Deuterium and helium desorption/retention properties of low-activation vanadium alloys possible for reuse in a short time in fusion reactors
Yuji Yamauchi, Yusuke Tanoue, Kazuki Keta, Takuya Nagasaka, Jingjie Shen, Satoshi Tomioka, Yutaka Matsumoto
Nuclear Materials and Energy, 2024年06月
研究論文（学術雑誌）

Deuterium permeation of low-activation vanadium alloys possible for reuse in a short time in fusion reactors
Aoi Yoshizawa, Yuji Yamauchi, Takuya Nagasaka, Satoshi Tomioka, Yutaka Matsumoto, Naoki Higashi
Fusion Engineering and Design, 2024年05月
研究論文（学術雑誌）

Control and Application of Ultrahigh Hydrogen Flux in Materials
Makoto I. KOBAYASHI, Yuji HATANO, Masanori HARA, Yasuhisa OYA, Yuji YAMAUCHI, Teppei OTSUKA, Takuya NAGASAKA
Plasma and Fusion Research, 18, 2105073, 2105073, Japan Society of Plasma Science and Nuclear Fusion Research, 2023年08月17日
研究論文（学術雑誌）

Deuterium retention in tungsten irradiated by high-dose neutrons at high temperature
M. Oya, M. Shimada, C. N. Taylor, M. Kobayashi, Y. Nobuta, Y. Yamauchi, Y. Oya, Y. Ueda, Y. Hatano
NUCLEAR MATERIALS AND ENERGY, 27, ELSEVIER, 2021年06月
英語, 研究論文（学術雑誌）, We investigated deuterium (D) retention in three W samples irradiated with MeV neutrons at high damage level of 0.39 similar to 0.74 displacements per atom (dpa) at high temperatures, 894 K, 1074 K and 1379 K. The W specimens were exposed to high-flux (similar to 1 x 1022 m(-2) s(-1)) and high-fluence (similar to 5 x 10(25) m(-2)) D plasma at 873 K in the Tritium Plasma Experiment. Broad desorption peaks extended from 900 K to 1200 K were observed for the neutron-irradiated W by thermal desorption spectroscopy (TDS). The retention in neutron-irradiated specimens was much larger than for an un-irradiated specimen. The highest D retention was obtained for a specimen irradiated at 894 K. With increasing neutron irradiation temperature, the retention was reduced about by half at 1074 K and further increase of the temperature (1379 K) resulted in comparable retention. In addition, one-dimensional diffusion calculations (D desorption in TDS and D depth distribution in plasma exposure) were performed to derive retention parameters (the detrapping energy, the depth occupied by D atoms and D/W ratio) from experimental D retention properties of neutron-irradiated W. By TDS simulation calculation, simple dependences of the peak temperature, height and width of TDS peaks on the retention parameters were obtained with total retention in the orders of 10(19) similar to 10(22) m(-2). The calculation of the depth distribution of trapped D atoms made a relationship between the D/W ratio and the depth occupied by D atoms after plasma exposure at relevant conditions. By comparing the relationship (the D/W and the depth) with that obtained from the experimental results, we estimate each retention parameters for the specimens irradiated by high-dose neutrons at the high temperatures. And, we discussed the neutron-irradiation temperature dependence of D retentions.

D retention and depth profile behavior for single crystal tungsten with high temperature neutron irradiation
Y. Oya, F. Sun, Y. Yamauchi, Y. Nobuta, M. Shimada, C. N. Taylor, W. R. Wampler, M. Nakata, L. M. Garrison, Y. Hatano
JOURNAL OF NUCLEAR MATERIALS, 539, ELSEVIER, 2020年10月
英語, 研究論文（学術雑誌）, Single crystalline W (tungsten) samples irradiated at 633, 963 and 1073 K by neutrons to a damage level of 0.1 dpa were exposed to a high-flux D (deuterium) plasma at 673, 873 and 973 K, respectively, in TPE (Tritium Plasma Experiment) at INL (Idaho National Laboratory). Deuterium desorption was analyzed by TDS (Thermal Desorption Spectroscopy), and D depth profiles were determined by NRA (Nuclear Reaction Analysis) at SNL (Sandia National Laboratories). HIDT (Hydrogen Isotope Diffusion and Trapping) simulation code was applied to evaluate D behavior for neutron-damaged W at higher temperature.The D retention at depths up to 3 mu m for the neutron-damaged sample at 673 K was two orders of magnitude larger than that for undamaged tungsten, and its D desorption spectrum had a single broad stage at around 900 K. As the neutron irradiation/plasma exposure temperature increased, D retention was largely reduced, and the desorption temperature was shifted to higher temperatures above 1100 K. The D depth profiles by NRA also showed D migration toward bulk by higher temperature irradiation, compared to undamaged W.The HIDT simulation indicated that the major binding energy of D was changed from 1.43 eV to 2.07 eV at higher neutron irradiation and plasma exposure temperatures, suggesting that some vacancies and small vacancy clusters would aggregate to form larger voids, or depopulation of weak traps at high D plasma exposure temperatures. It can be said that more stable trapping sites played dominant roles in the D retention at higher neutron irradiation and plasma exposure temperature. The binding energy by HIDT simulation was almost consistent with the reported value by TMAP, but the consideration of not only total D retention measured by TDS but also D depth profile by NRA led to the more accurate D behavior in neutron-damaged W. (C) 2020 Elsevier B.V. All rights reserved.

Deuterium retention in neutron-irradiated single-crystal tungsten
M. Shimada, Y. Oya, W. R. Wampler, Y. Yamauchi, C. N. Taylor, L. M. Garrison, D. A. Buchenauer, Y. Hatano
Fusion Engineering and Design, 136, 1161, 1167, Elsevier Ltd, 2018年11月, ［査読有り］
英語, 研究論文（学術雑誌）, © 2018 Elsevier B.V. Six single crystal tungsten specimens were neutron irradiated to a dose of 0.1 displacements per atom (dpa) at three different irradiation temperatures (633 K, 963 K, and 1073 K) at the High Flux Isotope Reactor in Oak Ridge National Laboratory under the US-Japan PHENIX project. A pair of neutron-irradiated tungsten specimens was exposed to deuterium (D) plasma to D ion fluence of 5.0 × 1025 m−2 at three different exposure temperatures (673 K, 873 K, and 973 K) at the Tritium Plasma Experiment in Idaho National Laboratory. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (Tritium Migration Analysis Program, TMAP) were used to understand D behavior in neutron-irradiated tungsten. A broad D desorption spectrum from the plasma-exposure temperature up to 1173 K was observed. Total D retention up to 1.9 × 1021 m−2 and near-surface D concentrations up to 1.7 × 10−3 D/W were experimentally measured from the 0.1 dpa neutron-irradiated single crystal tungsten. Trap density up to 2.0 × 10−3 Trap/W and detrapping energy ranging from 1.80 to 2.60 eV were obtained from the TMAP modeling.

Effects of modified surfaces produced at plasma-facing surface on hydrogen release behavior in the LHD
Y. Nobuta, S. Masuzaki, M. Tokitani, N. Ashikawa, D. Nagata, N. Yoshida, Y. Oya, M. Yajima, G. Motojima, H. Kasahara, M. Miyamoto, N. Ohno, Y. Yamauchi
Nuclear Materials and Energy, 12, 483, 487, Elsevier Ltd, 2017年08月, ［査読有り］
英語, 研究論文（学術雑誌）, © 2017 The Authors In the present study, an additional deuterium (D) ion irradiation was performed against long-term samples mounted on the helical coil can and in the outer private region in the LHD during the 17th experimental campaign. Based on the release behavior of the D and hydrogen (H) retained during the experimental campaign, the difference of release behavior at the top surface and in bulk of modified surfaces is discussed. Almost all samples on the helical coil can were erosion-dominant and some samples were covered with boron or carbon, while a very thick carbon films were formed in the outer private region. In the erosion-dominant area, the D desorbed at much lower temperatures compared to that of H retained during the LHD plasma operation. For the samples covered with boron, the D tended to desorb at lower temperatures compared to H. For the carbon deposition samples, the D desorbed at much higher temperatures compared to no deposition and boron-covered samples, which was very similar to that of H. The D retention capabilities at the top surface of carbon and boron films were 2–3 times higher than no deposition area. The results indicate that the retention and release behavior at the top surface of the modified layer can be different from that of bulk substrate material.

Deuterium retention in helium and neutron irradiated molybdenum
C. N. Taylor, Y. Yamauchi, M. Shimada, Y. Oya, Y. Hatano
Fusion Science and Technology, 71, 4, 491, 495, AMER NUCLEAR SOC, 2017年05月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, © American Nuclear Society. Understanding and managing D retention in plasma facing components is essential for tritium safety in fusion reactors. Neutron irradiated and virgin low carbon arc cast (LCAC) Mo, as well as Mo foil samples with and without He pre-irradiation, were used to investigate D retention. D and He retention were investigated simultaneously in thermal desorption spectroscopy using a high resolution residual gas analyzer. Results show a significant increase in D retention with He pre-irradiation. Vacancies and vacancy clusters are found to retain D in LCAC samples, but neutron irradiated Mo retains more D in vacancy clusters.

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年03月25日
日本語

まとめと今後の研究計画　　　　　　　　　　　　　　　
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
J. Plasma Fusion Res., 2017年03月
日本語

トリチウム挙動および中性子照射効果 (プロジェクトレビュー 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)
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, プラズマ・核融合学会編集委員会, 2017年03月
日本語

Effect of impurity deposition layer formation on D retention in LHD plasma exposed W
Y. Oya, H. Fujita, C. Hu, Y. Uemura, S. Sakurada, K. Yuyama, X. Li, Y. Hatano, N. Yoshida, H. Watanabe, Y. Nobuta, Y. Yamauchi, M. Tokitani, S. Masuzaki, T. Chikada
Nuclear Materials and Energy, 9, 84, 88, Elsevier Ltd, 2016年12月01日, ［査読有り］
英語, 研究論文（学術雑誌）, © 2016 The Authors Effect of carbon based mixed-material deposition layer formation on hydrogen isotope retention was studied. The tungsten (W) samples were placed at four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (higher heat load area), and ER (erosion dominated area) during 2013 plasma experimental campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼ 4000 shots of hydrogen plasma in a 2013 plasma experimental campaign. Most of the sample surface except for ER was covered by a mixed-material deposition layer formed by plasma experimental campaign, which consisted of carbon, but some metal impurities were contained. For ER sample, He bubbles were formed due to long term He discharge cleaning and He plasma experiments during the plasma experimental campaign. The additional 1 keV D2+ implantation was performed to evaluated the D retention enhancement by plasma exposure. It was found that both of H and D retentions were clearly increased. In particular, the H retention was controlled by the thickness of the carbon-dominated mixed-material deposition layer, indicating most of the H was trapped by this mixed-material deposition layer. It is concluded that the accumulation of low-Z mixed-material layer on the surface of the first wall is one of key issues for the determination of hydrogen isotope retention in first wall.

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, 109-111, 104, 108, ELSEVIER SCIENCE SA, 2016年11月01日, ［査読有り］
英語, 研究論文（学術雑誌）, © 2016 Elsevier B.V. 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 °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 °C and 0.1–1.0 atm D2 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 (< 10×) at lower temperatures, but the permeability converges to that of the ITER tungsten at 1000 °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.

