Masato NAKAYAMA¹⁾, Masanori HARA¹⁾, Masao MATSUYAMA¹⁾, and Kiyokazu HIROKAMI<sup>2)

1)</sup>Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama
²⁾Radioisotope Laboratory, Research and Development in Natural Sciences Center, Administration Center for Promotion of Research, Organization for Promotion of Research, University of Toyama, 3190 Gofuku Toyama, Toyama Pref. 930-8555, Japan

Abstract
A modified integral counting method (MICM) with various quenched samples (MICM-VQ) has been investigated for its applicability for different scintillators using β emitters, ¹⁴C and ³⁵S. To assess the influence of scintillators, three sets of ¹⁴C quenched standards and two ³⁵S cocktail series were prepared. Two sets of ¹⁴C quenched standards were used for the toluene-compatible scintillator, the other for the Ultima Gold^TM scintillator. Sulfur-35 cocktail series were prepared with either EcoscintTM XR or Ultima Gold^TM AB. The radioactivity of these samples was determined using the MICM-VQ, with the results conforming to assayed values. Hence the MICM-VQ can assay the radioactivity of sample cocktails with various scintillators and requires no standard sample.

Keywords: liquid scintillation counting, modified integral counting method, no quenched standard sets, different scintillators, point of convergence
Accepted: 27 December 2016

M. Tokitani^a, M. Miyamoto^b, S. Masuzaki^a, Y. Fujii^b, R. Sakamoto^a, Y. Oya^c, Y. Hatano^d, T. Otsuka^e, M. Oyaidzu^f, H.Kurotaki^f, T. Suzuki^f, D. Hamaguchi^f, K. Isobe^f, N. Asakura^f, A. Widdowson^g, M. Rubel^h, JET Contributors<sup>1

a</sup>National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan
^bShimane University, Matsue, Shimane 690-8504, Japan
^cShizuoka University, Shizuoka 422-8529, JapandUniversity of Toyama, Toyama 930-8555, Japan
^dUniversity of Toyama, Toyama 930-8555, Japan
^eKindai University, Higashi-Osaka, Osaka, 577-8502, Japan
^fNational Institute for Quantum and Radiological Science and Technology (QST), Rokkasho Aomori 039-3212, Japan
^gEUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB, UK
^hRoyal Institute of Technology (KTH), 100 44 Stockholm, Swede

Abstract

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

Keywords: JET ILW divertor, TEM observation, Deposition, Erosion
Accepted: 4 January 2017

上田良夫，波多野雄治¹⁾，横峯健彦²⁾，檜木達也²⁾，長谷川晃³⁾，大矢恭久⁴⁾，室賀健夫⁵⁾UEDA Yoshio, HATANO Yuji¹⁾, YOKOMINE Takehiko²⁾, HINOKI Tatsuya²⁾, HASEGAWA Akira³⁾,OYA Yasuhisa⁴⁾and MUROGA Takeo⁵⁾
大阪大学，¹⁾富山大学，²⁾京都大学，³⁾東北大学，⁴⁾静岡大学，⁵⁾核融合科学研究所

Keywords: divertor, DEMO reactors, plasma facing component, tungsten, neutron irradiation, tritium retention, helium coolant

Accepted: 20 December 2016

檜木達也，長谷川 晃¹⁾，福田 誠¹⁾，田中照也²⁾，大矢恭久³⁾，波多野雄治⁴⁾，上田良夫⁵⁾
HINOKI Tatsuya, HASEGAWA Akira¹⁾, FUKUDA Makoto¹⁾, TANAKA Teruya²⁾, OYA Yasuhisa³⁾,HATANO Yuji⁴⁾and UEDA Yoshio⁵⁾

京都大学，¹⁾東北大学，²⁾核融合科学研究所，³⁾静岡大学，⁴⁾富山大学，⁵⁾大阪大学

Abstract

米国のオークリッジ国立研究所にある研究用原子炉HFIR を用いたタングステン材料の中性子照射試験を 行った．熱中性子による核変換の影響を抑制するため，熱中性子遮蔽を検討しガドリニウムによる熱中性子遮蔽 キャプセルを開発した．ガドリニウムによる熱中性子遮蔽効果や照射キャプセルの熱分布のモデリングを行い， これまでに知見が不足している高温領域で比較的高い線量での中性子照射試験を実施することができた．

Keywords: neutron irradiation, tungsten, thermal neutron shielding, gadolinium, high temperature, high fluence
Accepted: 20 December 2016

大矢恭久，波多野雄治¹⁾，片山一成²⁾，山内有二³⁾，信太祐二³⁾，大塚哲平⁴⁾， 近田拓未，原 正憲¹⁾，大宅 諒⁵⁾，上田良夫⁵⁾，外山 健⁶⁾
OYA Yasuhisa, HATANO Yuji¹⁾, KATAYAMA Kazunari²⁾, YAMAUCHI Yuji³⁾, NOBUTA Yuji³⁾, OTSUKA Teppei¹⁾, CHIKADA Takumi, HARA Masanori¹⁾, OYA Makoto, UEDA Yoshio⁵⁾ and TOYAMA Takeshi⁶⁾

