
トリカイ ユウジ鳥養 祐二教授Yuji Torikai
■研究者基本情報
委員歴
- 2022年04月 - 現在, 核融合炉工学委員会委員, 国立研究開発法人量子科学技術研究開発機構量子エネルギー部門
- 2022年04月 - 現在, 原型炉設計合同特別チーム, 国立研究開発法人量子科学技術研究開発機構量子エネルギー部門
- 2021年06月 - 現在, APLS処理水に係る海域モニタリング専門家会議, 環境省
- 2021年05月 - 2022年03月, 令和3年度核融合炉工学研究委員会委員, 文部科学省(国立研究開発法人量子科学技術研究開発機構)
- 2021年05月 - 2022年03月, 令和3年度核融合炉工学研究委員会委員, 国立研究開発法人量子科学技術研究開発機構
- 2021年04月 - 2022年03月, 原型炉設計合同特別チーム, 国立研究開発法人量子科学技術研究開発機構核融合エネルギー部門
- 2020年05月 - 2021年03月, 令和2年度核融合炉工学研究委員会委員, 文部科学省(国立研究開発法人量子科学技術研究開発機構)
- 2020年04月 - 2021年03月, 原型炉設計合同特別チーム, 国立研究開発法人量子科学技術研究開発機構核融合エネルギー部門
- 2019年04月 - 2020年03月, 原型炉設計合同特別チーム, 国立研究開発法人量子科学技術研究開発機構核融合エネルギー部門
■研究活動情報
受賞
- 2022年11月, プラズマ学融合学会賞 第27回技術進歩賞, 微粒子に蓄積するトリチウムの測定技術開発とJETで生成されたダスト分析への応用, 一般社団法人プラズマ核融合学会
芦川直子、大塚哲平、鳥養祐二、朝倉伸幸、増崎貴
国内学会・会議・シンポジウム等の賞 - 2022年11月, 第27回技術進歩賞, 微粒子に蓄積するトリチウムの測定技術開発とJETで生成されたダスト分析への応用, 一般社団法人プラズマ核融合学会
芦川直子、大塚哲平、鳥養祐二、朝倉伸幸、増崎貴
国内学会・会議・シンポジウム等の賞 - 2021年03月, 第3回日本原子力学会材料部会Best Figure賞, JET-ILWダストに対する燃焼法による残留トリチウム測定。酸化反応による錫燃焼の瞬間, 一般社団法人日本原子力学会
国内学会・会議・シンポジウム等の賞
論文
- 〔主要な業績〕Overview of tritium retention in divertor tiles and dust particles from the JET tokamak with the ITER-LIKE WALL
Y. Torikai; G. Kikuchi; A. Owada1; S. Masuzaki; T. Otsuka; N. Ashikawa; M. Yajima; M. Tokitani; Y. Oya; S. E. Lee; Y. Hatano; N. Asakura; T. Hayashi; M. Oyaidzu; J. Likonen; A. Widdowson; M. Rubel and JET Contributors, 筆頭著者, Divertor tiles after JET-ILW campaigns and dust collected after JET-C and JET-ILW operation were examined by a set of complementary techniques (full combustion and radiography) to determine the total, specific and areal tritium activities, poloidal tritium distribution in the divertor and the presence of that isotope in individual dust particles. In the divertor tiles, the majority of tritium is detected in the surface region and, the areal activities in the ILW divertor are in the 0.5 - 12 kBq cm-2 range. The activity in the ILW dust is associated mainly with the presence of carbon particles being a legacy from the JET-C operation. The total tritium activities show significant differences between the JET operation with ILW and the earlier phase with the carbon wall (JET-C) indicating that tritium retention has been significantly decreased in the operation with ILW., IOP Publishing
Nuclear Fusion, 2023年12月05日, [査読有り] - 核融合プラズマ対向材の残留トリチウム測定法
菊 地 絃 太 *, 芦 川 直 子 , 鳥 養 祐 二, ラスト(シニア)オーサー, 核融合原型炉の実現に向け,核融合炉内に蓄積したトリチウムの定量測定法の確立が必須である.本研究で は,迅速かつ簡便なトリチウムの定量測定法としてオートクレーブ法を提案する.核融合プラズマ曝露試料に対 する測定手法を確立するため,大型ヘリカル装置(LHD)で曝露された試料中に含まれるトリチウムを実測した.,その結果,ボロンや炭素の堆積がある LHD 試料内から得られた残留トリチウム量は1.3-10.3 Bq であり, また全量測定が可能であることを示した.堆積層が十分に薄く無視できる試料では拡散モデルに基づく理論値と オートクレーブ法による測定値で放出挙動が一致したが,堆積層が形成されている試料では堆積の特徴の違いに より理論値と測定値で放出挙動が異なることを明らかにし,これは残留トリチウムが存在する深さに依存するこ とを示唆した., プラズマ・核融合学会
プラズマ・核融合学会誌, 2023年06月, [査読有り] - Tritium distributions in castellated structures of Be limiter tiles from JET-ITER-like wall experiments
