発表論文 2017年


Modified Integral Counting Method with Various Quenched Samples for Different Scintillators

Masato NAKAYAMA1), Masanori HARA1), Masao MATSUYAMA1), and Kiyokazu HIROKAMI2)

1)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

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 GoldTM scintillator. Sulfur-35 cocktail series were prepared with either EcoscintTM XR or Ultima GoldTM 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


Micro-/nano-characterization of the surface structures on the divertor tiles from JET ITER-like wall

M. Tokitania, M. Miyamotob, S. Masuzakia, Y. Fujiib, R. Sakamotoa, Y. Oyac, Y. Hatanod, T. Otsukae, M. Oyaidzuf, H.Kurotakif, T. Suzukif, D. Hamaguchif, K. Isobef, N. Asakuraf, A. Widdowsong, M. Rubelh, JET Contributors1

aNational 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


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


日米科学技術協力事業PHENIX 計画 -前半の成果と後半の研究計画-
Japan‐US Joint Research Project PHENIX‐Accomplishments in the First 3 Years and Research Plans in the Second Half‐
PHENIX 計画の概要
1. Overview of PHENIX Project

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)

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

Accepted: 20 December 2016


日米科学技術協力事業PHENIX 計画 -前半の成果と後半の研究計画-
Japan‐US Joint Research Project PHENIX‐Accomplishments in the First 3 Years and Research Plans in the Second Half‐
3. Task 2 (1):Neutron Irradiation Plan

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)



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


日米科学技術協力事業PHENIX 計画 -前半の成果と後半の研究計画-
Japan‐US Joint Research Project PHENIX‐Accomplishments in the First 3 Years and Research Plans in the Second Half‐
5.タスク3 トリチウム挙動および中性子照射効果
5. Task 3: Tritium Behavior and Neutron Irradiation Effect

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)



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


日米科学技術協力事業PHENIX 計画 -前半の成果と後半の研究計画-
Japan‐US Joint Research Project PHENIX‐Accomplishments in the First 3 Years and Research Plans in the Second Half‐
6. Summary and Future Plan

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)


Accepted: 20 December 2016


Development of H, D, T Simultaneous TDS Measurement System and H, D, T Retention Behavior for DT Gas Exposed Tungsten Installed in LHD Plasma Campaign

Yasuhisa Oyaa, Cui Hub, Hiroe Fujitaa, Kenta Yuyamaa, Shodai Sakuradaa, Yuki Uemuraa, Suguru Masuzakic, Masayuki Tokitanic, Miyuki Yajimac, Yuji Hatanod, and Takumi Chikadaa

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


 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


Effect of helium irradiation on deuterium permeation behavior in tungsten

Yuki Uemuraa, Shodai Sakuradaa, Hiroe Fujitaa, Keisuke Azumaa, Quilai Zhoua, Yuji Hatanob, Naoaki Yoshidac, Hideo Watanabec, Makoto Oyaizud, Kanetsugu Isobed, Masashi Shimadae, Dean Buchenauerf, Robert Kolasinskif, Takumi Chikadaa, Yasuhisa Oyaa

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


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


Design of a tritium gas cell for beta-ray induced X-ray spectrometryusing Monte Carlo simulation

Masanori Haraa, Shinsuke Abea, Masao Matsuyamaa, Tsukasa Asob, Katsuyoshi Tatenumac, Tomohiko Kawakamic, Takeshi Itoc

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


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


Tritium Counting Using a Europium Coordination Complex

Masanori Haraa, Haruna Sakaguchia, Masato Nakayamaa, Shinsuke Abea, Masao Matsuyamaa, Takayuki Abea, and Tsukasa Asob

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


The luminescence of Eu(DPA)33- 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)33-. The formation of Eu (DPA)33- was confirmed through spectrometry. Tritiated water was added to the prepared solution of Eu (DPA)33-. 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


Investigation of irradiation effects on highly integrated leading-edge electronic components of diagnostics and control systems for LHD deuterium operation

