M. Shoji^a,⁎, T. Aso^b, M. Hara^c, R. Benii^d, Y. Kato^e, T. Furusawa^e, T. Yoshimura^e
a Division of Radioisotope and Radiation Research, Life Science Research Center, University of Toyama, Sugitani 2630, Toyama 930-0194, Japan
^b Electronics and Computer Engineering, National Institute of Technology, Toyama College, Ebie-neriya 1-2, Imizu city, Toyama 933-0293, Japan
^c Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^d Advanced Course, National Institute of Technology, Toyama College, Ebie-neriya 1-2, Imizu city, Toyama 933-0293, Japan
^e Hitachi, Ltd., Mure 6-22-1, Mitaka-city, Tokyo 181-8622, Japan

Abstract
The ¹²⁵I pulse-height spectra via a liquid scintillation counter (LSC) displayed notable variations. The counting efficiencies of higher and lower energy peaks increased and decreased, respectively, with the enhancement of the amount of high atomic numbered elements within the cocktails. This tendency was ascribed to the increasing probability of the interaction of photons with the scintillation cocktail. Moreover, it was noted that the shape of a ¹²⁵I spectrum strongly depends on the amount of high atomic numbered elements.

Keywords: liquid scintillation counting, ¹²⁵I ,Electron capture decay, Geant4 Monte Carlo simulation
Accepted: 30 April 2018

M. Rubel^a,⁎, A. Widdowson^b, J. Grzonka^c,d, E. Fortuna-Zalesna^c, Sunwoo Moon^a, P. Petersson^a,
N. Ashikawa^e, N. Asakura^f, D. Hamaguchi^f, Y. Hatano^g, K. Isobe^f, S. Masuzaki^e, H. Kurotaki^f,
Y. Oya^h, M. Oyaidzu^f, M. Tokitani^e, JET Contributors<sup>1,2

a</sup>Royal Institute of Technology (KTH), 10044, Stockholm, Sweden
^bCulham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, UK
^cWarsaw University of Technology, 02-507, Warsaw, Poland
^dInstitute of Electronic Materials Technology, 01-919, Warsaw, Poland
^eNational Institute for Fusion Science, Toki, 509-5292, Japan
^fNational Institutes for Quantum, Radiological Science and Technology, Rokkasho, 039-3212, Japan
^gToyama University, Hydrogen Isotope Research Center, Gofuku, 3190, Toyama, 930-8555, Japan
^hShizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529, Japan

Abstract
Operation of the JET tokamak with beryllium and tungsten ITER-like wall provides unique opportunity for detailed studies on dust generation: quantity, morphology, location, etc. The programme carried out in response to ITER needs for safety assessment comprises: (i) remotely controlled vacuum cleaning of the divertor; (ii) local sampling of loosely bound matter from plasma-facing components (PFC); (iii) collection of mobilized dust on various erosion-deposition probes located in the divertor and in the main chamber. Results of comprehensive analyses performed by a number of complementary techniques, e.g. a range of microscopy methods, electron and ion spectroscopy, liquid scintillography and thermal desorption, are summarized by following points:
(a) Total amount of dust collected by vacuum cleaning after three campaigns is about 1–1.4 g per campaign (19.1–23.5 h plasma operation), i.e. over 100 times smaller than in JET operated with carbon walls (i.e. in JETC).
(b) Two major categories of Be dust are identified: flakes of co-deposits formed on PFC and droplets (2–10 μm in diameter). Small quantifies, below 1 g, of Be droplets and splashes are associated mainly with melting of beryllium limiters.
(c) Tungsten dust occurs mainly as partly molten flakes originating from the W-coated tiles.

Keywords: JET tokamak, ITER-like wall, Dust, Beryllium, Tungsten
Accepted: 9 March 2018

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

a</sup> National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan
^b Shimane University, Matsue, Shimane 690-8504, Japan
^c Shizuoka University, Shizuoka 422-8529, Japan
^d University of Toyama, Toyama 930-8555, Japan
^e Kindai University, Higashi-Osaka, Osaka, 577-8502, Japan
^f QST, Rokkasho, Aomori 039-3212, Japan
^g EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB, UK
^h University of Helsinki, PO Box 64, FI-00560 Helsinki, Finland
ⁱ Royal Institute of Technology (KTH), 100 44 Stockholm, Sweden

Abstract
Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified (“geological-like”) mixed-material deposition layer which mainly included Be and Ni with the thickness of ∼2 μm. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper.

