M. Kobayashi¹, K. Kawasaki¹, K. Tatenuma³, M. Hara³, M. Matsuyama³, T. Fujii⁴, H. Yamana⁴, Y. Oya¹, K. Okuno^1
1Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Japan
²Kaken Co. Ltd., 1044, Hori, Mito-city, Ibaraki, 310-0903, Japan
³Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan
⁴Research Reactor Institute: Kyoto University, Kumatori, Japan

Abstract

    The release kinetics of tritium for Li_2+xTiO₃ (x = 0, 0.2, 0.4) and Li₄TiO₄ were studied by means of Thermal Desorption Spectroscopy (TDS). Tritium-TDS spectrum at the heating rate of 0.5 K / min for Li_2+xTiO₃ irradiated with thermal neutron consisted of two release stages at 480 K and 580 K, namely Peaks 1 and 2, respectively. The activation energy of Peak 1 was estimated to be around 0.37 eV, while that of Peak 2, about 0.63 eV. The latter corresponded to the activation energy of tritium diffusion in Li₂TiO₃. For Li₄TiO₄ exposed to tritium-deuterium mixture gas, two tritium release stages named as Peaks A and B were also observed at 450 K and 600 K, respectively, in tritium-TDS spectrum at the heating rate of 0.5 K / min. As the release temperature regions of Peak 1 and Peak A were almost the same, tritium releases of these peaks were considered to be originated from a same process. O-T bonds formed on the surface of Li₄TiO₄ were decomposed as Peak B in TDS spectra.

JK. Hashizume¹⁾, K. Ogata¹⁾, S. Akamaru²⁾, Y. Hatano²⁾

¹⁾Interdisciplinary Graduate School of Science and Engineering, Kyushu University, Hakozaki 6-10-1, Fukuoka 812-8581, Japan
²⁾ Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

    The hydrogen solubility in zirconia (yttria stabilized zirconia, YSZ) using a tritium imaging plate (IP) technique has been measured in a temperature region ranging from 673K to 973K. The tritium distribution in the zirconia specimens was nearly uniform suggesting an equilibrated dissolution. The hydrogen solubility in specimens were rather low (around ppm or below) and showed Arrhenius type temperature dependence, while it seemed to be independent of Y contents.

Keywords: Zirconia, YSZ, Tritium Gas Exposure, Imaging Plate Technique, Hydrogen Solubility

H. Yoshida-Ohuchi, Y. Hatato¹⁾, A. Mohammadi²⁾,T. Kawano³⁾

Radioisotope Research and Education Center, Graduate School of Pharmaceutical Science, Tohoku University, Sendai 980-8578, Japan
¹⁾Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
²⁾Radiation Research Center, Shiraz University, Shiraz 85111-71946, Iran
³⁾National Institute for Fusion Science, Toki-city 509-5292, Japan^a, Y. Edao^a, K. Sato^a

Abstract

    A combined technique of an imaging plate (IP) and thin absorbers was applied to tritium in nickel and vanadium specimens using copper, aluminum, and gold foil as the absorber. Copper and aluminum foil are used as a K-edge filter with X-ray absorption at 9.0 keV and 1.56 keV, respectively. Gold has L-edges X-ray absorption at around 13 keV. With this technique, photostimulated luminescence (PSL) decay curves are obtained by changing absorber’s thickness. In the nickel specimen, the difference in PSL decay curves between for the copper and gold absorber was clearly observed 20 days after loading, however, all curves became similar single pattern after 388 days. The same curve pattern was obtained in vanadium for all absorbers. The cross section images and depth profiles, which were taken at 468 days and 3.9 years after loading for the nickel and vanadium specimen, respectively, show no significant inclination of tritium concentration for both specimens. These results indicate that uniform tritium distribution in the specimen provides the similar single PSL decay curves pattern for the copper or aluminum and gold absorber.

Keywords: tritium, imaging plate, thin absorber, bremsstrahlung X-rays, depth distribution, non-destructive measurement

M. Matsuyama¹⁾, S. Abe¹⁾, K. Nishimura²⁾, Y. Ono³⁾, Y. Oya⁴⁾, K. Okuno⁴⁾, T. Hino⁵⁾, A. Sagara²⁾

¹Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama, 930-8355,Japan

²The Southwestern Institute of Physics, Chengdu 610041, Sichuan, China

³National Institute for Science, Toki 509-5292, Japan

⁴Radiochemistry Research Laboratory, Shizuoka University, Shizuoka 422-8529, Japan

⁵Laboratory of Plasma Physics and Engineering,Hokkaido University, Sapporo 060-8628, Japan

Abstract

    Retention behavior of tritium by stainless steel exposed to plasmas in the Large Helical Device (LHD) has been studied by-ray-induced X-ray spectrometry (BIXS) and an imaging plate (IP) technique. Plasma-exposed stainless steel samples were prepared by exposing the small plates at the location of 6.5L in LHD. X-ray photoelectron spectroscopy (XPS) was also applied to analyze deposition layers on the surface of a sample. XPS analyses showed that the deposition layers contain B, C, O and Ti as well as constituent elements of stainless steel. Chemical form of metallic elements was mainly oxides, but a part was metallic state. The plasma-exposed sample was exposed to tritium gas along with a bare stainless steel plate for comparison. The BIXS and IP measurements showed that the tritium retention largely increased by plasma exposure and it strongly depended on sample temperatures during vacuum heating and tritium exposure. In addition, it was seen from the IP images that non-uniform tritium distribution on the surface was formed even tritium exposure at room temperature.

Keywords: tritium retention, plasma exposure, BIXS, imaging plate, plasma-facing materials

Y. Oya¹⁾, M. Kobayashi¹⁾, N. Yoshida²⁾, N. Ashikawa³⁾, A. Sagara³⁾, Y. Hatano⁴⁾, K. Okuno¹⁾

¹⁾Radioscience Research Laboratory, Faculty of Science, Shizuoka University, 836, Ohya, Suruga-ku, Shizuoka 422-8529, Japan.
²⁾ Institute for Applied Mechanics, Kyushu University, Fukuoka 816-8580, Japan.
³⁾National Institute for Fusion Science, Gifu 509-5292, Japan.
⁴⁾Hydrogen Isotope Research Center, University of Toyama, 3190, Gofuku, Toyama 930-8555, Japan.