S. Lee; Y. Hatano; S. Masuzaki; Y. Oya; M. Tokitani; M. Yajima; T. Otsuka; N. Ashikawa; Y. Torikai; N. Asakura; H. Nakamura; H. Kurotaki; T. Hayashi; T. Nozawa; A.M. Ito; J. Likonen; A. Widdowson; M. Rubel and JET Contributors, Tritium retention in the castellated structure of beryllium limiters used in JET with the ITER-like wall (ILW) during the first (ILW1), third (ILW3) and all three (ILW1-3) campaigns were examined and evaluated. Tritium was deposited on the surfaces inside the castellation grooves together with deuterium, beryllium, oxygen, carbon and small amounts of metallic impurities such as nickel, copper and tungsten. The tritium content after the ILW1 campaign was greater than after the ILW3 campaign. This is attributed to the steadily decreasing amount of carbon impurities in JET from campaign to campaign. The majority of tritium was retained in shallow regions in the grooves, up to 2 mm from the entrance to the gap. It was comparable on all sides of the castellation, i.e. no difference has been detected between the toroidal and poloidal gaps. Secondly, the tritium retention in the gaps was similar on all specimens independent of their position in the tokamak, while the retention on the plasma-facing surfaces clearly depended on the tile position. The tritium deposition patterns in the castellation were also compared with the deuterium distribution determined in earlier studies., IOP Science, IAEA
Nuclear Fusion, 2023年03月08日, [査読有り] - Positron annihilation study of tungsten exposed to low-energy deuterium plasma
Tetsuya Hirade; Hikaru Furuta; Yuji Torikai; Yuki Fujimura; Koji Michishio, Positron annihilation lifetime spectroscopy (PALS) measurements by use of a positron source of 22Na were performed for the polycrystalline tungsten samples that were exposed to low-energy deuterium (D) plasma. The energy of D plasma was low and then it was expected that it would affect just the near-surface region. However, we obtained the longer mean positron annihilation lifetime in the tungsten samples exposed to the low-energy D plasma than the virgin tungsten sample. Moreover, the same longer values were obtained even on the other (no exposed) side of the samples, although the thickness of the samples was 1 mm or 2 mm. The elongation of the positron lifetime probably indicates defects formation in W. Ohsawa, et al. predicted that the high concentration hydrogen creates high concentration vacancy-hydrogen clusters by the first-principles calculations. We also performed positron annihilation age-momentum correlation (AMOC) measurements to know if there are some D’s at the positron trapping sites such as vacancies and found that some of the positrons annihilate with electrons in D., J-STAGE
JJAP Conference Proceedings, 2023年02月13日, [査読有り] - An overview of tritium retention in dust particles from the JET-ILW divertor.