Keywords: JET-ILW,Divertor,Erosion-deposition,Fuel inventory
Accepted: 22 January 2018

Masanori Hara^a,⁎, Ryota Uchikawa^a, Yuji Hatano^a, Masao Matsuyama^a, Tsukasa Aso^b, Tomohiko Kawakami^c, Takeshi Ito^c
a Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama City, Toyama 930-8555, Japan
^b Electronics and Computer Engineering, National Institute of Technology, Toyama College, 1-2 Ebie-neriya, Imizu City, Toyama 933-0293, Japan
^c KAKEN Company Limited, 1044 Horimachi, Mito City, Ibaraki 310-0903, Japan

Abstract
Tritium beta-ray induced X-ray spectra in various gas mediums were simulated by Monte Carlo simulation using Geant4 tool kit. The simulated beta-ray induced X-ray spectrum (s-BIX spectrum) was composed of the bremsstrahlung component and characteristics X-rays from constituent elements. The total number of photons in s-BIX spectrum decreased with increasing pressure of medium except argon. In argon medium, the characteristics X-ray of argon was generated by beta particles from tritium decay, and the contribution of Ar-Kα and -Kβ compensated the reduction of bremsstrahlung generated by solid matter with increasing argon pressure. At 0.001 atm of medium pressure, the total counts in s-BIX spectrum was independent from gas medium. Therefore, the gas medium dependence in BIXS at low pressure (less than 0.001 atm) was not serious issue.

Keywords: Tritium beta-ray induced X-ray spectrometry, Bremsstrahlung, Monte Carlo simulation, Geant4
Accepted: 19 April 2018

V. Kh Alimov ^{a, b, c}, O.V. Ogorodnikova ^{a, *}, Y. Hatano ^b, YuM. Gasparyan ^a, V.S. Efimov ^a, M. Mayer ^d, Z. Zhou ^e, M. Oyaizu ^f, K. Isobe ^f, H. Nakamura ^f, T. Hayashi <sup>f

a</sup> National Research Nuclear University, MEPhI (Moscow Engineering Physics Institute), Moscow 115409, Russia
^b Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan
^c A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow 119071, Russia
^d Max-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany
^e University of Science and Technology Beijing, Beijing 100083, China
^f National Institutes for Quantum and Radiological Science and Technology, Rokkasho 039-3212, Japan

Abstract
Surface topography of and deuterium (D) retention in reduced activation ferritic-martensitic Eurofer’97 and ferritic oxide dispersion strengthening ODS-16Cr steels have been studied after exposure at 600 K to low-energy (70 and 200 eV), high-flux (~1022 D/m2s) pure D and D-10%He plasmas with D fluence of 2×10²⁵ D/m². The methods used were scanning electron microscopy, energy-scanning D(³He,p)⁴He nuclear reaction, and thermal desorption spectroscopy. As a result of the plasma exposures, nano-sized structures are formed on the steel surfaces. After exposure to pure D plasmas, a significant fraction of D is accumulated in the bulk, at depths larger than 8 mm. After exposures to D-He plasmas, D is retained mainly in the near-surface layers. In spite of the fact that the He fluence was lower than the D fluence, the He retention in the steels is one order of magnitude higher than the D retention.

Keywords: Deuterium retention, EUROFER'97 and 16Cr-ODS steels, Pure deuterium and helium-seeded deuterium plasmas, Surface morphology
Accepted: 31 January 2018

Mitsuhiro Inoue ^a, Yuta Takahashi^b, Takayuki Abe^a, Minoru Umeda ^b,*
a Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Gofuku 3190,Toyama 930-8555, Japan
^b Department of Materials Science and Technology, Graduate School of Engineering, Nagaoka University of Technology, Kamitomioka 1603-1, Nagaoka, Niigata 940-2188, Japan

Abstract
In the present study, the reaction selectivity of methanol oxidation and O₂ reduction at Pte RueC co-sputtering samples has been described. The samples were prepared using the polygonal barrel-sputtering method and tested by linear sweep voltammetry in 0.5 mol/L H₂SO₄ containing methanol and O₂. When sputtering was performed at a constant Ar gas pressure of 3 Pa, methanol oxidation became highly selective for the samples prepared at AC powers of 30 and 50 W. This characteristic was likely induced by the existence of O₂ reduction intermediates, which can enhance the oxidation of CO, a methanol oxidation intermediate. The samples displaying O₂ reduction selectivity were obtained at Ar gas pressure ranging between 0.4 and 3 Pa (AC power: 100 W). The O₂ reduction selectivity may be attributed to the deposition of the electron-rich metallic Ru, which exhibits poor methanol oxidation performance. Based on these results, PteRueC samples that show reaction selectivity in methanol oxidation and O₂ reduction are independently prepared by the polygonal barrel-sputtering method.

