V.Kh. Alimov ^a,^b, B. Tyburska-Püschel^c, Y. Hatano^a, J. Roth^c, K. Isobe^b, M. Matsuyama^a, T. Yamanishi^b

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

Abstract

    Samples prepared from polycrystalline ITER-grade tungsten were damaged by irradiation with 20 MeV W ions at room temperature to a fluence of 1.4 × 10¹⁸ W/m². Due to the irradiation, displacement damage peaked near the end-of-range, 1.35 μm beneath the surface, at 0.89 displacements per atom. The damaged as well as undamaged W samples were then exposed to low-energy, high-flux (10²² D/m² s) pure D and helium-seeded D plasmas to an ion fluence of 3 × 10²⁶ D/m² at various temperatures. Trapping of deuterium was examined by the D³(He,p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV allowing determination of the D concentration at depths up to 6 μm. It has been found that (i) addition of 10% helium ions into the D plasma at exposure temperatures of 440-650 K significantly reduces the D concentration at depths of 0.5-6 μm compared to that for the pure plasma exposure; (ii) generation of the W-ion-induced displacement damage significantly increases the D concentration at depths up to 2 μm (i.e., in the damage zone) under subsequent exposures to both pure D and D-He plasmas.

V.Kh. Alimov^a, ^bB. Tyburska-Püschel^c, S. Lindig^c, Y. Hatano^a, M. Balden^c, J. Roth^c, K. Isobe^b, M. Matsuyama^a, T. Yamanishi^b

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

Abstract

    Surface topography and deuterium retention in polycrystalline ITER-grade tungsten have been examined after exposure to a low-energy (38 eV/D), high-flux (10²² D/m² s) deuterium plasma with ion fluences of 10²⁶ and 10²⁷ D/m² at various temperatures. The methods used were scanning electron microscopy equipped with focused ion beam, thermal desorption spectroscopy, 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 320 to 815 K, small blisters of size in the range from 0.2 to 5 μm, depending on the exposure temperature and ion fluence, are formed on the W surface. At an ion fluence of 10²⁷ D/m², the deuterium retention increases with the exposure temperature, reaching its maximum value of about 10²² D/m² at 500 K, and then decreases below 10¹⁹D/m² at 800 K.

S. Fukada^a, Y. Edao^a, K. Sato^a, T. Takeishi^a, K. Katayama^a, K. Kobayashi^b, T. Hayashi^b, T. Yamanishi^b, Y. Hatano^c, A. Taguchi^c, S. Akamaru^c

^aDepartment of Advanced Energy Engineering Science, Kyushu University, Hakozaki, Higashi-ku, Fukuoka 812-8581, Japan
^bTritium Process Laboratory of Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan
^cHydrogen Isotope Research Center, University of Toyama, Gofuku, Toyama 930-8555, Japan

Abstract

    An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (i) how fast HTO penetrates through concrete walls, (ii) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (iii) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The experiment is comparatively carried out using disks of cement paste which W (water)/C (cement) weight ratio is 0.6:1 with or without hydrophobic paints, and mortar disks which W/C/sand ratio is 0.6:1:2 with or without the paints. The hydrophobic paints tested in the present study are an epoxy polymer resin paint and an acrylic-silicon polymer resin one. After T exposure during specified time under a constant HTO vapor pressure in an acrylic box, the amount of water-soluble HTO on/in the disks is determined using a technique of H₂O dissolution during specified time. The results obtained here are summarized as follows: (1) HTO penetration in porous concrete can be correlated in terms of the effective diffusivity. (2) Its value in porous cement without coating is 1.2 × 10^-11 m²/s at 25 ℃ (3) HTO penetrates only through pores in cement, and there is no path for HTO transfer in non-porous sand. (4) Rates of sorption and dissolution of HTO in disks of cement and mortar coated with the epoxy resin paint are correlated in terms of the effective diffusivity through the paint film which value is D_T = 1.0 × 10^-16 m²/s. The rate-determining step is diffusion through the paint. (5) The epoxy resin paint works more effectively as an anti-HTO diffusion coating. (6) Another acrylic-silicon resin paint does not work well as anti-HTO diffusion coating. This may be because the hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H₂O vapor or liquid. (7) The HTO uptake inside the epoxy paint is greater than that of the silicon one. (8) The permeation reduction factor (PRF) of HTO for the epoxy paint at steady-state is expected large, if HTO vapor only contributes to diffusion. However, when concrete surfaces coated with the epoxy paint are under wet conditions, the PRF value becomes smaller. All those results can be used to estimate the effect of HTO soaking in concrete walls in case of accidental T release in a fusion reactor room and to decontaminate wastes of tritiated concrete.

