Hiroaki Nakamura¹, Hisanori Miyanishi², Takuo Yasunaga³, Susumu Fujiwara⁴ , Tomoko Mizuguchi⁴ , Ayako Nakata⁵ , Tsuyoshi Miyazaki⁵ , Takao Otsuka⁶ , Takahiro Kenmotsu⁷ ,Yuji Hatano<sup>8

1</sup> Department of Helical Plasma Research, National Institute for Fusion Science, Toki, Gifu, JAPAN
² Department of Energy Engineering and Sciences, Nagoya University, Nagoya, Aichi, JAPAN
³ Kyushu Institute of Technology, IIzuka, Fukuoka, JAPAN
⁴ Kyoto Institute of Technology, Kyoto, JAPAN
⁵ National Institute for Materials Science, Tsukuba, Ibaraki, JAPAN
⁶ RIKEN Center for Biosystems Dynamics Research, Kobe, Hyogo, JAPAN
⁷ Doshisha University, Kyotanabe, Kyoto, JAPAN
⁸ University of Toyama, Toyama, Toyama, JAPAN
⁹ Institute for Molecular Science, Okazaki, JAPAN

Abstract
Using molecular dynamics, we simulate the structural change of a telomeric DNA by β-decay of substituted tritium to helium-3. The configuration of the telomeric DNA is obtained by removing TRF2 protein from the TRF2-Dbd-DNA complex (PDBID: 3SJM). We assume that hydrogens of guanines in the DNA are replaced to helium-3. The charge distribution in the MD simulation for the modified guanine is obtained by DFT simulation. We adopt, as the MD simulation, Nanoscale Molecular Dynamics code with CHARMM36 force field and Langevin thermostat algorithm. Changing both the number of replaced guanine N and the temperature T, we calculate the root mean square deviation to quantify the durability of DNA. It is found that as N or T becomes larger, the RMSD of the DNA becomes also larger. Namely, as the intensity of the β-decays becomes larger or as the temperature is increased, the DNA structure becomes more fragile.

Accepted :19 September 2019

N. Ashikawa^a,b,, Y. Torikai^c, N. Asakura^d, T. Otsuka^e, A. Widdowson^f, M. Rubel^g, M. Oyaizu^d,M. Hara^h, S. Masuzaki^a, K. Isobe^d, Y. Hatano^h, K. Heinolaⁱ, A. Baron-Wiechec^f, S. Jachmich^f, T. Hayashi<sup>d

a</sup> National Institute for Fusion Science, Toki, 509-5292 Japan
^b SOKENDAI, Toki, 509-5292 Japan
^c Ibaraki University Mito, 310-8512 Japan
^d National Institute for Quantum and Radiological Science and Technology, Rokkasho, 039-3212 Japan
^e Kindai University, Higashi-Osaka, 577-8502, Japan
^f CCFE, Culham Science Centre, Abingdon, OX14 3DB, UK
^g KTH Royal Institute of Technology, 100 44 Stockholm, Sweden
^h University of Toyama, Toyama, Japan
ⁱ University of Helsinki, Helsinki, Finland

Abstract
Quantitative tritium inventory in dust particles from campaigns in the JET tokamak with the carbon wall (2007-2009) and the ITER-like wall (ILW 2011-2012) were determined by the liquid scintillation counter and the full combustion method. A feature of this full combustion method is that dust particles were covered by a tin (Sn) which reached 2100 K during combustion under oxygen flow. The specific tritium inventory for samples from JET with carbon and with metal walls was measured and found to be similar. However, the total tritium inventory in dust particles from the ILW experiment was significantly smaller in comparison to the carbon wall due to the lower amount of dust particles generated in the presence of metal walls.

Keywords: Dust, JET, ITER-Like Wall, Tritium, Liquid scintillography, Full combustion method
Accepted: 9 April 2019

Masanori Hara, ^*a Miki Shoji,^b and Tsukasa Aso<sup>c

a</sup>University of Toyama, Organization for Research Promotion, Hydrogen Isotope Research Center, 3190 Gofuku, Toyama City, Toyama 930-8555, Japan
^bUniversity of Toyama, Organization for Research Promotion, Life Science Research Center, 2630 Sugitani, Toyama City, Toyama 930-0194, Japan
^cToyama College, National Institute of Technology, Electronics and Computer Engineering, 1-2 Ebie-neriya, Imizu City, Toyama 933-0293, Japan

Abstract
Liquid scintillation counters (LSCs) have been widely used for low-level tritium measurements. To obtain an accurate tritium activity using a LSC, a quenching correction is required. The quenching occurs from interruptions to the scintillation process (chemical quenching) and by absorption of scintillation photons by colored substances (color quenching). There is no common method for the correction of color quenching. Here, two-dimensional (2-D) scintillation spectra were measured with a conventional LSC connected to an external multichannel analyzer. The LSC had two photomultiplier tubes (PMTs). A 2-D spectrum was constructed from pulse heights from both PMTs. In a less-quenching cocktail, the 2-D scintillation spectra extended along a 45-deg line. However, the shape of the spectrum broadened with increasing color quenching and thus gave information about the color quenching. The effect of color quenching was qualitatively less significant in the relationship between the tritium counting efficiency and the quenching index parameter.