静岡大学，¹⁾富山大学，²⁾九州大学，³⁾北海道大学，⁴⁾近畿大学，⁵⁾大阪大学，⁶⁾東北大学

Abstract

原型炉プラズマ対向材料は極めて高いフラックスのD/Tプラズマに高温で長時間曝露されることから，大量 のトリチウムが材料中に蓄積するとともに，冷却材へ透過することが懸念される．PHENIX 計画タスク３では原 型炉を想定した高温下における中性子照射タングステン（W）のトリチウム滞留と透過挙動を明らかにすること を目的に研究を進めており，これまでに，高温で鉄イオンを照射した場合には，室温で照射したのち同じ温度で アニール（熱処理）した場合に比べ，重水素の滞留量が大きく減少することを明らかにした．これは，高温照射 中の欠陥のダイナミックな回復によるものである．水素同位体透過挙動では，温度によりトリチウムの拡散経路 が変化することが実験的に初めて示された．また，鉄イオン照射W やヘリウム（He）イオン照射W では低温で 透過率が低下することが示唆された．現在，オークリッジ国立研究所で中性子照射を行っており，今後は中性子 照射材中の水素同位体滞留・拡散・透過挙動をさらに詳細に明らかにしていく計画である．

Keywords: tritium, neutron irradiation, tungsten, plasma irradiation, retention, permeation, PWI

Accepted: 20 December 2016

上田良夫，波多野雄治¹⁾，横峯健彦²⁾，檜木達也²⁾，長谷川晃³⁾，大矢恭久⁴⁾，室賀健夫⁵⁾ UEDA Yoshio, HATANO Yuji¹⁾, YOKOMINE Takehiko²⁾, HINOKI Tatsuya²⁾, HASEGAWA Akira³⁾, OYA Yasuhisa⁴⁾and MUROGA Takeo⁵⁾

大阪大学，¹⁾富山大学，²⁾京都大学，³⁾東北大学，⁴⁾静岡大学，⁵⁾核融合科学研究所

Accepted: 20 December 2016

Yasuhisa Oya^a, Cui Hu^b, Hiroe Fujita^a, Kenta Yuyama^a, Shodai Sakurada^a, Yuki Uemura^a, Suguru Masuzaki^c, Masayuki Tokitani^c, Miyuki Yajima^c, Yuji Hatano^d, and Takumi Chikada^a

^aShizuoka University, Graduate School of Science and Technology, Shizuoka 422-8529, Japan
^bShizuoka University, Faculty of Science, Shizuoka 422-8529, Japan
^cNational Institute for Fusion Science, Gifu 509-5292, Japan
^dUniversity of Toyama, Hydrogen Isotope Research Center, Toyama 930-8555, Japan

Abstract

　All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s⁻¹ with Ar flow rate of 13.3 sccm.
　Using this HI-TDS system, D and T desorption behaviors for D₂⁺ implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trapped by the carbon-dominated mixed-material layer. By D₂⁺ implantation, major D desorption was found at ~900 K with a narrow peak due to energetic ion implantation. For gas exposure, H was preferentially replaced by D and T with a lower trapping energy. In addition, T replacement rate by additional H₂ gas exposure was evaluated. This fact indicates that the hydrogen replacement mechanism would be clearly changed by exposure methods.

Keywords: Simultaneous H, D, T measurement; thermal desorption spectroscopy; tungsten; Large Helical Device.

Accepted: 3 October 2016

Yuki Uemura^a, Shodai Sakurada^a, Hiroe Fujita^a, Keisuke Azuma^a, Quilai Zhou^a, Yuji Hatano^b, Naoaki Yoshida^c, Hideo Watanabe^c, Makoto Oyaizu^d, Kanetsugu Isobe^d, Masashi Shimada^e, Dean Buchenauer^f, Robert Kolasinski^f, Takumi Chikada^a, Yasuhisa Oya^a

^aGraduate School of Science ' Technology, Shizuoka University, 836 Ohya, Suruga, Shizuoka, 422-8529 Japan
^bHydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, 930-8555 Japan
^cInstitute for Applied Mechanics, Kyushu University, 6-1 Kasuga-koen, Kasuga, Fukuoka, 816-8580 Japan
^dNational Institutes for Quantum and Radiological Science and Technology, 2166 Obuchi, Rokkasho, Aomori, 039-3212 Japan
^eIdaho National Laboratory, 1955 N. Fremont Avenue, Idaho Falls, ID 83415 USA
^f Sandia National Laboratories, Chemistry, Combustion and Materials Center, Livermore, CA 94550 USA

Abstract

In this study, we measured deuterium (D) gas-driven permeation through tungsten (W) foils that had been pre-damaged by helium ions (He⁺). The goal of this work was to determine how ion-induced damage affects hydrogen isotope permeation. At 873 K, the D permeability for W irradiated by 3.0 keV He⁺ was approximately one order of magnitude lower than that for un-damaged W. This difference diminished with increasing temperature. Even after heating to 1173 K, the permeability returned to less than half of the value measured for un-damaged W. We propose that this is due to nucleation of He bubbles near the surface which potentially serve as a barrier to diffusion deeper into the bulk. Exposure at higher temperatures shows that the D permeability and diffusion coefficients return to levels observed for undamaged material. It is possible that these effects are linked to annealing of defects introduced by ion damage, and whether the defects are stabilized by the presence of trapped He.
Keywords: Tungsten, Hydrogen isotope permeation, Helium irradiation, Helium bubble

Accepted: 22 April 2017

Masanori Hara^a, Shinsuke Abe^a, Masao Matsuyama^a, Tsukasa Aso^b, Katsuyoshi Tatenuma^c, Tomohiko Kawakami^c, Takeshi Ito^c

^aHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama City, Toyama 930-8555, Japan
^bElectronics and Computer Engineering, National Institute of Technology, Toyama College, 1-2 Ebie-neriya, Imizu City, Toyama 933-0293, Japan
^cKAKEN Company Limited, 1044 Horimachi, Mito City, Ibaraki 310-0903, Japan

Abstract

One of the methods used for tritium gas analysis is beta-ray induced X-ray spectrometry (BIXS). Gascell design is important in this method. The structure of the gas cell for BIXS was optimized by MonteCarlo simulation of beta-ray induced X-ray spectra in various window geometries using the Geant4 toolkit (version 10.01.p02). The simulated spectrum from tritium decay fitted the observed one, and thesimulation model was used to obtain the cell parameters for BIXS. The optimum thickness of the goldlayer on a beryllium window was around 150 nm. This simulation model also considered the relationshipbetween self-absorption by hydrogen gas and the cell length. Self-absorption increased with increasingcell length and the relationship between the sample pressure and cell length was formulated.

Keywords: Tritium, Beta-ray induced X-ray spectrometry, Bremsstrahlung, Monte Carlo simulation, Geant4

Accepted: 13 April 2017

Masanori Hara^a, Haruna Sakaguchi^a, Masato Nakayama^a, Shinsuke Abe^a, Masao Matsuyama^a, Takayuki Abe^a, and Tsukasa Aso^b

^aUniversity of Toyama, Hydrogen Isotope Research Center, 3190 Gofuku, Toyama City, Toyama 930-8555, Japan
^bNational Institute of Technology, Toyama College, Electronics and Computer Engineering, 1-2 Ebie-neriya, Imizucity, Toyama 933-0293, Japan

Abstract

The luminescence of Eu(DPA)₃^3- induced by beta particles from tritium decay was measured. The solution of Eu³⁺ was prepared with europium(III) nitrate hexahydrate and was mixed with a DPA (2, 6- pyridinedicarboxylic acid or dipicolinic acid) solution of pH 11 to yield Eu(DPA)₃^3-. The formation of Eu (DPA)₃^3- was confirmed through spectrometry. Tritiated water was added to the prepared solution of Eu (DPA)₃^3-. The luminescence intensity is proportional to the amount of tritium. In this paper we demonstrate the potential of this Eu complex as an inorganic liquid scintillator.

Keywords — Luminescence, dipicolinic acid, tritiated water, liquid scintillation

Accepted: 1 August 2016

K. Ogawa^1,2, T. Nishitani¹, M. Isobe^1,2, I. Murata³, Y. Hatano⁴, S. Matsuyama⁵, H. Nakanishi^1,2, K. Mukai^1,2, M. Sato¹, M. Yokota¹, T. Kobuchi¹, T. Nishimura¹and M. Osakabe<sup>1,2

1</sup>National Institute for Fusion Science, National Institutes of Natural Sciences, Toki, 509-5292, Japan
²SOKENDAI (The Graduate University for Advanced Studies), Toki 509-5292, Japan
³Osaka University, Yamada-oka 2-1, Suita, Osaka 565-0871, Japan
⁴University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
⁵Tohoku University, Aramaki-Aza-Aoba 01, Aoba-ku, Sendai 980-8579, Japan

Abstract
High-temperature and high-density plasmas are achieved by means of real-time control, fast diagnostic, and high-power heating systems. Those systems are precisely controlled via highly integrated electronic components, but can be seriously affected by radiation damage. Therefore, the effects of irradiation on currently used electronic components should be investigated for the control and measurement of Large Helical Device (LHD) deuterium plasmas. For the precise estimation of the radiation field in the LHD torus hall, the MCNP6 code is used with the cross-section library ENDF B-VI. The geometry is modeled on the computer-aided design. The dose on silicon, which is a major ingredient of electronic components, over nine years of LHD deuterium operation shows that the gamma-ray contribution is dominant. Neutron irradiation tests were performed in the OKTAVIAN at Osaka University and the Fast Neutron Laboratory at Tohoku University. Gamma-ray irradiation tests were performed at the Nagoya University Cobalt-60 irradiation facility. We found that there are ethernet connection failures of programmable logic controller (PLC) modules due to neutron irradiation with a neutron flux of 3 × 10⁶ cm^-2 s^-1. This neutron flux is equivalent to that expected at basement level in the LHD torus hall without a neutron shield. Most modules of the PLC are broken around a gamma-ray dose of 100 Gy. This is comparable with the dose in the LHD torus hall over nine years. If we consider the dose only, these components may survive more than nine years. For the safety of the LHD operation, the electronic components in the torus hall have been rearranged.