中山将人、藤樫由佳、原　正憲、松山政夫

富山大学水素同位体科学研究センター
930-8555　富山県富山市五福3190

Abstract

    富山大学水素同位体科学研究センターにおいて、これまでトリチウム水を測定する際に使用していたLumaSafe Plusというシンテレ一夕が急きょ販売中止となり、よってその代替シンチレ一夕として、oscintXRとUltimaGoldABの特性を調べた｡ Ecoscint XRでは、測定値の不確かさがおよそ10%以内となったものの、酸等を含む試料水で計数効率の大幅な低下が見られた｡ 一方、Ultima Gold ABでは、測定値の不確かさはほぼ同程度であったが、計数効率の低下は見られなかったため、 当センターでの実験に使用する試料水に適した代替品であると判断された｡

Key Words: tritium, liquid scintillator, quenching, Ecoscint XR, Ultima Gold AB

Keywords: Tritium, trapping, Stalnless steel, Cu-Bealloy, BIXS

S. Akamaru¹, M. Inoue^1.2, Y. Honda^1.3, A. Taguchi¹, T. Abe¹

¹Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
²Department of Materials Science and Technology, Nagaoka University of Technology, Nagaoka, Niigata 940-2118, Japan
³YOUTEC Co., Ltd., Nagareyama, Chiba 270-0156, Japan

Abstract

    Ni nanoparticles were prepared on a submicron-sized Al₂O₃ powdery substrate by the polyhedral-barrel-sputtering technique and their microstructure and magnetic properties were investigated. By this technique, nanoparticles with an average diameter of 3.9-7.9nm were prepared on the Al₂O₃ particles surface, and the size distribution of the nanoparticles was reproduced using a log-normal formula. The deposited Ni nanoparticles were composed of a NiO layer, which was formed by air oxidation, and a Ni core. The magnetization curves suggested that the all the prepared samples exhibited superparamagnetic behavior, which can be explained by considering the size distribution of the Ni nanoparticles. An exchange-bias effect, which arising from the core-shell structure, was also observed. From the saturated magnetic moments of the Ni core, the thickness of the surface layer was estimated to be 1.0-1.3 nm, which was comparable with that of passivated oxide layer of bulk Ni.

S. Akamaru, M. Hara, and M. Matsuyama

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

Abstract

    An alternating current magnetic susceptometer for use as a hydrogen gauge for hydrogen-storage materials was designed and developed. The experimental system can simultaneously measure the hydrogen equilibrium pressure and the magnetic susceptibility of metal hydrides. The background voltage of the susceptometer was stabilized for a long period of time，without any adjustments，by attaching an efficient compensation circuit. The performance of the susceptometer at a static hydrogen concentration was demonstrated by measuring the magnetic susceptibility of a Pd-hydrogen system under equilibriurn conditions. The in situ measurement of the magnetic susceptibility of Pd during hydrogen absorption was carried out using the susceptometer. Since the in situ magnetic susceptibility obtained at a lower initial hydrogen pressure agreed with the magnetic susceptibility measured at a static hydrogen concentration，the susceptometer could be used to determine the hydrogen concentration in Pd in situ. At a higher initial hydrogen pressure，enhancement of the magnetic susceptibility was observed at the beginning of hydrogen absorption because the magnetic moments induced by the large temporary strain generated in the Pd affected the magnetic susceptibility.