Abstract

    Effects of energetic carbon implantation on deuterium retention behavior in tungsten were studied. The dislocation loops and vacancies were introduced in tungsten as irradiation damages by energetic carbon implantation. The density of irradiation damages was almost saturated by C⁺ implantation below 1 dpa according to TEM observations. The retentions of deuterium trapped by dislocation loops and vacancies were observed in TDS measurements, which were increased as increasing the carbon implantation energy. The deuterium desorption at higher temperature above 600 K was also found, corresponding to the desorption of deuterium retained in the intrinsic cavity in bulk region. However, this trapping was refrained by energetic C⁺ implantation, concluding that the retained carbon would play a role as the diffusion barrier of deuterium, which would prevent deuterium penetration through bulk.

Keywords: Tungsten, TDS, TEM, Carbon, Irradiation damage, Diffusion, Simultaneous implantation

Keywords: Alloying effect, ZrMn₂, Double plateau, Hydrogen occupation site

Y. Oya^a, Y. Hatano^b, M. Hara^b, M. Matsuyama^b, K. Okuno^a

^aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
^bHydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract

    Hydrogen isotope retention and desorption behaviors for Silicon carbide (SiC), Silicon nitride (Si₃N₄) and Silicon dioxide (SiO₂) were studied to elucidate the fundamental process of hydrogen isotope in Si related ceramics by means of T-IP (tritium imaging plate), thermal desorption spectroscopy (TDS) and X-ray photoelectron spectroscopy (XPS). The tritium gas exposure at 673 K showed that tritium was precipitated on the surface for SiO₂, although that for SiC was uniformly retained inside the bulk. The 0.2 keV D⁺₂ implantation revealed that the deuterium desorption stages for Si related ceramics were consisted of four desorption stages at around 450 K, 650 K, 800 K, and 950 K, attributing to the desorptions of deuterium trapped on the surface, retained in interstitial sites, trapped as Si-D bond and trapped as C/N/O-D bond, respectively. The retention enhancement of deuterium trapped by Si as Si-D bond and the reduction of deuterium trapped on the surface would be associated with the enhancement of covalent bond characteristics for Si related ceramics. These results indicate that the dangling bonds in the covalent ceramics have higher hydrogen isotope trapping efficiency to form chemical bond like Si-D bond. On the other hand, the surface adsorption of hydrogen isotope was enhanced for the higher ionicity ceramics by charge localization.

M. Matsuyama, S. Abe, Y. Ono¹⁾, K. Nishimura²⁾, N. Ashikawa²⁾, Y. Oya³⁾, K. Okuno³⁾, T. Hino⁴⁾, A. Sagara²⁾

Hydrogen Iosotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
¹⁾CRDNS, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
²⁾National Institute for Fusion Science, Toki 509-5292, Japan
³⁾Radiochemistry Research Laboratory, Shizuoka University, Shizuoka 422-8529, Japan
⁴⁾Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo 060-8628, Japan

Abstract

    Effects of plasma exposure for tritium retention have been studied using both samples of a plasma-exposed stainless steel and a bare stainless steel. It was seen that the plasma-exposed surface was covered with thin deposition layers. Constituent elements of the deposition layers mainly formed metallic oxides except for C and O, but a part of Fe and Mo was a metallic state. Both samples were simultaneously exposed to tritium gas at room temperature with no heat treatment, and a considerable increase in tritium retention was observed for the plasma-exposed sample. In addition, it was clarified that distribution of tritium retention on the plasma-exposed surface was non-uniform even exposure at room temperature. Increase in tritium retention and non-uniformity of tritium distribution were discussed from viewpoints of distribution and chemical states of constituent elements of the deposition layers.

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

^*This article is based on the presentation at the 22nd International Toki Conference (ITC22).

Y. Hatano¹, M. Shimada², Y. Oya³, G. Cao⁴, M. Kobayashi³, M. Hara¹, B. J. Merrill², K. Okuno³, M. A. Sokolov⁵, Y. Katoh⁵

¹Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
²Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
³Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
⁴Department of Engineering Physics, The University of Wisconsin, Madison, WI 53706, USA
⁵Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA

Abstract

    To investigate the effects of neutron irradiation on hydrogen isotope retention in tungsten, disk-type specimens of pure tungsten were irradiated in the High Flux Isotope Reactor in Oak Ridge National Laboratory followed by exposure to high flux deuterium (D) plasma in Idaho National Laboratory. The results obtained for low dose n-irradiated specimens (0.025 dpa for tungsten) are reviewed in this paper. Irradiation at coolant temperature of the reactor (around 50℃) resulted in the formation of strong trapping sites for D atoms. The concentrations of D in nirradiated specimens were ranging from 0.1 to 0.4 mol% after exposure to D plasma at 200 and 500℃ and significantly higher than those in nonirradiated specimens because of D-trapping by radiation defects. Deep penetration of D up to a depth of 50-100 μm was observed at 500℃. Release of D in subsequent thermal desorption measurements continued up to 900℃. These results were compared with the behaviour of D in ion-irradiated tungsten, and distinctive features of n-irradiation were discussed.

Keywords: plasma-facing material, tungsten, hydrogen isotope, irradiation, trapping effect

Keywords: Tritium, Tritiated water, Permeation, Chemical exchange, Durability, Corrosion

M. Hara^a, T. Yamamoto^b, K. Nishimura^b, S. Akamaru^a, K. Watanabe^a, M. Matsuyama^a

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555,Japan
^bFaculty of Engineering, Universityof Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

    Pressure-composition curves of ZrMn_x (X=1.9, 2.0)-Q2 (Q=H, D) were measured by a volumetric method. An inverse isotope effect of absorption pressure was found for the ZrMn_xQ_ysystem, i.e. the hydrogen absorption pressure was found to be larger than that of deuterium. On the other hand, the hydrogen desorption pressure was the same as that of deuterium. Thus, an inverse isotope effect merely appears in the absorption pressures. The degree of the inverse isotope effect decreased with increasing temperature. This effect can be explained by the hysteresis factor which is induced by hydrogen isotopes dissolved in ZrMn_1.9 or ZrMn_2.0.