Keywords: large helical device, neutron irradiation, gamma-ray irradiation, irradiation on electronic components
Accepted: 17 May 2017

Satoshi Akamaru ^a, Li Jin^b, Katsuhiko Nishimura^b, Masanori Hara^a, Takayuki Abe^a, Masao Matsuyama^{a a}Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bFaculty of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract
A new hydrogen sensor utilizing a ferromagnetic hydrogen absorbing alloy was developed. An optimum sensing element, Cu particles coated with Pd–Co hydrogen absorbing alloy was prepared by the barrel sputtering technique. The surface of prepared Cu particle was covered uniformly by Pd–Co thin layer constituted of aggregated nanoparticles. The sensitivity of the sensing element to H₂ concentrations under flowing dry N₂ and dry air gases was examined. The element has a reasonable sensitivity to the H₂ concentration of the range from 3.8％ to 0.2％, and the lower limit of detectable H₂ concentration was estimated to be less than 0.1％. In dry air, the water formation on the Pd–Co surface affected its sensing ability, because the temperature of the sensing element increased by the exothermic reaction. The effect of moisture on the H₂ sensing ability was also investigated. The moisture slightly degraded the output signal under flowing air. It could be ascribed to an additional consumption of hydrogen atoms by water molecules and oxygen atoms on the Pd–Co surface. This sensor takes advantage of magnetic susceptibility measurement, which requires no electrical wire between the sensing element and an electric circuit, leading to a safe evaluation system of H₂ concentration in air.

Keywords: Hydrogen sensor, Magnetic susceptibility, Pd–Co alloy
Accepted: 17 May 2017

V.Kh. Alimov^abc, Y. Hatano^a, M. Balden^d, M. Oyaizu^d, K. Isobe^e, H. Nakamura^e, T.Hayashi^e
aHydrogen Isotope Research Center, Organization for Promorion of Research, University of Toyama, Toyama 930-8555, Japan
^bA.N.Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russia
^cNational Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow 115409, Russia
^dMax-Planck-Institut für Plasmaphysik. D-85748 Garching, Germany
^eNational Institutes for Quantum and Radiological Science and Technology, Rokkasho 039-3212, Japan

Abstract
Four sets of targets were used in this study: (1) Fe targets surrounded with 304 type stainless steel composed of mid-Z elements: Fe, Cr, Ni, and Mn (designated as Fe[304SS] targets), (2) Fe targets surrounded with high-Z tungsten (designated as Fe[W] targets), (3) Cu targets surrounded with mid-Z copper (designated as Cu[Cu] targets), and (4) Cu targets surrounded with high-Z tungsten (designated as Cu[W] targets). The targets were exposed to low-energy (140 and 200 eV), high-flux (about 10²² D/m²s）deuterium (D) plasmas at various temperatures in the range from 355 to 740 K. The surface morphology of the Fe and Cu targets is found to be dependent strongly on atomic number of re-deposited species and on the exposure temperature. For the Fe[W] and Cu[WJ targets, due to formation of the W-enriched nano-sized structures on the target surfaces, the sputtering erosion yield is lower than that for the Fe[304SS] and Cu[Cu] targets, respectively. For the Fe[304SS], Fe[W], and Cu[W] targets, the sputtering erosion yield is increased distinctly as the exposure temperature rises from 355 to 740 K.

Keywords:Deuterium plasma, Iron, Copper, Tungsten, Surface morphology, Sputtering erosion
Accepted: 6 June 2017

Yutai Katoh^a, Daniel Clark^b, Yoshio Ueda^c, Yuji Hatano^d, Minami Yoda^e, Adrian S. Sabau^a, Takehiko Yokomine^f, Lauren M. Garrison^a, J. Wilna Geringer^a, Akira Hasegawa^g, Tatsuya Hinoki^f, Masashi Shimada^h, Dean Buchenauerⁱ, Yasuhisa Oya^j and Takeo Muroga^k
aOak Ridge National Laboratory, Oak Ridge, Tennessee
^bUnited States Department of Energy, Germantown, Maryland
^cOsaka University, Osaka, Japan
^dToyama University, Toyama, Japan
^eGeorgia Institute of Technology, Atlanta, Georgia
^fKyoto University, Kyoto, Japan
^gTohoku University, Sendai, Japan
^hIdaho National Laboratory, Idaho Falls, Idaho
ⁱSandia National Laboratory, Livermore, California
^jShizuoka University, Shizuoka, Japan
^kNational Institute for Fusion Science, Toki, Japan

Abstract
The PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the response to high heat loads and tritium interactions of irradiated and unirradiated materials and components. This paper provides an overview of the progress achieved during the first 3 years and discusses the plan for the remainder of the project.

Keywords: Plasma facing components, helium-cooled divertor, tungsten armor.
Accepted: December 30, 2016

Yasuhisa Oya¹, Keisuke Azuma¹, Akihiro Togari¹, Qilai Zhou¹, Yuji Hatano², Masashi Shimada³, Robert Kolasinski⁴and Dean Buchenauer<sup>4

1</sup>Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
²Hydrogen Isotope Research Center, Organization of Promotion of Research, University of Toyama, Gofuku, Toyama 930-8555, Japan
³Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
⁴Chemistry, Combustion and Materials Center, Sandia National Laboratory, Livermore, CA 94550, USA

Abstract
This paper reviews recent achievement of hydrogen isotope behavior for damaged tungsten. To demonstrate neutron irradiation, the irradiation damages were introduced into W by energetic Fe²⁺ irradiation and D retention behavior was examined by thermal desorption spectroscopy (TDS). The D trapping behavior was evaluated using Hydrogen Isotope Diffusion and Trapping (HIDT) code. It was found that D trapping states consisted of two-four stages with the trapping energy of 0.60 eV, 0.85 eV, 1.15-1.25 eV and 1.55 eV depending on the damage concentration and distribution. Based on these experimental results, the hydrogen isotope retention behavior in actual fusion condition was demonstrated. It was found that most of hydrogen isotope was retained in tungsten wall even if the wall temperature was kept at operation temperature.

