発表論文 2015年

[2015_01]

2D tritium distribution on tungsten tiles used in JET ITER-like wall project

Y. Hatanoa, A. Widdowsonb, N. Bekrisb, C. Ayresb, A. Baron-Wiechecb, J. Likonenc, S. Koivurantac, J. Ikonend, K. Yumizurua, JET-EFDA contributorse

aUniversity of Toyama, Gofuku 3190, Toyama 930-8555, Japan
bEURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, UK
cVTT Technical Research Centre of Finland, Association EURATOM-Tekes, P.O. Box 1000, FIN-02044 VTT, Finland
d University of Helsinki, Association EURATOM-Tekes, P.O. Box 43, FI-00014 University of Helsinki, Finland eJET-EFDA, Culham Science Centre, Abingdon OX14 3DB, UK

Abstract

Post-mortem measurements of 2-dimensional tritium (T) distribution using an imaging plate (IP) technique were performed for tungsten (W) divertor tiles (W-coated CFC) used in JET-ITER like wall (ILW) project. The observed T distributions were clearly inhomogeneous, and there were band-like regions with high T concentrations that extended in the toroidal direction on tiles 1, 3, 4 and 6. The concentrations of T in the band-like regions were higher by an order of magnitude than the concentrations in other parts. The inhomogeneous T distributions were explained by non-uniform co-deposition with other elements such as beryllium. The concentrations of T on the outboard vertical tiles (tiles 7 and 8) were low and relatively uniform in comparison with other tiles.

[2015_02]

Thermodynamics of hydrogen-induced superabundant vacancy in tungsten

K. Ohsawaa, F. Nakamorib, Y. Hatanoc, M. Yamaguchid

a Research Institute for Applied Mechanics, Kyushu University, Kasuga-koen 6-1, Kasuga-shi, Fukuoka 816-8580, Japan

bGraduate School of Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga 914-0055, Japan
cHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
dCenter for Computational Science and e-System, Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195, Japan

Abstract

We investigate superabundant vacancy formation induced by hydrogen in tungsten in terms of an equilibrium thermodynamic model to estimate hydrogen isotope retention in plasma facing materials. Vacancy-hydrogen cluster concentrations in the bulk tungsten are calculated as a function of the H concentration at finite temperature. A monovacancy in usual bcc transition metals is capable of accommodating six H atoms, while a maximum of 12 H atoms can be accommodated in a tungsten monovacancy, according to first-principle calculations. The present results provide thermodynamic profiles of vacancy-hydrogen clusters trapping more than six H atoms for the first time. In present work, configurational transitions of H atoms trapped in the monovacancy and activation energies for them are investigated by examining the transition paths in order to calculate configurational entropy. Vacancyhydrogen clusters trapping more than six H atoms exist in thermodynamic equilibrium. However, the major vacancy-hydrogen clusters are composed of six H atoms in a wide range of temperature and H concentration.

[2015_03]

Influence of helium on hydrogen isotope exchange in tungsten at sequential exposures to deuterium and helium-protium plasmas

N.P. Bobyra,, V.Kh. Alimovb,c, B.I. Khripunova, A.V. Spitsyna, M. Mayerb, Y. Hatanoc, A.V. Golubevaa,V.B. Petrova

aNRC"Kurcharov Institute", Ac. Kurcharov sq., 1/1, Moscow 123182, Russia
bMax-Planck-Institut fur Plasmaphysik, D-85748 Garching, Germany
cHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

Hydrogen isotopes exchange in tungsten was investigated after sequential exposures to low energy deuterium (D) and helium-seeded protium (He-seeded H) plasmas at sample temperatures of 403 and 533 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 8 μm. It was found that a significant part of the deuterium initially retained in tungsten after D plasma exposure was released during sequential exposure to a protium plasma. However, exposure of the D-plasma-exposed W samples to the He-seeded H plasma reduces the amount of released deuterium as compared to pure H plasma exposure.