Tritium desorption and tritium removal from tungsten pre-irradiated with helium
Yuji Nobuta, Yuji Hatano, Yuji Torikai, Masao Matsuyama, Shinsuke Abe, Yuji Yamauchi
Fusion Engineering and Design, 109-111, 1179, 1182, ELSEVIER SCIENCE SA, 2016年, ［査読有り］
英語, 研究論文（学術雑誌）, © 2015 Elsevier B.V. In this study, 1 keV DT+ ion irradiation was performed on tungsten pre-irradiated with helium. The thermal desorption behavior and the reduction of tritium retention during vacuum preservation at room temperature, as well as isochronal annealing were investigated using an IP technique, taking advantage of the fact that tritium detection is nondestructive and is highly sensitive. At a pre-irradiated helium fluence of 1 × 1017 He/cm2, retained tritium tended to be desorbed at higher temperatures when compared to no helium irradiation case. Tritium retention during preservation in vacuum and during isochronal annealing became smaller with increasing helium fluence up to 1 × 1017 He/cm2. At a helium fluence of 1 × 1018 He/cm2, the reduction of tritium retention was found to be greater compared to 1 × 1017 He/cm2. The results indicate that helium irradiation has a significant influence not only on the thermal desorption temperature of tritium but on longtime tritium reduction at room and elevated temperatures.

Deuterium retention on He-irradiated F82H with surface impurity layer
N. Shinoda, Y. Yamauchi, Y. Nobuta
Journal of Nuclear Materials, 463, 1001, 1004, ELSEVIER SCIENCE BV, 2015年07月22日, ［査読有り］
英語, 研究論文（学術雑誌）, © 2015 Elsevier B.V. Abstract The deuterium retention/desorption behavior of F82H has been studied to clarify the effect of radiation damage with respect to the impurity layer. To evaluate the effects of oxygen impurities, non-oxidized and thermally oxidized samples were prepared with the introduction of radiation damage through He⁺ irradiation. The amount of deuterium retained in the thermally oxidized samples was less than that of the non-oxidized samples. The amount of deuterium retained in the oxidized sample increased after He⁺ irradiation. The retention/desorption behaviors were discussed in the surface chemical state.

Helium irradiation effects on tritium retention and long-term tritium release properties in polycrystalline tungsten
Y. Nobuta, Y. Hatano, M. Matsuyama, S. Abe, Y. Yamauchi, T. Hino
Journal of Nuclear Materials, 463, 993, 996, ELSEVIER SCIENCE BV, 2015年07月22日, ［査読有り］
英語, 研究論文（学術雑誌）, © 2014 Elsevier B.V. Abstract DT⁺ ion irradiation with energy of 0.5 and 1.0 keV was performed on helium pre-irradiated tungsten and the amount of retained tritium and the long-term release of retained tritium in vacuum was investigated using an IP technique and BIXS. Tritium retention and long-term tritium release were significantly influenced by helium pre-irradiation. The amount of retained tritium increased until it reached 1 × 10¹⁷ He/cm², and at 1 × 10¹⁸ He/cm² it became smaller compared to 1 × 10¹⁷ He/cm². The amount of retained tritium in tungsten without helium pre-irradiation largely decreased after several weeks preservation in vacuum, and the long-term release rate during vacuum preservation was retarded by helium pre-irradiation. The results indicate that the long-term tritium release and the helium irradiation effect on it should be taken into account for more precise estimation of tritium retention in the long-term use of tungsten in fusion devices.

Deuterium retention properties of SiC/SiC composites as plasma facing materials for fusion reactors after deuterium irradiation at elevated temperatures
Yuji Nobuta, Tomoaki Hino, Yuji Yamauchi, Takashi Nozawa
Journal of the Vacuum Society of Japan, 58, 5, 173, 176, Vacuum Society of Japan, 2015年, ［査読有り］
英語, 研究論文（学術雑誌）, © 2015 Journal of the Vacuum Society of Japan. In this study, deuterium retention properties of SiC/SiC composites irradiated by deuterium ions at different temperatures in the range from 293 K to 923 K were investigated with thermal desorption spectroscopy, and the relationship between deuterium retention and the change of atomic composition was examined. Thermal desorption spectra of D2 showed two major peaks which was regarded as deuterium released from its Si-D and C-D bonds, respectively. At the irradiation temperature of 673 K, the ratio of the amounts of desorbed D2 at lower desorption peak (Si-D bond) increased compared to the room temperature case, owing to the decrease in carbon composition at the surface. On the other hand, at the irradiation temperatures at 823 and 923 K, the ratio of the amounts of desorbed D2 at higher desorption peak (C-D bond) increased, and the hydrocarbon desorption also increased. The present data suggests that the fuel hydrogen retention properties of SiC/SiC composites are significantly affected by the surface temperature.

Comparison of hydrogen isotope retention for tungsten probes in LHD vacuum vessel during the experimental campaigns in 2011 and 2012
Y. Oya, S. Masuzaki, M. Tokitani, M. Sato, K. Toda, H. Uchimura, N. Yoshida, H. Watanabe, Y. Yamauchi, T. Hino, M. Miyamoto, Y. Hatano, K. Okuno
Fusion Engineering and Design, 89, 7-8, 1091, 1095, ELSEVIER SCIENCE SA, 2014年10月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, To evaluate hydrogen isotope retention enhancement in W by plasma exposure, the stress relieved tungsten samples were placed at three or four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (Higher heat load area) and ER (erosion dominated area) during 2011 (15th) or 2012 (16th) plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼6700 shots of hydrogen plasma in a 2011 plasma experiment campaign and ∼5000 shots in a 2012 plasma campaign. Thereafter, additional 1.0 keV deuterium ion implantation was performed to evaluate the change of hydrogen isotope retention capacity by plasma exposure. It was found that more than 50 times of hydrogen retention enhancement for DP sample was derived compared to that for pure W. In especially, the carbon-dominated mixed-material layer would control the hydrogen isotope retention for all the area except for erosion-dominated area, indicating that the chemical structure for carbon-dominant mixed-material layer would govern the H and D retention enhancement for most area by long-term plasma exposure. Therefore, the surface area for carbon material would be one of key issues for the determination of hydrogen isotope retention in first wall, even if all tungsten first walls will be used. © 2013 Elsevier B.V.

Effect of induced damage on hydrogen isotope retention of F82H with impurity layer
N. Shinoda, Y. Yamauchi, Y. Nobuta, T. Hino
Fusion Engineering and Design, 89, 7-8, 921, 924, ELSEVIER SCIENCE SA, 2014年10月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, The effects of induced damage on hydrogen isotope retention in F82H with or without thermal oxidation were investigated using thermal desorption spectroscopy. To induce damage and modify the surface, glow discharge pre-irradiated Ar+ ions was examined. In non-oxidized samples, the amount of desorbed deuterium increased with Ar+ ion fluence. Oxygen depletion in the surface layer of non-oxidized samples from the Ar+ ion irradiation, which resulted in bulk diffusion of deuterium, is responsible for the increase in deuterium retention. A comparison between non-oxidized and oxidized samples clearly indicated that the surface oxide layer greatly influenced deuterium retention/desorption behaviors of F82H. © 2014 Elsevier B.V.

Tritium retention properties of tungsten, graphite and co-deposited carbon film
Y. Nobuta, Y. Hatano, M. Matsuyama, S. Abe, S. Akamaru, Y. Yamauchi, T. Hino, S. Suzuki, M. Akiba
Fusion Engineering and Design, 89, 7-8, 1516, 1519, ELSEVIER SCIENCE SA, 2014年10月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, DT+ ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both the amount of retained tritium and the reduction of retained tritium after preservation in vacuum were investigated using an IP technique and BIXS. In addition, the relationship between the retention properties of tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. The amount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1 keV of DT+ irradiation, graphite showed no reduction of the amount of retained tritium after six months preservation while that of carbon film decreased by approximately 20% after 40 days preservation. It was suggested that this difference might be associated with differences in the microstructure between graphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately half after 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decrease in tritium retention during preservation became slower. Tungsten's properties of releasing tritium while preserved in vacuum would be a useful tool for the reduction/removal of retained tritium. © 2014 Elsevier B.V.

Influence of surface condition on deuterium release from Li 2TiO3 pebble
Akihito Tsuchiya, Yuji Yamauchi, Yuji Nobuta, Tomoaki Hino, Masato Akiba, Mikio Enoeda
Fusion Engineering and Design, 89, 7-8, 1280, 1283, ELSEVIER SCIENCE SA, 2014年10月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, Lithium titanate (Li2TiO3) pebbles were irradiated with D3+ ions with energy of 5.0 keV, and the amounts of retained deuterium in the pebbles were measured by thermal desorption spectroscopy. In this research the irradiation/heating cycles were carried out repeatedly in order to investigate the influence of surface condition on deuterium release from Li2TiO3. The composition ratio of Li decreased with the increase of the number of the irradiation/heating cycle. Then, the desorption peaks of the gases contained deuterium atoms were shifted to higher temperature region, and the amount of desorbed gases in forms of water tended to increase. In addition, we carried out other experiments for the comparison. Comparing these results, we considered that the increase of the defects created by the irradiation was more responsible for the change in the desorption behavior by the irradiation/heating cycles than the lithium depletion. These results suggest that the tritium recovery efficiency would decrease with the increase of the defects and the damages especially at the low temperature region during the operation. © 2014 Elsevier B.V.

LHDプラズマに曝されたタングステンにおける水素同位体滞留能変化
大矢 恭久, 増崎 貴, 時谷 政行, 渡辺 英雄, 吉田 直亮, 波多野 雄治, 宮本 光貴, 山内 有二, 日野 友明, 奥野 健二
プラズマ・核融合学会誌, 90, 6, 319, 324, プラズマ・核融合学会, 2014年06月25日
日本語, 本レビューではプラズマ対向材料として注目されているタングステンの試料を,大型ヘリカル装置(LHD)の15サイクル実験(2011年)および16サイクル実験(2012年)期間中にLHD真空容器内に設置してプラズマ環境に曝露した際の水素同位体滞留能変化を調べた.これにより磁場閉じ込め装置の実機環境下での水素同位体滞留の基礎過程を理解するとともに水素同位体滞留能の変化がLHDにおける水素同位体挙動にどの程度影響を与えるのかを評価した.タングステン試料を約5000～6000ショット程度の水素プラズマ,300～800ショット程度のヘリウムプラズマおよび壁調整のためのグロー放電洗浄に曝した後に真空容器から取り出して観察したところ,表面には炭素堆積層が形成されており,その厚さは,堆積が多い場所の試料(DP試料)で4μm程度,熱負荷の高い場所の試料(HL試料)では100nm程度であった.これらの試料に重水素イオンを照射した後,昇温脱離ガス分析を行ったところ,重水素放出特性は純タングステンとは異なり,主要な放出ピークは800K-900Kに見られた.特にDP試料では900Kに大きな脱離ピークが見られた.純タングステンでは400K-600Kに脱離ピークが存在することから,LHDでプラズマに曝されたタングステン試料の重水素捕捉は純タングステンでの重水素捕捉とは状態が異なることが示唆された.LHDプラズマに134ショットだけ曝した試料の表面を観察すると,堆積層と表面近傍に転位ループが高密度に集積していることから,曝露初期には照射損傷導入と堆積層形成がダイナミックに進行し,これらに重水素を捕捉する能力が高いことが示唆された.

Improvement on infrared imaging video bolometer by laser calibration method
Muneji Itomi, Byron J. Peterson, Yuji Yamauchi, Ryuichi Sano, Kiyofumi Mukai, Shwetang N. Pandya
Plasma and Fusion Research, 9, SpecialIssue2, Japan Society of Plasma Science and Nuclear Fusion Research, 2014年, ［査読有り］
英語, 研究論文（学術雑誌）, � 2014 The Japan Society of Plasma Science and Nuclear Fusion Research. It is necessary to estimate and decrease energy losses while keeping a high performance plasma in a nuclear fusion reactor. Therefore radiation diagnostics are required to estimate the radiation loss as a part of the total energy loss. Our diagnostic, the InfraRed imaging Video Bolometer (IRVB), which has been used in Large Helical Device (LHD), JT-60U and KSTAR, measures the radiation loss from the plasma. First, a metal foil, which is apart of the diagnostic, absorbs radiation energy through an aperture and its temperature risesdue to this energy. An IR camera measures the 2D distribution of the temperature on the foil, then we obtain the irradiance intensity and distribution by solving the heat diffusion equation on the foil. The purpose of our study is the development of the foil calibration analysis with laser for measuring plasma radiation with an IRVB using a Finite Element Model (FEM). This paper shows the improvement from two viewpoints. The first one is from the spatial resolution, and second one isthe addition of the time dependent part in the heat diffusion equation with the IRVB calibration of the foil thermal diffusivity.