Keywords: First keyword, Second keyword, Third keyword
DOI: https://doi.org/10.1007/978-3-319-67459-9_6

Y Hatano¹, K Yumizuru¹, S Koivuranta², J Likonen², M Hara¹,M Matsuyama¹, S Masuzaki³, M Tokitani³, N Asakura⁴, K Isobe⁴, T Hayashi⁴, A Baron-Wiechec⁵, A Widdowson⁵ and JET contributors⁶
EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom

¹Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
²VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT, Finland
³National Institute for Fusion Science, Toki 509-5292, Japan
⁴National Institutes for Quantum and Radiological Science and Technology, Rokkasho 039-3212, Japan
⁵Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom

Abstract
Energy spectra of β-ray induced x-rays from divertor tiles used in ITER-like wall campaigns of the Joint European Torus were measured to examine tritium (T) penetration into tungsten (W) layers. The penetration depth of T evaluated from the intensity ratio of W(Lα) x-rays to W(Mα) x-rays showed clear correlation with poloidal position; the penetration depth at the upper divertor region reached several micrometers, while that at the lower divertor region was less than 500 nm. The deep penetration at the upper part was ascribed to the implantation of high energy T produced by DD fusion reactions. The poloidal distribution of total x-ray intensity indicated higher T retention in the inboard side than the outboard side of the divertor region.

Keywords: nuclear fusion, tritium, divertor, retention, tungsten, implantation, Joint European Torus
Accepted for publication : 30 August 2017

Motonori Watanabe^a,∗, Yuki Honda^b, Hidehisa Hagiwara^c, Tatsumi Ishihara^a,<sup>d

a</sup>International Institute for Carbon-Neutral Energy Research (I2CNER),Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan
^bDepartment of Chemistry, Biology and Environmental Science, Faculty of Science, Nara Women’s University, Kitauoyanishi-machi, Nara 630-8506
Japan
^cHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
^dDepartment of Applied Chemistry, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka, 819-0395, Japan

Abstract
This study focuses on [FeFe]-hydrogenase and its metallorganic mimics in terms of electronic and photophysical properties, which can be applied to the electrochemical and/or photochemical production of molecular hydrogen. Natural [FeFe]-hydrogenase, synthetic mimics of its active site and recent progresses in hybrid-type hydrogen production, for example, inorganic-combination photoelectrochemical and photochemical hydrogen production, are reviewed.

Keywords: [FeFe]-Hydrogenase, Artificial and bioengineering science, Hydrogen production, Photochemical reaction, Photo-electrochemical reaction
Accepted for publication : 15 September 2017

T. Toyama ^{a, *}, K. Ami ^b, K. Inoue ^a, Y. Nagai ^a, K. Sato ^c, Q. Xu ^c, Y. Hatano <sup>d

a</sup> Institute for Materials Research, Tohoku University, Ibaraki 311-1313, Japan
^b Graduate School of Science and Engineering for Education, University of Toyama, Toyama 930-8555, Japan
^c Research Reactor Institute, Kyoto University, Osaka 590-0494, Japan
^d Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan

Abstract
Deuterium trapping at irradiation-induced defects in tungsten, a candidate material for plasma facing components in fusion reactors, was revealed by positron annihilation spectroscopy. Pure tungsten was electron-irradiated (8.5 MeV at ~373 K and to a dose of ~1 ×10^-3 dpa) or neutron-irradiated (at 573 K to a dose of ~0.3 dpa), followed by post-irradiation annealing at 573 K for 100 h in deuterium gas of ~0.1 MPa. In both cases of electron- or neutron-irradiation, vacancy clusters were found by positron lifetime measurements. In addition, positron annihilation with deuterium electrons was demonstrated by coincidence Doppler broadening measurements, directly indicating deuterium trapping at vacancytype defects. This is expected to cause significant increase in deuterium retention in irradiated-tungsten.


Keywords:Hydrogen trapping at defects, Tungsten, Fusion materials, Positron annihilation spectroscopy
Accepted: 15 November 2017

C. N. Taylor^a*, Y. Yamauchi^b, M. Shimada^a, Y. Oya^c & Y. Hatano<sup>d

a</sup>Idaho National Laboratory, Fusion Safety Program, Idaho Falls, Idaho 83415
^bHokkaido University, Department of Nuclear Engineering, Sapporo, Japan
^cShizuoka University, Radioscience Research Laboratory, Faculty of Science, Shizuoka, Japan
^dUniversity of Toyama, Hydrogen Isotope Research Center, Toyama 930-8555, Japan

Abstract
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.
Keywords: Molybdenum, neutron damage, helium, deuterium retention
Accepted for Publication: August 5, 2016

Keisuke Azuma^a,∗, Yuki Uemura^a, Shodai Sakurada^a, Hiroe Fujita^a, Cui Hu^b, Yuji Hatano^c,Naoaki Yoshida^d, Masashi Shimada^e, Dean Buchenauer^f, Takumi Chikada^a,Yasuhisa Oya<sup>a

a</sup>Graduate School of Integrated Science and Technology, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529, Japan
^bFaculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529, Japan
^cHydrogen Isotope Research Center, University of Toyama, 3190 Gohuku, Toyama, 930-8555, Japan
^dResearch Institute for Applied Mechanics, Kyushu University, 6-1, Kasugakoen, Kasuga, Fukuoka, Japan
^eIdaho National Laboratory, Idaho Falls, ID, United States
^fSandia National Laboratories, Livermore, CA, United States