Accepted for Publication: August 16, 2019
Keywords: Liquid scintillation counter, color quenching, two-dimensional spectrum, counting efficiency

S. E. Lee, ^*a Y. Hatano,^b M. Hara,^b and M. Matsuyama<sup>b

a</sup>University of Toyama, Graduate School of Science and Engineering for Education, Toyama 930-8555, Japan
^bUniversity of Toyama, Organization for Promotion of Research, Hydrogen Isotope Research Center, Toyama 930-8555, Japan

Abstract
Nondestructive measurement of tritium (T) content in solid materials is important for safe and costeffective disposal of contaminated wastes, and beta-ray induced X-ray spectrometry (BIXS) has been developed for this purpose. A common way to obtain depth profiles of T in solids using BIXS is to perform simulation of X-ray spectra for assumed depth profiles and find a profile giving the best agreement with observation. A detailed understanding of attenuation of low-energy X-rays (≤18.6 keV) by detector components such as a window material is required for interpretation of measured spectra and simulation. In this study, BIXS spectra of a tungsten reference sample with known T depth profile were measured using two different semiconductor detectors and simulated using the Monte Carlo simulation toolkit Geant4. In the low-energy region (＜2 keV), the difference in internal structure resulted in a noticeable difference in the BIXS spectra. The disagreement between the measured and the simulated spectra was also significant at ＜2 keV. Nevertheless, at ＞ 2 keV, the BIXS spectra were insensitive to the internal structure of the detector, and the simulated spectra agreed well with the measured ones. The mechanism underlying the difference in the low-energy region was discussed.

Accepted: December 10, 2019
Keywords: Tritium, beta rays, X-rays, attenuation, Monte Carlo simulation

Qilai Zhou ^a,d,*, Akihiro Togari^b, Moeko Nakata^b, Mingzhong Zhao^, Fei Sun ^a, Miyuki Yajima ^c, Masayuki Tokitani ^c, Suguru Masuzaki ^c, Naoaki Yoshida ^e, Masanori Hara ^f, Yuji Hatano ^f, Yasuhisa Oya <sup>a

a</sup> Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
^b Graduate School of Science, Shizuoka University, Shizuoka 422-8529, Japan
^c National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan
^d School of Materials Science and Engineering, Wuhan University of Technology, Wuhan 430070, PR China
^e Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^f Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract
Tungsten (W) is one of the most promising candidates for plasma facing materials in the fusion reactor. Helium (He) in W can influence the retention of hydrogen isotopes. In the present study, W targets were simultaneously irradiated by He⁺-H₂⁺ or He⁺-D₂⁺ ion beams with the energies of 1 keV or 3 keV, at fixed temperatures in a range of room temperature(R.T.) to 1073 K. Mechanisms of He and hydrogen interaction in W were discussed, especially from the point of He retention, which was characterized by the high-temperature thermal desorption spectroscopy (TDS) and glow-discharge optical emission spectroscopy (GD-OES) measurements. It is found that He desorption shifts to a lower temperature range for the W simultaneously irradiated by 3 keV He⁺-1 keV H₂⁺ at 573 K, under He⁺ fluence up to 1×10²² He⁺m^-2. Transmission electron microscope (TEM) observation and annealing treatment at the temperature of 873-1073 K show that the increased He uptake is caused by the formation of dislocation. Enhanced retention amounts for the hydrogen isotopes were also confirmed. Amounts of the dislocation loops introduced by the H₂⁺-only irradiation can be reduced by annealing treatment at 873 K, while that introduced by He₂⁺ irradiation are quite stable, which grows larger at elevated temperatures. With an increase of H₂⁺ energy, Helium uptakes at both weak trapping sites and bubbles are increased, while the amounts of hydrogen retention are decreased. It suggests that hydrogen ion has a significant influence on the He trapping sites at the irradiation temperature up to 573 K, while the hydrogen retention is determined by the distribution of He bubbles and dislocation loops.

Keywords: Isotope,TDS,GD-OES,Dislocation loops,Deuterium,Retention
Accepted: 27 January 2020

F. Sun ^{a, *}, M. Nakata ^b, S.E. Lee ^c, M. Zhao ^b, T. Wada _b, S. Yamazaki _b, A. Koike _b, S. Kondo ^d,T. Hinoki ^e, M. Hara ^f, Y. Oya <sup>b

a</sup> Faculty of Science, Shizuoka University, Shizuoka, 422-8529, Japan
^b Graduate School of Science and Technology, Shizuoka University, Shizuoka, 422-8529, Japan
^c Graduate School of Science and Engineering for Education, University of Toyama, Toyama, 930-8555, Japan
^d Institute for Materials Research, Tohoku University, Sendai, 980-8577, Japan
^e Institute of Advanced Energy, Kyoto University, Kyoto, 611-0011, Japan
^f Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 930-8555, Japan

Abstract
In this study, energetic helium (He) ion irradiation was performed to obtain bulk He distribution in tungsten (W) materials, concurrent with damage introduction at high temperature. Then, deuterium (D) implantation and thermal desorption spectrometry were performed to evaluate D retention. At the same time, the surface tritium (T) concentration and depth distribution were evaluated by imaging plate (IP) and b-ray induced X-ray spectroscopy (BIXS) measurements after mixed D-T gas exposure. Numerical simulations were applied to evaluate changes in binding energies, diffusion depths, and trapping sites under different irradiation conditions. The results showed that weak trapping sites with higher concentration, such as vacancies, were produced during only energetic He⁺ irradiation events, leading to enhancement of D retention. Fe³⁺-He⁺ simultaneous irradiation promoted the formation of He_xV_y complexes, which reduced the concentration of vacancy trapping sites and changed the stress field around defects, leading to the suppression of D trapping behavior. From the reduced effects of D retention caused by He_xV_y complexes at higher temperatures, the results suggested that defect recovery was the dominant mechanism. With increasing damage level at higher temperatures, more weak trapping sites, such as dislocations and vacancies sites, were produced, leading to a more dominant influence on D retention than He_xV_y complex effects. It was also found that He_xV_y complexes prevented D diffusion to the bulk and that simulation results showed that the damage level had little impact on D diffusion depth.