K. Ogawa1,2, T. Nishitani1, M. Isobe1,2, I. Murata3, Y. Hatano4, S. Matsuyama5, H. Nakanishi1,2, K. Mukai1,2, M. Sato1, M. Yokota1, T. Kobuchi1, T. Nishimura1and M. Osakabe1,2

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

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


Hydrogen sensing ability of Cu particles coated with ferromagnetic Pd-Co layer

Satoshi Akamaru a, Li Jinb, Katsuhiko Nishimurab, Masanori Haraa, Takayuki Abea, Masao Matsuyamaa aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
bFaculty of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

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


Surface modification and sputtering erosion of iron and copper exposed to low-energy, high-flux deuterium plasmas seeded with metal species

V.Kh. Alimovabc, Y. Hatanoa, M. Baldend, M. Oyaizud, K. Isobee, H. Nakamurae, T.Hayashie
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

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


Progress in the U.S./Japan PHENIX Project for the Technological Assessment of Plasma Facing Components for DEMO Reactors

Yutai Katoha, Daniel Clarkb, Yoshio Uedac, Yuji Hatanod, Minami Yodae, Adrian S. Sabaua, Takehiko Yokominef, Lauren M. Garrisona, J. Wilna Geringera, Akira Hasegawag, Tatsuya Hinokif, Masashi Shimadah, Dean Buchenaueri, Yasuhisa Oyaj and Takeo Murogak
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

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


Interaction of Hydrogen Isotopes with Radiation Damaged Tungsten

Yasuhisa Oya1, Keisuke Azuma1, Akihiro Togari1, Qilai Zhou1, Yuji Hatano2, Masashi Shimada3, Robert Kolasinski4and Dean Buchenauer4

1Shizuoka 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

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


Tritium analysis of divertor tiles used in JET ITER-like wall campaigns by means of β-ray induced x-ray spectrometry

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

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


[FeFe]-Hydrogenase and its organic molecule mimics—Artificial and bioengineering application for hydrogenproduction

Motonori Watanabea,∗, Yuki Hondab, Hidehisa Hagiwarac, Tatsumi Ishiharaa,d

aInternational 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
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

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


Two-Dimensional TiO2 Nanosheets for Photo and Electro-Chemical Oxidation of Water: Predictions of Optimal Dopant Species from First-Principles

Namhoon Kim, Emily M. Turner, Yoonyoung Kim§,∥ Shintaro Ida, Hidehisa Hagiwara§,∥, Tatsumi Ishihara§,∥and Elif Ertekin*,†,∥

Department of Mechanical Science and Engineering, University of Illinois at Urbana−Champaign, 1206 W. Green Street, Urbana, Illinois 61801, United States
Department of Materials Science and Engineering, University of Florida, 1724 Gale Lemerand Drive, Gainesville, Florida 32603, United States
§Department of Applied Chemistry, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan
International Institute for Carbon Neutral Energy Research (WPI-I2CNER), Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan
Graduate School of Science and Technology, Kumamoto University, 2-39-1 Kurokami, Chuo, Kumamoto 860-8555, Japan

Due to a high surface area to volume ratio, twodimensional nanosheets have gained interest for photo and/or electrocatalytic water splitting. In particular, experimental Rh doping of lepidocrocite TiO2 nanosheets has significantly increased catalytic activity. We use first-principles density functional theory to consider the oxygen evolution reaction (OER) on both pristine and transition metal doped systems. While the undoped TiO2 nanosheets exhibit several limitations and require high overpotentials during the water splitting reaction, selected dopants modify the binding strength of reaction intermediates and can reduce rate limiting thermodynamic barriers and theoretical required overpotentials. We present an activity volcano for these nanosheets, with the full spectrum of 3d, 4d, and 5d transition metals as candidate dopants. Subsequent photocatalytic measurements of OER activity with selected dopants are carried out to validate the predictions, and the trends are found to be consistent. These results help describe how surface dopants affect reaction mechanisms and provide general design principles for high performance catalysts during the water splitting reaction.