Keywords:Electrocatalysts, Direct methanol fuel cells, Reaction selectivity, Polygonal barrel-sputtering method, Co-sputtering technique
Accepted:25 November 2018

Qilai Zhou ^{a, *}, Keisuke Azuma ^b, Akihiro Togari ^b, Miyuki Yajima ^c, Masayuki Tokitani ^c, Suguru Masuzaki ^c, Naoaki Yoshida ^d, Masanori Hara ^e, Yuji Hatano ^e, Yasuhisa Oya <sup>a

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

Abstract
The purpose of this study is to elucidate helium (He) retention behavior in tungsten (W) under simultaneous He and hydrogen (H) irradiation. Polycrystalline-W was irradiated by He⁺ and H₂⁺ simultaneously with the energy of 1.0 keV and 3.0 keV. He⁺ fluences were (0.5, 1.0, 10) ×10²¹ He⁺ m^-2 and H^₂+ fluence was 1.0 × 10²² H⁺ m^-2,respectively. After irradiation, He desorption behavior was investigated by high temperature thermal desorption spectroscopy (HT-TDS) in the temperature range of R.T.-1773 K. Micro-structure changes of W after irradiation were observed by TEM. It was found that simultaneous irradiation with different H₂⁺ energy significantly changed He retention behavior. 1.0 keV H₂⁺ suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H₂⁺ facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H₂⁺.

Keywords: Simultaneous irradiation, He retention, TDS, Tungsten, Hydrogen isotope
Accepted: 22 February 2018

V.Kh Alimov ^{a, b, c}, Y. Hatano ^{a, *}, K. Sugiyama ^d, M. Zibrov ^{d, e, f}, T. Schwarz-Selinger ^d, W. Jacob <sup>d

a</sup> Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama, 930-8555, Japan
^b A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow, 119071, Russia
^c National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, 115409, Russia
^d Max-Planck-Institut für Plasmaphysik, D-85748, Garching, Germany
^e Physik-Department E28, Technische Universitat München, D-85748, Garching, Germany
^f Department of Applied Physics, Ghent University, B-9000, Ghent, Belgium

Abstract
Eight samples of F82H ferritic/martensitic steel were irradiated at 300 K with 20 MeVW ions to the damage level of 0.54 displacements per atom at the damage peak. Three samples were afterwards annealed in vacuum at 423 K for 72 h and then at 373 K for 106 h. Three other samples were annealed in H₂ atmosphere at 100 kPa at the same annealing temperatures and durations. All samples were exposed at room temperature to D₂O vapor at the partial pressures in the range from 3 to 6 kPa for 365 to 1181 days. After termination of the D₂O vapor exposure, the surfaces of the samples were partly covered by small (≤1mm in diameter) drops of D₂O water. Thus, the damaged samples were exposed to a mixture of D₂O vapor and water. Trapping of deuterium at the ion-induced defects in the damage zone was examined by the D(³He, p)⁴He nuclear reaction. It has been found that the W-ion-induced defects generated in the F82H samples are decorated by deuterium diffusing from the surface.

Keywords:F82H steel, Hydrogen isotope retention, Radiation damage, Water vapor
Accepted: 19 April 2018

