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, Sweden

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^1,2

¹National 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

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