Tritium retention on the surface of stainless steel samples fixed on the plasma-facingwall in LHD
Masao Matsuyama, Shinsuke Abe, Kiyohiko Nishimura, Naoko Ashikawa, Yasuhisa Oya, Kenji Okuno, Yuji Yamauchi, Yuji Nobuta, Akio Sagara
Plasma and Fusion Research, 9, SpecialIssue2, Japan Society of Plasma Science and Nuclear Fusion Research, 2014年, ［査読有り］
英語, 研究論文（学術雑誌）, © 2014 The Japan Society of Plasma Science and Nuclear Fusion Research. Effects of pre-heating for retention and distribution of tritium have been studied using samples fixed on the wall of the Large Helical Device during a plasma campaign. The samples were fixed at four different locations. The plasma-facing surface of the samples was covered with deposition layers of different thickness in each sample. Retention behavior in deposition layers was observed using β-ray-induced X-ray spectrometry and imaging plate technique. Pre-heating of the samples in vacuum was changed in a temperature range from 300 to 623 K, and subsequent tritium exposure was carried out at 300K in every runs. Non-uniformity of tritium distribution clearly appeared even in the as-received samples which was not pre-heated. It is considered, therefore, that non-uniform adsorption sites of tritium have been produced during a formation process of deposition layers. In addition, it was seen that the amount of tritium retention increased with an increase in the pre-heating temperature, indicating that adsorption sites of tritium were newly formed in the deposition layers by heating in vacuum.

Enhancement of hydrogen isotope retention capacity for the impurity deposited tungsten by long-term plasma exposure in LHD
Yasuhisa Oya, Suguru Masuzaki, Masayuki Tokitani, Naoaki Yoshida, Hideo Watanabe, Yuji Yamauchi, Tomoaki Hino, Mitsutaka Miyamoto, Yuji Hatano, Kenji Okuno
Fusion Engineering and Design, 88, 9-10, 1699, 1703, ELSEVIER SCIENCE SA, 2013年10月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, The stress relieved tungsten samples were placed at three positions, PI (sputtering erosion dominated area), DP (deposition dominated area) and HL (Higher heat load area) during 15th plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼ 6700 shots of hydrogen plasma in a 15th long-term experiment campaign in LHD. Thereafter, the additional deuterium ion implantation to these tungsten samples was performed to evaluate the change of hydrogen isotope retention capacity in the samples by long-term plasma exposure. It was found that the carbon-dominant mixed-material layer with more than 100 nm thickness was formed on a wide area of the tungsten surface. The thicker mixed-material layer was formed on the DP sample, where the deuterium retention was about 21 times as high as that for pure W. The major desorption temperature of deuterium was shifted toward higher temperature side, which was comparable to the trapping characteristic of carbon or irradiation damages. © 2013 Elsevier B.V.

Effects of hydrogen mixture into helium gas on deuterium removal from lithium titanate
Akihito Tsuchiya, Tomoaki Hino, Yuji Yamauchi, Yuji Nobuta, Masato Akiba, Mikio Enoeda
Fusion Engineering and Design, 88, 9-10, 2298, 2301, ELSEVIER SCIENCE SA, 2013年10月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, Lithium titanate (Li2TiO3) pebbles were irradiated with deuterium ions with energy of 1.7 keV and then exposed to helium or helium-hydrogen mixed gas at various temperatures, in order to evaluate the effects of gas exposure on deuterium removal from the pebbles. The amounts of residual deuterium in the pebbles were measured by thermal desorption spectroscopy. The mixing of hydrogen gas into helium gas enhanced the removal amount of deuterium. In other words, the amount of residual deuterium after the helium-hydrogen mixed gas exposure at lower temperature was lower than that after the helium gas exposure. In addition, we also evaluated the pebbles exposed to the helium gas with different hydrogen mixture ratio from 0% to 1%, at 573 K. Although the amount of residual deuterium in the pebbles after the exposure decreased with increasing the hydrogen mixture ratio, the implanted deuterium partly remained after the exposure. These results suggest that the tritium inventory may occur at low temperature region in the blanket during the operation. © 2013 Elsevier B.V.

Deuterium Retention Properties of Co-Deposited Carbon Films Produced at Wall Gaps
NOBUTA Yuji, KANAZAWA Jun, YAMAUCHI Yuji, HINO Tomoaki, YOKOYAMA Kenji, SUZUKI Satoshi, EZATO Koichiro, ENOEDA Mikio, AKIBA Masato, AKAMARU Satoshi, HATANO Yuji
Plasma and Fusion Research, 8, 2405174, 2405174, The Japan Society of Plasma Science and Nuclear Fusion Research, 2013年
英語, Deuterium retention properties in co-deposited carbon film produced in gap and the relationship between this retention behavior and the crystal structure of carbon film were investigated. In the case of a wide gap, the atomic ratio of deuterium to carbon (D/C) in the film was almost constant at any depth in the gap, while in the case of a narrow gap the D/C ratio decreased with increasing distance from the gap entrance. The micro structure of carbon film tended to be more amorphous for the film produced at locations deeper in the gap. Thermal desorption spectra of D₂ in the film produced near the gap entrance showed one broad main peak at around 1100 K, while that in the film produced near the bottom showed very sharp peaks at around 950 K. This difference in desorption behavior was related with the differences of micro structure.

Evaluation of Plasma-Wall Interactions Using Material Probes in the Large Helical Device
SHIGEMURA Takuya, SHINODA Naoyuki, YAMAUCHI Yuji, NOBUTA Yuji, HINO Tomoaki, ASHIKAWA Naoko, SAGARA Akio, NISHIMURA Kiyohiko, LHD Experimental Group
Plasma and Fusion Research, 8, 2405162, 2405162, The Japan Society of Plasma Science and Nuclear Fusion Research, 2013年
英語, Control of plasma-wall interactions (PWIs) and understanding their mechanism are essential issues for realization of fusion reactor. In this study, plasma-wall interactions during FY2010 and FY2011 experiments in the Large Helical Device (LHD) were analyzed by using material probes. We installed material probes at the positions expected to have different deposits of materials such as boron and titanium and investigated the correlation between this impurity deposition and gas retention. The desorption behavior of hydrogen and helium depended on the level of impurity deposition. We observed a new peak for hydrogen desorption at the probe with large boron deposits. The amount of retained hydrogen was large in that probe; however the amount of retained helium was small in the same probe, which is opposite what we observe in the probe with titanium deposits and the probe with a thin deposition layer.

Deuterium Retention in Tungsten-Coated Reduced Activation Ferritic/Martensitic Steel
ARMANDO Marco, YAMAUCHI Yuji, TSUCHIYA Akihito, NOBUTA Yuji, HINO Tomoaki
Plasma and Fusion Research, 8, 2405103, 2405103, The Japan Society of Plasma Science and Nuclear Fusion Research, 2013年
英語, In order to evaluate tungsten (W) coating performance, thin W film was deposited onto reduced activation ferritic/martensitic steel (F82H) by using rf magnetron sputtering device, then was irradiated at room temperature by 1.7keV D⁺ ions. After the irradiation, the deuterium retention of the coated sample (W/F82H) was evaluated by thermal desorption spectroscopy (TDS) and compared with those of the F82H and pure W. During TDS measurements, D₂ was desorbed in all materials but HD was only found in significant amount in W/F82H and related to the increase of effective surface area due to the porous film structure. D retention increased with increasing fluence and was accompanied by growth of circular blisters. The size of blisters was limited by the F82H substrate structure.

Structure of C deposition layers under various deposition conditions
Y. Hamaji, T. Wada, H. T. Lee, Y. Ohtsuka, Y. Ueda, A. Kreter, V. Philipps, Y. Yamauchi, N. Ashikawa
Fusion Science and Technology, 63, 1T, 371, 373, 2013年, ［査読有り］
研究論文（学術雑誌）, © 2013, Taylor and Francis Inc. All rights reserved. The structure of deposited carbon layers formed under various conditions ranging from small scale laboratory to large scale magnetic confinement devices was characterized using Raman spectroscopy. By comparing ion beam and TEXTOR experiments, the deposition temperature is found to be the dominant factor in three dimensional disorder of sp2 sites and sp3 ratio at T>493 K. While, no clear temperature dependence on properties of aromatic rings was observed, indicating other parameters responsible for the aromatic rings present in carbon deposition layers. The carbon layers from JET differed significantly, indicating influence of Be compound formation on Raman parameters.

Deterioration of lithium ion battery performance due to film formation on graphite surfaces
Shota Takamaru, Tomoaki Hino, Yuji Yamauchi, Yuji Nobuta
Journal of the Vacuum Society of Japan, 56, 7, 277, 279, 2013年, ［査読有り］
英語, 研究論文（学術雑誌）, It is known that the performance of lithium ion batteries deteriorates as number of cycles increases. In order to investigate the reason for this the deterioration, the surface coverage offilm formation was measured for graphite electrodes with different stages of number of cycles. Simultaneously, the gas desorption amount was measured. As the number of cycles increased, so did the surface coverage. The gas desorption amount was roughly proportional to the surface coverage. A polymer-coating changes the characteristics of thefilm, battery performance and gas desorption. These results suggest that deterioration in the battery performance is owing tofilm formation because thefilm blocks lithium ion flow at the graphite surface. The expansion of the battery case is owing to gas desorption from the surfacefilm. In this study, the reasons for battery performance are clarified.

Hydrogen concentration of co-deposited carbon film produced in the vicinity of LHD divertor tiles
Y. Nobuta, K. Fukayama, N. Ashikawa, Y. Yamauchi, T. Hino, A. Sagara
Journal of Nuclear Materials, 438, SUPPL, S1040, S1043, ELSEVIER SCIENCE BV, 2013年, ［査読有り］
英語, 研究論文（学術雑誌）, In order to investigate the hydrogen retention and desorption behavior in co-deposited carbon film, a unique sample holder was mounted on the plasma-facing wall close to the graphite divertor tiles in the LHD during the 13th experimental campaign. The amount of deposited carbon for the sample with a large visual angle against the divertor tile was much larger than that with a small visual angle. The amount of retained hydrogen increased with the film thickness, but not proportional to the thickness. The carbon films contained boron caused by boronization and the retained hydrogen was trapped mainly by C-H and B-H bonds. Hydrogen concentration for the carbon film with a large visual angle against the graphite tile was much smaller than that with a small visual angle, possibly owing to the difference of the ratio of hydrogen influx to carbon influx. © 2013 Elsevier B.V. All rights reserved.

Effects of glow discharge cleanings on hydrogen isotope removal for plasma facing materials
Y. Yamauchi, A. Matsumoto, Y. Kosaka, Y. Kimura, K. Takeda, T. Hino, Y. Nobuta, K. Nishimura, Y. Ueda
Journal of Nuclear Materials, 438, SUPPL, S1146, S1149, ELSEVIER SCIENCE BV, 2013年, ［査読有り］
英語, 研究論文（学術雑誌）, The effect of the discharge cleanings on hydrogen isotope removal have been evaluated for graphite, stainless steel, tungsten, boron and titanium. For all materials, the helium glow discharge cleaning was the most effective on the hydrogen isotope removal among the inert gas discharges. High efficiency of energy transfer to target atom and deep projected range for helium ion might be responsible for the high removal fraction. The effect of argon glow discharge cleaning was small. The small removal fraction for the argon might be owing to re-deposition layer, which acted as a screening to the removal. The hydrogen isotope in the tungsten was hardly removed by the inert gas discharge cleanings. The small removal fraction for the tungsten might be owing to hydrogen isotope retention in deeper regions resulting from diffusion along with the grain boundary or the porous structure. Surface impurity and morphologies significantly influenced the deuterium removal effects. © 2013 Elsevier B.V. All rights reserved.