T Otsuka; S Masuzaki; N Ashikawa; Y Torikai; Y Hatano; M Tokitani; Y Oya; N Asakura; T Hayashi; H Tanigawa; Y Iwai; A Widdowson; M Rubel and JET Contributors,Abstract
Tritium (T) retention characteristics in dust collected from the divertor in JET with ITER-like wall (JET-ILW) after the third campaign in 2015–2016 (ILW-3) have been examined in individual dust particles by combining radiography (tritium imaging plate technique) and electron probe micro-analysis. The results are summarized and compared with the data obtained after the first campaign in 2011–2012 (ILW-1). The dominant component in ILW-1 dust was carbon (C) originating from tungsten-coated carbon fibre composite (CFC) tiles in JET-ILW divertor and/or legacy of C dust after the JET operation with carbon wall. Around 85% of the total tritium retention in ILW-1 dust was attributed to the C dust. The retention in tungsten (W) and beryllium (Be) dominated particles was 100 times smaller than the highest T retention in carbon-based particles. After ILW-3 the main component contributing to the T retention was W. The number of small W particles with T increased, in comparison to ILW-1, most probably by the exfoliation and fragmentation of W coatings on CFC tiles though T retention in individual W particles was smaller than in C particles. The detection of only very few Be-dominated dust particles found after ILW-1 and ILW-3 could imply stable Be deposits on the divertor tiles., IOP Publishing Ltd
Physica Scripta, 2022年02月, [査読有り] - Global distribution of tritium in JET with the ITER-like wall
S.E.Lee; Y. Hatano; M.Tokitani; S.Masuzaki; Y.Oya; T.Otsuka; N.Ashikawa; Y.Torikai; N.Asakura; H.Nakamura; K.Isobe; H.Kurotaki; D.Hamaguchi; T.Hayashi; A.Widdowson; S.Jachmich; J.Likonen; M.Rubel; JET contributors, Nondestructive analysis of tritium (T) distribution was performed by means of imaging plate technique on specimens cut from the Be limiters, W-coated carbon tiles and bulk W lamellae retrieved from the JET tokamak after the first and third experimental campaigns with the ITER-like wall. Afterwards, analyses were continued using X-ray photoelectron spectroscopy, microscopy techniques and thermal desorption spectroscopy. Co-deposits formed on the W-coated tiles in the 1st campaign showed large T retention because of high carbon content reaching up to 50 atomic %, while the carbon fraction in co-deposits after the 3rd campaign was distinctly lower. The T retention of the plasma-facing surface of the bulk W tile was smaller than that of the W-coated tiles by a factor of 20, while deposition of small amount of T was found at the side surfaces facing to the gaps in a lamella structure. The correlation of T distributions with surface morphology and the discharge conditions is discussed., ELSEVIA
Nuclear Materials and Energy, 2021年03月, [査読有り] - Investigation on tritium retention and surface properties on the first wall in the Large Helical Device
M. Yajima; S. Masuzaki; N. Yoshida; M. Tokitani; T. Otsuk; Y. Oya; Y. Torikai; G. Motojima; the LHD Experiment Groupa, In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD., ELSEVIA
Nuclear Materials and Energy, 2021年01月20日, [査読有り] - Tritium retention in displacement-damaged tungsten exposed to deuterium-tritium gas mixture at elevated temperatures
V.Kh.Alimov; Y.Torikaia; Y.Hatano; T.Schwarz-Selingerd, W samples, previously irradiated with 20 MeV W ions at room temperature to a displacement-damage level of 0.23 displacements per atom (dpa) at the peak of displacements, were exposed to a deuterium-tritium (D2 ― DT ― T2) gas mixture with a tritium content of 6% at a total pressure of 1.2 kPa and temperatures of 773 K and 973 K for 3 h. The concentration of tritium retained in the displacement-damage zone of these W samples was determined by a method combining chemical etching and subsequent analysis of tritium in the etching solution using liquid scintillation counting (CE-LSC)., ELSEVIER
Fusion Engineering and Design, 2020年11月18日, [査読有り] - Determination of retained tritium from ILW dust particles in JET
N. Ashikawa; Y. Torikai; N. Asakura; T. Otsuka; A. Widdowson; M. Rubel; M. Oyaizu; M. Hara; S. Masuzaki; K. Isobe; Y. Hatano; K. Heinola; A. Baron-Wiechec; S. Jachmich; T. Hayashi and JET, Quantitative tritium inventory in dust particles from campaigns in the JET tokamak with the carbon wall (2007–2009) and the ITER-like wall (ILW 2011–2012) were determined by the liquid scintillation counter and the full combustion method. A feature of this full combustion method is that dust particles were covered by a tin (Sn) which reached 2100 K during combustion under oxygen flow. The specific tritium inventory for samples from JET with carbon and with metal walls was measured and found to be similar. However, the total tritium inventory in dust particles from the ILW experiment was significantly smaller in comparison to the carbon wall due to the lower amount of dust particles generated in the presence of metal walls.
