S. Akamaru^a, T. Shimazaki^b, M. Kubo^c, T. Abe^a

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bAdvanced Institute for Computational Science, RIKEN, 7-1-26 Minatojima-minami-machi, Chuo-ku, Kobe 650-0047, Japan
^c Fracture and Reliability Research Institute, Graduate School of Engineering, Tohoku University, 6-6-1 Aoba, Aramaki, Aoba-ku, Sendai 980-8579, Japan

Abstract

    The methanation reaction of CO₂ on a Ru nanoparticle supported on TiO₂ catalyst has been investigated by density functional theory (DFT) using the generalized gradient approximation with periodic boundary conditions. Two plausible reaction paths were found for the transformation of CO₂to CH₄ on TiO₂-supported Ru nanoparticles. The origin of the high activity of the catalyst is discussed based on the overall reaction energy diagram obtained from DFT calculations. The CO₂ is readily and stably adsorbed on Ru cluster at moderate temperature as compared with that on bulk Ru surface. It is due to the difference of the Ru structure between the Ru nanoparticle and the bulk Ru surface. The elementary reactions of the hydrogenation of adsorbed CO and of the production of CH₄ are possible to become the rate-determining steps over the methanation reaction, because these two reactions have a higher potential energy barrier than that of other elementary reactions in the overall reaction path. These potential energy barriers for the hydrogenation of CO and the production of CH₄ on TiO₂-supported Ru nanoparticles were lower than those on bulk Ru surface, which explains the high activity of the Ru nanoparticle-loaded TiO₂ catalyst. The lowering of these potential energy barriers can be caused by weak charge transfer between Ru atoms and adsorbed species on the TiO₂-supported Ru nanoparticles. As the results, the catalytic activity of the Ru nanoparticles supported on TiO₂ catalyst is characterized by the structure of Ru nanoparticles and by the weak charge transfer between Ru atoms and adsorbed species.

Keywords: Density functional theory, CO₂ methanation, Ru/TiO₂, Nanoparticle

V. Kh. Alimov¹,², Y. Hatano², K. Sugiyama¹, M. Balden¹, T Höschen¹, M. Oyaidzu¹, J. Roth¹, J. Dorner¹, M. Fuβeder¹, T. Yamanishi³

¹Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
²Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
³Tritium Technology Group, Japan Atomic Energy Agency, Rokkasho 039-3212, Japan

Abstract

    Samples prepared from steels F82H and EUROFER97 were irradiated with 20 MeV W ions at 300K to 0.54 displacements per atom at the damage peak. Damaged and undamaged samples were exposed at elevated temperatures both to deuterium plasma at ion energies of 60 and 200 eV to a fluence of ≈ 10²⁶ Dm^-2 and to D₂ gas at a pressure of 100 kPa. The surface modification after plasma exposure was examined by scanning electron microscopy and Rutherford backscattering spectroscopy. Deuterium depth profiles were determined by the D(³He, p)⁴He nuclear reaction. In damaged steels loaded with deuterium, deuterium decorates the damage profile and the D concentration decreases with increasing temperature. After exposure of the F82H steel to the D plasma W-enriched near-surface layers are formed. The effective concentration of W in the near-surface steel layer depends on plasma exposure conditions.

Keywords: deuterium retention, ion irradiation, reduced activation steels, surface morphology

Y. Oya^a, S. Masuzaki^b, M. Tokitani^b, M. Sato^a, K. Toda^a, H. Uchimura^a, N. Yoshida^c,H. Watanabe^c, Y. Yamauchi^d, T. Hino^d, M. Miyamoto^e, Y. Hatano^f, K. Okuno^a

^aGraduate School of Science, Shizuoka University, Shizuoka, Japan
^bNational Institute for Fusion Science, Gifu, Japan
^cInstitute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
^dGraduate School of Engineering, Hokkaido University, Sapporo, Japan
^eDepartment of Material Science, Shimane University, Matsue, Shimane, Japan
^fHydrogen Isotope Research Center, University of Toyama, Toyama, Japan
^gHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

