されるトリチウムの定常濃度分布および定常濃度分布を形成するまでの緩和時間等を評価した。

核融合炉第一壁の候補材料であるフェライト系ステンレス鋼やタングステン合金は、水素溶解量が低く、また水素拡散が速いため、材料表面から導入された水素が材料中を拡散すると同時に、表面から脱離する過程（放出過程）が顕著である。しかしながら、材料表面に負荷された水素が、材料中を拡散していく過程と材料表面から放出されていく過程とを同時に測定した例は少ない。本研究では、表面からの水素放出が顕著である材料にラジオルミノグラフ法を適用し、材料中の拡散係数測定とこれに及ぼす表面放出挙動の影響を明らかにすることを目的としている。

核融合炉プラズマ対向壁の候補とされているタングステン(W)材料表面には水素イオン(H+)やヘリウムイオン(He+)照射によってブリスタが形成され、割れや剥が引き起こされる。特に水素誘起ブリスタでは、水素がその形成にどの様に関わるかについては十分な理解がなされていない。本研究では水素誘起ブリスタについて材料表面への水素侵入および材料中での拡散・集積がブリスタ形成に果たす役割を明確にすることを目的とし、鉄(Fe)、アルミニウム(Al)等の種々の金属表面に水素を注入しブリスタを形成させ、オートラジオグラフィを適用して水素の集積状況の観察を行った。金属による相違、温度効果など詳細にを調べ、ブリスタ形成機構の解明みた。

バナジウムは液体リチウムを冷却材、増殖材に用いたときのブランケット構造材料としての使用が考えられており、照射耐性、強度の観点から、同じく誘導放射能の減衰が早いCr、Tiを合金成分としたV-4Cr-4Tiが候補材として選定されている。現在、核融合科学研究所(NIFS)では酸素や窒素などの不純物を低減した高純度バナジウム合金、V-4Cr-4Ti (NIFS-HEAT-2)が開発されており、実用化に向けてそのデータベースの拡充が急務となっている。 核融合炉構造材料中のトリチウムの拡散挙動の解明は燃料粒子のプラズマ－壁間のリサイクリング、トリチウム透過、トリチウムインベントリーなどの観点から重要な問題であり、様々な金属、合金に対して研究が行われている。しかし炉心構造材料は厳しい温度勾配下での使用が予想されるため温度勾配を駆動力として物質の拡散が起きる熱拡散現象を考慮しなければならない。熱拡散現象の大きさと方向は輸送熱Q^*により与えられる。 熱拡散実験ではNIFS-HEAT-2板材(25×4×0.2mm)中にトリチウムを吸収させ均一濃度状態になった後アルミニウム棒に埋め込み上部にシースヒーター、下部に冷却水をセットし温度勾配をかけた。試料温度は上部で100℃、下部で83℃に保ち、1週間後試料を取り出し液体窒素温度でイメージングプレートに露出しトリチウム濃度分布を得た。その結果、本条件では濃度分布にほとんど変化は見られず、輸送熱Q^*の値は純バナジウムと同様に小さいことを示唆している。今後はさらに急峻な温度勾配下での実験を行いQ^*値を決定していく予定である。

イメージングプレート（IP）法およびトリチウムオートラジオグラフィ(TARG)により、Mo表面に形成された水素誘起ブリスター近傍のトリチウム分布を可視化した。

圧延および焼鈍した板状V-4Cr-4Ti（NIFS-HEAT-II）試料中に、トリチウムを含んだ水素ガスを注入しその後150℃から250℃で拡散焼鈍した。試料中のトリチウムの２次元分布をイメージングプレートを用いて測定した。得られた分布から圧延および焼鈍がトリチウムの拡散にどのような影響を及ぼすか調べた。また、拡散分布からトリチウムの拡散係数についても評価した。

これまでに、本研究グループにおいてジルコニウムあるいはジルカロイ平板試料を用いてのトリチウム分布をトリチウムイメージングプレート法にて測定し、その拡散係数を決定してきた。今回、ジルカロイ－４被覆管材から切り出された微小な試験片(肉厚0.65mm)を用いて、空間分解能に優れたイメージングプレート法（IP法）を使用し、拡散距離の短い肉厚方向(径方向)の拡散係数測定を測定方法を考えて行った。また、軸方向においては、過去の実験と同様の方法にて行い、被覆管中のトリチウムの軸方向と径方向の拡散係数を測定した。その結果、軸方向と径方向において拡散係数に明確な違いは見られず、過去に行われたジルコニウム平板との比較でもその値に大きな差は見られなかった。それより、イメージングプレートによる測定方法および測定結果の考察を行った。

