KINDAI UNIVERSITY


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OTSUKA Teppei

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FacultyDepartment of Electric and Electronic Engineering / Atomic Energy Research Institute
PositionAssociate Professor
Degree
Commentator Guidehttps://www.kindai.ac.jp/meikan/1491-otsuka-teppei.html
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Last Updated :2020/09/30

Research Activities

Research Areas

  • Energy, Nuclear engineering, Nuclear materials
  • Energy, Nuclear fusion, Nuclear fusion materials engineering
  • Nanotechnology/Materials, Structural and functional materials
  • Nanotechnology/Materials, Metallic materials

Research Interests

  • Tritium science and technology, Fusion Reactor Materials, Nuclear Materials

Published Papers

  • Li-rod structure in high-temperature gas-cooled reactor as a tritium production device for fusion reactors, Matsuura, Hideaki, Okamoto, Ryo, Koga, Yuki, Suganuma, Takuro, Katayama, Kazunari, Otsuka, Teppei, Goto, Minoru, Nakagawa, Shigeaki, Ishitsuka, Etsuo, Tobita, Kenji, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 146, 1077 - 1081, Sep. 2019 , Refereed
    Summary: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.
  • Tritium retention characteristics in dust particles in JET with ITER-like wall, T. Otsuka, S. Masuzaki, N. Ashikawa, Y. Hatano, Y. Asakura, Tatsuya Suzuki, Takumi Suzuki, K. Isobe, T. Hayashi, M. Tokitani, Y. Oya, D. Hamaguchi, H. Kurotaki, R. Sakamoto, Hiroyasu Tanigawa, M. Nakamichi, A. Widdowson, M. Rubel, Nuclear Materials and Energy, Nuclear Materials and Energy, 17, 279 - 283, Dec. 01 2018
    Summary:© 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.
  • Plasma-Wall Interaction on the Divertor Tiles of JET ITER-Like Wall from the Viewpoint of Micro/Nanoscopic Observations, Fusion Engineering and Design, Fusion Engineering and Design, 136, 199 - 204, Nov. 01 2018
    Summary:© 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.
  • Study on lithium rod test module and irradiation method for tritium production using high temperature gas-cooled reactor, Koga, Yuki, Matsuura, Hideaki, Ida, Yuma, Okamoto, Ryo, Katayama, Kazunari, Otsuka, Teppei, Goto, Minoru, Nakagawa, Shigeaki, Nagasumi, Satoru, Ishitsuka, Etsuo, Shimazaki, Yosuke, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 136, 587 - 591, Nov. 2018 , Refereed
    Summary: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 shot-peening on permeation and retention behaviors of hydrogen in alpha iron, Otsuka, Teppei, Goto, Kengo, Yamamoto, Akihiro, Hashizume, Kenichi, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 136, 509 - 512, Nov. 2018 , Refereed
    Summary: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.
  • Chemical forms of hydrogen desorbed by permeation through pure iron and oxide dispersion strengthened steels, Otsuka, Teppei, Goto, Kengo, Sakamoto, Kan, Hashizume, Kenichi, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 132, 107 - 109, Jul. 2018 , Refereed
    Summary: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.
  • Summary and Future Plan, UEDA Yoshio, HATANO Yuji, YOKOMINE Takehiko, HINOKI Tatsuya, HASEGAWA Akira, OYA Yasuhisa, MUROGA Takeo, Mar. 2017
  • Depth profiling of hydrogen in ferritic/martensitic steels by means of a tritium imaging plate technique, Otsuka, Teppei, Tanabe, Tetsuo, JOURNAL OF ALLOYS AND COMPOUNDS, JOURNAL OF ALLOYS AND COMPOUNDS, 580(Supplement 1), S44 - S46, Dec. 2013 , Refereed
    Summary: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.
  • Direct energy conversion from gamma ray to electricity using silicon semiconductor cells, K. Hashizume, H. Kimura, T. Otsuka, T. Tanabe, T. Okai, Materials Research Society Symposium Proceedings, Materials Research Society Symposium Proceedings, 1264, 223 - 228, Dec. 24 2010
    Summary: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.
  • Hydrogen diffusion and trapping process around MnS precipitates in alpha Fe examined by tritium autoradiography, Teppei Otsuka, Tetsuo Tanabe, JOURNAL OF ALLOYS AND COMPOUNDS, JOURNAL OF ALLOYS AND COMPOUNDS, 446, 655 - 659, Oct. 2007 , Refereed
    Summary: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.
  • Trace Elements wDiffusion of Helium in Water-Saturated, Compacted Sodium Montmorillonite Japan Ab Initio Calculations for the Substitutions of Al(OH)-4 and SO2-4 with Si Tetrahedra, Tomohiro Higashihara, Hiroshi Ohashi, Journal of Nuclear Science and Technology, Journal of Nuclear Science and Technology, 38, 577 - 580, Jan. 01 2001
  • Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall, S. Masuzakil, M. Tokitanii, T. Otsuka, Y. Oya, Y. Hatan, M. Miyamoto, R. Sakamoto, N. Ashikawa, S. Sakurada, Y. Uemura, K. Azuma, K. Yumizurus, M. Oyaizu, T. Suzuki, H. Kurotaki, D. Hamaguchi, K. Isobel, N. Asakura, A. Widdowson, K. Heinola, S. Jachmich, M. Rubel, PHYSICA SCRIPTA, PHYSICA SCRIPTA, T170, Dec. 2017
    Summary: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 from JET ITER-like wall, M. Tokitani, M. Miyamoto, S. Masuzaki, Y. Fujii, R. Sakamoto, Y. Oya, Y. Hatano, T. Otsuka, M. Oyaidzu, H. Kurotaki, T. Suzuki, D. Hamaguchi, K. Isobe, N. Asakura, A. Widdowson, M. Rubel, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 116, 1 - 4, Mar. 2017
    Summary: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.
  • Application of a Tritium Imaging Plate Technique to Depth Profiling of Hydrogen in Metals and Determination of Hydrogen Diffusion Coefficients, Teppei Otsuka, Tetsuo Tanabe, MATERIALS TRANSACTIONS, MATERIALS TRANSACTIONS, 58(10), 1364 - 1372, 2017 , Refereed
    Summary: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.
  • Study on Tritium Production Using a High-Temperature Gas-Cooled Reactor for Fusion Reactors: Evaluation of Tritium Outflow by Non-Equilibrium Diffusion Simulations, S. Nagasumi, H. Matsuura, K. Katayama, T. Otsuka, M. Goto, S. Nakagawa, FUSION SCIENCE AND TECHNOLOGY, FUSION SCIENCE AND TECHNOLOGY, 72(4), 753 - 759, 2017
    Summary: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 behavior of tritium in pure copper and its alloys into pure water at ambient temperature, Teppei Otsuka, Yusuke Ogawa, Hiroki Horinouchi, Kenichi Hashizume, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 113, 227 - 230, Dec. 2016 , Refereed
    Summary: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.
  • Gas-driven permeation of deuterium through tungsten and tungsten alloys, Dean A. Buchenauer, Richard A. Karnesky, Zhigang Zak Fang, Chai Ren, Yasuhisa Oya, Teppei Otsuka, Yuji Yamauchi, Josh A. Whaley, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 109(PA), 104 - 108, Nov. 2016 , Refereed