    To evaluate hydrogen isotope retention enhancement in W by plasma exposure, the stress relieved tung-sten samples were placed at three or four different positions, namely PI (sputtering erosion dominatedarea), DP (deposition dominated area), HL (Higher heat load area) and ER (erosion dominated area) dur-ing 2011 (15th) or 2012 (16th) plasma experiment campaign in Large Helical Device (LHD) at NationalInstitute for Fusion Science (NIFS), Japan and were exposed to ~6700 shots of hydrogen plasma in a 2011 plasma experiment campaign and ~5000 shots in a 2012 plasma campaign. Thereafter, additional 1.0 keV deuterium ion implantation was performed to evaluate the change of hydrogen isotope retention capacity by plasma exposure. It was found that more than 50 times of hydrogen retention enhancement for DPsample was derived compared to that for pure W. In especially, the carbon-dominated mixed-materiallayer would control the hydrogen isotope retention for all the area except for erosion-dominated area,indicating that the chemical structure for carbon-dominant mixed-material layer would govern the Hand D retention enhancement for most area by long-term plasma exposure. Therefore, the surface areafor carbon material would be one of key issues for the determination of hydrogen isotope retention infirst wall, even if all tungsten first walls will be used.

Keywords: Hydrogen isotope retention enhancement, Surface structure, Plasma wall interaction, TDS, LHD

C. N. Taylor¹ , M. Shimada¹ , B. J. Merrill¹ , M. W. Drigert² , D. W. Akers², Y. Hatano³

¹ Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
² Experimental Programs, Idaho National Laboratory, Idaho Falls, ID 83415, USA
³ Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

    Tungsten samples (6mm diameter and 0.2mm thick) were irradiated to 0.025 and 0.3 dpa with neutrons in the High Flux Isotope Reactor at Oak Ridge National Laboratory as part of the US/Japan Tritium, Irradiation and Thermofluids for America and Nippon (TITAN) collaboration. Samples were then exposed to deuterium plasma in Idaho National Laboratory's Tritium Plasma Experiment at 100, 200 and 500 ℃ C to a total fluence of 1×10²⁶ m^-2. Nuclear reaction analysis (NRA) and Doppler broadening positron annihilation spectroscopy (DB-PAS) were performed at various stages to characterize radiation damage and retention. We present the first results of neutron irradiated tungsten characterized by DB-PAS in order to study defect concentration. Two positron sources, ²²Na and ⁶⁸ Ge, probe 6˜ 58μm and through the entire 200μm thick samples, respectively. DB-PAS results reveal clear differences between the various irradiated samples. These results, and a correlation between DB-PAS and NRA data, are presented.

Keywords: neutron, defects, tungsten, Doppler broadening positron annihilation spectroscopy, plasma-facing components

J. Sun^a, W. Niu^a, A. Taguchi^b, T. Abe^b, Y. Yoneyama^a, N. Tsubaki^a

^a Department of Applied Chemistry, School of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

    Tuning hydrocarbon distribution in Fischer-Tropsch synthesis is greatly challenging. By employing three different pathways to deposit trace palladium on Co/H-ZSM5 catalyst, tunable isoparaffin and olefin selectivity was successfully achieved. The impregnated Pd showed poor promotion of Co dispersion and reducibility, producing a slight enhancement of FTS activity and isoparaffin selectivity. The unique mechanical stir during Pd sputtering induced re-dispersion of impregnated Co/H-ZSM5 particles and Pd was deposited with an intimate distance to Co species and with a weak interaction combining zeolite, due to which complete hydrogenation of olefins was achieved. The surface enriched Pd on presputtered Co catalyst was inclined to form Pd-Co nano-alloys, suppressing the chain growth activity by excessive hydrogenation process.