γ線をSi pn接合半導体に照射し、Si内で生成された電子－正孔対を電気エネルギーとして取り出すことが可能である。しかし、Si pn接合半導体にγ線照射した場合、γ線の透過性が高いため、Si pn接合半導体γ線のエネルギー付与は小さくなり、発生する電力も小さい。過去の報告から、Si pn接合半導体として太陽電池を使用した場合、太陽電池の前に鉛板を配置することで、出力が改善されることが分かっている。そこで本研究では、γ線照射によるSiへのエネルギー付与を増加させるためn型Si薄膜の厚さの異なるSi pn接合を作製し、その前後に鉛板を設置し、γ線照射し、Siへのエネルギー付与を増加させることで出力が向上することを確認し、出力増加のために鉛板の配置位置の最適化を行ってきた。具体的には試料であるn型Si薄膜の作製には高周波マグネトロン二極スパッタ装置を用いた。基板にp型Siウェーハ、ターゲット材にn型Siウェーハを用いた試料を作製した。スパッタガスとして、水素ガスを用いた。また、基板温度は150℃、スパッタ時間は2時間～19時間で、圧力は10Paで成膜を行った。また、γ線照射実験は九州大学量子線照射分析実験施設で行った。γ線源としては60Co(0.3kGy/h）を用いた。その結果、n型Si薄膜の膜厚が大きいほど出力が向上することが分かった。また、鉛板をγ線源に対してシリコンの後に設置することで出力が向上することを確認した。さらに、MCNPコードを用いて様々な半導体材料についてシミュレーションを行い、その結果、γ線照射に対してSiよりもそれぞれ5倍、3倍程度大きなエネルギー付与を持つことが分かったInPとGaAsについて、pn接合半導体とショットキー接合半導体を作製し、γ線照射し、出力を測定し、Siとの比較を行うことを検討中である。

ステンレス鋼からのトリチウム放出挙動の解明は、トリチウム化されたステンレス鋼の中間貯蔵や処分にかかるハンドリング時の安全性を確保するうえで重要である。 金属表面酸化膜が水素透過を阻害することは良く知られた事実である。ところが最近、室温付近におけるSUS316からのトリチウム放出は、バルク中の拡散が律速しているとの報告がなされた[1]。一方では、表面には極めて高濃度のトリチウムが存在していることもわかっており、そのトリチウム放出挙動に及ぼす役割は明確にされていない。そこで本研究では、ステンレス鋼の微細組織を変化させて、表面のトリチウム濃度の変化およびトリチウム放出の変化を調べ、両者の対応から表面トリチウムの役割を明らかにすることを試みた。 実験は、フェライト系(8Cr-2W)およびオーステナイト系(SUS316) の2種類のステンレス鋼について行った。これら試料鋼への水素導入法・温度を変化させることにより鋼中トリチウム濃度を変化させて、液体シンチレーション測定法によるトリチウム放出実験を行った。その結果、トリチウム導入時の温度・濃度により異なる表面酸化膜バリア効果およびバルク内トラップ効果が、放出挙動を大きく変化させることが分かった。発表では、放出挙動に対応した表面トリチウム濃度測定結果についても報告する。また、これらの結果を関連づけて放出挙動を数値モデル化し、トリチウム放出の機構を検討する。[1] R-D. Penzhorn et al., J. Nucl. Mater. 353 (2006) 66-74

近年、核融合炉の燃料ガスやブランケットシステムからのトリチウム回収あるいはトリチウム分析用の水素ポンプとして酸化物プロトン導電体が注目されている。本研究では酸化物プロトン導電体のひとつであるBa2In2O5セラミックス試料を、トリチウムガスあるいはトリチウム水蒸気中で加熱処理し、その表面および切断面のトリチウム分布をトリチウムイメージングプレートにて測定し、表面へのトリチウムの吸着挙動および試料内部でのトリチウムの溶解量及び拡散係数を測定した。

核融合炉ブランケット構造材の候補材料である低放射化バナジウム合金中の水素同位体の拡散挙動を、トリチウムイメージングプレートを用いて調べた。バナジウム合金試料としてNIFS-HEAT-IIの引抜き加工材および焼鈍材を用い、トリチウムをグロー放電注入した後、拡散焼鈍し、トリチウムイメージングプレートにて拡散後の分布を調べ、トリチウムの拡散係数を求めた。拡散係数の温度依存性から合金元素や加工による欠陥がトリチウムの拡散に及ぼす影響を議論する。