Nuclear Materials and Energy, 2020年01月, [査読有り] - Influence of plasma impurities on the fuel retention in tungsten
A. Kreter; D. Nishijima; R.P. Doerner; M. Freisinger; Ch. Linsmeier; Y. Martynova; S. Möller; M. Rasinski; M. Reinhart; A. Terra; Y. Torikai and B. UnterbergHayashi and JET Contributors
Nuclear Fusion, 2019年06月 - Measurement of tritium in tungsten deposition layer by imaging plate technique after exposure to gaseous tritium
M. Noguchi; K. Katayama; Y. Torikai; N. Ashikawa; A. Taguchi; S. Fukada, © 2017 Elsevier B.V. It is important to understand tritium desorption behavior from plasma-facing materials of a fusion reactor in order to discuss effective tritium recovery method from in-vessel components. However, basic behavior of hydrogen isotopes in W deposition layer is not understood completely. In this study, characterized tungsten deposition layer formed by hydrogen plasma sputtering was exposed to gaseous tritium at 300 °C or 500 °C and tritium desorption behavior by vacuum heating was investigated by the imaging plate technique. For comparison, bare tungsten substrates were exposed to gaseous tritium in the same condition. Initial tritium activity in the deposition layer was much higher than that in the bare substrate. Tritium desorption behavior from tungsten deposition layer was different by the temperature of the layer during tritium exposure process. By heating at 500 °C for 1 h, 97.5% of tritium was desorbed from the layer exposed to tritium at 300 °C. On the other hand, by heating at 500 °C for 2 h, only 44.6% of tritium was desorbed from the layer exposed to tritium at 500 °C. To recover most tritium from W deposition layer and W substrate, heating at above 700 °C is required.
Fusion Engineering and Design, 2017年11月, [査読有り] - The microstructure of tungsten exposed to D plasma with different impurities
M. Rasinski; A. Kreter; Y. Torikai; Ch. Linsmeier
Nuclear Materials and Energy, 2017年08月, [査読有り] - Comparison of TEM and positron annihilation lifetime spectroscopy on tungsten exposed to mixed D plasmas
M. Rasinski; A. Kreter; Y. Torikai; G. P. Karwasz
European Microscopy Congress 2016, 2016年12月, [査読有り] - Desorption behavior of tritiated water from organic functionalized mesoporous silica
A. Taguchi; Y. Kato; R. Akai; Y. Torikai, The desorption and enrichment of tritiated water from organic functionalized mesoporous SBA-15 were investigated. The desorption behavior of tritiated water was dependent on the nature of the organic functional group present i.e., -COOH, -SO3H, and -NH2. Enriched tritiated water was obtained by thermal desorption from-SO3H and-NH2 grafted SBAs after spontaneous desorption of tritiated water at 25 degrees C. Although the volumetric amount of concentrated tritiated water was small, the concentration ratios of tritiated water thermally desorbed and at 25 degrees C were 1.2 and 1.3 for -SO3H and -NH2 grafted SBA, respectively. In contrast, -COOH grafted SBA showed a lower T concentration after thermal desorption. Fourier transform infrared spectroscopy studies using deuterated water (D2O) indicated the presence of strong interactions between -COO- groups and D. (C) 2016 Elsevier B.V. All rights reserved., ELSEVIER SCIENCE SA
Fusion Engineering and Design, 2016年12月, [査読有り] - Tritium desorption and tritium removal from tungsten pre-irradiated with helium
Y. Nobuta; Y. Hatano; Y. Torikai; M. Matsuyama; S. Abe; Y. Yamauchi, © 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.