Enhancement of hydrogen isotope retention in tungsten exposed to LHD plasmas
Yasuhisa Oya, Suguru Masuzaki, Tetsuo Fujishima, Masayuki Tokitani, Naoaki Yoshida, Hideo Watanabe, Yuji Yamauchi, Tomoaki Hino, Mitsutaka Miyamoto, Yuji Hatano, Kenji Okuno
Journal of Nuclear Materials, 438, SUPPL, S1055, S1058, ELSEVIER, 2013年, ［査読有り］
英語, 研究論文（学術雑誌）, Tungsten (W) samples, undamaged and damaged by irradiation with 2.4 MeV Cu, were placed flush to the first wall of Large Helical Device (LHD) nearby the divertor target, and were exposed to 134 shots of hydrogen plasma discharges. Thereafter, to evaluate the enhancement of hydrogen isotope retention for W due to LHD hydrogen (H) plasma exposures, the W samples with and without LHD H plasma exposures were implanted with 1.0 keV D2+. For W samples without LHD H plasma exposures, the D retention in the damaged W was by a factor of ∼6 higher than that in the undamaged W. After LHD H plasma exposures, the carbon layers ∼4 nm in thickness were deposited on the W surfaces. Due to trapping of D atoms in the carbon layers, the D retention enhancement for undamaged and damaged W samples exposed to LHD H plasmas was significantly higher than that for the W samples without LHD H plasma exposures. © 2013 Elsevier B.V. All rights reserved.

Tritium absorption of co-deposited carbon films
Yuji Nobuta, Yuji Yamauchi, Tomoaki Hino, Satoshi Akamaru, Yuji Hatano, Masao Matsuyama, Satoshi Suzuki, Masato Akiba
Fusion Engineering and Design, 87, 7-8, 1070, 1073, ELSEVIER SCIENCE SA, 2012年08月, ［査読有り］
英語, 研究論文（学術雑誌）, Co-deposited carbon film with different deuterium concentration, D/C, were exposed to tritium gas at the temperature of 423 K, and then the atomic ratio of absorbed tritium to carbon, T/C, was evaluated. The obtained data were discussed with crystal structure of the carbon film. The T/C increased with decreasing D/C of carbon film. The carbon film with low D/C had more defective structure. The reduction of D/C by the heating before tritium exposure led to the increase of absorption amount. These results suggest that carbon film with more defective structure and low D/C film could absorb large amount of tritium. The hydrogen isotope concentration in the present experiment was saturated below the orders of 10 -4, which was 3-4 orders of magnitude smaller than that of co-deposited carbon film with hydrogen isotope. © 2012 Elsevier B.V.

Effects of Helium Ion Irradiation on Deuterium Retention Behavior of SiC/SiC Composites as Plasma Facing Materials for Fusion Reactors (小特集 第52回真空に関する連合講演会プロシーディングス(2))
NOBUTA Yuji, YAMAUCHI Yuji, HINO Tomoaki, YOON Han-Ki
真空, 55, 4, 164, 166, 日本真空協会, 2012年
英語, 研究論文（国際会議プロシーディングス）, SiC/SiC composites are promising materials for the plasma-facing walls in a fusion reactor. In order to investigate the effect of helium ion irradiation on the deuterium retention behavior of SiC/SiC composites, helium ion irradiation followed by deuterium ion irradiation was performed on SiC/SiC composites using an ECR ion source. After the irradiation experiments, deuterium retention behavior was measured using thermal desorption spectroscopy technique. The total amount of retained deuterium was enhanced with helium pre-irradiation. Thermal desorption spectra of D 2 showed two main peaks which could be explained as deuterium released from its Si-D and C-D bonds, respectively. These two peaks temperatures were shifted to low temperatures with helium pre-irradiation. The atomic composition of silicon at the sample surface increased after helium irradiation. Due to this silicon composition increase, the amount of deuterium trapped due to the Si-D bond increased greatly. This data indicates a large effect of helium irradiation on the deuterium retention behavior of SiC/SiC composites. Copyright © 2002 - 2012 Hogrefe Publishing.

Removal of deuterium retained in B, Ti, and TiO2by neon glow discharge
Y. Yamauchi, Y. Kosaka, Y. Nobuta, T. Hino, K. Nishimura
Fusion Science and Technology, 62, 1, 66, 70, American Nuclear Society, 2012年
英語, 研究論文（国際会議プロシーディングス）, The removal of deuterium retained in boron, titanium, and titanium oxide films by neon glow discharge was investigated. The films were exposed to deuterium glow plasma to retain the deuterium and subsequently exposed to neon glow plasma. The temperature of the exposures was room temperature. The residual deuterium was estimated by thermal desorption spectroscopy. The removal ratio of deuterium by neon glow discharge largely depended on the material. Namely, the ratios for boron, titanium, or titanium oxide were 14%, 2%, or 40%, respectively. The ratios for the boron and the titanium oxide roughly agreed with the estimation from SRIM code calculations, while the ratio for the titanium did not agree with the estimation. These results suggest that the reduction of the deuterium retention is owing to the etching and the ion impact desorption of neon ions in the cases of boron and titanium oxide, and the prompt retrapping of deuterium by titanium atoms might occur in the case of titanium. The comparison between titanium and titanium oxide clearly shows that the removal effect by glow discharge largely depended on the surface conditions, such as oxygen impurity.

Foil calibration for IR imaging bolometer by laser irradiation
Ryuichi Sano, Byron J. Peterson, Evgeny A. Drapiko, Dongcheol Seo, Yuji Yamauchi, Tomoaki Hino
Plasma and Fusion Research, 7, SPL.ISS.1, 2012年, ［査読有り］
英語, 研究論文（学術雑誌）, The IR imaging video bolometer (IRVB) provides the power distribution of plasma radiation. The radiation distribution is obtained from the temperature distribution on the bolometer foil. It is necessary to calibrate between the temperature distribution and the incident radiation power on the bolometer foil. This paper describes a new calibration technique for the foil which we have developed. The bolometer foil was irradiated with a He-Ne laser and the temperature distribution was measured by an IR camera while changing the irradiation position. The temperature distribution measured was analyzed by the comparison with the results calculated by FEM. We repeated this comparison while changing the parameters such as effective foil thickness and effective emissivity in the calculation until the calculated distribution converged to the measured one. The temperature distribution calculated by the FEM agreed well with the measured one, so the calibration between the radiation power and the temperature profile can be suitably conducted by this technique. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

Toroidal distributions of amounts of retained gases and deposited impurities during 13th experimental campaign in the LHD
Naoyuki Shinoda, Naoko Ashikawa, Yuji Nobuta, Yuji Yamauchi, Tomoaki Hino, S. Suguru Masuzaki, Akio Sagara
Plasma and Fusion Research, 7, SPL.ISS.1, 2012年, ［査読有り］
英語, 研究論文（学術雑誌）, The gas desorption behavior, and the amounts of retained gases and deposited impurities in the first wall during the 13th experimental campaign in the Large Helical Device (LHD) have been evaluated using a material probes technique. The desorption behavior of hydrogen and helium depended on the atomic composition of deposited impurities. The amount of retained hydrogen was large in the vicinity of the anodes used for the glow discharge conditioning (GDC). The amount of retained helium was small near the anodes in this campaign. The toroidal dependence on the amount of retained gas corresponded to the total time for the GDC and the gas species used for glow discharge in the final stage. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

Deuterium retention and desorption behavior of co-deposited carbon film produced in gap
Yuji Nobuta, Kenji Yokoyama, Jun Kanazawa, Yuji Yamauchi, Tomoaki Hino, Satoshi Suzuki, Koichiro Ezato, Mikio Enoeda, Masato Akiba
Plasma and Fusion Research, 7, SPL.ISS.1, 2012年, ［査読有り］
英語, 研究論文（学術雑誌）, Co-deposition of deuterium with carbon in an opening on a plasma-facing surface, a so-called 'gap', was simulated by using a deuterium arc discharge with carbon electrodes. The carbon deposition distribution and deuterium retention/desorption behavior of the carbon film were investigated. The amount of deposited carbon decreased exponentially with an increase of the distance from the gap entrance and more rapidly decreased with an increase in discharge gas pressure. The deuterium concentration in the carbon film increased with discharge gas pressure. At a high discharge gas pressure of 36 Pa, the atomic ratio of D/C in the carbon film reached as high as 0.9. Deuterium retained in the film desorbed mainly in the forms of D2, HD, CD4 and C2D4. The desorption behavior of retained deuterium depended on D/C. In a film with a high D/C ratio, desorption of D2 started at lower temperatures. The amount of desorbed hydrocarbons (CD4 and C2D4) increased with D/C. Carbon film with high D/C tended to contain a polymer-like structure, which could be related to the desorption behavior of the retained deuterium. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

Hydrogen isotope retention and desorption in tungsten during glow discharges
Y. Yamauchi, Y. Kimura, Y. Kosaka, Y. Nobuta, T. Hino, K. Nishimura, Y. Ueda
Fusion Science and Technology, 60, 4, 1475, 1478, AMER NUCLEAR SOC, 2011年11月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, Hydrogen isotope retention and desorption behaviors in tungsten during glow plasma exposures using hydrogen isotope or inert gas were investigated. The rapid pressure drop of D2 and the simultaneous pressure rises of HD and H2 were observed during deuterium discharge. The net amount of retained deuterium was several times larger than that of stainless steel The desorption of HD or H2was several times smaller than that of stainless steel. Compared with stainless steel, the amount of desorbed deuterium during subsequent inert gas discharge was several times smaller.

Deuterium retention in F82H after low energy hydrogen ion irradiation
T. Ito, Y. Yamauchi, T. Hino, T. Shibayama, Y. Nobuta, K. Ezato, S. Suzuki, M. Akiba
Journal of Nuclear Materials, 417, 1-3, 1147, 1149, ELSEVIER SCIENCE BV, 2011年10月01日, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, The influence of 50 eV hydrogen ion irradiation on the deuterium retention and desorption behavior in the reduced activation ferritic-martensitic steel F82H was investigated by thermal desorption spectroscopy. The amount of retained deuterium in the hydrogen irradiated F82H was up to 10 times larger than without the hydrogen irradiation. In the F82H irradiated by the low energy hydrogen ions at 300 and 523 K, the desorption peak of deuterium was shifted to lower temperatures than for samples unirradiated or irradiated at 773 K. In the F82H irradiated by hydrogen at 773 K, the amount of retained deuterium decreased compared to materials irradiated at 300 or 523 K. © 2010 Elsevier B.V. All rights reserved.

Exchange and reduction of retained hydrogen isotope by glow discharges
Yukihiko Kimura, Tomoaki Hino, Kiyohiko Nishimura, Yuji Yamauchi, Yuji Nobuta
Journal of Nuclear Materials, 417, 1-3, 896, 899, ELSEVIER, 2011年10月01日, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, Following hydrogen glow discharge of stainless steel, a deuterium glow discharge was conducted to exchange the retained hydrogen in the stainless steel for deuterium. During the deuterium discharge, about 60% of the retained hydrogen desorbed. The amount of retained deuterium was 1.5 times larger than that of desorbed hydrogen. After the deuterium discharge, glow discharge cleaning with inert gas was conducted and the amount of the removed hydrogen isotope was evaluated. The removal ratio of hydrogen isotopes in the stainless steel was the largest during He glow discharge, and the smallest during Ar glow discharge. The removal ratio of retained hydrogen isotope by neon discharge cleaning did not change although sequential discharges of deuterium and neon were conducted repeatedly. © 2011 Elsevier B.V. All rights reserved.