Kazuto Tokumitsu ^{a, *}, Masao Matsuyama ^b, Kazuki Morita ^a
a Department of Material Engineering, The University of Tokyo, Hongo 7-3-1, Tokyo 113-8656, Japan
^b Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract
Hydrogen storage and permeation alloys have been studied based on their hydrogen affinities. The hydrogen uptake and release behavior of intermetallic compounds is important to understand the mechanism of trapping hydrogen atoms and for the advance to next-generation hydrogen energy materials. Here, we propose the site-controlled model by constitutional vacancies. We studied the uptake of hydrogen atoms in nonstoichiometric B2-type Fe-Al compounds by electrochemically charging hydrogen in an alkali water solution. For the Fe-poor nonstoichiometric alloys Fe₄₆Al₅₄ and Fe₄₈Al₅₂, the release of taken-up hydrogen began immediately after switching off the current and continued for 1 h, whereas no hydrogen was released from Fe₅₀Al₅₀ and Fe₅₂Al₄₈. The maximum amount of hydrogen released from Fe₄₆Al₅₄ charged at 10 V was about 6.30 cm³ per gram of sample, corresponding to an atomic ratio of H/M = 2.26 × 10^-2. This hydrogen concentration is 10⁶ times higher than the solubility of hydrogen in pure a-iron. Such a significant uptake of hydrogen is concluded to be peculiar to the Fe-poor nonstoichiometric compositions, which contain constitutional vacancies. This is the first report to adequately explain the high absorption capacity of Fe-poor Fe-Al intermetallic compounds.

Keywords:Hydrogen absorbing materials, Intermetallics, Point defects, Vacancy formation
Accepted: 21 May 2018

Yasuhisa Oya¹, Keisuke Azuma¹, Akihiro Togari¹, Qilai Zhou¹, Yuji Hatano², Masashi Shimada³, Robert Kolasinski⁴, and Dean Buchenauer^4
1 Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan {oya.yasuhisa,azuma.keisuke.16,togari.akihiro.17, zhou.qilai}@shizuoka.ac.jp
² Hydrogen Isotope Research Center, Organization of Promotion of Research, University of Toyama, Gofuku, Toyama 930-8555, Japan hatano@ctg.u-toyama.ac.jp
³ Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA masashi.shimada@inl.gov
⁴ Chemistry, Combustion and Materials Center, Sandia National Laboratory, Livermore, CA 94550, USA {rkolasi,dabuche}@sandia.gov

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.

Qilai Zhou^a,⁎, Keisuke Azuma^b, Akihiro Togari^b, Miyuki Yajima^c, Masayuki Tokitani^c,
Suguru Masuzaki^c, Naoaki Yoshida^d, Masanori Hara^e, Yuji Hatano^e, Yasuhisa Oya^a
a Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
^b Graduate School of Science and Technology, Shizuoka University, Shizuoka 422-8529, Japan
^c National Institute for Fusion science, Oroshi, Toki, Gifu 509-5292, Japan
^d Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^e Hydrogen Isotope Research Center, University of Toyama, 930-8555, Japan

Abstract
Poly-crystalline tungsten (W) samples were simultaneously irradiated with Helium (He) and Deuterium (D) ions using the triple-ion implantation device. He effect on D retention and transportation was studied using different combination of ion energies and He/D flux ratios in the simultaneous implantation. The experimental results show that D trapping at dislocation loops is significantly reduced in the case of 3 keV He⁺–3 keV D₂⁺ at He/D flux ratios over 0.6. D trapping by stronger trapping sites such as vacancies and vacancy clusters showed less dependence on the flux ratio. On the contrary, the D retention increases at each He/D flux ratio in the case of 3 keV He⁺–1 keV D₂ +compared to only D₂⁺ implantation even the He/D flux ratio reaches a value of 1.0. TEM observations confirmed that dense dislocation loops are formed rather than He bubbles, which is responsible for the enhanced D retention in W.

Keywords: Simultaneous implantation, D retention, Helium, Flux ratio, Transportation, Thermal desorption spectroscopy
Accepted: 15 June 2018

M. Matsuyama^a,⁎, H. Zushi^b, K. Tokunaga^b, A. Kuzmin^c, K. Hanada^b
a Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^b RIAM, Kyushu University, 6-1 Kasuga-Koen, Kasuga, Fukuoka 816-8580, Japan
^c National Institute for Fusion Science, Toki, Gifu 509-5292, Japan

Abstract
Dependence of heating temperature on tritium retention behavior in stainless steel type 316 L (SS316L) has been examined about each process of degassing and tritium exposure. Two kinds of SS316L samples were employed: bare SS316L and plasma-exposed SS316L. The amount of tritium retained in surface layers of a sample was nondestructively measured by β-ray-induced X-ray spectrometry, and changes in the surface state with heating in vacuum were analyzed by X-ray photoelectron spectroscopy (XPS). Significant increase in tritium retention in bare SS316L samples appeared in a degassing temperature region above 600 K. Similar tendency of tritium retention was observed for the plasma-exposed sample. It was seen that the degassing process prior to tritium exposure significantly influenced to the tritium retention behavior. Furthermore, it was suggested from surface analysis by XPS that chemical states of SS316L surface at high temperatures play an important role for tritium retention behavior.