Fusion Engineering and Design, 2016年01月, [査読有り] - Tritium burning in inertial electrostatic confinement fusion facility
M. Ohnishi; Y. Yamamoto; H. Osawa; Y. Hatano; Y. Torikai; I. Murata; K. Kamakura; M. Onishi; K. Miyamoto; H. Konda; K. Masuda; E. Hotta, © 2015 Elsevier B.V. An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.
Fusion Engineering and Design, 2015年11月, [査読有り] - Gaseous tritium uptake by C deposition layer on tungsten
Y. Hamaji; H.T. Lee; Y. Torikai; K. Sugiyama; Y. Ueda, Abstract Tritium (T) uptake by exposure to gaseous T on deuterated C deposition layer on W was investigated. The C deposition layer were prepared by mixed D and C ion irradiation. The specimens were exposed to D and T mixed gas at 423 and 573 K. The additional T retention by gas exposure at 573 K was about 4 times higher than that by gas exposure at 423 K. Further heating from 573 to 673 K in vacuum after the gas exposure, resulted in more than 70% of retained T stayed after heating at 673 K. That should be due to changing of nature of trap sites in C deposition layer during experiments. These results suggest that additional T trapped in such trap sites require higher temperature to remove retained T in such trap sites.
Journal of Nuclear Materials, 2015年08月, [査読有り] - Tritium release from SS316 under vacuum condition
Y. Torikai and R.-D. Penzhorn, 筆頭著者
Fusion Science and Technology, 2015年04月, [査読有り] - Measurement of Uptake and Release of Tritium by Tungsten
M. Nakayama; a; Y. Torikai; M. Saito; R. -D. Penzhorn; K. Isobe; T. Yamanishi; H. Kurishita
Fusion Science and Technology, 2015年04月, [査読有り] - Tritium Trapping on the Plasma Irradiated Tungsten Surface
Y. Torikai; V. Kh. Alimov; K. Isobe; M. Oyaidzu; T. Yamanishi; R.-D. Penzhorn; Y. Ueda; H. Kurishita; V. Philipps; A. Kreter; M. Zlobinski, 筆頭著者, Tungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.
Fusion Science and Technology, 2015年04月, [査読有り] - Tritium Removal from Tritiated Water by Organic Functionalized SBA-15
A. Taguchi; Y. Kato; R. Akai; Y. Torikai; M. Matsuyama, Mesoporous silicas (SBA-15) were modified by -COOH, -SO3H and -NH2 groups and their tritium adsorption ability from tritiated water under solid-liquid sorption was investigated. The adsorption abilities and separation factor of organic functionalized SBAs were comparable to those of bare SBA. The desorption of water from bare SBA and -COOH functionalized SBA were studied by Fourier transform infra-red spectroscopy using D2O as a probe molecule. An interaction was observed for D2O with -COOH group where the hydrogen bonds became weaker than D2O with bare SBA., AMER NUCLEAR SOC
Fusion Science and Technology, 2015年04月, [査読有り] - Behavior of Tritium in the Vacuum Vessel of JT-60U
Y. Torikai; M. Saito; V. Alimov; N. Miya and Y. Ikeda
Fusion Science and Technology, 2015年03月, [査読有り] - Current status of nanostructured tungsten-based materials development