    Summary: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 retention in individual metallic dust particles examined by a tritium imaging plate technique, T. Otsuka, Y. Hatano, PHYSICA SCRIPTA, PHYSICA SCRIPTA, T167, Feb. 2016 , Refereed
    Summary: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.
  • Tritium retention in individual metallic dust particles examined by a tritium imaging plate technique, T. Otsuka, Y. Hatano, Physica Scripta, Physica Scripta, 2016(167), Jan. 25 2016
    Summary:Tritium imaging plate technique (TIPT) has been applied to examine tritium (T) retention in individual particles made of titanium (Ti) with 30 and 100 μm in diameter and tungsten (W) with 50 μm 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 β-ray and thus accurate quantification of the amount of T in the individual particle.
  • Change of chemical states of niobium in the oxide layer of zirconium-niobium alloys with oxide growth, Kan Sakamoto, Katsumi Une, Masaki Aomi, Teppei Otsuka, Kenichi Hashizume, JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 52(10), 1259 - 1264, Oct. 2015 , Refereed
    Summary: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.
  • Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten, Masashi Shimada, Masanori Hara, Teppei Otsuka, Yasuhisa Oya, Yuji Hatano, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 463, 1005 - 1008, Aug. 2015 , Refereed
    Summary: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 tritium loaded near the surface region of metals by gas absorption and plasma implantation, T. Otsuka, Y. Ogawa, M. Higaki, Y. Ishitani, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 463, 1029 - 1032, Aug. 2015
    Summary: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.
  • Hydrogen solubility and diffusivity in a barium cerate protonic conductor using tritium imaging plate technique, K. Yamashita, T. Otsuka, K. Hashizume, SOLID STATE IONICS, SOLID STATE IONICS, 275, 43 - 46, Jul. 2015
    Summary: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.
  • BEHAVIOR OF TRITIUM PERMEATION INDUCED BY WATER CORROSION OF ALPHA IRON AROUND ROOM TEMPERATURE, Teppei Otsuka, Kenichi Hashizume, FUSION SCIENCE AND TECHNOLOGY, FUSION SCIENCE AND TECHNOLOGY, 67(3), 511 - 514, Apr. 2015 , Refereed
    Summary: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.
  • Determination of hydrogen diffusion coefficients in F82H by hydrogen depth profiling with a tritium imaging plate technique, M. Higaki, T. Otsuka, K. Tokunaga, K. Hashizume, K. Ezato, S. Suzuki, M. Enoeda, M. Akiba, Fusion Science and Technology, Fusion Science and Technology, 67(2), 379 - 381, Mar. 01 2015
    Summary: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.
  • Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten, Masashi Shimada, G. Cao, T. Otsuka, M. Hara, M. Kobayashi, Y. Oya, Y. Hatano, NUCLEAR FUSION, NUCLEAR FUSION, 55(1), Jan. 2015
    Summary: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).
  • Hydrogen permeation in iron and nickel alloys around room temperature, T. Otsuka, M. Shinohara, H. Horinouchi, T. Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 442(1-3), S726 - S729, Nov. 2013
    Summary: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.
  • Retention and release mechanisms of tritium loaded in plasma-sprayed tungsten coatings by plasma exposure, T. Otsuka, T. Tanabe, K. Tokunaga, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 438, S1048 - S1051, Jul. 2013
    Summary: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.
  • Deuterium trapping at defects created with neutron and ion irradiations in tungsten, Y. Hatano, M. Shimada, T. Otsuka, Y. Oya, V. Kh. Alimov, M. Hara, J. Shi, M. Kobayashi, T. Oda, G. Cao, K. Okuno, T. Tanaka, K. Sugiyama, J. Roth, B. Tyburska-Pueschel, J. Dorner, N. Yoshida, N. Futagami, H. Watanabe, M. Hatakeyama, H. Kurishita, M. Sokolov, Y. Katoh, NUCLEAR FUSION, NUCLEAR FUSION, 53(7), Jul. 2013
    Summary: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.
  • Material behavior on heat loading and hydrogen penetration of vacuum plasma spray tungsten coatings on reduced activation ferritic/martensitic steel, Kazutoshi Tokunaga, Tomohiro Hotta, Teppei Otsuka, Akira Kobayashi, Kuniaki Araki, Yoshio Miyamoto, Tadashi Fujiwara, Makoto Hasegawa, Kazuo Nakamura, Koichiro Ezato, Satoshi Suzuki, Mikio Enoeda, Masato Akiba, Takuya Nagasaka, Ryuta Kasada, Akihiko Kimura, Yosetsu Gakkai Ronbunshu/Quarterly Journal of the Japan Welding Society, Yosetsu Gakkai Ronbunshu/Quarterly Journal of the Japan Welding Society, 31(4), 2013
    Summary: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.
  • Determination of hydrogen diffusion and permeation coefficients in pure copper at near room temperature by means of tritium tracer techniques, H. Horinouchi, M. Shinohara, T. Otsuka, K. Hashizume, T. Tanabe, Journal of Alloys and Compounds, Journal of Alloys and Compounds, 580(1), S73 - S75, 2013
    Summary: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 Φ = (2.8 ± 0.4) × 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. © 2013 Elsevier B.V. All rights reserved.
  • Behaviour of tritium in plasma-sprayed tungsten coating on steel exposed to tritium plasma, T. Otsuka, T. Tanabe, K. Tokunaga, PHYSICA SCRIPTA, PHYSICA SCRIPTA, T145, Dec. 2011 , Refereed
    Summary: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.
  • OVERVIEW OF RECENT TRITIUM EXPERIMENTS IN TPE, Masashi Shimada, T. Otsuka, R. J. Pawelko, P. Calderoni, J. P. Sharpe, FUSION SCIENCE AND TECHNOLOGY, FUSION SCIENCE AND TECHNOLOGY, 60(4), 1495 - 1498, Nov. 2011
    Summary: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.
  • BEHAVIOR OF TRITIUM NEAR SURFACE REGION OF METALS EXPOSED TO TRITIUM PLASMA, T. Otsuka, M. Shimada, T. Tanabe, J. P. Sharpe, FUSION SCIENCE AND TECHNOLOGY, FUSION SCIENCE AND TECHNOLOGY, 60(4), 1539 - 1542, Nov. 2011
    Summary: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.
  • APPLICATION OF TRITIUM TRACER TECHNIQUE TO DETERMINATION OF HYDROGEN DIFFUSION COEFFICIENTS AND PERMEATION RATE NEAR ROOM TEMPERATURE FOR TUNGSTEN, T. Ikeda, T. Otsuka, T. Tanabe, FUSION SCIENCE AND TECHNOLOGY, FUSION SCIENCE AND TECHNOLOGY, 60(4), 1463 - 1466, Nov. 2011
    Summary: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.
  • Thermomigration of tritium in V-4Cr-4Ti alloy, Kenichi Hashizume, Kazuhiro Ogushi, Teppei Otsuka, Tetsuo Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 417(1-3), 1175 - 1178, Oct. 2011 , Refereed
    Summary: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.
  • Hydrogen permeation in metals near room temperature by a tritium tracer technique, Takahiro Ikeda, Teppei Otsuka, Tetsuo Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 417(1-3), 568 - 571, Oct. 2011 , Refereed