Hydrogen isotope retention and desorption in tungsten during glow discharges
Y. Yamauchi, Y. Kimura, Y. Kosaka, Y. Nobuta, T. Hino, K. Nishimura, Y. Ueda
Fusion Science and Technology, 60, 4, 1475, 1478, American Nuclear Society, 2011年
英語, 研究論文（国際会議プロシーディングス）, Hydrogen isotope retention and desorption behaviors in tungsten during glow plasma exposures using hydrogen isotope or inert gas were investigated. The rapid pressure drop of D2 and the simultaneous pressure rises of HD and H2 were observed during deuterium discharge. The net amount of retained deuterium was several times larger than that of stainless steel The desorption of HD or H2was several times smaller than that of stainless steel. Compared with stainless steel, the amount of desorbed deuterium during subsequent inert gas discharge was several times smaller.

Fuel retention in LHD
S. Masuzaki, M. Kobayashi, M. Tokitani, N. Ashikawa, T. Hino, Y. Yamauchi, Y. Nobuta, N. Yoshida, M. Miyamoto, R. Sakamoto, J. Miyazawa, T. Morisaki, N. Ohyabu, H. Yamada, A. Komori
Fusion Science and Technology, 58, 1, 321, 330, American Nuclear Society, 2010年
英語, 研究論文（学術雑誌）, A global particle balance study has been investigated in the Large Helical Device (LHD) in which the first wall and the divertor tiles are made of stainless steel (SUS-316L) and carbon, respectively. The carbon area is less than 10% of the stainless steel area. The analyzed discharges have been conducted under an intrinsic helical divertor (HD) or a local island divertor (LID). The HD is an open divertor at this stage, and the LID is a closed divertor equipped with a baffle structure and a pump system. In the HD configuration, fuel retention up to 75% of injected hydrogen was observed, and the retained hydrogen affected the plasma density control. On the other hand, almost all fueled hydrogen was evacuated by the pumps in the LID configuration. After each experimental campaign, detailed analyses of the invessel material probes (SUS-316L stainless steel) and a divertor tile exposed to various plasma discharges during each experimental campaign were conducted. The areal density of the retained hydrogen both in the material probes and the divertor tile was in the range 1021 to 1022 H/m2, and it corresponded to the averaged areal density that was observed after an experimental day with high-density discharges.

Wall conditioning in LHD
S. Masuzaki, N. Ashikawa, K. Nishimura, M. Tokitani, T. Hino, Y. Yamauchi, Y. Nobuta, N. Yoshida, M. Miyamoto, A. Sagara, N. Noda, H. Yamada, A. Komori
Fusion Science and Technology, 58, 1, 297, 304, American Nuclear Society, 2010年
英語, 研究論文（学術雑誌）, Wall conditioning in the Large Helical Device (LHD) has been conducted successively since the first experimental campaign in 1998. The effects of wall conditioning on the vacuum condition, the plasma performance, and the surface modification of the plasma-facing components have been analyzed by both macroscopic and microscopic observations such as residual gas analysis and transmission electron microscope observation, respectively. The main tools for the conditioning are mild baking (95°)
glow discharges with hydrogen, helium, and neon
and wall coating with titanium and boron. Though the baking temperature is lower than in other fusion devices, it reduces impurity gases well just after the start of vacuum pumping, and it reduces retained hydrogen in plasma-facing components during the experimental campaign. Helium glow discharge was revealed to cause heavy damage on the surfaces of metallic components and the contamination of the hydrogen discharges with helium released from wall. Neon glow discharge has been conducted since it causes much less damage and hastens the conditioning of the wall. Boronization is very effective to reduce oxygen impurity in plasma, and the effects last for the whole experimental campaign in LHD.

4.核融合炉用プラズマ対向機器構造材料の燃料水素の吸蔵に関する研究
日野友明, 鈴木哲, 信太祐二, 山内有二, 金澤潤
日本原子力研究開発機構JAEA-Review, 12-14,63-66, 2009年11月
日本語

Electrostatic linear ion trap for low-energy ion beam storage
T Suzuki, Y Yamauchi
NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION A-ACCELERATORS SPECTROMETERS DETECTORS AND ASSOCIATED EQUIPMENT, 562, 1, 53, 56, ELSEVIER SCIENCE BV, 2006年06月, ［査読有り］
英語, 研究論文（学術雑誌）, An electrostatic linear ion trap was constructed and applied to the storage of an He ion beam of kinetic energy 208 eV. The time-of-flight detection of the stored ions confirms the validity of its storage performance. It is shown that the interaction with the residual gas plays a crucial role in the loss mechanism of the stored ions. The lifetime of the stored ions, which was measured to be 5.25 +/- 0.65 mS in the present study, with modest vacuum conditions, is expected to lengthen substantially with ultra-high vacuum conditions. (c) 2006 Elsevier B.V. All rights reserved.

Deuterium retention and desorption behavior of boron-titanium as first wall material of fusion experimental device
T. Hino, Y. Hashiba, Y. Yamauchi, Y. Hirohata, K. Nishimura, N. Ashikawa, S. Masuzaki, A. Sagara, N. Noda, N. Ohyabu, A. Komori, O. Motojima
Fusion Engineering and Design, 81 A, 1-4, 127, 131, ELSEVIER SCIENCE SA, 2006年02月, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, Boron-titanium (B-Ti) was deposited on stainless steel substrate by evaporation of titanium and boron followed by annealing. The boron concentration at the surface region was comparable with the titanium concentration. Thermal desorption spectrum of deuterium was obtained after deuterium ion irradiation at room temperature, and the amount of retained deuterium and desorption temperatures of deuterium were measured. For comparison, titanium (Ti) and boron (B) films were deposited on the stainless steel substrate and the thermal desorption spectra were similarly obtained. The amount of retained deuterium in B-Ti was approximately 1/3 or 1/2 of that in Ti or B, respectively. The peak temperatures in B-Ti were 470 and 620 K, which were lower than those in B, 500 and 720 K. These results suggest that the use of B-Ti may reduce fuel hydrogen retention and recycling in fusion experimental devices, compared with the case of boron. © 2005 Elsevier B.V. All rights reserved.

TDS measurement of hydrogen released from stainless steel oxidized in H2O-containing atmospheres
Akira Yamauchi, Yuji Yamauchi, Yuko Hirohata, Tomoaki Hino, Kazuya Kurokawa
Materials Science Forum, 522-523, 163, 170, TRANS TECH PUBLICATIONS LTD, 2006年, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, Hydrogen dissolved in the Cr2O3 scale formed on the stainless steel in the H2O-containing atmospheres is observed by TDS (thermal desorption spectroscopy) measurements. The amount of dissolved hydrogen in the Cr2O3 scale reaches a maximum about 0.32 mol% when the H2O concentration in the gas reaches 20%. It was found from GDS (glow discharge spectroscopy) measurements that hydrogen may exist at the oxide scale / substrate interface or in Cr2O3 scale bounded that interface. However, results from the Vickers hardness and the observation of scale morphology by SEM (scanning electron microscopy), hydrogen dissolved in the Cr2O3 scale would have little effect on a decrease in the mechanical property of the Cr2O3 scale. Therefore, hydrogen dissolved in the Cr2O3 scale may not be main factor of the deterioration of the Cr2O3 scale.

Axial elongation of water droplet by application of electric field for bio-chemical reaction system
T. Hino, E. Hyashishita, Y. Yamauchi, Y. Hirohata, M. Nishikawa
Physica Scripta T, T118, 193, 195, 2005年, ［査読有り］
英語, 研究論文（学術雑誌）, In bio-micro devices, numerous functions of liquid flow such as elongation, expansion and transportation are required. Electro-wetting is one of the methods to control the liquid flow. In the present study, a water droplet was elongated in a channel based upon electro-wetting, and the effect of electrode diameter and shape of electrode immersed in a water droplet on the elongation was examined. The elongation increased as a diameter of electrode, and significantly increased when the electrode with a T-shape was employed. The electric field at the edge of water droplet is kept relatively large when the rod electrode with a large diameter or the T-shape electrode is employed. The present result can be explained by this effect, the change of local electric energy at the edge of the droplet. © Physica Scripta 2005.

Surface flatness of polycrystalline copper after argon ion etching followed by annealing
T. Hino, T. Taguchi, Y. Yamauchi, Y. Hirohata, M. Nishikawa
Journal of Vacuum Science and Technology B: Microelectronics and Nanometer Structures, 22, 6, 2632, 2634, A V S AMER INST PHYSICS, 2004年11月, ［査読有り］
英語, 研究論文（学術雑誌）, Oblique injection of argon ions with energy of 1 keV was conducted for etching of polycrystalline copper. The surface became rough owing to the formation of blisters. The average diameter and height of blisters was approximately 150 and 20-30 nm, respectively. The polycrystalline copper irradiated by argon ions was annealed to rupture the blisters. The retained argon desorbed in the temperature range from 500 to 800 K. The height of blisters significantly reduced to 10-15 nm although the diameter and surface density of blisters roughly remained the same. © 2004 American Vacuum Society.

Preliminary results on development of a thin GSO scintillator for neutron science
JH Kaneko, N Susa, S Tsuchida, M Watanabe, S Miura, T Mizuno, Y Yamauchi, M Hashiba, T Oku, A Homma, F Fujita, T Ino, M Furusaka, T Sawamura, HM Shimizu, Y Kiyanagi
NUCLEAR INSTRUMENTS & METHODS IN PHYSICS RESEARCH SECTION A-ACCELERATORS SPECTROMETERS DETECTORS AND ASSOCIATED EQUIPMENT, 529, 1-3, 307, 309, ELSEVIER SCIENCE BV, 2004年08月
英語, 研究論文（学術雑誌）, A thin GSO scintillator was developed aiming at applications for the neutron science. In order to fabricate a large and thin GSO scintillator in the future, fabrication was started from two types of GSO powders. Although several GSO films were fabricated by evaporation techniques, these films did not have luminescence in photo luminescence (PL) spectroscopy. Thus, these samples were annealed up to 1550degreesC. However, it resulted in no luminescence observation. In response to these results, GSO powders were directly annealed; a phenomenon looking like melting and sintering were observed for Gd2Si2O7:Ce at a temperature higher than 1500degreesC. This sample was quasi-transparent and had luminescence at 405 nm in PL spectrum. In addition, this sample was tougher than a GSO single crystal. In conclusion, the prospect of fabrication of a thin GSO scintillator by combination of annealing and mechanical polishing was obtained. (C) 2004 Elsevier B.V. All rights reserved.

Enhancement of reactive species density in nitrogen plasma by mixture of helium and nitridation experiment for silicon
T. Hino, Y. Yamauchi, J. Ono, Y. Hirohata
Vacuum, 74, 3-4 SPEC. ISS., 467, 471, PERGAMON-ELSEVIER SCIENCE LTD, 2004年06月07日, ［査読有り］
英語, 研究論文（学術雑誌）, In order to increase the electron plasma density and/or temperature in electron cyclotron resonance (ECR) nitrogen plasma, electron acceleration has to be enhanced by reducing the electron collision frequency. One method is to add gas species with a lower electron collision cross-section, compared with the case of nitrogen gas. For this purpose, the helium gas was added into the ECR nitrogen plasma, where the total gas pressure and the absorption power of microwave were kept the same. The emission intensities of nitrogen molecular ion and excited nitrogen molecular were measured as a function of the ratio of helium pressure to nitrogen gas pressure. The densities of both excited nitrogen molecular and nitrogen molecular ion showed maxima to the ratio of helium gas pressure to nitrogen gas pressure. Hence, the densities of such reactive species can be maximized by the addition of helium gas into nitrogen plasma. Based upon the present results, the nitridation experiment for silicon was conducted by changing the ratio of helium gas pressure. In this irradiation experiment, the substrate was negatively biased and the temperature was kept 900 K. The degree of nitridation well corresponded to the change of density of reactive species. © 2004 Elsevier Ltd. All rights reserved.