Y. Hatano¹, M. Shimada², T. Otsuka³, Y. Oya⁴, V.Kh. Alimov¹, M. Hara¹, J. Shi¹, M. Kobayashi⁴, T. Oda⁵, G. Cao⁶, K. Okuno⁴, T. Tanaka⁷, K. Sugiyama⁸, J. Roth⁸, B. Tyburska-Pü uschel⁸, J. Dorner⁸, N. Yoshida⁹, N. Futagami⁹, H. Watanabe⁹, M. Hatakeyama¹⁰, H. Kurishita¹⁰, M. Sokolov¹¹,Y. Katoh¹¹

¹ Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan
² Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID, USA
³Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Kasuga, Japan
⁴Faculty of Science, Shizuoka University, Shizuoka, Japan
⁵Department of Nuclear Engineering and Management, The University of Tokyo, Japan
⁶Department of Engineering Physics, The University of Wisconsin, Madison, WI, USA
⁷National Institute for Fusion Science, Toki, Japan
⁸Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching, Germany
⁹Research Institute for Applied Mechanics, Kyushu University, Kasuga, Japan
¹⁰Institute for Materials Research, Tohoku University, Oarai, Japan
¹¹Oak Ridge National Laboratory, Oak Ridge, TN, USA

Abstract

    The effects of neutron and ion irradiations on deuterium (D) retention in tungsten (W) were investigated. Specimens of pure W were irradiated with neutrons to 0.3 dpa at around 323K and then exposed to high-flux D plasma at 473 and 773 K. The concentration of D significantly increased by neutron irradiation and reached 0.8 at % at 473K and 0.4 at % at 773 K. Annealing tests for the specimens irradiated with 20MeV W ions showed that the defects which play a dominant role in the trapping at high temperature were stable at least up to 973 K, while the density decreased at temperatures equal to or above 1123 K. These observations of the thermal stability of traps and the activation energy for D detrapping examined in a previous study (≈1.8 eV) indicated that the defects which contribute predominantly to trapping at 773K were small voids. The higher concentration of trapped D at 473K was explained by additional contributions of weaker traps. The release of trapped D was clearly enhanced by the exposure to atomic hydrogen at 473 K, though higher temperatures are more effective for using this effect for tritium removal in fusion reactors.

M. J. Downa^a, J. Tennyson^a, M. Hara^b, Y. Hatano^b, K, Kobayashi^c

^aDepartment of Physics and Astronomy, University College London, London WC1E 6BT, UK
^bHydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama-City, Toyama 930-8555, Japan
^cDepartment of Physics, University of Toyama, 3190 Gofuku, Toyama-City, Toyama 930-8555, Japan

Abstract

    A tritium enhanced water absorption spectrum previously recorded in the 7200-7245 cm^-1 region is analysed. Variational calculations for HTO predict absorption to be dominated by the 2ν₃ vibrational band in this region. New assignment are made for HTO based on this line list with a band origin measured to be at 7236.03 cm^-1. A calculated T₂O line list predicts absorption in this region to be below the experimental detection limit despite the large quantity of tritium present. From 170 lines observed 37 known H₂¹⁶O lines are identified and 111 new HTO assignments are made.

V.Kh. Alimov, Y. Hatano, K. Sugiyama, J. Roth, B. Tyburska-Püschel, J. Dorner, J. Shi, M. Matsuyama, K. Isobe, T. Yamanishi

^aHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^bMax-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
^cTritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan

Abstract

    Samples of polycrystalline ITER-grade W and recrystallized W were irradiated at room temperature with 20 MeV Wions to displacement damage up to 0.5 dpa. The damaged Wsamples were then exposed to (i) D neutrals at 403?573 K and (ii) D₂ gas at 673-1073 K and pressures of 1.2 and 100 kPa. Trapping of deuterium in the damage zone was examined by the D(³He, p)⁴He nuclear reaction with ³He energies between 0.69 and 4.0 MeV allowing determination of the D concentration up to a depth of 6 μm. It has been found that generation of the W-ion-induced displacement damage leads to accumulation of deuterium in the damage zone up to concentration depending on the exposure temperature and, at temperatures ≥673 K, on the D₂ gas pressure. Thermal desorption spectra allowed a conclusion that deuterium is mainly accumulated in the form of D atoms bound to inner walls of vacancy clusters.

Y. Oya, S. Masuzaki, T. Fujishima, M. Tokitani, N. Yoshida, H. Watanabe, Y. Yamauchi, T. Hino, M. Miyamoto, Y. Hatano, K. Okuno

^aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
^bNational Institute for Fusion Science, Gifu 509-5292, Japan
^cResearch Institute for Applied Mechanics, Kyushu University, Fukuoka 816-8580, Japan
^dGraduate School of Engineering, Hokkaido University, Sapporo 060-8628, Japan
^eInterdisciplinary Faculty of Science and Engineering, Shimane University, Shimane 690-8504, Japan
^fHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

    Tungsten (W) samples, undamaged and damaged by irradiation with 2.4 MeV Cu²⁺, were placed flush to the first wall of Large Helical Device (LHD) nearby the divertor target, and were exposed to 134 shots of hydrogen plasma discharges. Thereafter, to evaluate the enhancement of hydrogen isotope retention for Wdue to LHD hydrogen (H) plasma exposures, theWsamples with and without LHD H plasma exposures were implanted with 1.0 keV D₂⁺. For W samples without LHD H plasma exposures, the D retention in the damaged W was by a factor of 6 higher than that in the undamaged W. After LHD H plasma exposures, the carbon layers~4 nm in thickness were deposited on theWsurfaces. Due to trapping of D atoms in the carbon layers, the D retention enhancement for undamaged and damaged W samples exposed to LHD H plasmas was significantly higher than that for the W samples without LHD H plasma exposures.