    Summary: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), and phi = (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.
  • Hydrogen behavior near surface regions in Mo and W studied by tritium tracer technique, Takamitsu Hoshihira, Teppei Otsuka, Ryusuke Wakabayashi, Tetsuo Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 417(1-3), 559 - 563, Oct. 2011 , Refereed
    Summary: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.
  • Application of tritium tracer techniques to observation of hydrogen on surface and in bulk of F82H, T. Otsuka, T. Tanabe, K. Tokunaga, N. Yoshida, K. Ezato, S. Suzuki, M. Akiba, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 417(1-3), 1135 - 1138, Oct. 2011
    Summary: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.
  • An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility, P. Calderoni, J. Sharpe, M. Shimada, B. Denny, B. Pawelko, S. Schuetz, G. Longhurst, Y. Hatano, M. Hara, Y. Oya, T. Otsuka, K. Katayama, S. Konishi, K. Noborio, Y. Yamamoto, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 417(1-3), 1336 - 1340, Oct. 2011
    Summary: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.
  • Determination of hydrogen diffusivity and permeability in W near room temperature applying a tritium tracer technique, T. Ikeda, T. Otsuka, T. Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 415(1), S684 - S687, Aug. 2011
    Summary: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.
  • Application of tritium imaging plate technique to examine tritium behaviors on the surface and in the bulk of plasma-exposed materials, T. Otsuka, M. Shimada, R. Kolasinski, P. Calderoni, J. P. Sharpe, Y. Ueda, Y. Hatano, T. Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 415(1), S769 - S772, Aug. 2011
    Summary: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.
  • Behaviour of tritium in plasma-sprayed tungsten coating on steel exposed to tritium plasma, T. Otsuka, T. Tanabe, K. Tokunaga, Physica Scripta T, Physica Scripta T, T145, 2011
    Summary: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.
  • Behavior of tritium accumulated on materials surface, Teppei Otsuka, Tetsuo Tanabe, FUSION ENGINEERING AND DESIGN, FUSION ENGINEERING AND DESIGN, 85(7-9), 1437 - 1441, Dec. 2010 , Refereed
    Summary: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.
  • Visualization of hydrogen depth profile by means of tritium imaging plate technique: determination of hydrogen diffusion coefficient in pure tungsten, T. Otsuka, T. Hoshihira, T. Tanabe, PHYSICA SCRIPTA, PHYSICA SCRIPTA, T138, Dec. 2009 , Refereed
    Summary: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.
  • Characterization of surface morphology and retention in tungsten materials exposed to high fluxes of deuterium ions in the tritium plasma experiment, R. D. Kolasinski, M. Shimada, D. A. Buchenauer, R. A. Causey, T. Otsuka, W. M. Clift, J. M. Shea, T. R. Allen, P. Calderoni, J. P. Sharpe, PHYSICA SCRIPTA, PHYSICA SCRIPTA, T138, Dec. 2009 , Refereed
    Summary: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.
  • A study of hydrogen blistering mechanism for Molybdenum by Tritium radio-luminography, T. Hoshihira, T. Otsuka, T. Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 390-91, 1029 - 1031, Jun. 2009
    Summary: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.
  • Hydrogen diffusion in Fe-Ni alloys around room temperature, Teppei Otsuka, Shinsuke Sasabe, Tetsuo Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 386-88, 884 - 887, Apr. 2009 , Refereed
    Summary: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.
  • Visualization of hydrogen distribution around blisters by tritium radio-luminography, T. Hoshihira, T. Otsuka, T. Tanabe, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 386-88, 776 - 779, Apr. 2009
    Summary: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.
  • Visualization of hydrogen depth profile by means of tritium imaging plate technique: Determination of hydrogen diffusion coefficient in pure tungsten, T. Otsuka, T. Hoshihira, T. Tanabe, Physica Scripta T, Physica Scripta T, T138, 2009
    Summary: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.
  • Characterization of surface morphology and retention in tungsten materials exposed to high fluxes of deuterium ions in the tritium plasma experiment, R. D. Kolasinski, M. Shimada, D. A. Buchenauer, R. A. Causey, T. Otsuka, W. M. Clift, J. M. Shea, T. R. Allen, P. Calderoni, J. P. Sharpe, Physica Scripta T, Physica Scripta T, T138, 2009
    Summary: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 from ferritic/martensitic stainless steel, Teppei Otsuka, Tetsuo Tanabe, FUSION SCIENCE AND TECHNOLOGY, FUSION SCIENCE AND TECHNOLOGY, 54(2), 541 - 544, Aug. 2008 , Refereed
    Summary: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.
  • Diffusional behavior of tritium in V-4Cr-4Ti alloy, K. Hashizume, J. Masuda, T. Otsuka, T. Tanabe, Y. Hatano, Y. Nakamura, T. Nagasaka, T. Muroga, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 367, 876 - 881, Aug. 2007
    Summary: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.
  • Diffusion and trapping of tritium in vanadium alloys, J. Masuda, K. Hashizume, T. Otsuka, T. Tanabe, Y. Hatano, Y. Nakamura, T. Nagasaka, T. Muroga, JOURNAL OF NUCLEAR MATERIALS, JOURNAL OF NUCLEAR MATERIALS, 363, 1256 - 1260, Jun. 2007
    Summary: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.
  • Profiling of hydrogen accumulation in a tempered martensite microstructure by means of tritium autoradiography, H Hanada, T Otsuka, H Nakashima, S Sasaki, M Hayakawa, M Sugisaki, SCRIPTA MATERIALIA, SCRIPTA MATERIALIA, 53(11), 1279 - 1284, Dec. 2005
    Summary: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.
  • Observation of hydrogen distribution around non-metallic inclusions in steels with tritium microautoradiography, T Otsuka, H Hanada, H Nakashima, K Sakamoto, M Hayakawa, K Hashizume, M Sugisaki, FUSION SCIENCE AND TECHNOLOGY, FUSION SCIENCE AND TECHNOLOGY, 48(1), 708 - 711, Jul. 2005
    Summary: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.
  • Charpy impact test of oxidized and hydrogenated Zircaloy using a thin strip specimen, T Otsuka, K Hashizume, M Sugisaki, JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 41(3), 247 - 251, Mar. 2004
    Summary: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.
  • Diffusion of helium in water-saturated, compacted sodium montmorillonite, S Sato, T Otsuka, Y Kuroda, T Higashihara, H Ohashi, JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 38(7), 577 - 580, Jul. 2001