Helium gas permeability of SiC/SiC composite developed for blanket component
Tomoaki Hino, Takahiro Jinushi, Yuko Hirohata, Masao Hashiba, Yuji Yamauchi, Yutai Katoh, Akira Kohyama
Fusion Science and Technology, 43, 2, 184, 190, American Nuclear Society, 2003年
英語, 研究論文（学術雑誌）, To employ SiC/S/C composite us blanket components of a fusion reactor, permeation behavior of helium gas has to be investigated since the helium is used as the coolant. For this purpose, a vacuum system consisting of upstream and downstream chambers was fabricated for the measurement of permeability, and the permeability was measured for several SiC/SiC composite materials recently developed. For the pressure range from 102 to 105 Pa in the upstream chamber, the pressure rise due to the permeation of helium in the downstream chamber linearly increased with the pressure of the upstream chamber. Then, the permeability was roughly constant for the pressure range of the upstream chamber. The permeabilities of SiC/SiC composites produced by polymer impregnation and pyrolysis (PIP), hot pressing (HP) and melt-infiltration-finished PIP were 5 × 10-5, 4 × 10-6, and 9 × 10-7 m2/s, respectively. In the matrix structure of the SiC/SiC composite made by the PIP method with a high permeability, cracking in the matrix and pores of micron size were observed. Compared to these materials, SiC/SiC composites produced by liquid phase sintering using submicron or nanopowder of β-SiC and the HP method had extremely low permeabilities. The permeability of the SiC/SiC composite made by using submicron or nanopowder of β-SiC became 1.5 × 10-9 or 4 × 10-11 m2/s. Based upon the present data, the helium gas flow was analyzed for a blanket module consisting of only SiC/SiC composite. If a vacuum pump is attached to the module, the helium leak into a plasma can be ignored, compared to the heliums produced by fusion reactions. Hence, the entire module can be made by only SiC/SiC composite, from a viewpoint of helium permeation.

Plasma surface interaction studies on low activation materials
T Hino, K Yamaguchi, T Jinushi, Y Hirohata, Y Yamauchi
MATERIALS FOR ADVANCED ENERGY SYSTEMS AND FISSION AND FUSION ENGINEERING, PROCEEDINGS, 129, 136, WORLD SCIENTIFIC PUBL CO PTE LTD, 2003年, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, In fusion demonstration reactors, use of low activation materials has been required for a safety of fusion plant. Major concern on low activation materials has been on the material damage due to neutron irradiation so far. However, these materials are also used as plasma facing components and then the plasma surface interactions have to be investigated. In the present study, interactions of ferritic steel with plasma ions, fuel hydrogen absorption of vanadium alloy and helium gas permeation in SiC/SiC composite were investigated to clarify the suitability of the low activation materials.
The ferritic steel had ion sputtering yield similar to stainless steel, although in the surface the fuel hydrogen retention differed with that of the stainless steel. The vanadium alloy causes the hydrogen embrittlement due to a large hydrogen absorption. In order to reduce the hydrogen absorption speed, the use of titanium oxide coating was found to be very effective. The SiC/SiC composite is employed as blanket material, and then the concern is permeation of helium gas coolant. The SiC/SiC composite recently developed showed extremely low permeability. Then, this material can be used for the blanket without a metal wrapping.

Silicon-carbide as plasma facing or blanket material
T Hino, T Jinushi, Y Yamauchi, M Hashiba, Y Hirohata, Y Katoh, A Kohyama
ADVANCED SIC/SIC CERAMIC COMPOSITES: DEVELOPMENTS AND APPLICATIONS IN ENERGY SYSTEMS, 144, 353, 361, AMER CERAMIC SOC, 2002年, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, Silicon-carbide is a candidate material for plasma facing components and blanket components in a next generation of fusion reactors. As the plasma facing material, one of important issues to be clarified is property associated with fuel hydrogen retention, which affects fuel hydrogen recycling and in-vessel tritium inventory. For SiC and SiC/SiC composite, deuterium retention properties were investigated using an ECR ion source and a technique of thermal desorption spectroscopy. The retained amount of deuterium was observed to be comparable to that of carbon fiber composite, CFC. However, the desorption temperature was approximately 150 K lower than that of CFC. In addition, the chemical erosion was extremely smaller than that of CFC. Thus, the use of silicon-carbide has advantages in terms of fuel hydrogen retention and chemical erosion, compared to the case of graphitic materials such as CFC.
As the blanket material, SiC/SiC composite may be employed with helium gas coolant. Therefore, one of major concerns is permeation of helium gas through the SiC/SiC composite. For several SiC/SiC composite materials recently developed, the helium gas permeability was measured using a vacuum system consisting of high and low pressure chambers, and fixing structure for sample. The permeability of the SiC/SiC composite made by NITE process using nano-powder of SiC was observed to be extremely low, approximately 4x10(-11) m(2)/s. The permeability of SiC made by using only nano-powder of SiC was (2-8)x10(-13) m(2)/S. Since the permeability is quite low, requirement for additional use of metal with SiC/SiC composite can be reduced, and then the engineering design for SiC/SiC blanket may become simple.

Electron emission from Na/Fe(100) surfaces by deexcitation of spin-polarized helium metastable atoms
Y Yamauchi, M Kurahashi, T Suzuki
SPIN 2000, 570, 972, 975, AMER INST PHYSICS, 2001年, ［査読有り］
英語, 研究論文（国際会議プロシーディングス）, A pulsed helium metastable atom beam was generated by the pulsed nozzle-skimmer discharge and a high. degree of spin polarization of the helium beam was obtained using the optical pumping method. Spin polarization of the outermost surface electron was detected for clean and sodium covered Fe(100) films deposited on MgO(100) by measuring the secondary electrons ejected by the irradiation of the spin-polarized helium metastable atom beam. Secondary electrons correspond to the Fermi level for a clean iron surface and those corresponding to the Na 3s level for the sodium-covered surface show a positive asymmetry, which indicates the negative polarization of electrons at these levels. The negative polarization of the Na 3s electrons rapidly decreases with the thickness increasing of the Na layer up to 0.5 ML but remains almost constant at a higher coverage (3ML).

トリチウム挙動および中性子照射効果 (プロジェクトレビュー 日米科学技術協力事業PHENIX 計画 : 6年間の成果と次期計画)
大矢 恭久, 波多野 雄治, 信太 祐二, 山内 有二, 小林 真, 大宅 諒, 片山 一成, 大塚 哲平, 上田 良夫, プラズマ・核融合学会誌 = Journal of plasma and fusion research, 96, 3, 140, 144, 2020年03月
プラズマ・核融合学会編集委員会, 日本語

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年
日本語

昇温脱離分析を用いた鉄鋼中の微量水素定量評価に向けた研究
小松裕人, 山内有二, 吉田肇, 日本金属学会講演概要(CD-ROM), 161st, ROMBUNNO.P181, 2017年08月23日
日本語

昇温脱離分析を用いた鉄鋼中の高精度微量水素評価法の確立に向けた研究
小松裕人, 山内有二, 吉田肇, 材料とプロセス(CD-ROM), 30, 1, ROMBUNNO.203, 2017年03月31日
日本語

Li₂TiO₃の組成変化及び欠陥生成が重水素脱離挙動に与える影響
鈴木祐太, 信太祐二, 山内有二, 星野毅, 日本原子力学会北海道支部研究発表会講演要旨集, 34th, 12‐13, 2017年
日本語

高温で水素に曝されたリチウムタイタネート(Li₂TiO₃)の原子組成及び重水素脱離挙動の評価
鈴木祐太, 信太祐二, 山内有二, 星野毅, 日本原子力学会春の年会予稿集(CD-ROM), 2016, ROMBUNNO.3L05, 2016年03月16日
日本語

アルゴンガスパフ下におけるタングステンの重水素保持挙動
笠間俊行, 山内有二, 信太祐二, 西村清彦, 日本原子力学会春の年会予稿集(CD-ROM), 2016, ROMBUNNO.1J04, 2016年03月16日
日本語

プラズマ装置からの排ガスにおける水素同位体化学種の定量的評価とその精製挙動
湯山健太, 植村有希, 桜田翔大, 藤田啓恵, HU Cui, 近田拓未, 大矢恭久, 松山政夫, 原正憲, 日野友明, 山内有二, 信太祐二, 田中将裕, 相良明男, 日本原子力学会秋の大会予稿集(CD-ROM), 2015, ROMBUNNO.N40, 2015年08月21日
日本語

高温下で重水素イオン照射した核融合炉用SiC繊維強化SiC複合材料の重水素保持挙動と損耗評価
信太祐二, 日野友明, 山内有二, 野澤貴史, 真空に関する連合講演会講演予稿集, 55th, 104, 2014年11月18日
日本語

プラズマ装置におけるトリチウム輸送ダイナミクスの解明
湯山健太, 小林真, 内村大道, 戸田健介, 松山政夫, 原正憲, 日野友明, 山内有二, 信太祐二, 相良明男, 田中将裕, 芦川直子, 大矢恭久, 奥野健二, 日本原子力学会春の年会予稿集(CD-ROM), 2014, ROMBUNNO.E12, 2014年03月10日
日本語

高温保存によるリチウムイオン2次電池負極材ガス放出挙動の変化
梅田基樹, 山内有二, 日野友明, 表面科学学術講演会講演要旨集, 33rd, 202, 2013年11月26日
日本語

Investigation of Tritium Inventory and Removal on Deposition Layer in LHD
Ashikawa N., Nishimura K., Torikai Y., Matsuyama M., Hong S-H., Hamaji Y., Ueda Y., Yamauchi Y., Nobuta Y., Hino T., 2013年11月

LHD16サイクルにてプラズマ曝露したタングステンにおける不純物堆積層と水素同位体滞留挙動の相関
佐藤美咲, 小林真, 内村大道, 増崎貴, 時谷政行, 吉田直亮, 渡邉英雄, 山内有二, 日野友明, 宮本光貴, 波多野雄治, 大矢恭久, 奥野健二, 日本原子力学会秋の大会予稿集(CD-ROM), 2013, ROMBUNNO.B35, 2013年08月20日
日本語

プラズマ排ガス中の不純物量と重水素回収効率に関する研究
田口僚久, 戸田健介, 松山政夫, 原正憲, 日野友明, 山内有二, 信太祐二, 相良明男, 田中将裕, 芦川直子, 大矢恭久, 奥野健二, 日本原子力学会春の年会予稿集(CD-ROM), 2013, ROMBUNNO.K12, 2013年03月11日
日本語

プラズマ排ガスからの重水素回収に及ぼす不純物影響に関する研究
田口 僚久, 芦川 直子, 大矢 恭久, 奥野 健二, 戸田 健介, 松山 政夫, 原 正憲, 日野 友明, 山内 有二, 信太 祐二, 相良 明男, 田中 将裕, 日本原子力学会 年会・大会予稿集, 2013, 0, 604, 604, 2013年
核融合炉に燃料として導入された水素同位体のほとんどは、プラズマ中で燃焼せずそのまま排気される。そのため水素同位体の燃料リサイクリングの観点から核融合炉の確立にむけ、水素同位体の回収・精製効率に関して詳細に議論する必要がある。当研究室では、実験室規模の知見から実機における水素同位体の移行挙動を明らかにするために、水素同位体ダイナミックス解明装置（EXPRESS）を設計・作成している。本研究では、プラズマ曝露部より排気された重水素ガスを回収、加熱した水素吸蔵合金（ST2002）を用いた重水素ガスの一次純化、Pd膜水素高純度精製装置による一次純化重水素ガスのさらなる高純化、水素吸蔵合金（ZrCo）による貯蔵、という一連の処置を行う装置を作成し、その装置の重水素純化の効率を測定し、重水素回収に及ぼす不純物影響に関して議論を行う。, 一般社団法人 日本原子力学会, 日本語