J. Sun^†, X. Li^‡, A. Taguchi^§, T. Abe^§, W. Niu^†, P. Lu^†, Y. Yoneyama^† and N. Tsubaki^†

^† Department of Applied Chemistry, School of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^§Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^‡Tianjin Key Laboratory of Applied Catalysis Science & Technology, School of Chemical Engineering & Technology, Tianjin University, The Collaborative Innovation Center of Chemistry and Chemical Engineering of Tianjin, Tianjin, 300072, People’s Republic of China

Abstract

    For coping with the increasing petroleum crisis, an efficient conversion of syngas (CO + H₂) to gasoline-ranged isoparafins has been paid more and more attention. Here, we report a metallic bifunctional catalyst for this conversion, consisting of highly dispersed Ru nanoparticles (NPs) and H-Beta zeolite support, prepared by a self-made polygonal barrel-sputtering process. The HRTEM and chemisorption results indicated that sputterd Ru NPs exhibited a high metal dispersion of 31.2% with a narrow diameter of 2-4 nm. These metallic Ru NPs were bonded with the acidic zeolite by a weakly physical force, clearly different from the conventional impregnated one. Without any reduction pretreatment, the Ru/H-Beta catalyst could be directly used in Fischer-Tropsch synthesis, showing a CO conversion of 1.6 times as much as the impregnated one. Furthermore, the short distance between sputtered Ru and acidic sites nearby was responsible for the enhanced C_iso/C_n ratio of 4.6, the highest value of gasoline-ranged hydrocarbons among the relevant reports.

Keywords: Fischer-Tropsch synthesis, ruthenium, sputter, isoparaffin, syngas

HATANO Yuji¹⁾, MIYAMOTO Mitsutaka²⁾, SHIMADA Masashi³⁾, UEDA Yoshio⁴⁾and TOKITANI Masayuki⁵

¹⁾Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan^2）Interdisciplinary Faculty of Science and Engineering of Shimane University, ³⁾Fusion Safety Program, Idaho National Laboratory, USA, ⁴⁾Graduate School and School of Engineering, Osaka University, ⁵⁾The Department of Helical Plasma Research, Institute for Fusion Science (NIFS)

Abstract

   *This paper is written in Japanese.

M. Matsuyama, S. Abe, K. Nishimura¹⁾, N. Ashikawa¹⁾, Y. Oya²⁾, K. Okuno²⁾, Y. Yamauchi³⁾, Y. Nobuta³⁾, A. Sagara¹⁾
Hydrogen Iosotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
¹⁾National Institute for Fusion Science, Toki 509-5292, Japan
²⁾Radiochemistry Research Laboratory, Shizuoka University, Shizuoka 422-8529, Japan
³⁾Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo 060-8628, Japan

Keywords: plasma-surface interaction, linear plasma device, tungsten, neutron irradiation, trap, helium, beryllium

V.Kh. Alimov^a, Y. Hatano^a, K. Sugiyama^b, M. Balden^b, M. Oyaidzu^c, S. Akamaru^a,K. Tada^d, H. Kurishita^e, T. Hayashi^c, M. Matsuyama^a

^aHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^b Max-Planck-Institut fur Plasmaphysik, D-85748 Garching, Germany
^cFusion Research and Development Directorate, International Fusion Energy Research Center, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japan
^dDepartment of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8579, Japan
^eInternational Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313, Japan

Abstract

    Surface topography and deuterium retention in polycrystalline hot-rolled W and W-5% Re have been examined after exposure to a low-energy (76 eV), high flux (around 10²² D/m² s) deuterium plasma to an ion fluence of 10²⁶ D/m² at various temperatures. The methods used were confocal laser scanning microscopy and the D(³He, p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 348 to 673 K, small blisters of size in the range from about 1 to about 15 lm, depending on the exposure temperature, are formed on the W and W-5%Re surfaces. In the W-5%Re, the deuterium retention demonstrates its maximum at exposure temperature of 463 K, while in the W this maximum is shifted to 523 K. A difference in the temperature dependence of the D retention for the W and W-5%Re is explained, as a rough approximation, by temperature dependences of the ductility of these materials. 