[2015_04]

Helium irradiation effects on tritium retention and long-term tritium release properties in polycrystalline tungsten

Y. Nobutaa,, Y. Hatanob, M. Matsuyamab, S. Abeb, Y. Yamauchib, T. Hinoa

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

Abstract

DT+ ion irradiation with energy of 0.5 and 1.0 keV was performed on helium pre-irradiated tungsten and the amount of retained tritium and the long-term release of retained tritium in vacuum was investigated using an IP technique and BIXS. Tritium retention and long-term tritium release were significantly influenced by helium pre-irradiation. The amount of retained tritium increased until it reached 1×1017 He/ cm2, and at 1×1018 He/cm2 it became smaller compared to 1×1017 He/cm2. The amount of retained tritium in tungsten without helium pre-irradiation largely decreased after several weeks preservation in vacuum, and the long-term release rate during vacuum preservation was retarded by helium pre-irradiation. The results indicate that the long-term tritium release and the helium irradiation effect on it should be taken into account for more precise estimation of tritium retention in the long-term use of tungsten in fusion devices.

[2015_05]

DYNAMICS FOR HT AND HTO RECOVERY THROUGH WATER BUBBLER AND CUO CATALYST

Y. Oya1, M. Sato1, K. Yuyama1, M. Hara2, Y. Hatano2, M. Matsuyama2, T. Chikada1

1Graduate School of Science, Shizuoka University: 836 Ohya, Suruga-ku Shizuoka, 422-8529 Japan
2Hydrogen Isotope Research Center, University of Toyama: 3190 Gofuku Toyama, 930-8555 Japan

Abstract

Dynamics of tritium recovery using CuO catalyst and water bubbler was studied as a function of gas flow rate and CuO temperature. The rate constant of tritiated water formation by CuO catalyst at the temperature above 500 K was determined to be k [s-1] = 5.4×105 exp (-0.65 eV / kBT). For the flow rate less than 50 sccm, it was found that the reaction rate will be controlled by the desorption rate of HTO on the surface of CuO. These results were applied for the design of tritium removal system at radiation-controlled area. It was concluded that the reactor tubing with 1.0 meter length at 600 K will be suitable to reduce the tritium concentration less than 1/1000 and the longer reactor tubing will be required if the operation temperature will be lower than 600 K.procedure. In tritium specified facility, tritium removal system is equipped and tritium is easily recovered [2-6]. By the way, for the radiation-controlled facility, many kinds of radioactive materials will be handled and it is quite unreasonable to prepare tritium specified recovery system and HEPA or activated charcoal filter will be used for radioactive materials removal, by which tritium cannot be recovered. Therefore, conventional tritium removal technique using the combination of water bubbler and CuO catalyst will be equipped for each experimental device [7, 8]. However, the detail mechanism and correlation of tritium recovery rate with temperature of catalyst and purge gas flow rate is not clearly understood. This study focuses on elucidation of reaction mechanism from tritium gas to tritiated water by CuO and evaluation of reaction rate constant.

[2015_06]

Irradiation effect on deuterium behaviour in low-dose HFIR neutron-irradiated tungsten

M. Shimada1, G. Cao2, T. Otsuka2, M. Hara4, M. Kobayashi5, Y. Oya5, Y. Hatano4

1 Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID, USA
2 Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI, USA
3Kyushu University, Interdisciplinary Graduate School of Engineering Science, Higashi-ku, Fukuoka, Japan
4 Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan
5 Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, Japan

Abstract

Tungsten samples were irradiated by neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory at reactor coolant temperatures of 50-70 ℃ to low displacement damage of 0.025 and 0.3 dpa. After cooling down, the HFIR neutron-irradiated tungsten samples were exposed to deuterium plasmas in the Tritium Plasma Experiment, Idaho National Laboratory at 100, 200 and 500 ℃ twice at the ion fluence of 5 × 1025 m-2 to reach the total ion fluence of 1 × 1026 m-2 in order to investigate the near-surface deuterium retention and saturation via nuclear reaction analysis. Final thermal desorption spectroscopy was performed to elucidate the irradiation effect on total deuterium retention. Nuclear reaction analysis results showed that the maximum near-surface (<5 μm depth) deuterium concentration increased from 0.5 at% D/W in 0.025 dpa samples to 0.8 at% D/W in 0.3 dpa samples. The large discrepancy between the total retention via thermal desorption spectroscopy and the near-surface retention via nuclear reaction analysis indicated the deuterium was trapped in bulk (at least 50 μm depth for 0.025 dpa and 35 μm depth for 0.3 dpa) at 500 ℃ cases even in the relatively low ion fluence of 1026 m-2.