Liイオン電池の充放電容量に与える負極被膜の影響
高丸翔太, 日野友明, 山内有二, 信太裕二, 真空に関する連合講演会講演予稿集, 53rd, 152, 2012年11月14日
日本語

水素・ヘリウム同時照射後の低放射化フェライト鋼の水素リテンション
小西徹, 芦川直子, 山内有二, 信太祐二, 日野友明, 核融合エネルギー連合講演会予稿集, 9th (CD-ROM), ROMBUNNO.29A-61P, 2012年
日本語

タングステン材料の重水素保持脱離挙動
二瓶直人, 日野友明, 山内有二, 信太祐二, 大矢恭久, 奥野健二, 上田良夫, 核融合エネルギー連合講演会予稿集, 9th (CD-ROM), ROMBUNNO.28A-84P, 2012年
日本語

LHDにおいて水素プラズマ曝露したタングステン試料中の水素同位体滞留能変化
大矢恭久, 増崎貴, 時谷政行, 渡辺英雄, 吉田直亮, 日野友明, 山内有二, 波多野雄治, 奥野健二, 核融合エネルギー連合講演会予稿集, 9th (CD-ROM), ROMBUNNO.29A-47P, 2012年
日本語

マテリアルプローブによるLHDのプラズマ壁相互作用の評価
重村卓哉, 山内有二, 信太祐二, 日野友明, 芦川直子, 相良明男, 西村清彦, 核融合エネルギー連合講演会予稿集, 9th (CD-ROM), ROMBUNNO.28A-15P, 2012年
日本語

スイープガス曝露によるリチウムタイタネート(Li₂TiO₃)からの重水素除去
信太祐二, 土谷晃史, 山内有二, 日野友明, 榎枝幹男, 秋場真人, 核融合エネルギー連合講演会予稿集, 9th (CD-ROM), ROMBUNNO.29A-95P, 2012年
日本語

各種タングステンの重水素保持脱離挙動
二瓶直人, ARMANDO Marco, 日野友明, 山内有二, 信太祐二, 大矢恭久, 奥野健二, 上田良夫, 日本原子力学会北海道支部研究発表会講演要旨集, 30th, 1, 2, 2012年
日本語

LHDにおける炭素堆積層評価および水素同位体保持特性に関する研究
芦川直子, 深山健介, 浜地志憲, 鳥養祐二, 山内有二, 信太祐二, 日野友明, 上田良夫, 松山政夫, 核融合エネルギー連合講演会予稿集, 9th (CD-ROM), ROMBUNNO.29A-48P, 2012年
日本語

大型ヘリカル装置第13サイクル実験におけるガス保持及び不純物堆積のトロイダル分布
篠田直幸, 芦川直子, 信太祐二, 山内有二, 日野友明, 増崎貴, 相良明男, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 3, 4, 2011年
日本語

黒鉛における重水素保持と希ガスグロー放電による除去
松本暁, 山内有二, 日野友明, 西村清彦, 上田良夫, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 5, 6, 2011年
日本語

重水素アーク放電によるスリット間で生成した同時堆積炭素膜の重水素濃度
金澤潤, 信太祐二, 山内有二, 日野友明, 横山堅二, 鈴木哲, 江里幸一郎, 榎枝幹男, 秋場真人, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 9, 10, 2011年
日本語

スイープガス曝露によるリチウムタイタネートの重水素除去
土谷晃史, 日野友明, 山内有二, 信太祐二, 秋場真人, 榎枝幹男, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 7, 8, 2011年
日本語

大型ヘリカル装置における炭素‐水素共堆積膜中の水素濃度
深山健介, 芦川直子, 信太祐二, 山内有二, 日野友明, 増崎貴, 西村清彦, 相良明男, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 11, 12, 2011年
日本語

レーザーを用いたイメージングボロメータ用金属箔のキャリブレーション
佐野竜一, PETERSON B. J, DRAPIKO E. A, SEO D. C, 山内有二, 日野友明, 日本原子力学会北海道支部研究発表会講演要旨集, 29th, 1, 2, 2011年
日本語

Neグロー放電によるB及びTi中の重水素保持量の低減
小坂祐太, 山内有二, 日野友明, 西村清彦, 日本原子力学会北海道支部研究発表会講演要旨集, 28th, 7, 8, 2010年
日本語

タングステン及び蒸着材の重水素リテンションとHeグロー放電による低減
松本暁, 山内有二, 日野友明, 西村清彦, 上田良夫, 日本原子力学会北海道支部研究発表会講演要旨集, 28th, 5, 6, 2010年
日本語

LHDの主放電及びグロー放電プラズマに曝露されたボロン膜及びチタン膜の不純物堆積とガス保持量
篠田直幸, 芦川直子, 信太祐二, 山内有二, 日野友明, 増崎貴, 相良明男, 日本原子力学会北海道支部研究発表会講演要旨集, 28th, 3, 4, 2010年
日本語

ヘリウムガス曝露によるリチウムタイタネートの重水素除去
土谷晃史, 山内有二, 信太祐二, 日野友明, 日本原子力学会北海道支部研究発表会講演要旨集, 28th, 9, 10, 2010年
日本語

LHD第12サイクルプラズマに曝されたマテリアルプローブ解析
松永陽子, 芦川直子, 信太祐二, 山内有二, 日野友明, 増崎貴, 相良明男, 日本原子力学会北海道支部研究発表会講演要旨集, 28th, 1, 2, 2010年
日本語

SiC／SiC複合材料の燃料水素保持特性
信太 祐二, 山内 有二, 日野 友明, Cho Ho-Jun, Yoon Han Ki, 表面科学学術講演会要旨集, 30, 0, 33, 33, 2010年
SiC／SiC複合材料は優れた高温強度特性と低放射化特性を有しており，核融合炉第一壁の候補材料となっている。第一壁における燃料水素の保持脱離挙動は燃料リサイクリングやトリチウムインベントリーの観点から重要である。本研究では，SiC／SiC複合材料に対し重水素照射実験を行い，燃料水素の保持脱離挙動を昇温脱離法で評価した。講演では，照射前後の組成変化や表面形態変化についても報告する予定である。, 公益社団法人 日本表面科学会, 日本語

核融合炉第一壁候補材料に対する水素同位体の保持脱離挙動
木村友基彦, 日野友明, 山内有二, 信太祐二, 日本原子力学会北海道支部研究発表会講演要旨集, 27th, 28, 29, 2009年
日本語

低エネルギー水素イオンを照射したF82H鋼の重水素保持脱離挙動
伊藤達哉, 信太祐二, 山内有二, 日野友明, 柴山環樹, 江里幸一朗, 鈴木哲, 秋場真人, 日本原子力学会北海道支部研究発表会講演要旨集, 27th, 26, 27, 2009年
日本語

LHD中に堆積した同時堆積炭素膜の水素濃度
濱竜太, 信太祐二, 山内有二, 日野友明, 芦川直子, 増崎貴, 相良明男, 西村清彦, 大藪修義, 小森彰夫, 本島修, 日本原子力学会北海道支部研究発表会講演要旨集, 27th, 30, 31, 2009年
日本語

ブランケット増殖材・リチウムタイタネイトの重水素保持脱離挙動
柴田博信, 佐藤大輔, 山内有二, 日野友明, 秋場真人, 鈴木哲, 日本原子力学会北海道支部研究発表会講演要旨集, 25th, 4, 5, 2007年12月10日
日本語

ボロン・タイテニアムの燃料水素保持・脱離挙動
犬飼宗志, 山内有二, 日野友明, 芦川直子, 西村清彦, 日本原子力学会北海道支部研究発表会講演要旨集, 25th, 8, 9, 2007年12月10日
日本語

酸素グロー放電によるステンレス鋼表面及び黒鉛からの炭素除去
野村拓也, 信太祐二, 山内有二, 日野友明, 日本原子力学会北海道支部研究発表会講演要旨集, 25th, 6, 7, 2007年12月10日
日本語

Arグロー放電によるポリカーボネイトの親水化
永井宏治, 山内有二, 日野友明, 日本原子力学会北海道支部研究発表会講演要旨集, 25th, 14, 15, 2007年12月10日
日本語

多結晶銅の斜めイオンエッチングにおける基板温度効果
小林達郎, 山内有二, 日野友明, 日本原子力学会北海道支部研究発表会講演要旨集, 25th, 16, 17, 2007年12月10日
日本語

LHDのローカル・アイランド・ダイバータ近傍で生成される共堆積炭素膜の評価
平田智也, 日野友明, 増崎貴, 芦川直子, 山内有二, 廣畑優子, 相良明男, 西村清彦, 大藪修義, 野田信明, 小森彰夫, 本島修, 日本原子力学会北海道支部研究発表会講演要旨集, 25th, 12, 13, 2007年12月10日
日本語

ネオングロー放電によるステンレス鋼の水素・ヘリウムの除去
武田建史郎, 日野友明, 山内有二, 日本原子力学会北海道支部研究発表会講演要旨集, 25th, 10, 11, 2007年12月10日
日本語

アーク放電で作製したカーボンダストの水素およびヘリウムリテンション
佐藤壮一朗, 山内有二, 広畑優子, 日野友明, 谷川尚, 鈴木哲, 秋場真人, 日本原子力学会北海道支部研究発表会講演要旨集, 24th, 38, 39, 2006年12月08日
日本語

SiC/SiC複合材料の酸化後のヘリウムガス透過率測定
大串裕介, 山内有二, 広畑優子, 日野友明, 香山晃, 日本原子力学会北海道支部研究発表会講演要旨集, 24th, 28, 29, 2006年12月08日
日本語

酸素グロー放電によるステンレス鋼内壁の不純物低減
野村拓也, 山内有二, 広畑優子, 日野友明, 日本原子力学会北海道支部研究発表会講演要旨集, 24th, 32, 33, 2006年12月08日
日本語

表面損傷を与えた低放射化フェライト鋼の燃料水素リテンション
片田洋介, 日野友明, 山内有二, 廣畑優子, 谷川尚, 江里幸一郎, 鈴木哲, 秋場真人, 日本原子力学会北海道支部研究発表会講演要旨集, 24th, 36, 37, 2006年12月08日
日本語

マテリアルプローブによるLHDプラズマ壁相互作用の解析
平田智也, 芦川直子, 日野友明, 山内有二, 廣畑優子, 増崎貴, 西村清彦, 相良明男, 大藪修義, 野田信明, 小森彰夫, 本島修, 日本原子力学会北海道支部研究発表会講演要旨集, 24th, 30, 31, 2006年12月08日
日本語

ブランケット増殖材,Li₂TiO₃の重水素保持・脱離挙動
佐藤大輔, 山内有二, 広畑優子, 日野友明, 鈴木哲, 秋場真人, 日本原子力学会北海道支部研究発表会講演要旨集, 24th, 34, 35, 2006年12月08日
日本語

低放射化バナジウム合金のヘリウム保持・脱離特性におよぼす不純物および熱処理温度の影響
奥大輔, 広畑優子, 山田剛, 山内有二, 日野友明, 長坂琢也, 室賀健夫, 核融合エネルギー連合講演会予稿集, 6th, 127, 2006年
日本語

ヘリウムイオン照射によるV-4Cr-4Ti合金のヘリウムの熱脱離挙動（_II_）:最終熱処理温度の影響
広畑 優子, 山田 剛, 山内 有二, 日野 友明, 長坂 琢也, 室賀 健夫, 日本原子力学会 年会・大会予稿集, 2005, 0, 627, 627, 2005年03月11日
低放射化バナジウム合金の最終熱処理温度を変化させて、高フラックス・高フルエンスでヘリウムイオンを室温で照射した。試料中に保持されたヘリウムの熱脱離挙動を昇温脱離法で調べた。最終熱処理後研磨した試料は、冷間圧延試料に比べて低温側（_から_500K）で脱離するHe量が増加し、脱離温度が高温側にシフトした。, 一般社団法人 日本原子力学会