H. Kurishita; S. Matsuo; H. Arakawa; T. Sakamoto; S. Kobayashi; K. Nakai; H. Okano; H. Watanabe; N. Yoshida; Y. Torikai; Y. Hatano; T. Takida; M. Kato; A. Ikegaya; Y. Ueda; M. Hatakeyama; T. Shikama, Nanostructured tungsten (W)-based materials offer many advantages for use as plasma facing materials and components exposed to heavy thermal loads combined with irradiation with high-energy neutron and low-energy ion. This paper first presents the recent progress in nanostructured toughened, fine grained, recrystallized W materials. Thermal desorption spectrometry apparatus equipped with an ion gun has been installed in the radiation controlled area in our Center at Tohoku University to systematically investigate the effects of displacement damage due to high-energy neutron irradiation on hydrogen isotope retention in connection with the nano- or micro-structures in W-based materials. In this paper, the effects of high-energy heavy ion irradiation on deuterium retention in W with different microstructures are described as a preliminary work with the prospective view of neutron irradiation effects., IOP Science
Physica Scripta, 2014年04月01日, [査読有り] - Surface erosion and modification of toughened, fine-grained, recrystallized tungsten exposed to TEXTOR edge plasma
Y. Ueda; M. Oya; Y. Hamaji; H. T. Lee; H. Kurishita; Y. Torikai; N. Yoshida; A. Kreter; J. W. Coenen; A. Litnovsky; V. Philipps, In order to evaluate the applicability of toughened, fine-grained, recrystallized (TFGR)-W to tokamak edge plasma environment, two TFGR-W specimens (TFGR-W 1.1wt%TiC and TFGR-W 3.3wt%TaC) were exposed to 31 identical ohmic discharges in the TEXTOR tokamak by means of a limiter lock system. The highest surface temperature reached was about 1300 °C. Under these temperature conditions, the bulk microstructure and dispersoids distribution of both TFGR-W remained intact, suggesting that these TFGR-W tungsten materials have sufficient stability under these plasma loading conditions. The erosion of TiC dispersoids on the surface was enhanced by plasma exposure above 1150 °C, while such enhanced erosion was not observed for TaC dispersoids probably due to the higher melting temperature of Ta than Ti. © 2014 EURATOM., IOP Sience
Physica Scripta, 2014年04月01日, [査読有り] - Tritium trapping behavior in tungsten pre-irradiated with D, He, Ar and N plasmas
Y Hamaji; Y Torikai; H T Lee; Y Otsuka; Y Ueda, Tritium (T) trapping in tungsten after plasma exposure (deuterium (D), helium (He), nitrogen (N), argon (Ar)) was studied using an imaging plate technique. Specimens were exposed to D and T mixed gas at 77 and 573 K to distinguish T trapped at the outermost surface and several tens of nanometers, respectively. He followed by N, Ar and D plasma exposures resulted in the largest increase in T trapping. This was interpreted to result from He bubble layers that can increase the surface area by formation of pores connected to the surface and/or an increase in surface trapping sites. T exposure at 77 K was found to be a very useful method to observe plasma-induced changes to T trapping at the outermost surface., IOP Science
Physica Scripta, 2014年04月01日, [査読有り] - Enrichment of tritiated water using mesoporous silica
A. Taguchi; Y. Kato; Y. Torikai; M. Matsuyama; S. Uchida, A. Taguchi, Y. Kato, Y. Torikai, M. Matsuyama, S. Uchida, Elsevia