Keywords: Hydrogen isotope, Helium, Irradiation damages, Simulation, Tungsten, Fusion
Accepted: 16 March 2020

Mitsuhiro Inoue^a, Asuka Shima^b, Kaori Miyazaki^a, Baowang Lu^a, Yoshitsugu Sone^c,d,Takayuki Abe<sup>a,*

a</sup> Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Gofuku 3190, Toyama, 930-8555, Japan
^b Japan Aerospace Exploration Agency, Jindaiji Higashi-machi 7-44-1, Chofu, Tokyo 182-8522, Japan
^c Japan Aerospace Exploration Agency, Yoshinodai 3-1-1, Chuo-ku, Sagamihara, Kanagawa 252-5210, Japan
^d SOKENDAI, Yoshinodai 3-1-1, Chuo-ku, Sagamihara, Kanagawa, 252-5210, Japan

Abstract
This study investigated the effects of sputtering conditions on the activities of high-performance CO₂ methanation catalysts prepared by a co-sputtering technique, employing a polygonal barrel apparatus. Average size of smaller Ru nanoparticles generated by co-sputtering Ru with TiO₂ or ZrO₂ varied with changes in the area ratio of the sputtering targets. The reaction temperature was decreased with decreases in the Ru particle size, and the most effective target area ratio was Ru:ZrO₂=1:0.5 when co-sputtering Ru and ZrO₂. For this optimized catalyst, increasing the sputtering time did not affect the Ru particle size but improved the catalytic activity. Small Ru particles were maintained even at a reaction temperature of 360 ℃, indicating that undesirable decreases in catalytic activity due to particle growth can be suppressed using this co-sputtering technique. These highly active co-sputtered catalysts would have applications in systems intended for the reduction of CO₂ emissions.

Keywords: CO₂methanation catalysts, Ru-metal oxide co-sputtering, Ru particle size, Sputtering conditions, Polygonal barrel-sputtering method, Dry process
Accepted: 7 April 2020

Mitsuhiro Inoue, Kaori Miyazaki, Baowang Lu, Chulho Song, Yoshitake Honda, Masazumi Arao,Tsukuru Ohwaki, Masashi Matsumoto, Hideto Imai, Asuka Shima, Yoshitsugu Sone, Ren Chung Peng,Toshiya Shibayanagi, and Takayuki Abe*

Abstract
This study elucidates the factors reducing the CO₂ methanation reaction temperature of TiO₂-supported Ru catalysts prepared via the polygonal barrel-sputtering method (Ru/TiO₂(BS)) to investigate the structure-sensitivity mechanism. The smaller nanoparticles deposited in Ru/TiO₂(BS) (＜4 nm) were amorphous RuO₂ because of air exposure after the preparation, and their surfaces were changed to island-shaped structures consisting of amorphous RuO₂ and amorphous Ru metal by H₂ exposure. In this case, dissociative hydrogen was also adsorbed in abundance on the amorphous Ru metals. Such hydrogen atoms were not observed in conventional Ru/TiO₂ catalysts. Under the supplied CO₂ + H₂ at a stoichiometric ratio of 1:4, these hydrogen atoms not only contributed to the generation of a unique CO intermediate (Ru-CO-Ru-H) from room temperature, but also reduced this CO adsorbate to methane even in low-temperature ranges (≤120 ℃). These reaction steps were completely different from the reported mechanisms. Accordingly, the formation of amorphous Ru metals and the adsorption of hydrogen atoms on them are essential for reducing the CO₂ methanation temperature. These are key factors of structure-sensitivity, which would also be useful for improving activities of various catalysts.

Published: April 15, 2020

Norihiro Ikemoto^a,b,*, Tomohiko Kawakami^a, Kazuo Yonehara^a, Yuri Natori^a,Katsuyoshi Tatenuma^a, Masanori Hara<sup>c

a</sup> KAKEN Incorporated, 1044 Horimachi, Mito City, Ibaraki 310-0903, Japan
^b Graduate School of Science and Engineering for Education, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
^c Hydrogen Isotope Research Center, Research Promotion Organization, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract
Molecular sieves (MSs) are a potentially useful means of hydrogen isotope recovery as well as the temporary storage and supply of such isotopes. In this study, the hydrogen adsorption isotherms for various MS samples were acquired after applying different water removal processes to activate the MSs. MS-3A sample was found not to adsorb hydrogen regardless of the activation conditions. MS-4A sample adsorbed hydrogen at 77 K after heating at 473, 513, and 673 K. However, the adsorption amount after heating at 473, 513 K was considerably less than that at 673 K. Consequently, heating at 673 K was required to activate this type of MS. In the case of an MS-5A sample, the presence of residual water had little effect on hydrogen isotope adsorption at 77 K. From these results, it is evident that MS-5A is the most suitable hydrogen adsorption material because it is readily activated and is less affected by water adsorption.

Keyword: DT fuel cycle, Molecular sieve, Cryogenic adsorption, Adsorption isotherm
Accepted: 13 April, 2020

Yoshinari Oshimi,^a Mayu Ohki,^a Misato Nagano,^a Takuyo Yasumatsu,^a* Masanori Hara,^b Satoshi Akamaru,^b Masato Nakayama,^b and Miki Shoji<sup>c

a</sup>Tokyo Power Technology Ltd., Nuclear Power Branch, Toyosu Arban Point Building, 4F 5-5-13 Toyosu, Koto-ku,Tokyo 135-0061, Japan
^bUniversity of Toyama, Organization for Research Promotion, Hydrogen Isotope Research Center, 3190 Gofuku,Toyama City, Toyama 930-8555, Japan
^cUniversity of Toyama, Organization for Research Promotion, Life Science Research Center, 2630 Sugitani, Toyama City, Toyama 930-0194, Japan

Abstract
For low-level tritium measurements using a liquid scintillation counter, scintillation vial selection is important. The applicability of polyethylene (PE) vials was studied. Three types of vials were tested: (1) 100-mL perfluoroalkoxy alkane (PFA) fluorine resin vials, (2) 100-mL PE vials, and (3) 145-mL PE vials. Ultima Gold LLTwas the reference liquid scintillator in this study. The background counts for these vials were found to be 2.5 counts per minute. Tritiated water of 1.5 Bq・mL^-1 was employed as an internal standard to determine the counting efficiency. The counting efficiencies for the 100-mL PFA, 100-mL PE, and 145-mL PE vials were estimated to be 17%, 16%, and 13%, respectively. The lower limits of detection of these vials for a counting time of 100 min were 1.45 Bq・L^-1 for 100-mL PFA vials, 1.54 Bq・L^-1 for 100-mL PE vials, and 1.47 Bq・L^-1 for 145-mL PE vials. Thus, these vials demonstrate similar performances for tritium measurements.