[2015_07]

Effect of hydrogen on fatigue crack propagation behavior of wrought magnesium alloy AZ61 in NaCl solution under controlled cathodic potentials

T. Kakiuchi1 , Y. Uematsu1, Y. Hatano2, M. Nakajima3, Y. Nakamura3, T. Taniguchi4

1Department of Mechanical and Systems Engineering, Gifu University, 1-1 Yanagido, Gifu 501-1193, Japan
2 Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
3Toyota National College of Technology, 2-1 Eisei-cho, Toyota, Aichi 471-8525, Japan
44Graduate Student, Course of Mechanical and Systems Engineering, Graduate School of Gifu University, 1-1 Yanagido, Gifu 501-1193, Japan

Abstract

Fatigue crack propagation (FCP) tests were performed in NaCl solution under controlled cathodic potentials to achieve hydrogen charged condition where anodic dissolution scarcely occurs to understand the effect of hydrogen on FCP behavior of wrought magnesium (Mg) alloy AZ61. FCP rates were accelerated under hydrogen charged conditions compared with dry air. Grazing incidence X-ray diffraction (GIXRD) and thermal desorption spectrometry (TDS) analyses revealed that FCP rates were independent of hydride compounds formed near the surface on crack wake, which indicates the acceleration could be attributed to hydrogen diffusion and hydrogen embrittlement is dominant in FCP behavior of Mg alloy.

[2015_08]

Thermal desorption behavior of deuterium for 6MeV Fe ion irradiated W with various damage concentrations

Y. Oyaa, X. Liaa, M. Satoa, K. Yuyamaa, L. Zhangb, S. Kondoc, T. Hinokic, Y. Hatanod, H. Watanabee, N. Yoshidae, T. Chikadaa

aGraduate School of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
bSouthwestern Institute of Physics, Chengdu, Sichuan 610041, China
cInstitute of Advanced Energy, Kyoto University, Uji, Kyoto 611-001, Japan
dHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
eResearch Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan

Abstract

W samples were irradiated at 300 K with 6 MeV Fe ion with damage concentrations in the range from 0.0003 to 1.0 displacements per atom (dpa) and then implanted at 300 K with 500 eV D ions to a fluence of 5 × 1021 D/m2. Deuterium retention in the damaged samples was examined in situ by thermal desorption spectrometry (TDS). Simulation of the TDS spectra was performed using the Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code to reveal the binding energies for deuterium captured by the ion-induced defects. It has been shown that the deuterium TDS spectra consist of two or three peaks (depending on the damage concentration) at about 400, 600 and 800 K, and can be simulated by the HIDT simulation code with the use of hydrogen-trap binding energies of 0.65, 1.25, and 1.55 eV.

[2015_09]

Tritium Retention in Reduced-Activation Ferritic/Martensitic Steels

Y. Hatanoa,, V. Kh. Alimova,b, A. V. Spitsync, N. P. Bobyrc, D. I. Cherkezc, S. Abea, O. V. Ogorodnikovab, N. S. Klimovd, B. I. Khripunovc, A. V. Golubeva,c V. M. Chernove, M. Oyaidzuf, T. Yamanishif, M. Matsuyamaa

aUniversity of Toyama,Gofuku 3190, Toyama 930-8555, Japan
bMax-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
cNRC Kurcharov Institute; Ac. Kurcharov sq., 1/1, Moscow RU-123182, Russia
d SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow Region, 142190, Russia
e JSC“A.A. Bochvar High-Technology Research Institute of Inorganic Materials”, Moscow, Russia
fJapan Atomic Energy Agency, 2-166 Oaza-Obuchi-Aza-Omotedate, Rokkasho-mura, Aomori 039-3212, Japan