M. Matsuyama, J. Saikawa, S. Abe, K. Nishimura¹, N. Ashikawa¹, Y. Oya²,
K. Okuno², T. Hino³, A. Sagara¹

Hydrogen Isotope Reseach Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
¹National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan
²Radiochemisty Research Laboratory, Shizuoka University, Sizuoka 422-8529, Japan
³Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo, 060-8628, Japan

Abstract

    Tritium retention of samples exposed to plasmas in the Large Helical Device (LHD) during each campaign in 12th, 13th and 14th cycles has been studied. Small sample plates made of stainless steel type 316L were fixed in advance at four different walls in LHD: location of a sample plate was I.5U, 5.5U, 6.5L and 9.5L. After plasma exposure in each cycle, these samples were exposed to tritium gas at a temperature of 300 or 623 K. Retention behavior of tritium in surface layers of each sample was mainly examined using β-ray-induced X-ray spectrometry (BIXS) and X-ray photoelectron spectroscopy (XPS). The energy spectra observed by BIXS and XPS showed the depositions of boron, carbon, titanium, chromium, iron, nickel and molybdenum with oxygen. Tritium retention of the samples exposed to plasma increased than that of a bare SS3 16L sample,althoughit was largely different inthe location of a sample. When the samples were exposed to tritium gas at 300K, the order of magnitude of tritium retention was as follows: 9.5L>>5.5U>6.5L>1.5U for 12th cycle, 6.5L>9.5>1.5U>5.5U for 13th cycle,and 6.5L>1.5U~1.5U>9.5L for 14th cycle.

Keywords: tritium, retention, plasma exposure, BIXS, stainless steel, plasma-facing material

Keywords: Tungsten, SS-316, Simultaneous implantation, Fuel retention

M. Hara, Y. Togashi

Hydrogen Isotope Reseach Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract

    The transport of tritiated water through polyimide film in the temperature range 278-313 K was measured utilizing a permeation measurement system designed by the authors. Basically, it consists of a flow system equipped with water bubblers. The amount of tritiatedwater permeated through a polymer is evaluated from the concentration rise of tritium in the bubblers. The permeability coefficient of tritiatedwater through polyimide decreased with increasing temperature. Pre-exponential factor and activation energy of the permeability coefficient were found to be (2 ± 1) × 10^-14 kmol m m^-2 s^-1 kPa^-1 and -5 ± 2 kJ mol^-1, respectively. This negative activation energy agrees with results published by Okamoto et al. The enthalpy of the tritiatedwater solubility was determined to be -34 ± 13 kJ mol^-1. Since permeation can be described by a one-dimensional diffusion model, the diffusion coefficient can be obtained from the quotient of permeability coefficient and solubility. The activation energy for the diffusion of water through polyimide was calculated to be 29 ± 13 kJ mol^-1.

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

R.-D. Penzhorn^a, Y. Torikai^a, K. Watanabe^a, M. Matsuyama^a, A. Perevezentsev^b

^aHydrogen Isotope Reseach Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
^bUKAEA JET Facilities, Abingdon, Oxon OX14 3EA, UK

Abstract

    Several type 316L stainless steel specimens of 6 mm thickness were charged with tritium at 473 K at Joint European Torus (JET) using five sets of conditions. Isothermal tritium release rates were investigated at Hydrogen Isotope Research Centre (HRC) over extended periods of time at 473,573, or 673 K constant temperature. The HTO/HT ratio of the liberated tritium was generally high, but decreased with decreasing release temperature. Nearly complete release of tritium required additional prolonged heating at 1073 K. Chemical etching and beta-ray-induced X-ray spectrometry measurements carried out at HRC provided complementary information on the tritium distribution in surface and bulk of thermally treated specimens. Whereas the thickness of the material and initial distribution of tritium in its bulk were found to play an important role for expedient thermal decontamination, the influence of the type of purge gas was only minor. Experimental evidence for tritium grain boundary diffusion is provided. Implications of the results for waste conditioning are discussed.