Keyword: Liquid scintillation counter, tritium measurement, 100-mL vial, lower limits of detection, counting efficiency.

Accepted: February 7, 2020

Yasunobu Arikawa, ^a* Yuki Iwasa,^a Kohei Yamanoi,^a Keisuke Iwano,^a Shinsuke Fujioka,^a Akifumi Iwamoto,^b Mitsuo Nakai,^a Yuji Hatano,^c Masanori Hara,^c Satoshi Akamaru,^c Takayoshi Norimatsu,^a and Ryosuke Kodama<sup>a

a</sup>Osaka University, Institute of Laser Engineering, 2-6 Yamadaoka, Suita, Osaka, Japan
^bNational Institute of Fusion Science, 322-6, Oroshi Cho, Toki city, Gifu, Japan
^cUniversity of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract
In inertial confinement fusion (ICF), a fuel target containing deuterium and tritium is used. In recent ICF experiments on the Gekko XII LFEX facility at the Institute of Laser Engineering at Osaka University (ILE-Osaka), a target comprised of a polystyrene capsule filled with D₂O liquid and a solution of X-ray tracer materials, such as copper, titanium, or chlorine, was developed. In this study, an additional T₂O doping technique by which tritium can be mixed uniformly has been developed. The T₂O is synthesized by T₂ gas using a CuO oxidation catalyst. The T₂O is agglutinated by cold trap and transferred to a target cell in which a D₂O-solution-filled target is placed. Because polystyrene is slightly permeable for T₂O and D₂O, D₂O is exchanged by T₂O and completely mixed. Thus, a uniform tritium-doped ICF target with various materials can be fabricated. The T₂O synthesizing and doping system is developed and tested using H₂ as a cold run. The H₂O is successfully doped to a D₂O prefilled target at approximately 50% doping. This scheme will be utilized in future fast ignition experiments at ILE-Osaka

Keyword:Inertial confinement fusion, tritium-contained target, fast ignition
Accepted:December 10, 2019

Akira Taguchi, ^a* Takumi Nakamori,^a Yuki Yoneyama,^a Takahiko Sugiyama,^b Masahiro Tanaka, ^c Kenji Kotoh,^d Yu Tachibana, ^e and Tatsuya Suzuki<sup>e

a</sup>University of Toyama, Hydrogen Isotope Research Center, 3190 Gofuku, Toyama 930-8555, Japan
^bNagoya University, Graduate School of Engineering, Furo-cho, Chikusa-ku, Nagoya 464-8603, Japan
^cNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
^dKyushu University, Graduate School of Engineering, 744 Motooka Nishi-ku, Fukuoka 819-0395, Japan
^eNagaoka University of Technology, Graduate School of Engineering, Nagaoka, Japan

Abstract
Using either single H₂ and D₂ or H₂-D₂ mixed gases, the sorption abilities of CHA (chabazite)-type zeolites ion-exchanged with K, Na, or Ca were studied at 77, 201, and 250 K. The LTA (Linde Type A) (3A)and FAU (faujasite)-type zeolites were also examined for comparison. The pore diameters in these materials were found to decrease on the order of FAU ＞ Ca-CHA ＞ [K-CHA, Na-CHA, and LTA(3A)]. The quantities of D₂ adsorbed on these zeolites were larger than the amounts of H₂. At higher temperatures, the CHA-type zeolites having smaller pores exhibited superior D₂/H₂ selectivity compared with the LTA(3A) and FAU, suggesting that hydrogen isotope separation using zeolites is affected by pore size.

Keyword: Zeolite, porous materials, hydrogen isotope separation
Accepted: December 10, 2019

波多野雄治
富山大学学術研究部理学系

Abstract
トリチウムは，将来の核融合炉の燃料として期待されるだけではなく，既に長寿命で安定した光源および電 源において活用されており，またニュートリノの質量決定や中性子過剰核の生成，負ミュオン顕微鏡の構築など 物理学の最先端分野で決定的な役割を果たしている．本章では工業的応用と最先端科学研究における活用の両面 から，トリチウム利用の現状と将来展望について述べる．

Keyword: tritium, radioluminescence, nuclear battery, neutron-rich nucleus, muon microscope
Accepted: 2019年10月18日

Yuji Nobuta^a,*, Yuji Hatano^b, Sun Eui Lee^c, Masato Nakayama<sup>b

a</sup> Graduate School of Engineering, Hokkaido University, Sapporo, 060-8628, Japan
^b Hydrogen Isotope Research Center, University of Toyama, Toyama, 930-8555, Japan
^c Graduate School of Science and Engineering, University of Toyama, Toyama, 930-8555, Japan

Abstract
In the present study, successive exposure to tritium gas and deuterium gas was performed at elevated temperatures for tungsten samples with irradiation defects created by helium ion irradiation to investigate the effects of hydrogen isotope exchange on tritium removal. It was found that, under the present heating conditions, heating treatment in deuterium gas was more effective for tritium removal than heating in a vacuum. In addition, heating in deuterium gas was effective to remove tritium retained in a deep region of sample comparing to the case of heating in a vacuum. The enhanced tritium release was explained by enhanced tritium detrapping due to trap occupation by deuterium and/or the increase in probability of surface recombination under deuterium gas exposure.