Publication Date (Web): August 15, 2017


Deuterium Retention in Helium and Neutron Irradiated Molybdenum

C. N. Taylora*, Y. Yamauchib, M. Shimadaa, Y. Oyac & Y. Hatanod

aIdaho 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

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


Effect of sequential Fe2+-C+implantation on deuterium retention in W

Keisuke Azumaa,∗, Yuki Uemuraa, Shodai Sakuradaa, Hiroe Fujitaa, Cui Hub, Yuji Hatanoc,Naoaki Yoshidad, Masashi Shimadae, Dean Buchenauerf, Takumi Chikadaa,Yasuhisa Oyaa

aGraduate 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

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


Overview of the JET results in support to ITER

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

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


Recent R&D Results on Fusion Nuclear Technology for ITER and DEMO Reactor in Japan

Toshihiko Yamanishia*, Norikiyo Koizumia, Masataka Nakahiraa, Yoshihiko Nunoyaa, Satoshi Suzuki,Hiroyuki Tobaria, Mieko Kashiwagia, Takaaki Isonoa, Takashi Inouea, Makoto Sugimotoa, Yoshinori Kusamaa, Yoshinori Kawamuraa,Hiroyasu Tanigawaa,Masaru Nakamichia,Takashi Nozawaa, Tsuyoshi Hoshinoa, Yoshio Uedab, Yuji Hatanoc, Takeo Murogad, and Satoshi Fukadae
aNational Institutes for Quantum and Radiological Science and Technology
bOsaka University
cUniversity of Toyama
dNational Institute for Fusion Science
eKyushu University

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


Development of a Compact Divertor Plasma Simulator for Plasma-Wall Interaction Studies on Neutron-Irradiated Materials

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

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


Hydrogen-Enriched Producer Gas Production and Chemical Conversion to Usable Gas Product Through Biomass Gasification Using NiO Nanoparticles Dispersed on SBA-15

Baowang Lu1, Yiwen Ju2, Takayuki Abe1, Katsuya Kawamoto3
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

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


Direct Scanning Electron Microscopy Observation of the Dispersion of Transition Metal Ion on Mesoporous Silica Support

Baowang Lu1, Yiwen Ju2, Takayuki Abe1, Katsuya Kawamoto3

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
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


The damage depth profile effect on hydrogen isotope retention behavior in heavy ion irradiated tungsten

H.Fujitaa, Y. Uemuraa,S.Sakuradaa,K.Azumaa,Q.Zhoua,T.Toyama,bN.Yoshidac, Y.Hatanod, T.Chikadaa,Y.Oyaa

cResearchInstitute for Applied Mechanics, KyushuUniversity,Fukuoka,8168580,Japan
dHydrogen IsotopeResearchCenter, University of Toyama,Toyama,9308555,Japan

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


Photoelectrochemical H2 evolution using TiO2- coated CaFe2O4 without an external applied bias under visible light irradiation at 470 nm based on device modeling

Shintaro Idaa,c, Kara Kearneyb,c, Takamitsu Futagamia, Hidehisa Hagiwaraa,c, Takaaki Sakaia, Motonori Watanabec, Angus Rockettcd and Tatsumi Ishiharaa,c

aDepartment of Applied Chemistry, Faculty of Engineering, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan. E-mail: s_ida@chem.kumamoto-u.ac.jp
bDepartment of Materials Science and Engineering, University of Illinois, 1304 West Green Street, Urbana, IL, 61801, USA
cInternational Institute for Carbon Neutral Energy Research (I2CNER), Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395, Japan
dDepartment of Metallurgy and Materials Engineering, Colorado School of Mines, 201 Hill Hall, 1500 Illinois Street, Golden, CO, 80401, USA