Abstract
Deuterium (D) retention behavior for the sequential 6 MeV iron (Fe) and 10 keV carbon (C) implantedtungsten (W) were evaluated by thermal desorption spectroscopy (TDS) and β-ray-induced X-ray spectroscopy (BIXS) to understand the synergetic effect of defect formation and C existence on D retention behavior for W under various damage distribution profiles. The experimental results indicated that reten-tion of D trapped by dislocation loops was controlled by 10 keV C⁺ implantation. The D retention wasreduced in the sequential Fe²⁺-C⁺implanted W with higher C⁺fluence in comparison to that with lower C⁺fluence due to the formation of C-W layer which suppressed D diffusion toward the bulk and densedefects at the surface which reduce effective D diffusion coefficient. On the other hand, the amount of Dtrapped by the defects in the deeper region than C⁺implantation region (50 nm) was increased due tothe formation of dense defects by 6 MeV Fe²⁺ implantation within the depth of 1.5 μm.

Keywords: Plasma facing components, Tungsten, Carbon, Hydrogen isotope retention
Accepted: 16 May 2017

X. Litaudon³⁵, Y. Hatano¹⁰⁸, et al
³⁵EUROfusion Programme Management Unit, Culham Science Centre, Culham, OX14 3DB, United Kingdom
¹⁰⁸University of Toyama, Toyama, 930-8555, Japan

Abstract
The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H  =  1 at β N ～1.8 and n/n GW ～ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed

Keywords: JET, plasma, fusion, ITER
Accepted for publication: 30 January 2017

Toshihiko Yamanishi^a*, Norikiyo Koizumi^a, Masataka Nakahira^a, Yoshihiko Nunoya^a, Satoshi Suzuki,Hiroyuki Tobari^a, Mieko Kashiwagi^a, Takaaki Isono^a, Takashi Inoue^a, Makoto Sugimoto^a, Yoshinori Kusama^a, Yoshinori Kawamura^a,Hiroyasu Tanigawa^a,Masaru Nakamichi^a,Takashi Nozawa^a, Tsuyoshi Hoshino^a, Yoshio Ueda^b, Yuji Hatano^c, Takeo Muroga^d, and Satoshi Fukada^e
aNational Institutes for Quantum and Radiological Science and Technology
^bOsaka University
^cUniversity of Toyama
^dNational Institute for Fusion Science
^eKyushu University

Abstract
Several key components, such as superconducting coils, remote handling equipment, heating systems, have been designed and manufactured by JADA (Japan Domestic Agency). These activities have been carried out in accordance with the agreed schedule; in collaboration with the ITER organization and other domestic agencies. As a significant technical program using ITER, to design and to manufacture the TBS (Test blanket system), some R&D and design activities have also been conducted in Japan. Under the IFERC (the International Fusion Energy Research Center) projects of BA (Broader Approach) activities, design and R&D activities on fusion DEMO reactor have been carried out. For the DEMO R&D activity, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. From 2007, the above projects produced a set of fruitful results. A series of advanced technologies for the DEMO blanket system has also been carried out by Universities in Japan. Some significant basic R&D studies have also been carried out under US-JA collaborative program.

Keywords : ITER, Broader Approach, fusion nuclear technology, blanket
Accepted for Publication：６ March 2017

Noriyasu OHNO, Tatsuya KUWABARA, Makoto TAKAGI, Ryo NISHIMURA, Miyuki YAJIMA¹⁾, Akio SAGARA¹⁾, Takeshi TOYAMA²⁾, Katuya SUZUKI²⁾, Hiroaki KURISHITA²⁾, Tatsuo SHIKAMA²⁾, Yuji HATANO³⁾ and Naoaki YOSHIDA⁴⁾
Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603, Japan
¹⁾National Institute for Fusion Science, Oroshi-cho, Toki, Gifu 509-5292, Japan
²⁾Institute for Materials Research, Tohoku University, Oarai-machi, Ibaraki 311-1313, Japan
³⁾Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
⁴⁾Research Institute for Applied Mechanics, Kyushu University, Kasuga-kouen, Kasuga, Fukuoka 816-8580, Japan

Abstract
We have developed a compact divertor plasma simulator (CDPS) that can produce steady-state deuterium and/or helium plasmas with densities above ～ 10¹⁸ m⁻³ for Plasma-Wall Interaction (PWI) studies of neutronirradiated materials. The maximum particle flux is about 10²²m⁻²s⁻¹. The CDPS was installed and is being operated in the radiation-controlled area of the International Research Center for Nuclear Materials Science, Institute forMaterials Research, Tohoku University. We are able to control sample temperature within uncertainty of 5 ◦C during plasma exposure by adjusting the cooling air flow rate to the sample holder. The CDPS has a sample-carrier system, which makes it possible to transfer a plasma-irradiated sample from the sample holder to an infrared heater for analysis using thermal desorption spectroscopy (TDS) without exposing it to the air. This avoids the oxidation of the sample and minimizes the time between the end of plasma exposure and TDS analysis. An ITER-like tungsten (W) sample (A.L.M.T. Corp.), which has been irradiated by neutrons to 0.06 dpa in a fission reactor, was exposed to a deuterium plasma in the CDPS. The experimental results clearly show that the total deuterium retention in the neutron-irradiated W sample increases significantly in comparison with a pristine W, as demonstrated by broadening of the TDS spectrum at high temperatures.