Conference Activities & Talks

  • 高温ガス炉用LiロッドのT閉じ込め特性にZr水素吸蔵性能が与える影響, 岡本亮, 松浦秀明, 井田祐馬, 古賀友稀, 片山一成, 大塚哲平, 後藤実, 中川繁昭, 石塚悦男, 長住達, 日本原子力学会春の年会予稿集(CD-ROM),   2018 03 08
  • 管状ジルコニウムにおける水素透過挙動, 片山一成, 泉野純逸, 松浦秀明, 大塚哲平, 深田智, 日本原子力学会春の年会予稿集(CD-ROM),   2018 03 08
  • JET‐ITER like wallから発生したダスト粒子のトリチウム保持特性, 大塚哲平, 増崎貴, 芦川直子, 波多野雄治, 朝倉伸幸, 鈴木卓美, 鈴木達也, 磯部兼嗣, 林巧, 時谷政行, 大矢恭久, 濱口大, 黒滝宏紀, 坂本隆一, 谷川博康, 中道勝, WIDDOWSON A, RUBEL M, 日本原子力学会春の年会予稿集(CD-ROM),   2018 03 08
  • ODS鋼中のトリチウム透過挙動, 浦部雄大, 橋爪健一, 大塚哲平, 坂本寛, 日本金属学会九州支部・日本鉄鋼協会九州支部・軽金属学会九州支部合同学術講演大会講演概要集,   2018
  • 改良ステンレス鋼燃料被覆管のBWR装荷に向けた研究開発(2)(6)トリチウム透過特性・水蒸気酸化特性, 高橋克仁, 坂本寛, 大塚哲平, 鵜飼重治, 平井睦, 山下真一郎, 日本原子力学会秋の大会予稿集(CD-ROM),   2017 08 29
  • ジルコニウム中の炭素拡散係数測定, 大塚哲平, 松本剛, 日本原子力学会秋の大会予稿集(CD-ROM),   2017 08 29
  • 超高温ガス炉用LiロッドにおけるZrを用いたトリチウム閉じ込め法の検討~高温(700~850°C)条件下におけるZr水素吸蔵特性~, 岡本亮, 松浦秀明, 井田祐馬, 古賀友稀, 片山一成, 大塚哲平, 後藤実, 中川繁昭, 石塚悦男, 長住達, 島崎洋祐, 日本原子力学会秋の大会予稿集(CD-ROM),   2017 08 29
  • 日米科学技術協力事業PHENIX計画―前半の成果と後半の研究計画―5.タスク3 トリチウム挙動および中性子照射効果, 大矢恭久, 波多野雄治, 片山一成, 山内有二, 信太祐二, 大塚哲平, 近田拓未, 原正憲, 大宅諒, 上田良夫, 外山健, プラズマ・核融合学会誌,   2017 03 25
  • 金属ナノコンタクトへの低温水素吸蔵と共鳴トンネル現象の観測, 高田弘樹, 家永紘一郎, 瀬尾優太, ISLAM Md. S, 稲垣祐次, 辻井宏之, 橋爪健一, 大塚哲平, 河江達也, 日本物理学会講演概要集(CD-ROM),   2017 03 21
  • 高温ガス炉を用いたT生産Li装荷ロッドの照射試験体及び試験法の検討~Zr層を考慮した試験体の評価~, 井田祐馬, 松浦秀明, 長住達, 古賀友稀, 岡本亮, 片山一成, 大塚哲平, 後藤実, 中川繁昭, 石塚悦男, 日本原子力学会春の年会予稿集(CD-ROM),   2017 03 09
  • Tritium Behavior and Neutron Irradiation Effect, 大矢 恭久, 波多野 雄治, 片山 一成, 山内 有二, 信太 祐二, 大塚 哲平, 近田 拓未, 原 正憲, 大宅 諒, 上田 良夫, 外山 健, プラズマ・核融合学会誌 = Journal of plasma and fusion research,   2017 03
  • 純鉄の水素透過挙動に及ぼすショットピーニングの影響, 後藤健吾, 大塚哲平, 橋爪健一, 材料とプロセス(CD-ROM),   2016 09 01
  • 高温ガス炉トリチウム生産と閉じ込め手法の検討, 片山一成, 松浦秀明, 大塚哲平, 深田智, 後藤実, 中川繁昭, 日本原子力学会秋の大会予稿集(CD-ROM),   2016 08 26
  • 処分環境下におけるジルカロイの腐食挙動(2)酸化膜の性状評価, 池田陽子, 大塚哲平, 桜木智史, 吉田誠司, 日本原子力学会秋の大会予稿集(CD-ROM),   2016 08 26
  • 改良ステンレス鋼燃料被覆管のBWR装荷に向けた研究開発(5)トリチウム透過試験, 平井睦, 坂本寛, 鵜飼重治, 木村晃彦, 草ヶ谷和幸, 大塚哲平, 山下真一郎, 日本原子力学会秋の大会予稿集(CD-ROM),   2016 08 26
  • 酸素溶解ジルコニウム中のトリチウム拡散, 森玉貴也, 橋爪健一, 大塚哲平, 加藤修, 建石剛, 日本原子力学会秋の大会予稿集(CD-ROM),   2016 08 26
  • 処分環境下におけるジルカロイの腐食挙動(3)諸要因の検討, 大塚哲平, 橋爪健一, 加藤修, 建石剛, 吉田誠司, 桜木智史, 日本原子力学会秋の大会予稿集(CD-ROM),   2016 08 26
  • 高温ガス炉を用いたトリチウム生産と研究の概要, 松浦秀明, 片山一成, 大塚哲平, 後藤実, 中川繁昭, 日本原子力学会秋の大会予稿集(CD-ROM),   2016 08 26
  • Vナノコンタクトへの低温での水素・重水素吸蔵現象と電気伝導特性変化, 高田弘樹, 家永紘一郎, 上野友輔, ISLAM Md. S, 稲垣祐次, 辻井宏之, 橋爪健一, 大塚哲平, 河江達也, 日本物理学会講演概要集(CD-ROM),   2016 03 22
  • 金属ナノコンタクトへの低温水素注入による電子状態制御, 高田弘樹, 家永紘一郎, ISLAM Md. S, 稲垣祐次, 辻井宏之, 橋爪健一, 大塚哲平, 河江達也, 日本物理学会講演概要集(CD-ROM),   2016 03 22
  • 高温ガス炉におけるLi装荷法とトリチウム生産性能の検討, 長住達, 中屋裕行, 松浦秀明, 片山一成, 大塚哲平, 後藤実, 中川繁昭, 日本原子力学会春の年会予稿集(CD-ROM),   2016 03 16
  • 室温近傍の水腐食によって生成したジルコニウム酸化物の結晶構造, 大塚哲平, 橋爪健一, 加藤修, 建石剛, 吉田誠司, 桜木智史, 日本原子力学会春の年会予稿集(CD-ROM),   2016 03 16
  • 21aPS-84 Control of electrical properties by hydrogen loading into metallic nanocontacts at low temperatures, Takata H, Ienaga K, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan,   2016
  • Resonant tunneling of hydrogen atoms in Nb nanocontacts, Takata H, Ienaga K, Kajiwara Y, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan,   2016
    Summary:

    水素は最も軽い元素であり非常に強い量子性を示す。この強い量子性を反映して、水素吸蔵金属内にある水素原子は、低温ではトンネル効果によって内部を拡散すると考えられている。我々はこれまで、ブレークジャンクション法を用いて作成した金属Pd、Vナノコンタクトへの液体水素中における水素吸蔵現象を、電気伝導特性変化から追跡してきた。これまでの実験からは、熱的拡散が抑制された低温においても水素吸蔵が進行することが明らかになっている。更に今回、Nbナノコンタクトへの水素吸蔵実験を行い、電気伝導特性に金属内水素の移動を反映していると考えられる異常を観測した。本講演ではこれら実験で得られた結果について報告する。

  • 液体水素温度における金属ナノコンタクトへの水素吸蔵現象と電気伝導特性変化, 上野友輔, 高田弘樹, 家永紘一郎, ISLAM Md. S, 稲垣祐次, 辻井宏之, 橋爪健一, 大塚哲平, 河江達也, 日本物理学会講演概要集(CD-ROM),   2015 09 24
  • 液体水素温度におけるV,Nbナノコンタクトへの水素吸蔵現象の研究, 高田弘樹, 家永紘一郎, 上野友輔, ISLAM Md. S, 稲垣祐次, 辻井宏之, 橋爪健一, 大塚哲平, 河江達也, 日本物理学会講演概要集(CD-ROM),   2015 09 24
  • 水腐食により純鉄に取り込まれた水素の観察へのトリチウムトレーサー技術の応用, OTSUKA TEPPEI, 材料とプロセス(CD-ROM),   2015 09 01
  • シリコン半導体素子のガンマ線電池利用, HASHIZUME KEN'ICHI, MATSUBARA KEISUKE, OTSUKA TEPPEI, NAGATA HIROAKI, 日本原子力学会秋の大会予稿集(CD-ROM),   2015 08 21
  • 液体水素温度における金属ナノコンタクトへの水素・重水素吸蔵現象の研究, TAKATA HIROKI, IENAGA KOICHIRO, UENO YUSUKE, ISLAM MD. S, KAWASAKI YOSUKE, INAGAKI YUJI, TSUJII HIROYUKI, HASHIZUME KEN'ICHI, OTSUKA TEPPEI, KAWAE TATSUYA, 日本物理学会講演概要集(CD-ROM),   2015 03 24
  • 低温における金属内への水素および重水素の吸蔵・拡散現象の研究, KAWASAKI YOSUKE, TAKATA HIROKI, ISLAM MD. S, NISHIMURA NAOTO, INAGAKI YUJI, IENAGA KOICHIRO, TSUJII HIROYUKI, HASHIZUME KEN'ICHI, OTSUKA TEPPEI, KAWAE TATSUYA, 日本物理学会講演概要集(CD-ROM),   2015 03 24
  • 超伝導ナノコンタクトに対する水素不純物効果の研究, UENO YUSUKE, TAKATA HIROKI, ISLAM MD. SAIFUL, IENAGA KOICHIRO, INAGAKI YUJI, TSUJII HIROYUKI, HASHIZUME KEN'ICHI, OTSUKA TEPPEI, KAWAE TATSUYA, 日本物理学会講演概要集(CD-ROM),   2015 03 24
  • 22pAA-11 Experimental study on hydrogen/deuterium absorption into metallic nanocontact at liquid hydrogen temperature, Takata H, Ienaga K, Ueno Y, Islam Md. S, Kawasaki Y, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, 日本物理学会講演概要集,   2015 03 21
  • 21pPSB-47 Hydrogen/Deuterium absorption into metallic nanocontact at liquid hydrogen temperature, Kawasaki Y, Takata H, Islam Md. S, Nishimura N, Inagaki Y, Ienaga K, Tsujii H, Hashidume K, Otsuka T, Kawae T, 日本物理学会講演概要集,   2015 03 21
  • ダスト粒子中のトリチウム定量技術の開発, OTSUKA TEPPEI, HATANO YUJI, 日本原子力学会春の年会予稿集(CD-ROM),   2015 03 05
  • SA時BWR用制御棒損傷に関する基礎試験(2)(1)B4Cの水蒸気酸化挙動, HASHIZUME KEN'ICHI, OTSUKA TEPPEI, HORIUCHI RYO, SAKAMOTO KAN, 日本原子力学会春の年会予稿集(CD-ROM),   2015 03 05
  • パラジウム金属およびバナジウム金属へのトンネル効果による水素吸蔵現象の研究, KAWAE TATSUYA, TAKADA HIROKI, IENAGA KOICHIRO, INAGAKI YUJI, HASHIZUME KEN'ICHI, OTSUKA TEPPEI, 日本金属学会講演概要(CD-ROM),   2015 03 04
  • 18aPS-49 Hydrogen absorption into metallic point-contact and electrical conductivity at liquid hydrogen temperature, Ueno Y, Takata H, Ienaga K, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan,   2015
  • 21pPSB-48 Hydrogen impurity effects of superconducting metallic nano-contacts, Ueno Y, Takata H, Islam Md. Saiful, Ienaga K, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan,   2015
  • 17pCS-4 Experimental study on hydrogen absorption into metallic V, Nb point-contact at liquid hydrogen temperature, Takata H, Ienaga K, Ueno Y, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan,   2015
  • トリチウムを用いたBaCe0.9Y0.1O3-α中の水素挙動の解明, YAMASHITA KENTA, OTSUKA TEPPEI, HASHIZUME KEN'ICHI, 固体イオニクス討論会講演要旨集,   2014 11 16
  • 9pPSA-132 Hydrogen Absorption into Vanadium by Quantum Effect, Takata H, Ienaga K, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting abstracts of the Physical Society of Japan,   2014 08 22
  • 10aBE-5 Hydrogen Absorption into Vanadium by Quantum Effect, Takata H, Ienaga K, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting abstracts of the Physical Society of Japan,   2014 08 22
  • 金属バナジウムへのトンネル効果による水素吸蔵現象の研究, TAKATA HIROKI, IENAGA KOICHIRO, INAGAKI YUJI, TSUJII HIROYUKI, HASHIZUME KEN'ICHI, OTSUKA TEPPEI, KAWAE TATSUYA, 日本物理学会講演概要集,   2014 08 22
  • 純鉄の水腐食によるトリチウムの取り込みと透過挙動, 大塚哲平, 橋爪健一, 日本原子力学会春の年会予稿集(CD-ROM),   2014 03 10
  • フェライト鋼にプラズマ注入したトリチウムの除去過程, 檜垣誠, 大塚哲平, 橋爪健一, 日本原子力学会春の年会予稿集(CD-ROM),   2014 03 10