H. Ohuchi-Yoshida^a, Y. Hatano^b, Y. Kino^c, Y. Kondo

^aGraduate School of Pharmaceutical Sciences, Tohoku University, 6-3 Aramaki-Aoba, Aoba-ku, Sendai 980-8578, Japan
^bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^c Graduate School of Science, Tohoku University, Aramaki-Aoba, 6-3 Aoba-ku, Sendai 980-8578, Japan
^d Ishinomaki Senshu University, 1 Shinmito Minamisakai Ishinomaki 986-8580, Japan

Abstract

    A method to detect tritium non-destructively in regions deeper than the escape depth of beta rays is being developed using bremsstrahlung induced by beta rays with an imaging plate (IP). An IP made of europium-doped BaFBr(I), a photostimulated luminescence (PSL) material, is a two-dimensional radiation sensor. The bremsstrahlung energy spectrum is a continuum with photon energies, varying based on the atomic number and thickness of the target (or absorbing) material. When tritium migrates into matter, the bremsstrahlung energy spectrum distribution would change. The PSL intensity of the IP is affected by this energy spectrum variation. In order to quantify the amount of tritium in deeper regions with the IP technique, a tritium depth profile is required. In this study, a new method of obtaining a tritium depth profileusing the combined technique of the IP and thin absorbers is presented.

Y. Oya^a, M. Kobayashi^a, J. Osuo^a, M. Suzuki^a, A. Hamada^a, K. Matsuoka^a, Y. Hatano^b, M. Matsuyama^b, T. Hayashi^c, T. Yamanishi^c, 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
^cJapan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan

Abstract

    Effect of surface oxide layer on the hydrogen isotope permeation was studied. Iron oxide was uniformly formed in the oxide layer, although chromium was limited at the interface between the oxide layer and bulk SS-316. The permeation behavior of deuterium for oxidized SS-316 was compared with that for unoxidized SS-316 at temperature range of 333-673 K. The deuterium permeability for the oxidized SS- 316 was reduced 1/10-1/20 times as high as that for unoxidized one. However, the activation energy of deuterium permeation as gas form for oxidized SS-316 was almost the same as that for unoxidized SS-316 and was 0.64 eV, which was almost consistent with the sum of activation energies for diffusion and solubility. This fact indicates that the deuterium permeation is diffusion limited. The permeability of deuterium as water form was almost constant even if heating temperature is high, showing that the deuterium was permeated through bulk SS-316 and react with oxygen at the oxide layer as water desorption, which is controlled by the permeation flux of deuterium and oxygen concentration on the surface of oxide layer in downstream side.

Y. Edao^a, S. Fukada^a, Y. Nishimura^a, K. Katayama^a, T. Takeishi^a, Y. Hatano^b, A. Taguchi^b

^aDepartment of Advanced Energy Engineering Science, Kyushu University, Fukuoka 812-8581, Japan
^bHydrogen Isotope Research Center, University of Toyama, Gofuku, Toyama 930-8555, Japan

Abstract

    The effects of hydrophobic paint coating on a concrete material of cement paste on the tritium transport are investigated. The cement paste is coated with two kinds of paints, acrylic-silicon resin paint and epoxy resin paint. We investigated the amount of tritium trapped in the samples exposed to tritiated water vapor by means of sorption and release. It was found that both the hydrophobic paints could reduce effectively tritium permeation during 50 days exposure of tritiated water vapor. The effect of tritium reduction of the epoxy paint was higher than that of silicon while the amount of tritium trapped in the epoxy paint was larger than that of silicon due to difference of the structure. Based on an analysis of a diffusion model, the rate-determining step of tritium migration through cement paste coated with the paints is diffusion through the paints respectively. It was found that tritium was easy to penetrate through silicon because there were many pores or voids in the silicon comparatively. In the case of tritium released from the epoxy paint, it is considered that tritium diffusion in epoxy is slow due to retardation by isotope exchange reaction to water included in epoxy paint.