T. Taguchi, M. Kobayashi, K. Kawasaki, Y. Miyahara, N. Ashikawa, A. Sagara, N. Yoshida, M. Miyamoto, K. Ono, Y. Hatano, Y, Oya, K. Okuno

^aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
^bNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi, Gifu 509-5292, Japan
^cInstitute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^dDepartment of Material Science, Shimane University, Matsue, Shimane 690-8504, Japan
^eHydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract

    Dynamics of deuterium recycling, including retention and sputtering behaviors was studied for C+ implanted tungsten. The amount of deuterium trapped by irradiation damages was clearly increased in the C⁺ implantation sample because the irradiation damages in the C⁺ implanted sample were formed more than those in the only D₂⁺ implanted one. In addition, the deuterium diffusion toward the bulk would be refrained by the formation of W-C mixed layer, which would work as the deuterium diffusion barrier. The in situ sputtered particle measurement system has been established and revealed that the formation of hydrocarbons such as CD₄ was directly observed during D₂⁺ implantation into the C⁺ implanted tungsten. In the lower deuterium fluence, the CD₄ sputtering rate was enhanced with increasing the deuterium fluence. It was considered that the sputtering rate of CD₄ would be controlled by the concentration of deuterium on the top surface of the W-C mixed layer.

M, Yajima, Y. Hatano, S. Kajita, J. Shi, M. Hara, N. Ohno

^aDepartment of Energy Engineering and Science, Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya-shi, Aichi, Japan
^bHydrogen Isotope Research Center, Toyama University, 3190 Gofuku, Toyama-shi, Toyama, Japan
^cEcoTopia Science Institute, Furo-cho, Chikusa-ku, Nagoya-shi, Aichi, Japan

Abstract

    Tungsten specimens with fiberform nanostructure (Nano-W) were prepared by exposure to helium plasma in the divertor plasma simulator NAGDIS-II with different amounts of helium fluence. For comparison, tungsten specimens with smooth surface (Polished-W) were also prepared. Surface area of Nano-W was measured by using Brunauer, Emmet and Teller (B.E.T.) method. Tritium retention of Nano-W and Polished-W was investigated by an imaging plate (IP) and b-ray induced X-ray spectrometry (BIXS) technique exposure to mixture gas of deuterium and tritium. It was found that surface area of Nano-W was significantly larger than that of Polished-W and increased in proportion to the amount of helium irradiation. On the other hand, tritium retention showed a saturation trend when the helium fluence was higher than 5.0 ×10²⁵ m^-2.

V.Kh. Alimov, Y. Hatano, B. Tyburska-Püschel, K. Sugiyama, I. Takagi, Y. Furuta, J. Dorner, M. Fuβeder, K. Isobe, T. Yamanishi, M. Matsuyama

^aHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^bMax-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
^cDepartment of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan
^dTritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan

Abstract

    W samples were irradiated at 300 and 573 K with 4.8 and 20 MeV W ions to displacement damage levels in the range from 0.022 to 50 displacements per atom at the damage peak. 50 μm thick W samples were exposed to high flux D plasma at 550 K on the side opposite to the damaged one, whereas 2 mm thick W samples were exposed to low flux D plasma at 403 K on the damaged side. Trapping of deuterium at displacement damage was examined by the D(³He, p)⁴He nuclear reaction with ³He energies between 0.69 and 4.0 MeV allowing determination of the D concentration up to a depth of 6 μm. It was found that (i) at the damage level above 0.1 dpa, the concentration of the W-ion-induced defects responsible for trapping of diffusing D atoms

S. Akamaru, M. Inoue, T. Abe
Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

    In this study, the surfaces of NaCl particles were modified with metal films using the polygonal barrel-sputtering method. When Pt was sputtered on NaCl particles, the individual particles changed from white to metallic. Characterization of the treated samples indicated that thin Pt metal films were uniformly deposited on the NaCl particles. Immersion of the treated NaCl particles in water revealed that they floated to the surface of the water with the increase in the immersion time, although their original cubic shapes remained unchanged. The floating phenomenon of the Pt-coated NaCl particles, as mentioned above, suggests that NaCl was dissolved by the permeation of water through invisible defects such as grain boundaries in the Pt films, leading to the formation of hollow particle-like materials. It should be noted that uniform film deposition on the NaCl particles could also be achieved by sputtering with Au or Cu. Based on the obtained results, our sputtering method allows uniform surface modification of water-soluble and water-reactive powders that cannot be treated by conventional wet process using water.

Keywords;Particle Surface Modification, Ionic Crystal, Dry Process, Sputtering Technique, NaCl

T. Nozaki, Y. Hatano

Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

    Dense hafnium nitride (HfN) layers were prepared between Pd protection films and a Ta substrate in a composite hydrogen separation membrane to prevent a reaction between the Pd and the substrate at high temperatures. No significant reduction in hydrogen permeation rate was observed for the membrane with 50-nm-thick HfN layers at 873 K through at least 35 h, whereas the specimen without HfN layers rapidly deteriorated within 5 h. Hydrogen permeability of the former specimen was 4 × 10-9 mol m-1 s-1 Pa-0.5 at 873 K at steady state. This value was smaller than the initial permeability of Pd-covered Ta before deterioration by an order of magnitude. The measurements of pressure-composition isotherms by using a HfN powder specimen showed that the hydrogen solubility in HfN was sufficiently high and comparable with the solubility in Ta. Therefore, the low permeability observed with the HfN intermediate layers was ascribed to low hydrogen diffusivity in HfN.

Keywords:Hydrogen permeation, Group 5 metals, Hafnium nitride, Intermediate layer, Pressure-composition isotherms

V.Kh. Alimov¹, B.I. Khripunov², A.V.

¹A.N. Frumkin Institute of Physical Chemistry and Electrochemistry RAS, Moscow, Russia
²NRC ≪Kurchatov Institute, Moscow, Russia
³Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

Hydrogen isotopes exchange in tungsten was investigated after sequential irradiations by low energy deuterium (D) and mixed heliumprotium H-He-plasmas at sample temperatures of 400 and 530 K. Deuterium depth profiles were measured by the D(³He, p)⁴He nuclear reaction with ³He⁺ energies between 0.69 and 4.5 MeV allowing determination of the D concentration up to a depth of 7 μm. It was found that significant part of deuterium initially retained in tungsten after deuterium plasma exposure was released during sequential exposure to protium plasma. However, exposure of the D-plasma-exposed W samples to helium-protium plasma reduces the amount of released deuterium as compared to pure protium plasma irradiation.