  • Material behavior on heat loading and hydrogen penetration of vacuum plasma spray tungsten coatings on reduced activation ferritic/martensitic steel, Kazutoshi Tokunaga, Tomohiro Hotta, Teppei Otsuka, Akira Kobayashi, Kuniaki Araki, Yoshio Miyamoto, Tadashi Fujiwara, Makoto Hasegawa, Kazuo Nakamura, Koichiro Ezato, Satoshi Suzuki, Mikio Enoeda, Masato Akiba, Takuya Nagasaka, Ryuta Kasada, Akihiko Kimura, Yosetsu Gakkai Ronbunshu/Quarterly Journal of the Japan Welding Society,   2013 12 25
    Summary: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.
  • 日米科学技術協力事業TITANプロジェクト 3.第一壁・ブランケットの物質熱輸送に関する研究 3.1 第一壁トリチウム・物質移行, 徳永和俊, 宮本光貴, 大塚哲平, 梶田信, 大野哲靖, 上田良夫, プラズマ・核融合学会誌,   2013 11 25
  • 日米科学技術協力事業TITANプロジェクト 4.照射複合効果に関する研究 4.1 照射・トリチウム複合効果, 波多野雄治, 大矢恭久, 原正憲, 小田卓司, 大塚哲平, 佐藤紘一, ZHANG Kun, プラズマ・核融合学会誌,   2013 11 25
  • 4.1 Irradiation-Tritium Synergism(4. Irradiation Synergism,Japan-US Joint Research Project TITAN), HATANO Yuji, OYA Yasuhisa, HARA Masanori, ODA Takuji, OTSUKA Teppei, SATO Koichi, ZHANG Kun, Journal of plasma and fusion research,   2013 11 25
    Summary: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.
  • 3.1 Tritium and Mass Transfer in First Wall(3. Transport of Materials and Heat in First Walls and Blankets,Japan-US Joint Research Project TITAN), TOKUNAGA Kazutoshi, MIYAMOTO Mitsutaka, OTSUKA Teppei, KAJITA Shin, OHNO Noriyasu, UEDA Yoshio, Journal of plasma and fusion research,   2013 11 25
    Summary: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.
  • トリチウムトレーサー技術による水素化ジルカロイの腐食挙動の解明, 大塚哲平, 田辺哲朗, 橋爪健一, 西村務, 加藤修, 建石剛, 桜木智史, 日本原子力学会秋の大会予稿集(CD-ROM),   2013 08 20
  • ジルコニウム中の水素の溶解,析出に対する酸素の影響, 橋爪健一, 森聡史, 大塚哲平, 日本原子力学会秋の大会予稿集(CD-ROM),   2013 08 20
  • トリチウムイメージングプレート法による低放射化フェライト・マルテンサイト鋼F82H中の水素拡散係数測定, 檜垣誠, 大塚哲平, 橋爪健一, 徳永和俊, 江里幸一郎, 鈴木哲, 榎枝幹男, 秋場真人, 日本原子力学会春の年会予稿集(CD-ROM),   2013 03 11
    Summary: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。
  • ジルコニウム合金表面酸化膜の成長にともなう特性変化, 坂本寛, 宇根勝己, 橋爪健一, 大塚哲平, 青見雅樹, 日本原子力学会春の年会予稿集(CD-ROM),   2013 03 11
    Summary: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。
  • ヘリウム照射したタングステンのトリチウム蓄積挙動, 大塚哲平, 橋爪健一, 島田雅, 徳永和俊, 日本原子力学会春の年会予稿集(CD-ROM),   2013 03 11
    Summary: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。
  • W中の水素の拡散と捕獲, 田辺哲朗, 池田隆博, 星平貴光, 大塚哲平, KURRI KR (CD) (CD-ROM),   2013
  • 7. Tritium Permeation, Contamination and Decontamination(Tritium Science and Technology for Fusion Reactor), HATANO Yuji, TORIKAI Yuji, OYA Yasuhisa, ODA Takuji, TANAKA Satoru, NAKAMURA Hirofumi, ASAKURA Yamato, OHUCHI Hiroko, OTSUKA Teppei, KOBAYASHI Kazuhiro, Journal of plasma and fusion research,   2009 10 25
  • Application of IP technique to determine diffusion coefficients of tritium in Zr, Saruwatari Yuki, Hirano Tomohisa, Otsuka Teppei, Hashizume Kenichi, Tanabe Tetsuo, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan,   2005
    Summary: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。
  • Charpy Impact Test of Oxidized and Hydrogenated Zircaloy Using a Thin Strip Specimen, OTSUKA Teppei, HASHIZUME Kenichi, SUGISAKI Masayasu, Journal of Nuclear Science and Technology,   2004 03 25
    Summary: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...
  • Wettability of Zirconium Alloys' Oxide films, OTSUKA Teppei, NAKAYAMA Kenji, ISOTANI Takenori, HASHIZUME Kenichi, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan,   2004
    Summary: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。
  • ICONE11-36307 CHARPY IMPACT TEST OF OXIDIZED AND HYDROGENATED ZIRCALOY USING A THIN STRIP SPECIMEN, OTSUKA Teppei, HASHIZUME Kenichi, SUGISAKI Masayasu, Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE,   2003
  • Diffusion of Helium in Water-Saturated, Compacted Sodium Montmorillonite, SATO Seichi, OTSUKA Teppei, KURODA Yasuhiro, HIGASHIHARA Tomohiro, OHASHI Hiroshi, Journal of Nuclear Science and Technology,   2001 07 25
  • Diffusion of helium and estimated diffusion coefficients of hydrogen dissolved in water-saturated, compacted Ca-montmorillonite (Feature Articles "Japan-China Workshop on Nuclear Waste Management and Reprocessing"), HIGASHIHARA T, Otsuka Teppei, Sato Seichi, Journal of nuclear fuel cycle and environment,   2001 03