2004年春の年会(E-30)において、純水中でガンマ線照射を行ったジルカロイ酸化膜表面は、大気中にて同積算照射量の照射を行った場合より高い親水性を示す（水の濡れ性が良くなる）ことを報告した。これは、ジルカロイ酸化膜表面と放射線分解した水との相互作用、または、水中における酸化膜表面とγ線との相互作用により酸化膜表面状態が変化し、濡れ性を良くする方向に働いたことを示唆している。しかし、水中でγ線照射したジルカロイ酸化膜表面の物理的・化学的状態を直接観察または測定する手段は乏しく、表面状態がどのように濡れ性に影響を及ぼすかは明らかではない。本報告では、水中にてジルカロイ酸化膜、単結晶および多結晶ジルコニア、金属類にγ線を照射し、濡れ性の変化と表面状態との関連について検討した。

トリチウムの燃料被覆管中への蓄積や環境への漏出を評価するためにジルコニウム合金中のトリチウムの拡散係数を把握しておくことは重要である。金属中に溶解したトリチウムの拡散係数を測定する手段として、トリチウムを金属中で拡散させ、その放射能を測定することにより、拡散深さや拡散プロファイルを決定する方法がよく用いられている。今回、トリチウムの濃度プロファイル測定のための手段としてイメージングプレート（IP）を用い、トリチウムの拡散係数を決定した。その結果を既存のデータと比較して、この方法の金属中に溶解したトリチウムの濃度分布決定法としての妥当性を検討した。

The impact properties of an oxidized and a hydrogenated Zircaloy have been studied with an instrumented Charpy machine by using a strip Charpy V-notch specimen (1 mm thick by 4 mm wide). Fracture processes such as crack initiation and propagation were examined using load-displacement curves obtained in this study. In the case of the hydrogenated specimen containing preferentially oriented hydrides, an appreciable decrease in the absorbed energy was observed in the crack propagation rather than in the crack initiation. From results of fractographs of the specimen, it was suggested that the r...

ジルコニウム合金酸化膜の濡れ性を接触角測定により調べた。また、純水中および大気中でガンマ線照射を行い、ジルコニウム合金酸化膜の親水化機構を検討した。 ジルコニウム合金酸化膜は、水との接触角が親水性および疎水性の境界である90度付近にあり、結晶構造および表面粗さなどの要因により親水性および疎水性を示し得ることが分かった。また、純水中および大気中でジルコニウム合金酸化膜にガンマ線照射を行った場合に、酸化膜の濡れ性の変化には大きな違いが見られた。

原型炉初期保有及び炉工学試験用トリチウム（Ｔ）調達法として高温ガス炉を用いたＴ製造法を検討している．HTTRを用いたＴ製造実証試験に向け，試験体，試験法，試験後Ｔ計測法を明確にすることを目的に下記(1)～(4)を実施した． (1)アルミナ円柱キャプセルの製作：水素吸収用にニッケル（Ni）被覆ジルコニウム（Zr）粒子及び高密度アルミナ円柱キャプセルを試作した．試作したアルミナ容器に，リチウムアルミニウム酸化物（LiAlO2）及びNi被覆Zr粒子を装荷した状態で600℃に加熱して３日間保持した後，アルミナ容器外部の水素圧力を観測することで，その封止性能を確認した． (2) Ni被覆チタン（Ti）粒子の水素吸収性能に関する検討：LiAlO2共存下においてNi被覆したTi粒子による水素吸蔵特性を確認した．Ni被覆厚さ（0.3～3ミクロン）には依存せず，Ni被覆Ti粒子中の見かけの水素拡散係数は純Tiの水素拡散係数の1/100程度であることがわかった．また，LiAlO2を予水素処理することにより，Ni被覆Ti粒子の水素吸蔵特性を改善できることがわかった． (3)照射後試料Ｔ測定法の検討：水バブラーと酸化銅を組み合わせることで，水蒸気状Ｔ（HTO）と水素状Ｔ（HT）を弁別測定可能なＴ測定装置を作製した．LiAlO2の模擬試料としてLi2TiO3を準備し，京大炉で中性子照射した後，九大にてＴ放出実験を行った．Arガスパージでは，ほとんどのＴがHTOとして放出されることがわかった． (4)HTTR炉心中性子マップの作成： HTTRを用いたＴ製造試験の検討に資するために，HTTR炉内の中性子束空間分布を示す中性子束マップを作成した．一般的に，原子炉内の中性子束空間分布は燃焼に伴い変化するため，HTTRの3次元炉心モデルを作成して炉心燃焼計算を行い，燃焼日数別にHTTRの炉内中性子束マップを作成した．