Key words: tungsten, detritiation, deuterium plasma, helium-protium plasma.

松山政夫

富山大学水素同位体科学研究センター

Abstract

    核融合炉システムにおけるトリチウムの閉じ込め性能およびその健全性を監視し，施設内の作業者および公 衆の安全性を確保するためには，種々のトリチウム測定装置の設置が必要となる．そこでトリチウムの化学形や 物理的状態，濃度範囲および測定環境などの条件を考慮した測定装置や測定法の例示と共に注意事項を記す.

Keywords: tritium, measurement, β-ray, D-T reactor, environment

S. Akamaru^a, T. Matsumoto^b, Masanori Hara^a, K. Nishimura^b, N. Nunomura^c, M. Matsuyama^a

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bFaculty of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^cInformation Technology Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

    In this study, the magnetic susceptibility of the Pd-Co-H system was simultaneously measured with pressure-composition isotherms at ambient temperature. The magnetic susceptibility of the system abruptly decreased as the hydrogen content increased to 0.02 for all Pd-Co alloys under study. In the plateau region, the susceptibility decreased linearly with an increase in hydrogen uptake. The single metal hydride phase exhibited weak magnetic susceptibility. These behaviours were qualitatively interpreted in terms of the electronic structure of the Pd-Co-H system. The unoccupied down-spin band of Co just above the Fermi level gradually decreased with increasing hydrogen uptake; thus, the magnetic moment arising from uncompensated spin reduced, and the ferromagnetic transition temperature decreased.

Key words: Hydrogen absorbing materials; Pd-Co alloy; Magnetic measurements; Electronic band structure

M. Hara^a, H. Fujinami^b, S. Akamaru^a, N. Nunomura^c, K. Watanabe^a, K. Nishimura^a, M. Matsuyama^a

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bFaculty of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^cInformation Technology Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

    The electronic structures of palladium, palladium-silver alloys (Pd-Ag) and their hydrides were calculated to evaluate the terminal composition in the plateau region in the pressure-composition isotherm. The electronic structures were calculated by the Korringa-Kohn-Rostoker coherent potential approximation (KKR-CPA), MACHIKANEYAMA 2002 developed by Akai. The Fermi level of palladium at various hydrogen contents moved to the edge of the d-band with increasing the hydrogen content and the Fermi level of PdH0.6 reached the edge of the d-band. This hydrogen composition, [H]/[Pd] = 0.6, agreed well with the terminal composition in the plateau region. The Fermi level of Pd-Ag was similarly lifted up to the edge of the d-band with increasing the hydrogen content. The terminal compositions in isotherms of Pd-Ag alloys were estimated from the Fermi level position and the density of states at the Fermi level. The estimated compositions agreed with those experimentally obtained. Therefore, the terminal composition in the plateau region in the pressure-composition isotherms of Pd-Ag alloys could be evaluated by using with the KKR-CPA method.

Keywords:Pd-Ag, KKR-CPA, Plateau region, Terminal composition

対馬勝年¹、松山政夫²、阿部信介²

¹富山大学理学部
²富山大学水素同位体科学研究センター

Abstract なし

波多野雄治、大矢恭久^1）、 原 正憲、 小田卓司^2）、大塚哲平^3）、佐藤紘一^4）、張　鯤

富山大学水素同位体科学研究センター
¹⁾静岡大学大学院理学研究科
²⁾東京大学大学院工学系研究科
³⁾九州大学大学院総合理工学研究院
⁴⁾京都大学原子炉案験所

Abstract

    プラズマ対向材料中のトリチウム挙動に及ぼす中性子照射の影響を明らかにするため，候補材であるタング ステンをオークリッジ国立研究所の研究炉 High Flux Isotope Reactor (HFIR) で中性子照射した上で、アイダホ 国立研究所の線型プラズマ装置 Tritium Plasma Experiment (TPE) にて同位体である重水素の高フラックスプラ ズマにばく露し、捕獲重水素濃度と昇温脱離挙動を調べた.照射欠陥の捕獲効果により水素同位体滞留量が著しく増大すると共に、加熱処理による除去が困難となるため，同位体交換法等の新たなトリチウム除去技術の開発が必要であることが示された.

Keywords:neutron irradiation, tritium inventory, hydrogen isotope, plasma-facing material, tungsten, radiation damage, trap, retention, diffusion

上田良夫、 波多野雄治¹⁾

大阪大学大学院工学研究科
¹⁾富山大学水素同位体科学研究センター

Abstract

    プラズマ対向材料として現在最も有力視されているタングステンと、水冷却に比べ社会的受容性の観点から 多くの利点を有するヘリウムガス冷却によるプラズマ対向機器(PlasmaFacing Component: PFC)を主たる対象 として、核融合炉複雑環境下におけるPFCの健全性と安全性の評価、およびこれらの評価結果に基づく、高い信 頼性を有する原型炉ダイパータを実現するための課題の明確化をめざし、 2013年 (平成23) 年4月から新たに PHENIX (PFC evaluation by tritium Plasma. Heat and Neutron Irradiation eXperiments) 計画が開始された. 本章ではPHENIX計画の概要を紹介する.

Keywords:plasma facing compone, tntungsten, neutron-irradiation, high heat flux, thermo-mechanicalpr operty, tritium inventory

S. Akamaru¹, F. Ishikawa², K. Nishimura², T. Abe¹, M. Matsuyama¹

¹Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
²Faculty of Engineering, University of Toyama, Toyama 930-8555, Japan

Abstract

    MgB₂ grains were coated with metal nanoparticles (Ag, Zn and Sn) using the barrel sputtering technique, and transport and superconducting properties were evaluated. Almost all MgB₂ grains were uniformly coated with metal nanoparticles with average diameters of less than 20 nm. The electrical resistivity of the coated MgB₂ decreased as the amount of metal coating increased. The critical current densities of almost all coated MgB₂ were enhanced compared to those of bare MgB₂. These results explained the improvement of intergranular connectivity between MgB₂ grains by the insertion of metal nanoparticle in grain boundaries. After annealing at 473 K, only MgB₂ coated with Sn nanoparticles showed a decrease in electrical resistivity and the enhancement of the critical current density. These results can be understood by the effect of improvement of intergranular connectivity between MgB₂ grains by annealing.