S. Akamaru, M. Hara, M. Matsuyama

Abstract

    Pressure-composition isotherms and the magnetic susceptibilities of Pd-TM-H (TM = Cu, Au, Pt, and Ir) systems were measured at ambient temperature, and the effects of alloying between Pd and transition metals on the hydrogen storage capability of these Pd-TM alloys were investigated by considering their electronic band structures. All of the magnetic susceptibilities for the Pd-TM-H systems decreased linearly with hydrogen uptake. For the Pd-Cu alloy, the magnetic susceptibility was nearly zero at the terminal composition of hydrogen in the plateau region obtained from the pressure-composition isotherm, and the terminal composition decreased with increasing Cu substitution. These results indicated that the hydrogen-storage capability was proportional to the amount of unoccupied d states in the electronic band structure of the Pd-Cu alloy. The Pd-Au-H system exhibited substantially the same behavior as the Pd-Cu- H system. For the Pd-Pt and Pd-Ir alloys, the magnetic susceptibility at the terminal composition of hydrogen in the plateau exhibited a finite positive value, indicating that the unoccupied d states in the Pd-Pt and Pd-Ir alloys were not filled when the maximum quantity of hydrogen was stored in the alloys. These finite magnetic susceptibilities at the terminal composition of hydrogen in the plateau region were explained by the structural modification of the unoccupied d states in the electronic band structures due to alloying. 

Keywords: Pd binary alloy, Hydrogen storage capability, Magnetic susceptibility, Electronic band structure

Abstract

    Effects of pre-heating for retention and distribution of tritium have been studied using samples fixed on the wall of the Large Helical Device during a plasma campaign. The samples were fixed at four different locations. The plasma-facing surface of the samples was covered with deposition layers of different thickness in each sample. Retention behavior in deposition layers was observed using β-ray-induced X-ray spectrometry and imaging plate technique. Pre-heating of the samples in vacuum was changed in a temperature range from 300 to 623 K, and subsequent tritium exposure was carried out at 300K in every runs. Non-uniformity of tritium distribution clearly appeared even in the as-received samples which was not pre-heated. It is considered, therefore, that non-uniform adsorption sites of tritium have been produced during a formation process of deposition layers. In addition, it was seen that the amount of tritium retention increased with an increase in the pre-heating temperature, indicating that adsorption sites of tritium were newly formed in the deposition layers by heating in vacuum.

Keywords: tritium retention, plasma-facing material, deposition layer, β-ray-induced X-ray spectrometry

H. Kurishita, M. Hatakeyama, M. Narui, S. Matsuo, T. Shikama, K. Tada, N. Ohno¹⁾, T, Takeuchi²⁾, Y. Hatano³⁾, M. Nishiura⁴⁾, Y. Nakashima⁵⁾, H. Watanabe⁶⁾, K. Tokunaga⁶⁾, T. Hino⁷⁾, Y. Ueda⁸⁾, T. Muroga⁴⁾, A. Sagara⁴⁾, O. Kaneko⁴⁾

International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Ibaragi 311-1313, Japan
¹⁾Graduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan
²⁾National Institute for Materials Science, Tsukuba 305-0047, Japan
³⁾Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
⁴⁾National Institute for Fusion Science, Toki 509-5292, Japan
⁵⁾Plasma Research Center, University of Tsukuba, Tsukuba 305-8577, Japan
⁶⁾Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580, Japan
⁷⁾Graduate School of Engineering, Hokkaido University, Sapporo 060-8628, Japan
⁸⁾Graduate School of Engineering, Osaka University, Suita 565-0871, Japan