Keywords : plasma-wall interaction, divertor plasma simulator, neutron-irradiated material, hydrogen isotope retention
Accepted: 30 August 2017

Baowang Lu¹, Yiwen Ju², Takayuki Abe¹, Katsuya Kawamoto^3
1Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama, 930-8555, Japan
²Key Laboratory of Computational Geodynamics, Chinese Academy of Sciences, College of Earth Science, University of Chinese Academy of Sciences, Beijing 100049, China
³Graduate School of Environmental and Life Science, Okayama University, 3-1-1 Tsushima-naka, Kita-ku, Okayama-shi, Okayama, 700-8530, Japan

Abstract
NiO/SBA-15 with nano NiO particles and cavities, has high catalytic ability for gas conversion and favors the decomposition of large tar molecules derived from gasification due to its larger pore sizes. Therefore, in this work the bench-scale cedar gasification was achieved using NiO/SBA-15 as a reforming catalyst to produce H₂ rich producer gas. NiO/SBA-15 appeared to be suitable for H₂ rich (over 50 v/v % (N₂ free)) producer gas production, as well CH₄, CnHm and tar reduction. Although the amount of NiO did not affect the gas composition, tar removal was decreased when the amount of NiO was considerably increased. In addition, the conversion of the producer gas was also carried out at high and low temperatures in the presence or absence of steam, using NiO/SBA-15 as a gas conversion catalyst. Regardless of the conversion temperature, conversion of the producer gas was largely affected by steam. At 750 ℃ and no steam ≈14% CO₂ was converted to CO, whereas no CO₂ conversion occurred in the presence of steam. At low temperatures, the maximum CH₄ yield in the absence of steam was 23%, which was higher than that in the presence of steam (15%).
Keywords: Hydrogen Production, Biomass Gasification, Chemical Conversion, Mesoporous Silica, Nano NiO Particle
Accepted: 18 July 2017

Baowang Lu¹, Yiwen Ju², Takayuki Abe¹, Katsuya Kawamoto<sup>3

1</sup>Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama, 930-8555, Japan
²Key Laboratory of Computational Geodynamics, Chinese Academy of Sciences; College of Earth Science, University of Chinese Academy of Sciences, Beijing 100049, China
³Graduate School of Environmental and Life Science, Okayama University, 3-1-1 Tsushima-naka, Kita-ku, Okayama-shi, Okayama, 700-8530, Japan
Abstract
A technique for analysing transition metal ion dispersion on mesoporous silica support using SEM observation was developed. In the presence of metal ion dispersion, no white point caused by metal oxide could be observed in YAGBSE, BE and COMPO mode images. By contrast, a white point could be clearly observed.

Keywords : Dispersion, Transition Metal Ion, Mesoporous, Mesoporous Silica, Nano NiO Particle, SEM Observation
Accepted: 12 November 2016

H.Fujita^a, Y. Uemura^a,S.Sakurada^a,K.Azuma^a,Q.Zhou^a,T.Toyama,^bN.Yoshida^c, Y.Hatano^d, T.Chikada^a,Y.Oya<sup>a

a</sup>GraduateSchoolofScience&Technology,ShizuokaUniversity,Shizuoka,4228529,Japan
^bInstituteforMaterialsResearch,TohokuUniversity,Ibaraki,3111313,Japan
^cResearchInstitute for Applied Mechanics, KyushuUniversity,Fukuoka,8168580,Japan
^dHydrogen IsotopeResearchCenter, University of Toyama,Toyama,9308555,Japan

Abstract
To evaluate the damage depth profile effect on hydrogen isotope retention in tungsten (W), combination usage of 0.8 MeV and 6.0 MeV Fe ions were implanted into W with the damage concentrations between 0.03 and 0.1 dpa. Thereafter, 1.0 keV deuterium ion (D₂⁺) implantation was performed with the flux of 1.0 × 10¹⁸ D⁺ m⁻² s⁻¹ up to the fluence of 1.0 × 10²² D⁺ m⁻², and the D retention behavior was evaluated by thermal desorption spectroscopy (TDS). The experimental results indicated that 6.0 MeV Fe ion irradiation would introduce vacancies and voids into bulk that were clearly controlled by the damage concentration, and the voids would become the most stable D trapping sites. It was found that D de-trapping from irradiation defects at lower temperature would be enhanced by the accumulation of defect near the surface due to 0.8 MeV Fe ion irradiation..

Keywords : Plasma facing materials, Tungsten, Hydrogen isotope retention behavior, Heavy ion irradiation
Accepted: 31 May 2017

S Masuzaki^1,2, M Tokitani^1,2, T Otsuka³, Y Oya⁴, Y Hatano⁵, M Miyamoto⁶, R Sakamoto^1,2, N Ashikawa^1,2, S Sakurada⁴, Y Uemura⁴, K Azuma⁴, K Yumizuru⁵, M Oyaizu⁷, T Suzuki⁷, H Kurotaki⁷, D Hamaguchi⁷, K Isobe⁷, N Asakura⁷, A Widdowson⁸, K Heinola⁹, S Jachmich¹⁰, M Rubel¹¹ and JET contributors¹²

 

¹ National Institute for Fusion Science, Oroshi 322-6, Toki 509-5292, Japan

² SOKENDAI (The Graduate University for Advanced Studies), Oroshi 322-6, Toki 509-5292, Japan