A tritium (T) production method for fusion reactors using high-temperature gas-cooled reactor has been developed. In this period of research, we focus our attention particularly on the Li-loading rod, in which stable T confinement is required. A basic characteristics of Zr for the T confinement, i.e., influences of chemical form of T produced and hydrogen absorption capability of the Zr when coexisting with oxides, was examined. In addition, experimental procedure and the test module structure have been investigated assuming that the future irradiation test in High Temperature engineering Test Reactor (HTTR). A method of the irradiation test and the structure of the test module have been clarified. A method for chemical evaluation of impurity concentration in He gas in HTTR has also been developed, and it has been shown that the influence of exposure of radiation due to T production is negligible.

Heading on the demonstration of the tritium production using high-temperature gas-cooled reactor, tritium permeation experiment using the mockup of the Li-loading rod, which structured by Zr and Al2O3 tubes, was performed at 700 ℃ temperature. Tritium was kept being contained in the rod during 10 hours, which shows that the Li-loading rod has excellent tritium-containment performance. Other experiment showed that tritium-absorption performance of Zr is reduced in coexistence state of Zr and LiAlO2, but the performance can be recovered by using Ni coating on the Zr rod. An experimental procedure and test module were examined assuming future irradiation test in High Temperature engineering Test Reactor (HTTR). It was shown that almost 30 g of tritium can be produced in HTTR during 1 year operation.

ITER will use Be as a main chamber wall material and W as a divertor material. To examine the performances of these materials in tokamak environment, JET in EU has performed ITER-Like Wall (ILW) experimental campaigns with Be main chamber tiles and W-coated CFC divertor tiles. In this study, tritium (T) distributions on these tiles were examined using imaging plate (IP) technique and beta-ray induced X-ray spectrometry. Two retention mechanisms were found: (1) co-deposition with Be and other impurities, and (2) implantation into material bulk. T concentrations in Be deposition layers were lower than those in carbon deposition layers formed in previous campaigns with carbon tiles. Far less co-deposition of T on the sides of tiles was observed for ILW tiles compared with carbon tiles. Significantly reduced T retention was expected with Be and W walls in comparison with C walls. Technique to measure T retention in an individual dust particle using IP was also developed.

Models of transport of tritium through metal/water or water/metal interfaces are proposed as follows, (1) Tritium entry in metals by water corrosion, (2) Tritium release into water by oxidation, (3) effects of residual stress on tritium permeation behaviors (1) and (2). A part of tritium produced by water corrosion of metals dissolves and enters in interstitials of metals. The rate of permeation of tritium is determined by fugacity of tritium dissolution and diffusivity of hydrogen in metals. Atomic tritium permeated through the metals is released into water by oxidation of atomic tritium to water form. The effects of stress or strain on tritium permeation is very small.

In order to clarify mechanisms of hydrogen accumulation and release in/from metals, tritium tracer techniques are developed to measure amounts of released or permeated hydrogen from metals and distribution of hydrogen retained in metals. Applying these techniques, hydrogen diffusion and permeation coefficients in metals at low temperatures around room temperature are successfully determined for the first time.