Abstract

    International Research Center for Nuclear Materials Science of the Institute for Materials Research, Tohoku University (hereafter the "Center") was founded in 1969 and has been playing a vital role as the joint-use research center in Japan to assess the dynamic and static effects of neutron irradiation on the physical and mechanical properties of a variety of structural and functional materials through the use of nuclear reactors in Japan and overseas. The Center is now also open to researchers overseas. As a new initiative, the Center started an interactive joint research scheme on nuclear fusion reactor engineering with the NIFS in fiscal 2010. The interactive joint research aims at pioneering inter-disciplinary fields that connect neutron reactor engineering with other nuclear fusion sciences, and at conducting activities primarily on the key research subjects through inter-research-center collaboration. For this, a TDS (Thermal Desorption Spectrometer) with an ion gun (IG-TDS) has been installed in the radiation controlled area at the Center. Development of a compact divertor plasma simulator (C-DPS) system that will be integrated with the IG-TDS apparatus is in progress. It is prospected that the Center could play a leading role in international collaborative studies of neutron irradiation effects on plasma material interaction, along with other major research institutes over the world.

Keywords: interactive joint research, fusion reactor engineering, neutron irradiation, hydrogen isotope retention, plasma facing material, plasma material interaction, thermal desorption spectrometer, compact divertor plasma simulator

Y. Nobuta^a,*, Y. Hatano^b, M. Matsuyama^b, S. Abe^b, S. Akamaru^b, Y. Yamauchi^a,T. Hino^a, S. Suzuki^c, M. Akiba^c

^aLaboratory of Plasma Physics and Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628, Japan
^bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^cJapan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193, Japan

Abstract

    DT⁺ ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both theamount of retained tritium and the reduction of retained tritium after preservation in vacuum wereinvestigated using an IP technique and BIXS. In addition, the relationship between the retention propertiesof tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. Theamount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1 keVof DT⁺irradiation, graphite showed no reduction of the amount of retained tritium after six monthspreservation while that of carbon film decreased by approximately 20% after 40 days preservation. Itwas suggested that this difference might be associated with differences in the microstructure betweengraphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately halfafter 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decreasein tritium retention during preservation became slower. Tungsten’s properties of releasing tritium whilepreserved in vacuum would be a useful tool for the reduction/removal of retained tritium.

Keywords: Tritium retention, Tungsten, Graphite, Co-deposited carbon film, Ion irradiationa

H. Kurishita¹, S. Matsuo¹, H. Arakawa¹ T. Sakamoto², S. Kobayashi², K. Nakai², H. Okano³, H. Watanabe³, N. Yoshida³, Y. Torikai⁴, Y. Hatano⁴, T. Takida⁵, M. Kato⁵, A. Ikegaya⁵, Y. Ueda⁶, M. Hatakeyama¹, T. Shikama¹

¹International Research Center for Nuclear Materials Science, IMR, Tohoku University, Oarai, Ibaraki 311-1313, Japan
²Department of Materials Science and Biotechnology, Ehime University, Matsuyama 790-8577, Japan
³Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
⁴Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
⁵ALMT Corporation, Toyama 931-8543, Japan
⁶Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871, Japan

Abstract

    Nanostructured tungsten (W)-based materials offer many advantages for use as plasma facing materials and components exposed to heavy thermal loads combined with irradiation with high-energy neutron and low-energy ion. This paper first presents the recent progress in nanostructured toughened, fine grained, recrystallized W materials. Thermal desorption spectrometry apparatus equipped with an ion gun has been installed in the radiation controlled area in our Center at Tohoku University to systematically investigate the effects of displacement damage due to high-energy neutron irradiation on hydrogen isotope retention in connection with the nano- or micro-structures in W-based materials. In this paper, the effects of high-energy heavy ion irradiation on deuterium retention in W with different microstructures are described as a preliminary work with the prospective view of neutron irradiation effects.

Keywords: tungsten, nanostructures, titanium carbides, tantalum carbide, hydrogen isotope, retention, displacement damage, neutron irradiation