Yuji Nobuta^a,Yuji Yamauchi^a, Tomoaki Hino^a, Satoshi Akamaru^b, Yuji Hatano^b, Masao Matsuyama^b, Satoshi Suzuki^c, Masato 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

    Co-deposited carbon film with different deuterium concentration, D/C, were exposed to tritium gas at the temperature of 423 K, and then the atomic ratio of absorbed tritium to carbon, T/C, was evaluated. The obtained data were discussed with crystal structure of the carbon film. The T/C increased with decreasing D/C of carbon film. The carbon film with low D/C had more defective structure. The reduction of D/C by the heating before tritium exposure led to the increase of absorption amount. These results suggest that carbon film with more defective structure and low D/C film could absorb large amount of tritium. The hydrogen isotope concentration in the present experiment was saturated below the orders of 10^-4, which was 3-4 orders of magnitude smaller than that of co-deposited carbon film with hydrogen isotope.

Y. Hatano^a, J. Shi^a, N. Yoshida^b, N. Futagami^b, Y. Oya^c, H. Nakamura^d

^aHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^bResearch Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^cRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
^dTritium Technology Group, Japan Atomic Energy Agency, Tokai-mura 319-1195, Japan

Abstract

    To examine the resolution of isotope analysis of hydrogen with glow-discharge optical emission spectroscopy (GDOES), depth profiles of hydrogen and deuterium in a H-containing Ta/D-containing Ti/Ni layered structure were measured. The depth profiles of deuterium could be measured with sufficient resolution in the presence of relatively large amounts of hydrogen and vice versa. In addition, measurements of depth profiles of He implanted in W at room temperature were also performed with Ne plasma. The intensity of the He emissions was sufficiently high at a fluence of 1020 He m^-2 or higher. The depth profiles of He measured in this manner were in good agreement with the results of cross-sectional observations using a transmission electron microscope. Therefore, it was concluded that GDOES with Ne plasma is a promising technique for the depth profile analysis of plasma-facing materials and deposited layers formed on them.

H. Nakamura^a, Y. Hatano^b, T. Yamanishi^a

^aTritium Technology Group, Japan Atomic Energy Agency, Tokai-mura 319-1195, Japan
^bHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

    Deuterium behavior into the metals soaked in hot heavy water has been investigated in order to understand the oxidation initiated tritium infiltration in the fusion reactor. Disks of SS304, F82H and Ni and gold plated SS304 and F82H as the tritium permeation coating were soaked in an autoclave at 573 K. After the soaking, residual deuterium in the specimen was measured by the thermal desorption method and elemental depth distribution in the specimen was measured by a glow discharge optical elemental spectroscopy method. As the results, the oxide thickness has grown with the soaking time, and infiltrated deuterium amount also increases with oxidation time. Deuterium exists at the interface of the oxide and metals. Deuterium in the gold plated specimens was less than that in the bare SS304 about 1/5. Deuterium in nickel was less than that in the SS304 by one orders magnitude and oxide thickness was also thinner than SS304. Those results indicate that deuterium infiltration into the material would be initiated by the deuterium gas generated at the oxidation process of metal. Gold plating as the permeation reduction coating could be relatively effective to prevent deuterium infiltration into the metal in SS304.

Y. Hatano^a, M. Hara^a, H. Ohuchi-Yoshida^b, H. Nakamura^c, T. Yamanishi^c

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bGraduate School of Pharmaceutical Sciences, Tohoku University, 6-3 Aramaki-Aoba, Aoba-ku, Sendai 980-8578, Japan
^cTritium Technology Group, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195, Japan

Abstract

    Concentration of tritium in water (4-400 kBq cm^-3) was measured by exposing an imaging plate without protection layer (Fujifilm, BAS-IP TR) to vapor for 2-48 h. It was found that tritium gradually penetrated into Eu-doped BaFBr phosphor and induced sufficiently intense photostimulated luminescence (PSL) even at the concentration of 4 kBq cm^-3. The intensity of PSL was proportional to tritium concentration in water. In addition, tritium absorbed in phosphor was reversibly released by keeping IP in air, and IP was able to be used repeatedly if total duration of exposure was ca. 24 h or less. The contamination of IP with tritium was not serious. It was concluded that IP technique has potential to measure tritium concentration in water without direct handling of tritiated water and with a minimum amount of radioactive waste.