Keyword: Tritium removal,Bake-out, Hydrogen isotope exchange,Tungsten,Irradiation defect
Accepted:13 April 2020

Mingzhong Zhao,^a Moeko Nakata^b, Fei Sun^c, Yuji Hatano^d, Yoji Someya^e, Kenji Tobita^e,Yasuhisa Oya<sup>b

a</sup>Shizuoka University, Graduate School of Science and Technology
^bShizuoka University, Graduate School of Integrated Science and Technology
^cShizuoka University, Faculty of Science
^dUniversity of Toyama, Organization for Promotion of Research, Hydrogen Isotope Research Center
^eNational Institutes for Quantum and Radiological Science and Technology

Abstract
The deuterium (D) permeation behavior for 1 displacement per atom Fe²⁺ damaged tungsten (W) was studied by the gas-driven permeation method and compared with undamaged W. The results of thermal desorption spectroscopy showed that dislocation loops and voids were formed in damaged W. It was found that the D permeation behavior in W was affected by irradiation defects. The effective diffusivity and permeability in the damaged W were lower than that in undamaged W. However, the difference in effective diffusivity and permeability between the undamaged sample and the damaged sample was reduced with increasing the heating temperature. Under 965 K, which was enough for D detrapping from voids, the permeability for damaged W was consistent with that for undamaged W.

Keyword:Tungsten, irradiation damage, gas-driven permeation
Accepted: December 10, 2019

M. Khalid Hossain, ^a,b* Kenichi Hashizume,^a Shinnosuke Jo,^a Kaname Kawaguchi,^a and Yuji Hatano<sup>c

a</sup>Kyushu University, Interdisciplinary Graduate School of Engineering Science, Department of Advanced Energy Engineering Science, Kasuga, Fukuoka 816-8580, Japan
^bBangladesh Atomic Energy Commission, Atomic Energy Research Establishment, Dhaka 1349, Bangladesh
^cUniversity of Toyama, Organization for Promotion of Research, Hydrogen Isotope Research Center, 3190 Gofuku, Toyama 930-8555, Japan

Abstract
Hydrogen release behavior from rare earth oxides (REOs) (Y₂O₃, Sm₂O₃, Eu₂O₃, Gd₂O₃, Dy₂ O₃, Er₂O₃, and Yb₂O₃) exposed to 133 Pa of deuterium (D₂) gas or 2 kPa of heavy water (D₂O) vapor at 873 K for 5 h was examined using thermal desorption spectroscopy. Hydrogen solubility and diffusivity in Y₂O₃, Gd₂O₃, Dy₂O₃, Er₂O₃, and Yb₂O₃ exposed to a deuterium-tritium gas mixture (5% to 7% T, 133 Pa) at 873 K and 973 K for 5 h were determined using a tritium imaging plate method. The structural and morphological properties of sintered disk specimens of those REOs were evaluated using an X-ray diffractometer and a scanning electron microscope. From the obtained results, the REO materials were clearly categorized into two kinds in terms of their crystal structure and hydrogen solubility: Monoclinic specimens of Sm₂O₃, Eu₂O₃, and Gd₂O₃ had relatively high hydrogen solubility and diffusivity, while cubic Y₂O₃, Dy₂O₃, Er₂O₃, and Yb₂O₃ had lower ones. The present study suggests that the cubic REOs could be suitable in a nuclear fusion reactor as the tritium barrier materials.

Keyword: Rare-earth oxide, D₂ and D₂O exposure, tritium exposure, TIP and TDS methods, hydrogen dissolution and diffusion.

Accepted: January 28, 2020

Shota Nakazawa ^a, Kazuki Nakamura ^a, Hikari Fujita ^b, Hans Maier ^c, Thomas Schwarz-Selinger ^c, Yuji Hatano ^d, Naoko Ashikawa ^{e, f}, Wataru Inami ^a,Yoshimasa Kawata ^a, Takumi Chikada <sup>a, *

a</sup> Shizuoka University, Shizuoka, Japan
^b The University of Tokyo, Tokyo, Japan
^c Max-Planck-Institut fur Plasmaphysik, 85748, Garching, Germany
^d University of Toyama, Toyama, Japan
^e National Institute for Fusion Science, Gifu, Japan
^f The Graduate University for Advanced Studies, Gifu, Japan

Abstract
Tritium permeation and retention are serious problems in D-T fusion reactors from the viewpoint of fuel efficiency and radiological safety. Functional ceramic coatings have been intensively studied for the development of tritium permeation barriers for several decades, while reports about tritium retention in the ceramic coatings are scarce. Moreover, irradiation may affect tritium retention in fusion materials,which is important to precisely evaluate tritium inventory in the reactor. In this study, the gamma-ray irradiation effect on deuterium retention in reduced activation ferritic/martensitic steel and three kinds of ceramic coatings were investigated through deuterium exposure, gamma-ray irradiation using cobalt-60 gamma-ray sources and deuterium depth profile measurements. The amount of deuterium retention in yttrium oxide, silicon carbide, and zirconium oxide coatings decreased after the irradiation in the dose rate of 2.43 Gy s^-1, while no clear change in the retention was observed at the lower dose rate. From these results, the gamma-irradiation effect on deuterium retention would have a threshold dose rate. Diffusion and desorption of deuterium would be accelerated by excitation of deuterium via energy transfer from electrons generated by Compton scattering.