Keywords: MgB₂, intergranular connectivity, metal nanoparticles, barrel sputtering technique

R.-D. Penzhorn, Y. Hatano, M. Matsuyama, Y. Torikai

Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

    Stainless steel exposed to gaseous tritium characteristically shows a firmly trapped fraction of tritium in the surface layer, which is not fully removable by water at ambient temperature. Prolonged thermal treatment of tritium-loaded specimens at 443 K causes substantial depletion of the bulk but almost no depletion of the surface layer. For complete removal of hydrogen isotopes from the bulk and the surface, temperatures exceeding 573 K are necessary. Upon chemical etching virtually all tritium trapped in the surface layer appears in the etching solution as tritiated water. Following removal of the layer by chemical etching, the tritium-rich layer reappears after months of aging at ambient temperature with nearly the original tritium activity. Comparison of chronic tritium release rates into liquid water before and after etching reveals that the surface layer only marginally influences the rate. X-ray photoelectron spectroscopy provides evidence that during prolonged aging the surface layer continues to grow while at the same time trapping a fraction of bulk tritium released at ambient temperature. Experimental results suggest different mechanisms of hydrogen uptake and release by the bulk and surface layers. Inference of tritium activity in the bulk of aged or heat-exposed stainless steel material from surface activity measurements may depart significantly from reality.

Keywords: tiritium interaction, stainless steel, aging
Note: Some figures in this paper are in color only in the electronic version.

K. Hashizume^a, K. Ogata^a, M. Nishikawa^a, T. Tanabe^a, S. Abe^b, S. Akamaru^b, Y. Hatano^b

^a Interdisciplinary Graduate School of Engineering Science, Kyushu Univ., Fukuoka, Japan
^b Hydrogen Isotope Research Center, Univ. of Toyama, Toyama, Japan

Abstract

    Using a tritium imaging plate technique, kinetics of tritium dissolution and its solubility in several oxides were examined. Mirror-polished single crystals of alumina, spinel and zirconia were used as specimens, which were exposed to 133 Pa of a tritium(T)?deuterium(D) gas mixture (T/(T + D) ~0.17) at temperatures ranging from 673 to 973 K for 1-5 h. The T surface activity on the specimens increased with increasing temperature and exposure time, it almost saturated at 873 K and reached 2 × 10⁵ Bq/cm² (1 × 10¹⁴ T/cm²), and no clear difference appeared among the types of specimens. The T activity in the oxide bulk also increased with temperature, in which there was a trend for the oxides: spinel ≧ zirconia ≧ alumina. In the T dissolution process for all oxides, the concentration gradient due to its diffusion was not observed even for short exposure times: the T density was almost uniform over the specimens in transition states and increased with exposure time up to the saturated value. These experimental results suggested that the rate-controlling process of T dissolution in the temperature region should be not its diffusion in the oxides but dissociation of hydrogen molecules (T-D mixture in this case) into atoms, its adsorption on the surface and/or T penetration from the surface into the bulk.

R. Miura^a, T. Fujishima^a, H. Uchimura, K. Toda^a, M. Kobayashi^a, N. Ashikawa^b, A. Sagara^b, N. Yoshida^c, Y. Hatano^d, Y. Oya^a, K. Okuno^a

^aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
^bNational Institute for Fusion Science, Gifu, JapancResearch Institute for Applied Mechanics, Kyushu University, Fukuoka, Japan
^cHydrogen Isotope Research Center, University of Toyama, Toyama, Japana

Abstract

    The D₂⁺ fluence dependence on deuterium (D) retention was studied to clarify the D retention mechanism in tungsten. The additional D desorption stage was observed around 660 K in the TDS spectrum for a sample implanted with D₂⁺ up to the fluence of 1023 D⁺ m^-2, which desorption stage was not observed the D₂⁺ implanted sample with the fluence less than 10²² D⁺ m^-2. The TEM observation showed that the highly dense voids were formed in tungsten by D₂⁺ implantation with the fluence of 1023 D⁺ m^-2, considering that the D would be trapped by voids. To understand the D trapping by voids in C⁺ implanted tungsten, C⁺-D₂⁺ sequential implantation experiments at various C⁺ implantation temperatures were performed. It was found that the amount of D desorbed around 560 K was increased by increasing the C⁺ implantation temperature. The formation of the voids was observed with increasing the C⁺ implantation temperature by TEM, indicating that the increase of D desorption around 560 K was caused by the formation of voids. However, the desorption temperature of D trapped by voids in C⁺ implanted sample was lower than that in D₂⁺ implanted one. TEM observation and XPS measurement indicated that this difference was caused by the increase of void size and/or the presence of implanted carbon.

Keywords: Deuterium retention behavior, Tungsten, Carbon, Voids, TDS, TEM

Y. Jin^a, R. Yang^b, Y. Mori^a, J. Sun^a, A. Taguchi^c, Y. Yoneyama^a, T. Abe^c, N. Tsubaki^a

^aDepartment of Applied Chemistry, School of Engineering, University of Toyama, Gofuku 3190, Toyama, 930-8555, Japan
^bSchool of Biological and Chemical Engineering, Zhejiang University of Science & Technology Hangzhou, 310023, PR China
^cHydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