Keywords: Tritium retention, Degassing temperature, Tritium exposure temperature, BIXS, Surface analysis, SS316L
Accepted: 29 May 2018

Yasuhisa Oya^a,⁎, Shodai Sakurada^a, Keisuke Azuma^a, Qilai Zhou^a, Akihiro Togari^a, Sosuke Kondo^b, Tatsuya Hinoki^b, Naoaki Yoshida^c, Dean Buchenauer^d, Robert Kolasinski^d, Masashi Shimada^e, Chase N. Taylor^e, Takumi Chikada^a, Yuji Hatano^f
a Graduate School of Science and Technology, Shizuoka University, Shizuoka 422-8529, Japan
^b Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011, Japan
^c Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^d Energy Innovation Department, Sandia National Laboratories, Livermore, CA 94550, USA
^e Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
^f Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract
To evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He⁺ was introduced into W bulk by 201 – 1000 keV He⁺, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W.

Keywords:He in damaged, W Hydrogen isotope retention behavior, Damaged W Fusion
Accepted: 26 June 2018

V. Kh. Alimov ^{a, b, c, d}, Y. Hatano ^{a, *}, N. Yoshida ^e, N.P. Bobyr ^d, M. Oyaidzu ^f, M. Tokitani ^g, T. Hayashi <sup>f

a</sup> Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama 930-8555, Japan
^b A.N. Frumkin Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow, 119071, Russia
^c National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow 115409, Russia
^d National Research Center “Kurchatov Institute”, Moscow, 123182, Russia
^e Research Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580, Japan
^f National Institutes for Quantum and Radiological Science and Technology, Rokkasho, 039-3212, Japan
^g National Institute for Fusion Science, Toki, 509-5292, Japan

Abstract
Targets of Reduced Activation Ferritic Martensitic (RAFM) steel F82H were exposed to low-energy (200 eV), high flux (about 10²² D/m²s) deuterium (D) plasma at 623-773 K to various D fluences in the range from 1 ×10²⁵ to ^2.5×10²⁶ D/m². The surface morphology of the plasma-exposed targets was examined with a field-emission scanning electron microscope. Cross-sectional observations of nanostructures formed on the F82H target surfaces were performed using a transmission electron microscope equipped with an energy dispersive X-ray spectrometer. It has been shown that nano-sized fiberlike layers are formed on the target surfaces under D plasma exposure. Micro-sized surface morphology pattern depends on the D fluence. As the D fluence increases, clusters of the fiber-like layers begin to be formed and organized into ordered structure.

Keywords:Deuterium plasma, F82H steel, Self-assembly, Surface morphology
Accepted: 17 August 2018

Keisuke Azuma^a,⁎, Akihiro Togari^a, Qilai Zhou^b, Yuji Hatano^c, Naoaki Yoshida^d, Masashi Shimada^e, Chase N. Taylor^e, Dean Buchenauer^f, Robert Kolasinski^f, Takumi Chikada^a,Yasuhisa Oya<sup>a

a</sup> Graduate School of Integrated Science and Technology, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529, Japan
^b Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka, 422-8529, Japan
^c Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, 930-8555, Japan
^d Research Institute for Applied Mechanics, Kyushu University, 6-1, Kasugakoen, Kasuga, Fukuoka, Japan
^e Idaho National Laboratory, Idaho Falls, ID, United States
^f Sandia National Laboratories, Livermore, CA, United States

Abstract
Deuterium (D) retention behaviors for the 3 keV Helium (He⁺) implanted damaged-Tungsten (W) and 10 keV Carbon (C⁺) - 3 keV He⁺ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He⁺ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He⁺ fluence. The D retention in the deeper region was reduced by He⁺ implantation with higher He⁺ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C⁺ - He⁺ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C⁺ - He⁺ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

Keywords:Plasma facing components, Tungsten, Carbon, Helium, Hydrogen isotope retention
Accepted: 7 August 2018