Keyword: Hydrogen isotope, Retention, Gamma-ray irradiation, Reduced activation ferritic/martensitic Ceramics coating
Accepted: 10 June 2020

Y. Oya ^{a, *}, F. Sun ^a, Y. Yamauchi ^b, Y. Nobuta ^b, M. Shimada ^c, C.N. Taylor ^c, W.R. Wampler ^d, M. Nakata ^a, L.M. Garrison ^e, Y. Hatano <sup>f

a</sup> Shizuoka University, Shizuoka, Japan
^b Hokkaido University, Sapporo, Japan
^c Idaho National Laboratory, Idaho Falls, ID, USA
^d Sandia National Laboratories, Albuquerque, NM, USA
^e Oak Ridge National Laboratory, Oak Ridge, TN, USA
^f University of Toyama, Toyama, Japan

Abstract
Single crystalline W (tungsten) samples irradiated at 633, 963 and 1073 K by neutrons to a damage level of 0.1 dpa were exposed to a high-flux D (deuterium) plasma at 673, 873 and 973 K, respectively, in TPE (Tritium Plasma Experiment) at INL (Idaho National Laboratory). Deuterium desorption was analyzed by TDS (Thermal Desorption Spectroscopy), and D depth profiles were determined by NRA (Nuclear Reaction Analysis) at SNL (Sandia National Laboratories). HIDT (Hydrogen Isotope Diffusion and Trapping) simulation code was applied to evaluate D behavior for neutron-damaged W at higher temperature. The D retention at depths up to 3 mm for the neutron-damaged sample at 673 K was two orders of magnitude larger than that for undamaged tungsten, and its D desorption spectrum had a single broad stage at around 900 K. As the neutron irradiation/plasma exposure temperature increased, D retention was largely reduced, and the desorption temperature was shifted to higher temperatures above 1100 K. The D depth profiles by NRA also showed D migration toward bulk by higher temperature irradiation, compared to undamaged W.The HIDT simulation indicated that the major binding energy of D was changed from 1.43 eV to 2.07 eV at higher neutron irradiation and plasma exposure temperatures, suggesting that some vacancies and small vacancy clusters would aggregate to form larger voids, or depopulation of weak traps at high D plasma exposure temperatures. It can be said that more stable trapping sites played dominant roles in the D retention at higher neutron irradiation and plasma exposure temperature. The binding energy by HIDT simulation was almost consistent with the reported value by TMAP, but the consideration of not only total D retention measured by TDS but also D depth profile by NRA led to the more accurate D behavior in neutron-damaged W.

Keyword: Single crystal tungsten, Deuterium retention,Neutron irradiation, TDS, NRA, HIDT,Nuclear fusion
Accepted: 10 June 2020

原　正憲

概要
トリチウムの量は放射性核種としてはBqの単位，水素同位体としてみるのであればmolの単位あるいは分子 数で議論される．この二つの観点からの数字を実感することは，安全な実験計画，取り扱いにつながる．そこで， この二つの側面から，化学物質としてのトリチウム，トリチウムによる被ばくについて簡略化した仮想的な系で 述べた．併せて，トリチウムを使用する際に有用と思われる数値をまとめた．

Keyword: tritium, hydrogen isotope, range, isotope exchange, internal dose
Accepted: 2020年5月1日

M Tokitani¹, M Miyamoto², S Masuzaki¹, Y Hatano³, S E Lee³, Y Oya⁴, T Otsuka⁵, M Oyaidzu⁶, H Kurotaki⁶, T Suzuki<sup>6

1</sup> National Institute for Fusion Science, National Institutes of Natural Sciences, Oroshi, Toki, Gifu 509-5292, Japan
² Shimane University, Matsue, Shimane 690-8504, Japan
³ University of Toyama, Toyama 930-8555, Japan
⁴ Shizuoka University, Shizuoka 422-8529, Japan
⁵ Kindai University, Osaka 577-8502, Japan
⁶ QST, Rokkasho Aomori 039-3212, Japan

Abstract
Surface characterization of bulk tungsten tiles (W lamellae) used during the first campaign of JET with the ITER-Like Wall (JET-ILW) was performed by means of microscopy and tritium imaging techniques. This is the first report regarding very detailed structural studies of W lamellae from the JET-ILW divertor. A special feature of the W lamellae installed in JET is the intrinsic network of micro-cracks detected on surfaces of the as-manufactured material. Analyses of different W lamellae samples on the plasma-facing surface reveal two types of surface structures caused by plasma impact: areas with strong erosion and regions of mild plasma interaction. In regions of strong erosion, a thin modified layer (thickness of ~20 nm) with a high density of defects including bubble-like structures has been formed. In addition, features indicating melting along edges of micro-cracks with the micro-scale plastic deformation have been identified.

Published: 26 February 2020

大塚哲平，波多野雄治^１） OTSUKA Teppei and HATANO Yuji^１）
近畿大学理工学部，^１）富山大学学術研究部理学系

概要
トリチウムを利用するうえで理解しておくべき法令について簡単にまとめ，国内の主要トリチウム実験施設である富山大学および量子科学研究開発機構，欧州および米国の実験室規模を紹介する．また，施設における共同研究利用の申請方法や，トリチウムを含む試料の購入方法について解説する．

Keyword: radioisotope, tritium, facility, regulatory requirements
Accepted: ２０２０年７月６日

大塚哲平，原　正憲^１） OTSUKA Teppei and HARA Masanori^１）
近畿大学理工学部，^１）富山大学学術研究部理学系

概要
トリチウム実験を行う際の実験室への入室から，測定の実際，実験中の安全対策および個人被ばく防護のための装具の着用，実験後の後片付けについて紹介する．

Keyword: safety, personal protection equipment, smear, contamination
Accepted: ２０２０年７月６日

S.E. Lee ^a,*, Y. Hatano ^a, M. Hara ^a, S. Masuzaki ^b, M. Tokitani ^b, M. Oyaizu ^c, H. Kurotaki ^c,D. Hamaguchi ^c, H. Nakamura ^c, N. Asakura ^c, Y. Oya ^d, J. Likonen ^e, A. Widdowson ^f, S. Jachmich ^g, K. Helariutta ^h, M. Rubel ⁱ, JET Contributors<sup>1

a</sup> University of Toyama, Toyama 930-8555, Japan
^b National Institute for Fusion Science, Oroshi 322-6, Toki 509-5292, Japan
^c National Institutes for Quantum and Radiological Science and Technology (QST), Rokkasho Aomori 039-3212, Japan
^d Shizuoka University, Shizuoka, 422-8529, Japan
^e VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Finland
^f Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX 14 3DB, UK
^g Association Euratom-Etat Belge, ERM-KMS, Brussels Belgium
^h University of Helsinki, P.O. Box 55, FI-00014 University of Helsinki, Finland
ⁱ KTH Royal Institute of Technology, 100 44 Stockholm, Sweden

Abstract
Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced Xray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ～ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.

