発表論文 2017年

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[2017_01]

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
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 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

[2017_02]

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, 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

[2017_03]

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)
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

[2017_04]

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.タスク2(1)中性子照射計画
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)

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

[2017_05]

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)

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

[2017_06]

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.まとめと今後の研究計画
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)

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

[2017_07]

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


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

[2017_08]

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

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

[2017_09]

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


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

[2017_10]

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


Abstract

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

[2017_11]

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

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

[2017_12]

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

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

[2017_13]

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

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

[2017_14]

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

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

[2017_15]

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

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

[2017_16]

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

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