Mitsuhiro Inoue¹ · Asuka Shima² · Kaori Miyazaki¹ · Baowang Lu¹ · Takayuki Abe¹ · Yoshitsugu Sone^3
1 Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, Toyama 930‑8555, Japan
²Japan Aerospace Exploration Agency (JAXA), 7‑44‑1 Jindaiji‑higashi‑machi, Chofu, Tokyo 182‑8522, Japan
³ Japan Aerospace Exploration Agency (JAXA), SOKENDAI, 3‑1‑1 Yoshinodai, Chuo‑ku, Sagamihara, Kanagawa 252‑5210, Japan

Abstract
CO₂methanation catalysts were prepared by co-sputtering with Ru and metal oxides such as TiO₂ and ZrO₂ using the polygonal barrel-sputtering method. The co-sputtering technique not only resulted in the decrease in the reaction temperature but also maintained the deposition of smaller Ru particles during the reaction at higher temperature.

Keywords:Heterogeneous catalysis,Electron microscopy, CO₂ methanation catalysts, Ru–metal oxide co-sputtering, Ru particle size, Polygonal barrel-sputtering method
Accepted:18 March 2018

Baowang Lu, Mitsuhiro Inoue and Takayuki Abe

Abstract
Traditional low-temperature H₂ production from methane needs high operating temperatures (above 500 ℃), generates CO₂ emissions, and shows poor stability. This study focused on more low-temperature stable H₂ production from methane using a carbon-supported Ru catalyst (BS-Ru/C). This catalyst could stably generate H₂ without CO_x at 200 ℃ over 15 days. The temperature for clean H₂ production in conjunction with carbon fixation was found to be 200–250 ℃, far lower than around 550 ℃ for the conventional lowtemperature methane catalytic decomposition (MCD). The BS-Ru/C catalyst enabled the occurrence of steam reforming of methane (SRM) at 260 ℃, which far exceeded other previous low-temperature catalysts reported at around 500 ℃. Although CO₂ was produced during the SRM process from 260 to 500 ℃, no CO was produced. This lower-temperature stable H₂ production technology using the BS-Ru/C catalyst with outstanding ability and stability will provide a helping hand to build a H₂ society in near future.

Accepted:10th January 2018

M. Shimada^a, Y. Oya^b, W.R. Wampler^c, Y. Yamauchi^d, C.N. Taylor^a, L.M. Garrison^e, D.A. Buchenauer^f, Y. Hatano^g
a Idaho National Laboratory, Idaho Falls, ID, USA
^b Shizuoka University, Shizuoka, Japan
^c Sandia National Laboratories, Albuquerque, NM, USA
^d Hokkaido University, Sapporo, Japan
^e Oak Ridge National Laboratory, Oak Ridge, TN, USA
^f Sandia National Laboratories, Livermore, CA, USA
^g University of Toyama, Toyama, Japan

Abstract
Six single crystal tungsten specimens were neutron irradiated to a dose of 0.1 displacements per atom (dpa) at three different irradiation temperatures (633 K, 963 K, and 1073 K) at the High Flux Isotope Reactor in Oak Ridge National Laboratory under the US-Japan PHENIX project. A pair of neutron-irradiated tungsten specimens was exposed to deuterium (D) plasma to D ion fluence of 5.0×10²⁵m⁻² at three different exposure temperatures (673 K, 873 K, and 973 K) at the Tritium Plasma Experiment in Idaho National Laboratory. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (Tritium Migration Analysis Program, TMAP) were used to understand D behavior in neutron-irradiated tungsten. A broad D desorption spectrum from the plasma-exposure temperature up to 1173 K was observed. Total D retention up to 1.9×10²¹m⁻² and near-surface D concentrations up to 1.7×10⁻³ D/W were experimentally measured from the 0.1 dpa neutron-irradiated single crystal tungsten. Trap density up to 2.0×10⁻³ Trap/W and detrapping energy ranging from 1.80 to 2.60 eV were obtained from the TMAP modeling.

Keywords:Tungsten, Single crystal ,Plasma-facing components, Plasma-material interaction, Tritium retention, Neutron-irradiation

Accepted:24 April 2018

Yuji Nobuta^a,⁎, Yuji Hatano^b, Yuji Torikai^c, Masato Nakayama^b
a Graduate School of Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628, Japan
^b Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^c Graduate School of Science and Engineering, Ibaraki University, Nakanarusawa 4-12-1, Ibaraki 316-8511, Japan

Abstract
Tritium-containing polycrystalline tungsten, which was pre-irradiated with helium prior to tritium exposure, was isothermally heated under a vacuum or deuterium gas in order to investigate the effect of the deuterium on tritium removal from the tungsten. It is found that in the case of 3 h of baking at temperatures varying from 423 K and 573 K, tritium retention after baking in deuterium gas became smaller than that in a vacuum. At the baking temperature of 573 K, tritium removal was hastened in the case of baking in deuterium atmosphere. The results indicate that the baking in deuterium atmosphere is effective to remove tritium from tungsten. The acceleration of diffusion of tritium in bulk and recombination of tritium with deuterium at the surface might be possible mechanisms.