Keywords: Tritium analysis, Beryllium, Radiography, Joint European Torus, ITER-like wall
Accepted: 14 August 2020

V.Kh. Alimov^1,2,3, Y. Hatano⁴, T. Kuwabara⁵, T. Toyama¹, Y. Someya⁶ and A.V. Spitsyn<sup>2

1</sup> Institute for Materials Research, Tohoku University, Oarai, Ibaraki, 311-1313, Japan
² National Research Centre ‘Kurchatov Institute’, Moscow 123182, Russian Federation
³ A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russian Federation
⁴ Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan
⁵ Graduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan
⁶ National Institutes for Quantum and Radiological Science and Technology, Rokkasho 039-3212, Japan

Abstract
We examine the effect of neutron irradiation on the release of deuterium from tungsten at 573 K to understand the efficiency of tritium removal by baking out at moderate temperatures. Tungsten samples, undamaged and neutron-irradiated to a damage level of approximately 0.016 displacements per atom, are exposed to low-energy (108 eV), high-flux (3.0 × 10²¹ to 9.4 × 10²¹ m⁻² s⁻¹) deuterium plasma at temperatures ranging from 573 to 773 K to an ion fluence of 1.1 × 1025 m−2. At each exposure temperature, two undamaged and two neutron-irradiated tungsten samples are exposed to plasma. The deuterium content in the tungsten samples is measured by thermal desorption spectrometry soon after the plasma exposure and after post-plasma annealing at 573 K for 30 h. It is found that: (i) the deuterium retention in the neutron-irradiated tungsten samples is significantly higher than that in the undamaged tungsten samples; (ii) annealing at 573 K of undamaged tungsten samples pre-exposed to deuterium plasma at 573–773 K leads to an almost complete (60%–99%) release of deuterium from the samples; (iii) annealing at 573 K of neutron-irradiated tungsten samples pre-exposed to deuterium plasma at 573–773 K leads to a significant (8%–20%) release of deuterium from the samples.

Accepted:6 July 2020

Takeshi TOYAMA¹⁾, Miyuki YAJIMA²⁾, Noriyasu OHNO³⁾, Tatsuya KUWABARA³⁾, Vladimir Kh. ALIMOV^1,4,5) and Yuji HATANO<sup>6)

1)</sup>Institute for Materials Research, Tohoku University, Oarai 311-1313, Japan
²⁾National Institute for Fusion Science, Toki 509-5292, Japan
³⁾Graduate School of Engineering, Nagoya University, Nagoya 464-8063, Japan
⁴⁾National Research Centre “Kurchatov Institute”, Moscow 123182, Russia
⁵⁾A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russia
6)Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan

Abstract
This overview presents recent results regarding hydrogen isotope absorption and emission dynamics in neutron-irradiated tungsten (W) using our recently developed Compact Diverter Plasma Simulator (CDPS), a linear plasma device in a radiation-controlled area. Neutron irradiation to 0.016 - 0.06 displacement per atom resulted in a significant increase in deuterium (D) retention due to trapping effects of radiation-induced defects.We analyzed the dependency of D retention on the D plasma fluence by exposing neutron-irradiated pure W to D plasma at 563K over a range of D fluence values. The total retention was revealed to be proportional to the square root of D fluence, indicating that the implanted D atoms first occupy the defects caused by neutron-irradiation near the surface and then the defects located in deeper regions. We further investigated the effects of post-plasma annealing on D emission; neutron-irradiated pure W was exposed to D plasma at 573K and was then annealed at the same temperature for 30 hours. Approximately 10% of the absorbed D was released by annealing, suggesting that a heat treatment of the plasma-facing component of a fusion reactor at moderately elevated temperatures could contribute to the removal of accumulated hydrogen isotopes. The experimental results obtained in this study were only available by investigating neutron-irradiated specimens with the CDPS system, which will be essential for future studies of material behavior and plasma-wall interactions in the fusion reactor environment.

Keyword: tungsten, neutron irradiation, TDS, deuterium
Accepted: 29 September 2020

M. Khalid Hossain ^a,b,1,*, Kenichi Hashizume ^a, Yuji Hatano <sup>c

a</sup> Department of Advanced Energy Engineering Science, Interdisciplinary Graduate School of Engineering Science, Kyushu University
^b Atomic Energy Research Establishment, Bangladesh Atomic Energy Commission
^cHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama

Abstract
Proton-conducting oxides have potential applications in hydrogen sensors, hydrogen pumps, and other electrochemical devices including the tritium purification and recovery systems of nuclear fusion reactors. Although the distribution of hydrogen (H) in such oxide materials is an important aspect, its precise measurement is difficult. In the present study, the hydrogen solubility and diffusivity behavior of BaZr_0.9Y_0.1O_2.95 (BZY),BaZr_0.955Y_0.03Co_0.015O_2.97(BZYC), and CaZr_0.9In_0.1O_2.95 (CZI) were studied using tritiated heavy water vapor i.e., DTO (～2 kPa, tritium (T) = 0.1%) by converting the photostimulated luminescence (PSL) values of the imaging plate (IP). The samples were exposed to DTO vapor at 673 K for 2 h or at 873 K for 1 h. The disc-shaped oxide specimens (diameter～7.5 mm; thickness～ 2.3 mm; theoretical density (TD) greater than 98%) were prepared by conventional powder metallurgy. The IP images of the specimen surfaces of all the three materials T-exposed revealed that BZY showed the most uniform T distribution with the highest tritium activity. The cross-sectional T concentration profiles of the cut specimens showed that T diffused deeper into BZY and BZYC than into CZI. The hydrogen solubility and diffusivity in the CZI specimen were lower than that in the BZY and BZYC specimens.This suggested that barium zirconates were more favorable proton conductors than calcium zirconates.