CaFe2O4 (CFO) can be used as a photocathode to evolve H2 from water in a photoelectrochemical cell. However, CFO degrades during operation and an external voltage is necessary for PEC H2 evolution because the onset potential is less than the potential required for water oxidation considering the overpotential at the counter electrode. In order to develop a reliable CFO electrode with a greater onset potential, improvement of chemical stability and suppression of surface recombination is necessary. In this study, a chemically stable electrode structure with a greater onset potential was achieved by coating the [00l]-oriented CFO with a thin layer of titanium dioxide (TiO2). A CFO|TiO2 electrode was designed using a device simulator. The simulation results predict that coating CFO with TiO2 produces a positive or negative shift in the onset potential under visible and ultraviolet light irradiation, respectively. The experimental onset potentials matched the simulation prediction. The observed onset potential for TiO2-coated CFO was around (1.6 V vs. RHE) under visible light (470 nm) and 0.9 V under ultraviolet light (300nm), compared to 1.2–1.3 V vs. RHE for a bare CFO electrode. The onset potential (1.6 V) under visible light irradiation is the most positive onset potential among the oxide photocathodes ever reported for PEC water splitting. Using the TiO2-coated CFO as the photocathode and RuO2-loaded Pt as the anode, stable photocurrent was observed under 470 nm excitation without an external voltage and evolution of H2 from the system was confirmed.

Accepted:28 February 2017


Hydrogen Isotope Retention and Permeation in Neutron-Irradiated Tungsten and Tungsten Alloys Under PHENIX Collaboration

Masashi Shimadaa Yasuhisa Oyab Dean A. Buchenauerc and Yuji Hatanod

aIdaho National Laboratory, Idaho Falls, Idaho
bShizuoka University, Shizuoka, Japan
cSandia National Laboratories, Livermore, California
dUniversity of Toyama, Toyama, Japan

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


Impact of Annealing on Deuterium Retention Behavior in Damaged W

Shodai Sakuradaa, Yuki Uemuraa, Hiroe Fujita,a,Keisuke Azuma,a, Takeshi Toyamab,Naoaki Yoshidac, Tatsuya Hinokid, Sosuke Kondod,Yuji Hatanoe,Masashi Shimada f,Dean Buchenauerg, Takumi Chikadaa and Yasuhisa Oyaa

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

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


Estimation of Tritium Permeation Rate to Cooling Water in Fusion DEMO Condition

Kazunari Katayamaa,Youji Someyab Kenji Tobitab, Hirofumi Nakamurac, Hisashi Tanigawab, Makoto Nakamurab, Nobuyuki Asakurab, Kazuo Hoshinob,Takumi Chikadad,Yuji Hatanoe and Satoshi Fukadaa

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

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


Effects of preparation condition on the photocatalytic activity of porphyrin-modified GaN:ZnO for water splitting

Hidehisa Hagiwaraa,b, Ryota Kakigia, Shuhei Takechia,Motonori Watanabeb, Satoshi Hinokumac, Shintaro Idaa,b Tatsumi Ishiharaa,b

a Department of Applied Chemistry, Kyushu University, Motooka 744, Nishi-ku, Fukuoka 819-0395, Japan
b International Institute for Carbon-Neutral Energy Research (I2CNER), Motooka 744, Nishi-ku, Fukuoka 819-0395, Japan
c Department of Applied Chemistry and Biochemistry, Graduate School of Science and Technology, Kumamoto University, 2-39-1 Kurokami, Chuo-ku, Kumamoto 860-8555, Japan

Effects of nitridation condition on gallium-zinc oxynitride solid solution (GaN:ZnO) was investigated to optimize the composition of GaN:ZnO for dye-modified photocatalysts. Gallium nitride (GaN) formed from Ga2O3 at 973 K, and GaN:ZnO was obtained over 1073 K under NH3 gas flow. Nitrogen content in GaN:ZnO increased with increasing nitridation temperature and time, while zinc content decreased because of evaporation. Although UV–vis absorption spectra of GaN:ZnO powders were not significantly changed in different compositions, the water splitting activities of the dye-modified GaN:ZnO photocatalysts depended on the composition of GaN:ZnO. The highest formation rates of H2 and O2 were achieved by the GaN:ZnO containing 15% of zinc and 73% of nitrogen.Finally, the nitridation condition was optimized at 1123 K 15 h under NH3 gas flow (200 ml/min) for preparation of the dye-modified GaN:ZnO powder as water splitting photocatalyst.
keywords : Photocatalysis, Water splitting , Nitridation , Solid solution , Porphyrin
Accepted:20 October 2016


Inorganic/whole-cell biohybrid photocatalyst for highly efficient hydrogen production from water