M. Shimada^a, Y. Hatano^b, Y. Oya^c, T. Oda^d, M. Hara^b, G. Cao^e, M. Kobayashi^c, M. Sokolov^f, H. Watanabe^g, B. Tyburska-Püschele^e,h, Y. Uedaⁱ, P. Calderoni^a, K. Okuno^c

^aFusion Safety Program, Idaho National Laboratory, Idaho Falls, ID, USA
^bHydrogen Isotope Research Center, University of Toyama, Toyama, Japan
^cRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, Japan
^dDepartment of Nuclear Engineering and Management, The University of Tokyo, Tokyo, Japan
^eDepartment of Engineering Physics, University of Wisconsin-Madison, Madison, WI, USA
^fOak Ridge National Laboratory, Oak Ridge, TN, USA
^gResearch Institute for Applied Mechanics, Kyushu University, Fukuoka, Japan
^h Institute fur Plasmaphysik, EURATOM Association, Garching, Germany
ⁱGraduate School of Engineering, Osaka University, Osaka, Japan

Abstract

    The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US-Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 ℃ to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of iondamaged tungsten via various ion species (2.8 MeV Fe²⁺, 20 MeV W²⁺, and 700 keV H^-) with that from neutron-irradiated tungsten to identify the similarities and differences among them.

M. Nose^a, W. A. Chiou^b, T. Kawabata^c, Y. Hatano^d, K. Matsuda^c

^aFaculty of Art and Design, University of Toyama, 180 Futakami, Takaoka-shi, Toyama 933‐8588, Japan
^bNISP Lab., NanoCenter, University of Maryland, College Park, MD 20742‐2831, USA
^cSchool of Science and Engineering, University of Toyama, Toyama-shi, Toyama 930‐8555, Japan
^dHydrogen Isotope Research Center, University of Toyama, Toyama-shi, Toyama 930‐8555, Japan

Abstract

    A CrAlN/18 vol.% BN nanocomposite film was deposited on substrate by reactive co-sputtering. The films showed an increase of about 30% in indentation hardness and achieved a maximum hardness of approximately 50 GPa after annealing at 800 ℃ in air. In contrast, the indentation hardness barely changed when the film sample was annealed at 800 ℃ in nitrogen and argon atmosphere. High-resolution transmission electron microscopy (HRTEM) images revealed that the uppermost layer was characterized by amorphous materials with embedded nanocrystalline particles (occurring at less than ~50 nm below surface). Energy dispersive X-ray spectroscopy (EDS) line profiles of cross-sectional thin films showed a high concentration of oxygen in the uppermost layer of the annealed sample. The indentation hardness of the air-annealed sample was measured by Ar⁺ ion sputtering before and after etching to the depth at 200 nm from the annealed surface. The hardness decreased from approximately 48 GPa to 43 GPa, which was the same level as the asdeposited films. These results indicate that oxidization of the film surface could be one of the factors responsible for the self-hardening of the CrAlN/BN film.

A. Hamada^a, M. Kobayashi^a, K. Matsuoka^a, M. Suzuki^a, J. Osuo^a, N. Ashikawa^a, A. Sagara^a, Y. Hatano^a, Y. Oya^a, K. Okuno^a

^aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, Japan
^bNational Institute for Fusion Science, Toki, Japan
^cHydrogen Isotope Research Center, University of Toyama, Toyama, Japan

Abstract

    The behavior of hydrogen retention and the change of chemical states of boron film exposed to hydrogen plasma in LHD were investigated. The sample was prepared in LHD, and atomic concentrations for the boron film after hydrogen plasma exposure were changed from 75% for boron, 15% for carbon and 8% for oxygen to 53%, 18% and 22%, respectively. B-C bond was a major chemical state of the boron film after hydrogen plasma exposure, although abundance of B-B bond was the highest before the plasma exposure. Total hydrogen retention measured by TDS was evaluated to be 1.7 × 1020 H m^-2, and the retentions of hydrogen as B-H-B, B-H and B-C-H bonds were, respectively, 4.8 × 10¹⁹, 7.2 × 10¹⁹ and 5.2 × 10¹⁹ H m^-2. It was concluded that the hydrogen retention could be estimated by taking account not only of chemical states of impurities, but also of hydrogen depth profile.