    The mm-sized capsule catalyst Co/SiO₂-Z with the H-type zeolite (HZSM-5) as the shell, and the Co/SiO₂-Z-Sp-Pd capsule catalyst with Pd loaded by sputtering method, as well as the Co/SiO₂-Z-IW-Pd capsule catalyst, loading the Pd by incipient wetness impregnation, were prepared and used for isoparaffin direct synthesis by Fischer-Tropsch synthesis reaction from syngas with H₂/CO ratio of 2/1 at 1.0 MPa and 533 K. The analysis results of XRD, SEM, EDS and NH₃-TPD showed that a compact HZSM-5 shell was formed on the Co/SiO₂ pellet, and the metallic Pd was well sputtered on the surface of the HZSM-5 shell for the Co/SiO₂-Z-Sp-Pd catalyst. The isoparaffin and olefin selectivity increased in the FTS reactions on the HZSM-5 capsule catalyst than on the Co/SiO₂ catalyst. The selectivity of isoparaffins increased, and the selectivity of olefins decreased when using the Co/SiO₂-Z-Sp-Pd catalyst, compared with the Co/SiO₂-Z catalyst or the Co/SiO₂-Z-IW-Pd catalyst, because metallic Pd introduced by sputtering on the zeolite shell could hydrogenate olefins efficiently. It was suitable to use the sputtering method for the preparation of the zeolite capsule catalyst loaded with metallic Pd, as conventional impregnation method could not reduce Pd cation due to the strong interaction from zeolite surface. The CO conversion, CH4 and CO₂ selectivity on the catalyst of Co/SiO₂-Z-Sp-Pd increased with temperature increasing from 513 K to 553 K. At elevated temperature, the isoparaffin selectivity also increased while the olefin selectivity was suppressed.

Keywords: Capsule catalyst, HZSM-5, Palladium (Pd), Sputtering, Fischer-Tropsch synthesis (FTS)

K. Matsubara, M. Danno, M. Inoue, H. Nishizawa, Y. Honda, T. Abe

Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

    In this study, tetrafluoromethane (CF₄) plasma-treatments of polymethylmethacrylate (PMMA) powder were performed using a polygonal barrel-plasma-treatment system to improve the PMMA's hydrophobicity. Characterization of the treated samples showed that the PMMA particle surfaces were fluorinated by the CF₄ treatment. The smooth surfaces of the particles changed into nano-sized worm-like structures after the plasma-treatment. The hydrophobicity of the treated PMMA samples was superior to that of the untreated samples. It was noted that the hydrophobicity of the treated samples and the surface fluorination level depended on the plasma-treatment time and radiofrequency (RF) power; high RF power increased the sample temperature, which in turn decreased the hydrophobicity of the treated samples and the surface fluorination because of the thermal decomposition of PMMA. The water-repellent effects were evaluated by using paper towels to show the application of the plasma-treated PMMA particles, with the result that the paper towel coated with the treated sample was highly water-repellent.

Keywords : Hydrophobicity, CF₄ plasma, Surface fluorination, Polymer particles

K. Matsubara, M. Danno, M. Inoue, H. Nishizawa, Y. Honda, T. Abe

Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

    In this study, polystyrene powder was subjected to CF₄ plasma-treatment using a polygonal barrel-plasmatreatment system to enhance the hydrophobicity of the particles. The characterization of the treated samples showed that the orange-peel-like surface of the particles was changed into rugged structures by the CF₄plasma-treatment. The plasma-treatment using our system also resulted in the fluorination of the particle surfaces. The contact angle of a water droplet loaded on the sample treated at a radio frequency (RF) power of 150Wfor 10 min (149.1°) was significantly larger than that on the untreated sample (123.6°): both the contact angle value and the fluorine content of sample depended on the RF power. These results indicate that the hydrophobicity of the polystyrene powder was improved by our system, allowing preparation of stable liquid marbles, including CuCl₂ solutions, which can be used as NH₃gas sensors.

Keywords : CF4 plasma, Hydrophobicity, Liquid marble, Polymer particles, Surface fluorination

K. Matsubara, M. Danno, M. Inoue, H. Nishizawa, Y. Honda, T. Abe

Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

    Generally direct synthesis of dimethyl ether (DME) from syngas (CO + H₂) containing CO₂ uses physical mixture catalyst composed of methanol synthesis catalyst and solid acidic catalyst for dehydration of methanol. But the distance between these two catalysts is far and easy to become separated. Here we report Cu/ZnO methanol synthesis catalyst which is directly loaded onto acidic zeolite catalyst, to accomplish direct synthesis of DME from syngas. The intimate contact between these two catalysts improves consecutive reaction efficiency, achieving high DME selectivity. Conventional wet impregnation method to prepare this kind of catalyst will make Cu cations adsorbed onto zeolite sites very strongly, which are almost impossible to be reduced due to the very strong interaction from zeolite surface. To solve this problem, we propose a new physical sputtering method here to directly load metallic Cu and Zn clusters onto acidic zeolite catalyst surface, to readily obtain highly selective and multifunctional Cu/ZnO/zeolite catalyst. This dry method does not produce waste water containing View the MathML sourceNO₃⁺ ion, which is very troublesome in the environment.

Keywords : Dimethyl ether, Physical sputtering, Zeolite, Cu/ZnO, Catalyst preparation

K. Matsubara, M. Danno, M. Inoue, Y. Honda, N. Yoshida, T. Abe

Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

    Metallic titanium (Ti) particles treated with N₂ plasma via the barrel-plasma-treatment system were thoroughly evaluated, to investigate changes in physical and chemical properties resulting from the treatment. The color of the Ti particles changed from gray to uniform brown upon plasma-treatment, indicating good surface uniformity. Depth profiling using X-ray photoelectron spectroscopy showed that the nitrogen implanted by plasma-treatment formed Ti-N bonds at or near the surface of the particles, resulting in the formation of a stoichiometric TiN layer. It was determined from cross-sectional transmission electron microscopy and nanobeam diffraction measurements that the resulting TiN layer had two different structures. Nanoindentation data showed that the surface of a treated sample was about five times harder than that of untreated particles.