Keywords:Tritium removal, Tungsten, Baking, Helium irradiation
Accepted:23 March 2018

Yuji Hatano^a,⁎, Yuna Konaka^b, Hiroto Shimoyachi^b, Takahiro Kenmotsu^c, Yasuhisa Oya^d, Hiroaki Nakamura^e
a Hydrogen Isotope Research Center, University of Toyama, Toyama, 930-8555, Japan
^b Faculty of Science, University of Toyama, Toyama, 930-8555, Japan
^c Faculty of Life and Medical Sciences, Doshisha University, Kyoto, 610-0321, Japan
^d Collage of Science, Shizuoka University, Shizuoka, 422-8529, Japan
^e National Institute for Fusion Science, Toki, 509-5292, Japan

Abstract
Double strand breaks (DSBs) of DNA molecules in tritiated water was examined under sterilized and non-sterilized conditions using a single molecule observation method. The genome DNA of bacteriophage T4 GT7 was immersed in sterilized tritiated water (5.2 MBq/cm³) and non-sterilized tritiated water (4.2 MBq/cm³) for 1, 7 and 14 day(s). Then the length of DNA molecules was measured using a fluorescence microscope after intercalation of fluorescent dye. The dose rate was 1.4–1.7×10⁻² Gy/h and the dose level was 0.41–5.8 Gy. The rate of DSBs induced by β-rays from tritium was successfully evaluated under the sterilized conditions and the value comparable with the DSB rate under γ-ray irradiation (Noda et al., Scientific Reports 7 (2017) 8557) was obtained. The length of DNA molecules in non-sterilized tritiated water was clearly shorter than that in the sterilized tritiated water. This observation suggested that the effects of tritium was far weaker than that of microorganisms (e.g. bacteria) and impurities in water even at the tritium concentration as high as 5.2 MBq/cm³.

Keywords:Tritium, β-rays, DNA, Radiation effects
Accepted:27 November 2018

Mingzhong Zhao ¹,Qilai Zhou², Moeko Nakata², Akihiro Togari², Fei Sun², Yuji Hatano⁴, Naoaki Yoshida⁵, Yasuisa Oya^3
1 Graduate School of Science and Technology, Shizuoka University, Shizuoka, Japan
² Faculty of Science, Shizuoka University, Shizuoka, Japan
³ Graduate School of Integrated Science and Technology, Shizuoka University, Shizuoka, Japan
⁴ Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, Toyama, Japan
⁵ Research Institute for Applied Mechanics, Kyushu University, Fukuoka, Japan

Abstract
Establishment of effective tritium removal method was one of important issues for the development of fusion reactor from the view of fuel recycle and safe operation. The deuterium (D) removal efficiency in tungsten (W) by energetic hydrogen (H) ions under room temperature and baking under 623 K were studied by thermal desorption spectroscopy (TDS). Iron (Fe) damaged W with various damage level by 6 MeV Fe²⁺ was adopted to simulate neutron irradiation damages. To understand the D removal behavior, the desorption of D₂ was measured in-situ by a quadrupole mass spectrometer (QMS) during H₂⁺ implantation and baking. The in-situ results showed that the desorption of D₂ started after H₂⁺ implantation and became slowly with the increment of H2⁺ implantation time. After H₂⁺ implantation, part of D trapped by dislocation loops, vacancy clusters and voids could be removed by hydrogen isotope exchange. However, the removal efficiency by hydrogen isotope exchange decrease obviously as the presence of irradiation damages. The D trapped by dislocation loops and vacancy clusters can be removed by baking with high efficiency. It is worth to note that the D trapped by voids cannot be removed by baking leading to the lower D removal efficiency for W with high damage level.

Keywords:Tungsten, Hydrogen, isotope, exchange, Irradiation damages
First Online: 19 August 2018