阿部孝之

富山大学 水素同位体科学研究センター

Abstract　なし

山西敏彦、波多野雄治¹⁾、赤丸悟士¹⁾、朝倉大和²⁾、磯部兼嗣、岩井保則、 　　　　　　　　　　　　　　　　奥野健二³⁾、小田卓司⁴⁾、大矢恭久³⁾、杉山貴彦⁵⁾、田中 知⁴⁾、鳥養祐二¹⁾、中村博文、林 巧、原 正憲¹⁾

日本原子力研究開発機構トリチウム工学研究グループ、
　　　　　　　　　　　　　　　　　　富山大学水素同位体科学研究センター^1)、
　　　　　　　　　　　　　　　　　　核融合科学研究所²⁾、
　　　　　　　　　　　　　　　　　　静岡大学理学部³⁾、
　　　　　　　　　　　　　　　　　　東京大学大学院工学系研究科⁴⁾、
　　　　　　　　　　　　　　　　　　名古屋大学大学院工学研究科⁵⁾、

Abstract　なし

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

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

Abstract

    A new plasma-treatment system was developed for preparing a functional powdery material. The nitrogen (N)-doped titanium dioxide (TiO₂) powder was prepared to determine the validity of this system. The physical properties of the plasma-treated TiO₂ powder agreed with those of a sample heated at 550 ℃ in the presence of ammonia, revealing the formation of the N-doped TiO₂. The absorbance in the visible region and the N contents of the plasma-treated samples depended on the plasma-treatment conditions that included the treatment time, RF power, and N² pressure. The specific surface areas of the samples were found to be 231-242 m²/g, which were similar to that of the untreated TiO₂ powder (236 m²/g).

Keywords: Plasma-treatment system, Nitrogen-doped TiO₂, Powdery material, Specific surface area

Y. Honda, S. Akamaru, M. Inoue, T. Abe

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

Abstract

    A novel polygonal barrel-plasma enhanced chemical vapor deposition (PB-PECVD) method was developed to prepare powdered materials with diamond-like carbon (DLC) films. To investigate the effects of frequency on the DLC film preparation, frequency power supplies of 250 kHz and 13.56 MHz were used and a DLC film was successfully formed on an SiO₂ glass plate at 250 kHz. The PB-PECVD process was further examined by using polymethyl methacrylate (PMMA) powder sample. It was determined that a DLC film could be uniformly deposited onto the PMMA particles by rotating a hexagonal barrel including the PMMA sample during the PB-PECVD procedure while operating at 250 kHz.

Keywords: Polygonal barrel-plasma enhanced chemical, vapor deposition (PB-PECVD) system, Film deposition, Polymer particles, Diamond-like carbon, Dry process

A. Shima¹, M. Sakurai², Y. Sone³, M. Ohnishi⁴, T. Abe ⁵

^1,2,3,4JAXA, Chofu-shi, Tokyo, 182-8511, Japan
⁵Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

    The Sabatier reaction catalyzed by titania-supported ruthenium (Ru/TiO₂) was investigated for the aim of practical water (H₂O) generation from carbon dioxide (CO2) reduction and hydrogen (H₂) at lower temperatures. Various Ru/TiO₂ catalysts in powder form were prepared by dry processing. Hydrogenation of CO₂ to methane successfully proceeded on the catalysts at temperatures below 300℃ without the formation of carbon monoxide (CO). It is noteworthy that some of the Ru/TiO₂ catalysts maintained their chemical properties after they were 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.