島 明日香¹、桜井 誠人¹、曽根 理嗣²、大西 充¹、米田 晶子³、阿部 孝之⁴

¹ (独)宇宙航空研究開発機構　研究開発本部未踏技研究開発本部未踏技術センター
²(独)宇宙航空研究開発機構 宇宙科学研究本部
³日本ピラー工業株式会社
⁴富山大学 富山大学 水素同位体科学研究センター

Abstract

    The Sabatier reaction catalyzed by titania-supported ruthenium (Ru/TiO₂) was investigated for the aim of practical water (H2O) generation from reduction of carbon dioxide (CO₂) with hydrogen (H₂) at lower temperatures. Various Ru/TiO₂ catalysts in powder form were prepared by a dry processing named “barrel-sputtering”. Hydrogenation of CO₂ to methane successfully proceeded on the catalysts at temperatures below 300°C without the formation of carbon monoxide (CO) even if pretreatment of the catalysts was carried out. It is noteworthy that catalytic activity of Ru/TiO₂ catalyst was enhanced when the catalyst was immobilized in a three-dimensional structure. In addition, use of the immobilized catalysts resulted in significant alleviation of not only catalyst weight but also temperature differences in the reactor.

Keyword(s): the Sabatier reaction, air-revitalization, titania-supported ruthenium catalyst, immobilization of catalyst

X.-G. Li^1,2,3, C. L.^1,2,3, J. Sun⁴, H. Xian^1,5, Y.-S. Tan⁶, Z. Jiang⁷, A. Taguchi⁸, M. Inoue⁸, Y. Yoneyama⁴, T. Abe⁸, N. Tsubaki^4,9

¹School of Chemical Engineering & Technology, Tianjin University, ²Tianjin Key Laboratory of Applied Catalysis Science & Technology, ³The Synergetic Innovation Center of Chemistry and Chemical Engineering of Tianjin,⁴Department of Applied Chemistry, School of Engineering,University of Toyama, ⁵Pei-Yang Distillation Engineering Limited Company,⁶State Key Laboratory of Coal Conversion, Institute of Coal Chemistry, Chinese Academy of Science, Taiyuan, ⁷Shanghai Synchrotron Radiation Facility, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, ⁸Hydrogen Isotope Research Center, University of Toyama

Abstract

    To substitute for petroleum, Fischer-Tropsch synthesis (FTS) is an environmentally benign process to produce synthetic diesel (n-paraffin) from syngas. Industrially, the synthetic gasoline (iso-paraffin) can be produced with a FTS process followed by isomerization and hydrocracking processes over solid-acid catalysts. Herein, we demonstrate a cobalt nano-catalyst synthesized by physical-sputtering method that the metallic cobalt nano-particles homogeneously disperse on the H-ZSM5 zeolite support with weak Metal-Support Interactions (MSI). This catalyst performed the high gasoline-range iso-paraffin productivity through the combined FTS, isomerization and hydrocracking reactions. The weak MSI results in the easy reducibility of the cobalt nano-particles; the high cobalt dispersion accelerates n-paraffin diffusion to the neighboring acidic sites on the H-ZSM5 support for isomerization and hydrocracking. Both factors guarantee its high CO conversion and iso-paraffin selectivity. This physical-sputtering technique to synthesize the supported metallic nano-catalyst is a promising way to solve the critical problems caused by strong MSI for various processes.

A. Aoki^a, M. Naruse^a, T. Abe^b

^a Materials Science & Engineering Graduate School of Engineering Nagoya Institute of Technology Tsukuri College, Gokiso, Showa-ku, Nagoya, 466-8555, Japan
^b Hydrogen Isotope Research Center, University of Toyama, 3790 Gofuku, Toyama 930-8555, Japan

Abstract

    A series circuit of bulk hetero-junction (BHJ) organic thin-film solar cells (OSCs) is investigated for electrolyzing water to gaseous hydrogen and oxygen. The BHJ OSCs applied consist of poly(3-hexylthiophene) as a donor and [6,6]-phenyl C61 butyric acid methyl ester as an acceptor. A series circuit of six such OSC units has an open circuit voltage (V_oc) of 3.4 V, which is enough to electrolyze water. The short circuit current (J_sc), fill factor (FF), and energy conversion efficiency (η) are independent of the number of unit cells. A maximum electric power of 8.86 mW cm^-2 is obtained at the voltage of 2.35 V. By combining a water electrolysis cell with the series circuit solar cells, the electrolyzing current and voltage obtained are 1.09 mA and 2.3 V under a simulated solar light irradiation (100 mW cm^-2, AM1.5G), and in one hour 0.65 mL hydrogen is generated.

Keyword: hydrogen evolution, series circuit, solar cells, thin films, water electrolysis

Y. Hatano^a, M. Shimada^a, V.Kh. Alimov^a, J. Shi^a, M. Hara^a, T. Nozaki^a a, Y. Oya^a c, M. Kobayashi^a c, K. Okuno^c, T. Oda^d, G. Cao^e, N. Yoshida^f, N. Futagami^f, K. Sugiyama ^g, J. Roth ^g, B. Tyburska-Püschel^g, J. Dorner^g, I. Takagi^h, M. Hatakeyamaⁱ, H. Kurishita ⁱ, M.A. Sokolov^a

^aHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^b Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
^c Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
^dDepartment of Nuclear Engineering and Management, The University of Tokyo, Tokyo 113-8656, Japan
^eDepartment of Engineering Physics, The University of Wisconsin, Madison, WI 53706, USA
^fResearch Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580, Japan
^gMax-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
^h Department of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan
ⁱ Institute for Materials Research, Tohoku University, Oarai 311-1313, Japan
^jOak Ridge National Laboratory, Oak Ridge, TN 37831, USA

Abstract

    Retention of D in neutron-irradiated W and desorption were examined after plasma exposure at 773 K. Deuterium was accumulated at a relatively high concentration up to a large depth of 50-100 μm due to the trapping effects of defects uniformly induced in the bulk. A significant D release in a vacuum continued to temperatures ≧1173 K because of the small effective diffusion coefficient and the long diffusion distance. Exposure of ion-irradiated W to D₂ gas showed a clear correlation between concentrations of trapped and solute D as determined by the trapping-detrapping equilibrium. These observations indicated that the accumulation of tritium in high concentrations is possible even at high temperatures if the concentration of solute tritium is high, and baking at moderate temperatures is ineffective for removal of tritium deeply penetrating into the bulk. Nevertheless, clear enhancement of D release was observed under the presence of solute H.