Keywords: Barium zirconate (BaZrO₃),Calcium zirconate (CaZrO₃),Photostimulated luminescence (PSL),Tritium imaging plate (TIP),Hydrogen solubility and diffusivity Fusion reactors materials
Accepted: 1 December 2020

Weizhe Gao,^[a] Lisheng Guo,^*[a] Yu Cui,^[a] Guohui Yang,^[a] Yingluo He,^[a] Chunyang Zeng,^[b]Akira Taguchi,^[c] Takayuki Abe,^[c] Qingxiang Ma,^[d] Yoshiharu Yoneyama,^[a] and Noritatsu Tsubaki<sup>*[a]

[a]</sup> W. Gao, Dr. L. Guo, Y. Cui, Dr. G. Yang, Y. He, Dr. Y. Yoneyama,Prof. N. Tsubaki Department of Applied Chemistry, School of Engineering University of Toyama
^[b] Dr. C. Zeng China Petroleum Chemical Industry Federation
^[c] Dr. A. Taguchi, Prof. T. Abe Hydrogen Isotope Research Center University of Toyama
^[d] Dr. Q. Ma State Key Laboratory of High-efficiency Coal Utilization and Green Chemical Engineering College of Chemistry and Chemical Engineering Ningxia University

Abstract
An oxide-zeolite (ZnCr₂O₄-ZSM-5) catalyst for directly converting CO₂ to aromatics was designed and developed. It showed high PX/X (the C-mol ratio of p-xylene to all xylene) and PX/aromatics (the C-mol ratio of p-xylene to aromatics) ratios, which reached 97.3 and 63.9 %, respectively.

Keywords: aromatics,CO₂,conversion,green chemistry,oxide-zeolite catalyst,ZSM-5 zeolite
Accepted manuscript online: October 29, 2020

Takuya Okazaki, Tatsuya Orii, Shin Yinn Tan, Tomoaki Watanabe, Akira Taguchi, Faidz A. Rahman, Hideki Kuramitz

Takuya Okazaki - Department of Environmental Biology and Chemistry, Graduate School of Science and Engineering for Research, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan; Department of Applied Chemistry, School of Science and Technology, Meiji University, 1-1-1, Higashimita, Tama-ku, Kawasaki, Kanagawa 214-8571, Japan;
Tatsuya Orii - Department of Environmental Biology and Chemistry, Graduate School of Science and Engineering for Research, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
Shin-Yinn Tan - Faculty of Engineering and Green Technology, Universiti Tunku Abdul Rahman, Jalan Universiti, Bandar Barat, 39100 Kampar, Malaysia
Tomoaki Watanabe - Department of Applied Chemistry, School of Science and Technology, Meiji University, 1-1-1, Higashimita, Tama-ku, Kawasaki, Kanagawa 214-8571, Japan
Akira Taguchi - Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
Faidz A. Rahman - Lee Kong Chian Faculty of Engineering and Science, Universiti Tunku Abdul Rahman, Sungai Long, 43000 Selangor, Malaysia
Hideki Kuramitz - Department of Environmental Biology and Chemistry, Graduate School of Science and Engineering for Research, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract
We present an electrochemical long period fiber grating (LPFG) sensor for electroactive species with an optically transparent electrode. The sensor was fabricated by coating indium tin oxide onto the surface of LPFG using a polygonal barrel-sputtering method. LPFG was produced by an electric arc-induced technique. The sensing is based on change in the detection of electron density on the electrode surface during potential application and its reduction by electrochemical redox of analytes. Four typical electroactive species of methylene blue, hexaammineruthenium(III), ferrocyanide, and ferrocenedimethanol were used to investigate the sensor performance. The concentrations of analytes were determined by the modulation of the potential as the change in transmittance around the resonance band of LPFG. The sensitivity of the sensor, particularly to methylene blue, was high, and the sensor responded to a wide concentration range of 0.001 mM to 1 mM.

https://dx.doi.org/10.1021/acs.analchem.0c01062
Accepted: June 19, 2020

F. Kodera^*,a, N. Yoshida^b, M. Inoue ^c, M. Umeda^d, and A. Miyakoshi^a

a Department of Materials Chemistry, National Institute of Technology, Asahikawa College, Shunkodai 2-2-1-6, Asahikawa, Hokkaido 071-8142, Japan
b Graduate School of Environmental Science, Hokkaido University, Kita-ku N10W5, Sapporo, Hokkaido 060-0810, Japan
c Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
d Department of Materials Science and Technology, Faculty of Engineering, Nagaoka University of Technology, Kamitomioka 1603-1, Nagaoka, Niigata 940-2188, Japan

Abstract
We developed a new method for the production of large amounts of carbon–nickel (C–Ni) powder, which contained multilayer graphene-covered Ni particles as the main component, using multimode microwave-assisted catalytic CH₄ decomposition.
Electrocatalytic hydrogen evolution reaction of the C–Ni powder in acidic solution was investigated using electrochemical method.
The C–Ni powder can be used as a noble metal-free electrocatalyst to implement the hydrogen evolution reaction with good electrochemical performance in the acidic solution.