Yuki Hondaa,Motonori Watanabea,Hidehisa Hagiwaraa,b,Shintaro Idaa,b,Tatsumi Ishiharaa,b

aInternationalInstituteforCarbon-NeutralEnergyResearch, KyushuUniversity, 744Motooka, Nishi-ku, Fukuoka8190395, Japan
bDepartmentofAppliedChemistry, FacultyofEngineering, KyushuUniversity, 744Motooka, Nishi-ku, Fukuoka819-0395, Japan

To obtain a clean hydrogen production system, we have developed an inorganic-bio hybrid photocat-alyst system based on the combination of anatase TiO2, methylviologen(MV) as an electron mediator, and a whole-cell biocatalystconsisting of[FeFe]-hydrogenase and maturase gene-harboring recombi-nant Escherichia coli; however, the apparent quantum yield at 300 nm(AQY300)for hydrogen production was low(0.3%).The system consists of a two-step reaction:(1)photocatalytic MV reduction by TiO2, and (2) hydrogen production with reduced MV using a biocatalyst. The enhancemen to of step 1 under biocatalyst-friendly conditions was investigated in an attempt to further improve the reaction efficiency. Among the condition tested, the use of 100 mMTris-HCl(pH7),150mM NaCl, and 5%(v/v) glycerol with P-25TiO2 especially enhanced the step 1 reaction by a 300-fold increase in the MV reduction rate compared with previously tested reaction condition(100 mMTris-HCl(pH7),150 mM NaCl,5%(v/v)glyc-erol, and 100mM ascorbate with anatase TiO2 ). Under the enhanced step 1 reaction, AQY300 and AQY350 for photocatalytic MV reduction reached 60.8 % and 52.2% , respectively. The enhanced step 1 reaction thus significantly improved the overall photocatalytic hydrogen productivity of the hybrid system and AQY300 and AQY350 reached 26.4% and 31.2%, respectively. The inorganic-whole-cell biohybrid system can therefore provide noble metal-free, efficient, and clean hydrogen production.

keywords:[Fe-Fe]-hydrogenase, Inorganic-bio hybrid, Photocatalysis, Water splitting, Whole-cell biocatalyst
Accepted:5 April 2017


Ni‐Fe Nitride Nanoplates on Nitrogen‐Doped Graphene as a Synergistic Catalyst for Reversible Oxygen Evolution Reaction and Rechargeable Zn‐Air Battery

Yuchi Fan, Shintaro Ida, Aleksandar Staykov, Taner Akbay, Hidehisa Hagiwara, Junko Matsuda, Kenji Kaneko and Tatsumi Ishihara

Obtaining bifunctional electrocatalysts with high activity for the oxygen evolution reaction (OER) and oxygen reduction reaction (ORR) is a main hurdle in the application of rechargeable metal-air batteries. Earth-abundant 3d transition metal-based catalysts have been developed for the OER and ORR; however, most of these are based on oxides, whose insulating nature strongly restricts their catalytic performance. This study describes a metallic Ni-Fe nitride/nitrogen-doped graphene hybrid in which 2D Ni-Fe nitride nanoplates are strongly coupled with the graphene support. Electronic structure of the Ni-Fe nitride is changed by hybridizing with the nitrogen-doped graphene. The unique heterostructure of this hybrid catalyst results in very high OER activity with the lowest onset overpotential (150 mV) reported, and good ORR activity comparable to that for commercial Pt/C. The high activity and durability of this bifunctional catalyst are also confirmed in rechargeable zinc-air batteries that are stable for 180 cycles with an overall overpotential of only 0.77 V at 10 mA−2.

Accepted: 16 May 2017


Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

S Masuzaki1,2, M Tokitani1,2, T Otsuka3, Y Oya4, Y Hatano5, M Miyamoto6, R Sakamoto1,2, N Ashikawa1,2, 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

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


Deuterium trapping at vacancy clusters in electron/neutron-irradiated tungsten studied by positron annihilation spectroscopy

T. Toyama a, *, K. Ami b, K. Inoue a, Y. Nagai a, K. Sato c, Q. Xu c, Y. Hatano d

a 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

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