³ Kindai University, 3-4-1 Kowakae, Higashiosaka, Osaka 577-8502, Japan

⁴ Shizuoka University, Shizuoka, 422-8529, Japan

⁵ University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

⁶ Shimane University, Matsue, Shimane 690-8504, Japan

⁷ National Institutes for Quantum and Radiological Science and Technology (QST), Rokkasho Aomori 039- 3212, Japan

⁸ Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom

⁹ University of Helsinki, Box 43, FI-00014 Helsinki, Finland

¹⁰ Association Euratom-Etat Belge, ERM-KMS, Brussels, Belgium

¹¹ Royal Institute of Technology (KTH), 100 44 Stockholm, Sweden

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

Keywords: JET, ITER-like wall, dust, XPS, divertor tiles, tritium retention, microstructure
Accepted:12 September 2017

Masashi Shimada^a Yasuhisa Oya^b Dean A. Buchenauer^c and Yuji Hatano<sup>d

a</sup>Idaho National Laboratory, Idaho Falls, Idaho
^bShizuoka University, Shizuoka, Japan
^cSandia National Laboratories, Livermore, California
^dUniversity of Toyama, Toyama, Japan

Abstract
Irradiation effects on heat-load and heat removal, thermo-mechanical properties, and tritium behavior in neutron-irradiated tungsten and tungsten alloy are being investigated under US-Japan PHENIX (Plasma facing components evaluation by tritium Plasma, HEat and Neutron Irradiation eXperiments) collaboration (2013–2018) to demonstrate feasibility and safety of helium-cooled divertor concept for a fusion demonstration (DEMO) and future fusion reactors. The PHENIX Task 3 is aimed at improved understanding of irradiation response on tritium retention and permeation in tungsten and tungsten alloys under divertor-relevant high-flux plasma for a Fusion Nuclear Science Facility (FNSF) and DEMO. This paper describes the challenge in elucidating tritium behavior in neutron-irradiated plasma facing components (PFCs), the PHENIX plans for neutron-irradiation and post irradiation examination, and progress in tritium behavior in neutron-irradiated tungsten.

keywords :Nertron-irradiation, tungsten, tritium
Accepted: 7 June 2017

Shodai Sakurada^a, Yuki Uemura^a, Hiroe Fujita,^a,Keisuke Azuma,^a, Takeshi Toyama^b,Naoaki Yoshida^c, Tatsuya Hinoki^d, Sosuke Kondo^d,Yuji Hatano^e,Masashi Shimada ^f,Dean Buchenauer^g, Takumi Chikada^a and Yasuhisa Oya^a

^aShizuoka University, Graduate School of Science ＆Technology, 836 Ohya, Suruga-ku, Shizuoka-shi, Shizuoka 422-8529, Japan
^bTohoku University, Institute for Materials Research, 2145-2 Narita-cho, Oarai-machi, Higashiibaraki-gun,Ibaraki 311-1313, Japan
^cKyushu University, Institute for Applied Mechanics, 6-1 Kasuga-koen, Kasuga-shi, Fukuoka 816-8580, Japan
^dKyoto University, Institute of Advanced Energy, Gokasho, Uji-shi, Kyoto 611-0011, Japan

^e University of Toyama, Hydrogen Isotope Research Center, 3190 Gofuku, Toyama-shi, Toyama 930-8555, Japan
^fIdaho National Laboratory, 1955 N. Fremont Avenue, Idaho Falls, Idaho 83415
^gSandia National Laboratories, 7011 East Avenue, Livermore, California 94550

Abstract
The annealing effects on deuterium (D) retention for 0.1–1.0 dpa iron (Fe) ion damaged W were studied as a function of annealing duration. The D₂ spectra for Fe damaged W with lower defect concentration showed that D trapped by vacancy clusters was clearly decreased as increasing annealing duration due to the recovery of vacancy clusters. On the other hand, at higher defect concentration, the desorption peak of D trapped by voids was shifted toward higher temperature side, which would be caused by aggregation of vacancies and vacancy clusters. It can be said that the recovery and aggregation behavior of defects are controlled by defect concentration. By disappearing of desorption of D trapped by vacancy clusters after annealing for longer duration, the desorption of D trapped by vacancies was increased, which could be explained by following two possibilities. One is that the retention of hydrogen isotope trapped by monovacancy was increased. The other is that number of vacancies during annihilation process of vacancy cluster were formed by annealing.

keywords:Hydrogen isotopes retention, heavy-ion irradiation, annealing, tungsten.
Accepted: 5 June 2017

Kazunari Katayama^a,Youji Someya^b Kenji Tobita^b, Hirofumi Nakamura^c, Hisashi Tanigawa^b, Makoto Nakamura^b, Nobuyuki Asakura^b, Kazuo Hoshino^b,Takumi Chikada^d,Yuji Hatano^e and Satoshi Fukada^a

^aDepartment of Advanced Energy Engineering Science, Kyushu University, Kasuga, Fukuoka, Japan
^bNational Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori, Japan
^cNational Institutes for Quantum and Radiological Science and Technology, Tokai-mura, Ibaraki, Japan
^dGraduate School of Science, Shizuoka University, Shizuoka, Japan
^eHydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract
The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle ＆ Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium twocomponent system. In the representative recent DEMO condition, the following tritium permeation\ rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day.

keywords : Photocatalysis, Tritium , Permeation, Blanket, Divertor, DEMO
Accepted:12 September 2016