Microporous and Mesoporous Materials, 2013年09月15日, [査読有り] - Tritium Interaction with Surface Layer and bulk of Type 316 Stainless Steel and Consequences of Aging
R.-D. Penzhorn; Y. Hatano; M. Matsuyama; Y. Torikai, ラスト(シニア)オーサー, Stainless steel exposed to gaseous tritium characteristically shows a firmly trapped fraction of tritium in the surface layer, which is not fully removable by water at ambient temperature. Prolonged thermal treatment of tritium-loaded specimens at <443 K causes substantial depletion of the bulk but almost no depletion of the surface layer. For complete removal of hydrogen isotopes from the bulk and the surface, temperatures exceeding 573 K are necessary. Upon chemical etching virtually all tritium trapped in the surface layer appears in the etching solution as tritiated water. Following removal of the layer by chemical etching, the tritium-rich layer reappears after months of aging at ambient temperature with nearly the original tritium activity. Comparison of chronic tritium release rates into liquid water before and after etching reveals that the surface layer only marginally influences the rate. X-ray photoelectron spectroscopy provides evidence that during prolonged aging the surface layer continues to grow while at the same time trapping a fraction of bulk tritium released at ambient temperature. Experimental results suggest different mechanisms of hydrogen uptake and release by the bulk and surface layers. Inference of tritium activity in the bulk of aged or heat-exposed stainless steel material from surface activity measurements may depart significantly from reality., ANS
Fusion Since and Technology, 2013年07月, [査読有り] - Tritium loading study of tungsten pre-exposed to TEXTOR plasmas
Y. Torikai; A. Taguchi; M. Saito; R.-D. Penzhorn; Y. Ueda; H. Kurishita; K. Sugiyama; V. Philipps; A. Kreter; M. Zlobinski; TEXTOR team, 筆頭著者, The uptake of tritium at 573 K by polycrystalline tungsten limiter material pre-exposed to TEXTOR plasmas was investigated by the imaging plate technique (IP) and found to be mostly non-homogeneous on the plasma facing surface. Particularly high concentrations of tritium were apparent in areas attributed to carbon deposition. The surface density of tritium outside of deposition zones was essentially comparable on both sides of the examined tungsten plate. Under a stream of argon at ambient temperature tritium was predominantly released as tritiated water. While tritium is initially liberated with rates in the (MBq/h) range, after a few days the rate drops to about 100 Bq/h, decreasing even further thereafter. Under atmospheric conditions the concentration of tritium on the surface remained virtually unchanged over a rather extended period of time, i.e. more than 500 d. Tritium in surface zones other than of "deposition" was also firmly trapped at ambient temperature. © 2013 Elsevier B.V. All rights reserved., Elsevia
Journal of Nuclear Materials, 2013年07月, [査読有り] - Deuterium retention in Toughened, Fine-Grained Recrystallized Tungsten