Misc

  • 地層処分模擬環境下で酸素・水素溶解ジルコニウムの腐食により生成した酸化皮膜の結晶構造解析, 大塚哲平, 橋爪健一, 加藤修, 建石剛, 吉田誠司, 桜木智史, 九州シンクロトロン光研究センター年報, 2016, 20‐22,   2018 03 , http://jglobal.jst.go.jp/public/201802281295934444
  • 高温中性子照射したタングステン中における水素同位体挙動, 大矢恭久, 小林真, 大塚哲平, 片山一成, 信太祐二, 山内有二, 原正憲, 波多野雄治, 島田雅, BUCHENAUER Dean, 加藤雄大, 核融合エネルギー連合講演会(CD-ROM), 12th, ROMBUNNO.28P‐52,   2018 , http://jglobal.jst.go.jp/public/201902259631890172
  • 日米科学技術協力事業PHENIX計画―前半の成果と後半の研究計画―5.タスク3 トリチウム挙動および中性子照射効果, 大矢恭久, 波多野雄治, 片山一成, 山内有二, 信太祐二, 大塚哲平, 近田拓未, 原正憲, 大宅諒, 上田良夫, 外山健, プラズマ・核融合学会誌, 93, 3, 139‐143,   2017 03 25 , http://jglobal.jst.go.jp/public/201702287697212647
  • Tritium Behavior and Neutron Irradiation Effect, 大矢 恭久, 波多野 雄治, 片山 一成, 山内 有二, 信太 祐二, 大塚 哲平, 近田 拓未, 原 正憲, 大宅 諒, 上田 良夫, 外山 健, プラズマ・核融合学会誌 = Journal of plasma and fusion research, 93, 3, 139, 143,   2017 03 , http://ci.nii.ac.jp/naid/40021166704
  • 21aPS-84 Control of electrical properties by hydrogen loading into metallic nanocontacts at low temperatures, Takata H, Ienaga K, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan, 71, 0,   2016 , http://ci.nii.ac.jp/naid/110010058225
  • Resonant tunneling of hydrogen atoms in Nb nanocontacts, Takata H, Ienaga K, Kajiwara Y, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan, 71, 0, 1493, 1493,   2016 , http://ci.nii.ac.jp/naid/130006244203
    Summary:<p>水素は最も軽い元素であり非常に強い量子性を示す。この強い量子性を反映して、水素吸蔵金属内にある水素原子は、低温ではトンネル効果によって内部を拡散すると考えられている。我々はこれまで、ブレークジャンクション法を用いて作成した金属Pd、Vナノコンタクトへの液体水素中における水素吸蔵現象を、電気伝導特性変化から追跡してきた。これまでの実験からは、熱的拡散が抑制された低温においても水素吸蔵が進行することが明らかになっている。更に今回、Nbナノコンタクトへの水素吸蔵実験を行い、電気伝導特性に金属内水素の移動を反映していると考えられる異常を観測した。本講演ではこれら実験で得られた結果について報告する。</p>
  • 22pAA-11 Experimental study on hydrogen/deuterium absorption into metallic nanocontact at liquid hydrogen temperature, Takata H, Kawae T, Ienaga K, Ueno Y, Islam Md. S, Kawasaki Y, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Meeting Abstracts of the Physical Society of Japan, 70, 0,   2015 , http://ci.nii.ac.jp/naid/110009989991
  • 21pPSB-47 Hydrogen/Deuterium absorption into metallic nanocontact at liquid hydrogen temperature, Kawasaki Y, Kawae T, Takata H, Islam Md. S, Nishimura N, Inagaki Y, Ienaga K, Tsujii H, Hashidume K, Otsuka T, Meeting Abstracts of the Physical Society of Japan, 70, 0,   2015 , http://ci.nii.ac.jp/naid/110009991877
  • 21pPSB-48 Hydrogen impurity effects of superconducting metallic nano-contacts, Ueno Y, Takata H, Islam Md. Saiful, Ienaga K, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan, 70, 0,   2015 , http://ci.nii.ac.jp/naid/110009991878
  • 17pCS-4 Experimental study on hydrogen absorption into metallic V, Nb point-contact at liquid hydrogen temperature, Takata H, Ienaga K, Ueno Y, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan, 70, 0,   2015 , http://ci.nii.ac.jp/naid/110010029183
  • 18aPS-49 Hydrogen absorption into metallic point-contact and electrical conductivity at liquid hydrogen temperature, Ueno Y, Takata H, Ienaga K, Islam Md. S, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan, 70, 0,   2015 , http://ci.nii.ac.jp/naid/110010029924
  • 9pPSA-132 Hydrogen Absorption into Vanadium by Quantum Effect, Takata H, Ienaga K, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan, 69, 0,   2014 , http://ci.nii.ac.jp/naid/110009874284
  • 10aBE-5 Hydrogen Absorption into Vanadium by Quantum Effect, Takata H, Ienaga K, Inagaki Y, Tsujii H, Hashidume K, Otsuka T, Kawae T, Meeting Abstracts of the Physical Society of Japan, 69, 0,   2014 , http://ci.nii.ac.jp/naid/110009875958
  • Tritium and Mass Transfer in First Wall, TOKUNAGA Kazutoshi, MIYAMOTO Mitsutaka, OTSUKA Teppei, KAJITA Shin, OHNO Noriyasu, UEDA Yoshio, プラズマ・核融合学会誌, 89, 11, 709, 713,   2013 11 , http://ci.nii.ac.jp/naid/110009685108
    Summary: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.
  • Irradiation-Tritium Synergism, HATANO Yuji, OYA Yasuhisa, HARA Masanori, ODA Takuji, OTSUKA Teppei, SATO Koichi, ZHANG Kun, プラズマ・核融合学会誌, 89, 11, 725, 730,   2013 11 , http://ci.nii.ac.jp/naid/110009685111
    Summary: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.
  • Tritium retention behavior in helium irradiated tungsten, Otsuka Teppei, Shimada Masashi, Tokunaga Kazutoshi, Hashizume Kenichi, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2013, 0,   2013 , http://ci.nii.ac.jp/naid/130004568657
    Summary: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。
  • Measurement of hydrogen diffusion coefficients in F82H by means of a tritium imaging plate technique, Higaki Makoto, Otsuka Teppei, Tokunaga Kazutoshi, Hasidume Kennichi, Esato Kouichirou, Suzuki Satoshi, Enoeda Mikio, Akiba Masato, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2013, 0,   2013 , http://ci.nii.ac.jp/naid/130004568700
    Summary: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。
  • Property change of oxide layer of Zr-based alloys with oxide growth, SAKAMOTO KAN, UNE KATSUMI, AOMI MASAKI, OTSUKA TEPPEI, HASHIZUME KENICHI, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2013, 0,   2013 , http://ci.nii.ac.jp/naid/130004569159