核融合実証炉の第一壁としてプラズマ対向面にタングステン(W)を被覆したフェライト・マルテンサイト鋼(F82H)を利用することが検討されている。W被覆層は放射性トリチウム(T)を含んだプラズマに曝されることになるので、炉内のTインベントリー評価ひいては安全性評価のために、TがW被覆層およびF82H基板にどのように進入し、蓄積(汚染)するのか、あるいは、加熱放出によってTがどのように除去(除染)されるのかを明らかにすることが重要である。本研究では、極微量のTを含んだ水素のDCグロー放電プラズマに、溶射法によりW被覆したF82Hを曝したのち、被覆層および基板中のトリチウム進入深さ分布をイメージングプレート法により測定し、その分布の温度変化および時間変化を調べることにより、両者への水素の進入・蓄積および放出機構を解明することを目的とした。 本年度の研究成果は、W被覆層中では主としてガス状水素の空隙(粒界)拡散が、F82H基板中では溶解水素の拡散が水素の進入または放出に大きく寄与しているという機構を構築できたことである。多孔質なW被覆層は、プラズマからの原子状、イオン状の水素の進入を表面(近傍)で阻み、ガス状水素の進入を空隙(粒界)に限定することにより、F82H基板への水素(T)の拡散進入・蓄積および透過を低減する役割を果たしていると考えられる。一方で、W被覆層では、プラズマ注入時の表面改質および表面近傍で生成した欠陥が水素捕獲サイトになり、これらに捕獲された水素が高濃度に局在化しており、内部では水素が各W粒子表面へ吸着し、さらに粒内への溶解するので、Tの蓄積量が増える可能性がある。表面近傍に局在した捕獲水素の捕獲エネルギーは極めて大きく、大気中保持や真空加熱を実施しても放出除去することが難しいこともわかった。

放射性トリチウムによって汚染された金属材料の保管・廃棄にかかるハンドリングや除染時において、外部被爆や経口摂取による内部被爆を防止する観点から、金属表面および内部のトリチウム挙動を理解することは極めて重要である。金属材料表面には極めて高密度にトリチウムが偏在することが知られているが、この表面に偏在したトリチウムが及ぼす材料内部からのトリチウム放出に及ぼす影響は必ずしも明らかになっていない。本研究ではトリチウムトレーサー技術を利用し、金属中のトリチウム吸・放出挙動に及ぼす表面に偏在したトリチウムの影響を解明することを目的としている。 水素溶解度や拡散係数が既知である各種金属にトレーサーレベルのトリチウムを含有した水素を高温気体吸収法(400℃,4kPa)により溶解し、これら金属の表面水素濃度をトリチウムイメージングプレート(IP)法により、また材料からの水素放出速度を液体シンチレーション計測法により測定し、両者の関係を調べた。この結果、表面に偏在した水素と内部に溶解した水素との存在量比によって、金属からの水素放出挙動が整理されることがわかった。ニッケル(Ni)のようなFCC金属では、内部に溶解した水素量が表面に偏在した水素量よりも多いため、金属からの水素放出は、見かけ上、内部に溶解した水素の拡散によるものである。表面に偏在した水素は、金属表面酸化膜に-OH基や吸着水として存在しており、深い捕獲サイトに捕獲されたものであると考えられる。このような表面捕獲水素の室温付近における脱捕獲速度は小さい。このため、内部に溶解した水素が放出されてしまえば、相対的に存在量が多くなった表面に偏在した水素の脱離が見えるようになる。同様に、表面酸化膜が水素を捕獲しやすい銅(Cu)やBCC金属からの水素放出は、内部に溶解した水素量が表面に偏在した水素量よりも少ないため、表面に偏在した水素の脱離によるものであるといえる。

In recent years, a special concern has been raised about the development and commercialization of fuel cell (FC) systems to solve both the global warming and energy problems. Under such circumstance, the role of this research project has been significantly increasing to ensure the safety use of FC systems in the near future. In this project, the effect of hydrogen on giga-cycle fatigue mechanism in high strength steels has been studied as well as the effect of hydrogen on fatigue properties of candidate materials for FC systems. The obtained results are as follows: (1) The evidences of interaction of hydrogen on giga-cycle fatigue failure have been shown by the fatigue tests of hydrogen-content-controlled specimens, the secondary ion mass spectrometry and the tritium autoradiography. (2) The giga-cycle fatigue mechanism taking the hydrogen interaction into consideration has been proposed. It has been shown that the giga-cycle fatigue strength can be improved by controlling hydrogen content in materials, inclusion size and inclusion type. (3) A fatigue design method in giga-cycle regime has been proposed based on the area parameter model, the statistics of extremes and the growth curve of the optically dark area (ODA). (4) A number of reliable fatigue data on the effect of hydrogen has been obtained about the candidate materials for FC systems. In addition, some important findings about the degradation mechanism due to hydrogen have been given, e.g. the slip localization due to hydrogen and the effect of phase transformations on the crack-growth acceleration, etc. Considering all the results in this project, the following two significant conclusions have been obtained: (I) Hydrogen does not cause so-called "embrittlement" of materials, but facilitates the dislocation mobility resulting in the slip concentration. (II) The role of hydrogen trapped by inclusions in giga-cycle fatigue mechanism is to cause the microscopic slip concentration even at the lower stress.