ヘリウムイオン照射によるV-4Cr-4Ti合金のヘリウムの熱脱離挙動
広畑 優子, 山内 有二, 山田 剛, 日野 友明, 長坂 琢也, 室賀 健夫, 日本原子力学会 年会・大会予稿集, 2004, 0, 704, 704, 2004年03月11日
低放射化バナジウム合金に高フラックス・高フルエンスでヘリウムイオンを室温で照射した。その後試料を加熱し、ヘリウムの脱離挙動を昇温脱離法で調べた。ほとんどのヘリウムは500Kおよび1200K近傍で脱離した。照射量が5×10<SUP>17</SUP>He/cm<SUP>2</SUP>で保持量が飽和傾向を示した。, 一般社団法人 日本原子力学会

LHD内壁ステンレス鋼のヘリウム保持特性
信太祐二, 山内有二, 広畑優子, 日野友明, 芦川直子, 相良明男, 増崎貴, 野田信明, 小森彰夫, 真空に関する連合講演会講演予稿集, 44th, 141, 2003年11月12日
日本語

LHD内壁材料の水素及びヘリウム保持・脱離特性―グロー放電装置による模擬実験―
山内有二, 佐藤慎吾, 広畑優子, 日野友明, 小森彰夫, 芦川直子, 相良明男, 野田信明, 大薮修義, 真空に関する連合講演会講演予稿集, 44th, 140, 2003年11月12日
日本語

JT‐60で生成されたカーボンダストの形態と水素保持量
吉田肇, 山内有二, 広畑優子, 日野友明, 秋場真人, 新井貴, 鈴木哲, 柳生純一, 宮直之, 真空に関する連合講演会講演予稿集, 44th, 143, 2003年11月12日
日本語

タングステンダストの重水素ガス吸収及び熱脱離特性
広畑 優子, 菊池 尚嗣, 吉田 肇, 山内 有二, 日野 友明, 谷口 正樹, 横山 堅二, 秋場 真人, 日本原子力学会 年会・大会予稿集, 2003, 0, 597, 597, 2003年08月07日
ITERのプラズマ対向壁材料であるタングステンがプラズマによって損耗し、ダストとして炉内に残存することを想定して、ダスト中のトリチウムインベントリーを見積もることを試みた。電子ビーム照射により作製したＷダストの重水素の溶解度は、1.32×10<SUP>-5</SUP>D/W（1000℃、1atm換算）で市販のＷ粉末の吸収重水素量よりも約２桁以上も小さかった。, 一般社団法人 日本原子力学会

水蒸気酸化で形成されたクロミア皮膜中の水素濃度と分布について
山内啓, 黒川一哉, 高橋英明, 山内有二, 広畑優子, 日野友明, 日本金属学会講演概要, 132nd, 363, 2003年03月27日
日本語

低放射化V‐4Cr‐4Ti合金の重水素イオン照射による重水素保持特性
広畑優子, 山田剛, 山内有二, 日野友明, 室賀健夫, 日本原子力学会春の年会要旨集, 41st, 263, 2003年03月05日
日本語

低放射化フェライト鋼F82Hの重水素リテンション
山内有二, 山口薫, 広畑優子, 橋場正男, 日野友明, 都筑和泰, 日本原子力学会春の年会要旨集, 41st, 262, 2003年03月05日
日本語

斜め入射による表面のスパッタリングの一様性
田口貴史, 山内有二, 広畑優子, 日野友明, 柳沢一郎, 西川正名, 新野高房, 真空に関する連合講演会講演予稿集, 43rd, 85, 86, 2002年10月16日
日本語

低放射化バナジウム合金の重水素イオン照射による重水素捕捉
広畑優子, 元嶋大, 山内有二, 日野友明, 室賀健夫, 真空に関する連合講演会講演予稿集, 43rd, 375, 376, 2002年10月16日
日本語

大型ヘリカル装置(LHD)第一壁の不純物堆積とガス保持特性
信太祐二, 広畑優子, 橋場正男, 山内有二, 日野友明, 増崎貴, 相良明男, 芦川直子, 本島修, 真空に関する連合講演会講演予稿集, 43rd, 15, 16, 2002年10月16日
日本語

グロー放電による大型ヘリカル装置内壁のコンディショニング
佐藤慎吾, 山内有二, 広畑優子, 日野友明, 小森彰夫, 真空に関する連合講演会講演予稿集, 43rd, 31, 32, 2002年10月16日
日本語

低放射化SiC繊維強化SiC複合材のヘリウムガス透過特性 (II)
地主孝広, 橋場正男, 山内有二, 広畑優子, 日野友明, 加藤雄大, 香山晃, 真空に関する連合講演会講演予稿集, 43rd, 339, 340, 2002年10月16日
日本語

低放射化フェライト鋼(F82H)の重水素イオン照射による重水素捕捉
山口薫, 山内有二, 広畑優子, 日野友明, 都築和泰, 真空に関する連合講演会講演予稿集, 43rd, 371, 372, 2002年10月16日
日本語

Cr₂O₃皮膜の劣化に及ぼす水蒸気の影響
山内啓, 黒川一哉, 高橋英明, 山内有二, 広畑優子, 日野友明, 材料とプロセス, 15, 6, 1096, 1099, 2002年09月01日
日本語

TPD法によるCr₂O₃皮膜からの水素放出挙動
山内啓, 黒川一哉, 高橋英明, 山内有二, 広畑優子, 日野友明, 日本金属学会講演概要, 130th, 178, 2002年03月28日
日本語

斜めイオン入射による銅のエッチング
山内有二, 田口貴史, 広畑優子, 日野友明, 新野高房, 柳沢一郎, 西川正名, 応用物理学関係連合講演会講演予稿集, 49th, 1, 159, 2002年03月27日
日本語

ヘリウムイオンによる多結晶銅のスパッタリング
山内有二, 広畑優子, 日野友明, 柳沢一郎, 西川正名, 新野高房, 真空に関する連合講演会講演予稿集, 42nd, 72, 73, 2001年10月17日
日本語

核融合炉ブランケット用SiC/SiC複合材のガス透過特性
橋場正男, 地主孝広, 山内有二, 広畑優子, 日野友明, 加藤雄大, 香山晃, 真空に関する連合講演会講演予稿集, 42nd, 44, 45, 2001年10月17日
日本語

プラズマプロセスによるバブル状酸化シリコンの作製
日野友明, 美保諭志, 山内有二, 広畑優子, 西川正名, 応用物理学関係連合講演会講演予稿集, 48th, 1, 529, 2001年03月28日
日本語

ヘリウムイオン照射によるポーラスシリコン発光のブルーシフト
山内有二, 櫻井敏春, 広畑優子, 日野友明, 西川正名, 応用物理学関係連合講演会講演予稿集, 48th, 3, 1405, 2001年03月28日
日本語

酸素/ヘリウムグロー放電によるシリコンの表面酸化とバブル生成
美保諭志, 山内有二, 広畑優子, 日野友明, 西川正名, 真空に関する連合講演会講演予稿集, 41st, 63, 64, 2000年10月25日
日本語

イオン照射によるシリコン発光特性の変化
山内有二, 桜井敏春, 広畑優子, 日野友明, 西川正名, 真空に関する連合講演会講演予稿集, 41st, 283, 284, 2000年10月25日
日本語

SiC/SiC複合材の重水素リテンション
山内有二, 森修平, 広畑優子, 日野友明, 加藤雄大, 香山晃, 真空に関する連合講演会講演予稿集, 41st, 155, 156, 2000年10月25日
日本語

カーボン・タングステン混合材料の重水素リテンション
平野史生, 山内有二, 広畑優子, 日野友明, 曽我部敏明, 黒田浩二, 真空に関する連合講演会講演予稿集, 40th, 228, 229, 1999年11月10日
日本語

酸素グロー放電によるシリコンの表面酸化
美保諭志, 山内有二, 広畑優子, 日野友明, 西川正名, プラズマ・核融合学会年会予稿集, 16th, 16, 133, 133, 1999年11月
プラズマ・核融合学会, 日本語

シリコン系材料のブリスター生成とエロージョン
山内有二, 広畑優子, 日野友明, 西川正名, プラズマ・核融合学会年会予稿集, 16th, 16, 187, 187, 1999年11月
プラズマ・核融合学会, 日本語

炭素・タングステン混合材料の重水素リテンションと化学エロージョン
平野史生, 山内有二, 広畑優子, 日野友明, 曽我部敏明, 黒田浩二, プラズマ・核融合学会年会予稿集, 15th, 195, 1998年11月
日本語

ヘリウムイオン照射によるシリコン表面のバブル形成
山内有二, 広畑優子, 日野友明, 西川正名, プラズマ・核融合学会年会予稿集, 15th, 275, 1998年11月
日本語

モリブデンの水素リテンション特性
望月寛夫, 山内有二, 広畑優子, 日野友明, 真空に関する連合講演会講演予稿集, 39th, 7, 8, 1998年
日本語

窒素プラズマに対する水素添加効果
土屋農人, 山内有二, 広畑優子, 日野友明, 真空に関する連合講演会講演予稿集, 39th, 77, 78, 1998年
日本語

モリブデンの水素リテンション
望月寛夫, 山内有二, 広畑優子, 日野友明, プラズマ・核融合学会年会予稿集, 14th, 130, 1997年11月
日本語

DIII‐DのDiMES実験によるプラズマ対向材の水素リテンション
山内有二, 広畑優子, 日野友明, 正木圭, 西堂雅博, 安東俊郎, WHYTE D, WONG C, プラズマ・核融合学会年会予稿集, 14th, 131, 1997年11月
日本語

熱粒子制御のためのプラズマ対向壁工学 8.プラズマ対向材料の水素リテンション特性
広畑優子, 山内有二, 日野友明, プラズマ・核融合学会誌, 73, 6, 600, 606, 1997年06月
日本語

リチウム膜の水素リテンション
金谷康, 山内有二, 広畑優子, 日野友明, 真空に関する連合講演会講演予稿集, 38th, 44, 45, 1997年
日本語

リチウム膜の水素リテンション特性
金谷康, 山内有二, 広畑優子, 日野友明, プラズマ・核融合学会秋季講演会予稿集, 13th, 205, 1996年09月
日本語

低Zプラズマ対向材料のヘリウムリテンション特性
山内有二, 日野友明, 広畑優子, 山科俊郎, 日本原子力学会春の年会要旨集, 34th, 462, 1996年03月
日本語

プラズマ対向材料の水素リテンション特性
日野友明, 山内有二, 小山和則, 広畑優子, 山科俊郎, プラズマ・核融合学会年会予稿集, 13th, 159, 1996年03月
日本語

核融合炉用プラズマ対向材料としてのタングステンの水素リテンション特性
小山和則, 山内有二, 広畑優子, 日野友明, 真空に関する連合講演会講演予稿集, 37th, 83, 84, 1996年
日本語

タングステンの水素リテンション特性
小山和則, 山内有二, 広畑優子, 日野友明, 山科俊郎, 核融合エネルギー連合講演会予稿集, 1st, 346, 347, 1995年11月
日本語

プラズマ対向材料中に保持されている水素に対するコンディショニング
山内有二, 小山和則, 金谷康, 広畑優子, 日野友明, 山科俊郎, 北海道大学工学部研究報告, 175, 45, 50, 1995年10月
北海道大学 = Hokkaido University, 日本語

B₄C転化材の水素脱離および水素リテンション特性
山内有二, 広畑優子, 日野友明, 山科俊郎, 秋場真人, 安藤俊郎, プラズマ・核融合学会年会予稿集, 11th, 161, 1994年03月
日本語

水素イオンによるB₄C転化材のエロージョンおよびリテンション特性
山内有二, 広畑優子, 藤田一郎, 日野友明, 山科俊郎, 安東俊郎, 秋場直人, 真空に関する連合講演会講演予稿集, 34th, 207, 208, 1993年
日本語