M. Oya; K. Uekita; H. T. Lee; Y. Ohtsuka; Y. Ueda; H. Kurishita; A. Kreter; J. W. Coenen; V. Philipps; S. Brezinsek; A. Litnovsky; K. Sugiyama; Y. Torikai, ラスト(シニア)オーサー, M. Oya, K. Uekita, H.T. Lee, Y. Ohtsuka, Y. Ueda, H. Kurishita, A. Kreter, J.W. Coenen, V. Philipps, S. Brezinsek, A. Litnovsky, K. Sugiyama, Y. Torikai, Elsevia
Journal of Nuclear Materials, 2013年07月, [査読有り]
講演・口頭発表等
- JET-ダイバータタイルのトリチウム分析
鳥養祐二、菊地絃太、○大和田篤志、増崎貴、大塚哲平、芦川直子、矢嶋美幸、時谷政行、大矢恭久、S.E.Lee、波多野雄治、朝倉伸行、林巧、小柳津誠、J.Likonen、A.Widdowson、M.Rubel
日本放射化学会第66回討論会()2022, 2022年09月16日, 日本放射化学会
20220915, 20220917 - 環境トリチウムの迅速測定法の開 発,( 魚中トリチウム濃 度の迅速なスクリーニング法の開 発
細根 孟留、 南 場 大 輝、川上 智 彦、柿内 秀 樹、鳥養 祐 二
日本原子力学会2022年秋の大会, 2022年09月08日, 日本原子力学会
20220907, 20220909 - 環境トリチウムの迅速測定法の開発,( 環境トリチウム測定の簡素化・迅速化
南場 大輝、細根 孟留、川上 智彦、柿内 秀樹、鳥養 祐二
日本原子力学会2022年秋の大会, 2022年09月08日, 日本原子力学会
20220907, 20220909 - 〔主要な業績〕海水中トリチウムの 測定法
鳥養 祐二; 本間 駿太
第3回 日本放射線安全管理学会 日本保健物理学会 合同大会, 2021年12月01日
20211201, 20211203 - Tritium retention in dust particles and divertor tiles of JET operated with the ITER-Like Wall
Y. TORIKAI; G. KIKUCHI; A. OWADA; S. MASUZAKI; T. OTSUKA; N. ASHIKAWA; M. YAJIMA; M. TOKITANI; Y. OYA; S. E. LEE; Y. HATANO; N. ASAKURA; T. HAYASHI; M. OYAIDZU; A. WIDDOWSON; M. RUBEL and JET Contributors
28th IAEA Fusion Energy Conference, 2021年05月14日
20210510, 20210515 - 核融合炉材料からの大気圧下でのトリチウムの放出
鳥養祐二
プラズマ・核融合学会 第37回年会, 2020年12月02日 - 原型炉のトリチウム管理基準の考え方
鳥養祐二
プラズマ核融合学会 第36回年会, 2019年11月30日
20191129, 20191202
共同研究・競争的資金等の研究課題
- プラズマ対向壁への水素同位体の蓄積量の定量手法の開発と定量評価
核融合科学研究所
2023年04月 - 2024年03月 - 魚介類トリチウム濃度の迅速測定法に関するフィジビリティスタディ
国立研究開発法人日本原子力研究開発機構
2022年06月 - 2023年03月 - 魚中のトリチウム濃度測定法の簡素化・最適化
2022年04月 - 2023年03月 - LHD真空容器内への水素同位体の蓄積量の定量手法の開発と定量評価
2022年04月 - 2023年03月 - 核融合実現に向けたトリチウム諸課題の検討
2022年04月 - 2023年03月 - LHD真空容器内への水素同位体の蓄積量の定量手法の開発と定量評価
2021年04月 - 2022年03月 - 真空下での効果的なトリチウム除染
2021年04月 - 2022年03月 - 核融合炉材料中のトリチウム移行挙動
2021年04月 - 2022年03月 - プラズマ対向材料におけるプラズマ駆動透過に及ぼす水素同位体効果
2021年04月 - 2022年03月 - LHDにおけるプラズマ・壁相互作用に関する研究会
2021年04月 - 2022年03月 - 水素同位体の挙動と機能および将来像
2021年04月 - 2022年03月 - 核融合実現に向けたトリチウム諸課題の検討
2021年04月 - 2022年03月 - JET ILW実験におけるプラズマ対向機器表面およびダストのトリチウム蓄積特性研究
2021年06月 - 2022年01月 - 金属壁の水素同位体置換に関する基礎実験
2020年04月 - 2021年03月 - 原型炉におけるトリチウムの課題
2020年04月 - 2021年03月 - 核融合炉材料中のトリチウム移行挙動
2020年04月 - 2021年03月 - 核融合原型炉二次冷却系におけるトリチウム挙動
2020年04月 - 2021年03月 - プラズマ対向材料におけるプラズマ駆動透過に及ぼす水素同位体効果
2020年04月 - 2021年03月 - LHDにおけるプラズマ・壁相互作用に関する研究会
2020年04月 - 2021年03月 - LHD真空容器内への水素同位体の蓄積量の定量評価
2020年04月 - 2021年03月 - JET ILW実験におけるプラズマ対向機器表面およびダストのトリチウム蓄積特性研究
2020年06月 - 2021年01月
社会貢献活動
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2022年11月01日 - 第1回 ALPS処理水モニタリングシンポジウム
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メディア報道
- トリチウムの迅速測定
NHK, NHK NEWS おはよう日本, 2023年08月08日, テレビ・ラジオ番組 - リスクを見つめる トリチウム水道水にも
読売新聞, 2023年06月27日, 新聞・雑誌 - 韓国で塩"品切れ"背景に「雨」と「原発」福島原発の処理水放出への”不安”
日本テレビ, ニュースゼロ, 2023年06月21日, テレビ・ラジオ番組 - ゴジてれChu!
福島中央テレビ, ゴジてれChu!, 2020年03月18日, テレビ・ラジオ番組 - ひるおび
TBSテレビ, ひるおび, 2019年10月03日, テレビ・ラジオ番組 - NEWS every
日本テレビ, NEWS every, 2019年10月03日, テレビ・ラジオ番組 - 深層NEWS
日本テレビ, 深層NEWS, 2019年09月26日, テレビ・ラジオ番組 - ミヤネ屋
読売テレビ, ミヤネ屋, 2019年09月25日, テレビ・ラジオ番組 - サタデーステーション
テレビ朝日, サタデーステーション, 2019年09月21日, テレビ・ラジオ番組