    Summary: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。
  • Tritium Permeation, Contamination and Decontamination, HATANO Yuji, TORIKAI Yuji, OYA Yasuhisa, ODA Takuji, TANAKA Satoru, NAKAMURA Hirofumi, ASAKURA Yamato, OHUCHI Hiroko, OTSUKA Teppei, KOBAYASHI Kazuhiro, Journal of plasma and fusion research, 85, 10, 726, 735,   2009 10 25 , http://ci.nii.ac.jp/naid/110007468385
  • 8^ International Conference on Tritium Science and Technology, 大塚 哲平, 日本原子力学会誌 = Journal of the Atomic Energy Society of Japan, 50, 3,   2008 03 01 , http://ci.nii.ac.jp/naid/10020188677
  • Wettability of Oxidized Zircaloy Surface gamma-irradiated in Water, OTSUKA Teppei, ISOTANI Takenori, Hashizume Kenichi, TANABE Tetsuo, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2005, 0, 405, 405,   2005 , http://ci.nii.ac.jp/naid/130007026447
    Summary: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。
  • Application of IP technique to determine diffusion coefficients of tritium in Zr, Saruwatari Yuki, Hirano Tomohisa, Otsuka Teppei, Hashizume Kenichi, Tanabe Tetsuo, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2005, 0, 401, 401,   2005 , http://ci.nii.ac.jp/naid/130007025361
    Summary: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。
  • Wettability of Zirconium Alloys' Oxide films, OTSUKA Teppei, NAKAYAMA Kenji, ISOTANI Takenori, HASHIZUME Kenichi, Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2004, 0, 458, 458,   2004 , http://ci.nii.ac.jp/naid/130007028659
    Summary: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。
  • ICONE11-36307 CHARPY IMPACT TEST OF OXIDIZED AND HYDROGENATED ZIRCALOY USING A THIN STRIP SPECIMEN, OTSUKA Teppei, HASHIZUME Kenichi, SUGISAKI Masayasu, The Proceedings of the International Conference on Nuclear Engineering (ICONE), 2003, 0,   2003 , http://ci.nii.ac.jp/naid/110002478801
  • RESEARCH ON THE SLIGHT CLIMATE BY THE LAND USE IN THE PROVINCIAL CITY OF THE INLAND PART : The actual measurement analysis of Kofu, Gifu, OTSUKA Teppei, TAKAGI Naoki, IWAI Kazuhiro, 日本建築学会北陸支部研究報告集, 45, 225, 228,   2002 06 23 , http://ci.nii.ac.jp/naid/110003884296
    Summary:In order to research the actual of the urban climate in Kofu and Gifu City, especially the authors analyzed what the land use influenced to the slight climate. The results wee shown in the following. (1) A difference in climate of every land use was seen by the fixed-point measurement. (2) A difference in climate in the whole of the city could be seen by the movement measurement.
  • Research on the slight climate by the land use in the provincial city of the land part : The actual measurement analysis of Kofu, Gifu(SUMMARIES OF TECHNICAL PAPERS OF ANNUAL MEETING ARCHITECTURAL INSTITUTE OF JAPAN 2002), OTSUKA Teppei, TAKAGI Naoki, IWAI Kazuhiro, Summaries of technical papers of Annual Meeting Architectural Institute of Japan. D-1, Environmental engineering I, Room acoustics and acoustic environment noise and solidborne sound environmental vibration light and color water supply and drainage water, 1, 653, 654,   2002 06 , http://ci.nii.ac.jp/naid/110004560407
  • Diffusion of helium and estimated diffusion coefficients of hydrogen dissolved in water-saturated, compacted Ca-montmorillonite (Feature Articles "Japan-China Workshop on Nuclear Waste Management and Reprocessing"), HIGASHIHARA T, Journal of nuclear fuel cycle and environment, 7, 1, 51, 55,   2001 03 , http://ci.nii.ac.jp/naid/40007139663

Research Grants & Projects

  • Japan Society for the Promotion of Science, Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (B), Study on tritium-production test and design of the Li-loading rod using HTTR for fusion reactors
  • Japan Society for the Promotion of Science, Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (B), Development of Li-loading rod of high-temperature gas-cooled reactor for tritium production, 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.
  • Japan Society for the Promotion of Science, Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (B), Measurement of Tritium Distributions on Divertor Tiles used in JET-ITER Like Wall Experiments, 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.
  • Japan Society for the Promotion of Science, Grants-in-Aid for Scientific Research Grant-in-Aid for Scientific Research (B), Tritium permeation induced by water corrosion of metals, 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.
  • Japan Society for the Promotion of Science, Grants-in-Aid for Scientific Research Grant-in-Aid for Young Scientists (B), Study on hydrogen accumulation mechanisms in metals by means of tritium profiling technique, 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.
  • Japan Society for the Promotion of Science, Grants-in-Aid for Scientific Research Grant-in-Aid for Specially Promoted Research, Elucidation of effect of hydrogen on giga-cycle fatigue mechanism and establishment of improvement method of fatigue strength reliability, 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.
  • Tritium autoradiography
  • Study on mechanical properties of Zircaloy