Abstract

The effects of displacement damage, plasma exposure and heat loads on T retention in reduced-activation ferritic/martensitic (RAFM) steels were investigated by exposing the steels to DT gas at 473 K. Despite enormous change in surface morphology, T retention in the heat-loaded specimen was comparable with that in the unloaded specimen. The exposure to plasma resulted in a drastic increase in T retention at the surface and/or sub surface. However, the T trapped at the surface/subsurface was easily removed by maintaining the specimens in air at ~300 K. Formation of radiation-induced defects led to a significant increase in T retention, and T trapped in the defects was not removed at ~300 K. These observations suggest that displacement damages have the largest effects on T retention at ~473 K.

[2015_10]

Behavior of Tritium in the Vacuum Vessel of JT-60U

K. Kobayashi, Y. Torikai, M. Saito, V. Alimov, N. Miya, Y. Ikeda

Abstract

Disassembly of the JT-60U torus was started in 2010 after 18 years D2 operations. In future the vacuum vessel will be treated as non-radioactive ones after the clearance procedure under the Japanese regulation depending on the tritium (T) contamination level. Note that the vessel was manufactured from Inconel 625 steel. Therefore, it was very important to study the hydrogen isotope behavior in Inconel 625 from viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D2 (92.8%)- T2 (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily. From these results, the behavior of tritium in the vacuum vessel of the JT-60U torus will be evaluated and discussed

[2015_11]

Isotope Effects on Desorption Kinetics of Hydrogen Isotopes Implanted Into Stainless Steel by Glow Discharge

M. Matsuyama, M. Kondo, N. Noda, M. Tanaka, K. Nishimura

Abstract

Desorption kinetics of hydrogen isotopes implanted into type 316L stainless steel by glow discharge have been studied by the experiment and numerical calculation. The temperature of a maximum desorption rate depended on glow discharge time and heating rate. Desorption spectra observed under various experimental conditions were successfully reproduced by numerical calculation which is based on a diffusion-limited process. It is suggested, therefore, that desorption rate of a hydrogen isotope implanted into the stainless steel is limited by a diffusion process of hydrogen isotope atoms in bulk. Furthermore, small isotope effects were observed for the diffusion process of hydrogen isotope atoms.

[2015_12]

Measurement of Uptake and Release of Tritium by Tungsten

M. Nakayamaa,, Y. Torikaia, M. Saitoa, R.-D. Penzhorna, K. Isobeb, T. Yamanishib, H. Kurishitac

a Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama-shi, Toyama Pref. 930-8555, Japan
b Tritium Technology Group, Japan Atomic Energy Agency, 2-4 Shirakata-Shirane, Tokai-mura, Naka-gun, Ibaraki Pref. 319-1195, Japan
c Institute for Material Research, Tohoku University, 2145-2 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki Pref. 311-1313, Japan
Abstract

The uptake of tritium by tungsten and its release behavior have been investigated. Specimens annealed at 773 K, 873 K, 973 K, 1,073 K and 1,173 K for 3 hours and loaded with tritium at 773 K for 3 hours accumulated 0.125 ppm, 0.068 ppm, 0.067 ppm, 0.038 ppm and 0.033 ppm, respectively (tritium solubilities were 3.6x 10-9 at.fr.Pa-1/2, 2.0x 10-9 at.fr.Pa-1/2, 1.9x 10-9 at.fr.Pa-1/2, 1.1x 10-9 at.fr.Pa-12/ and 9.7x 10-10 at.fr.Pa-1/2, respectively). The difference is attributed to the existence of trapping sites or oxide films.

[2015_13]

Tritium Removal from Tritiated Water by Organic Functionalized SBA-15

A. Taguchi, Y. Kato, R. Akai, Y. Torikai, M. Matsuyama

Abstract

Mesoporous silicas (SBA-15) were modified by -COOH, -SO3H and -NH2 groups and their tritium adsorption ability from tritiated water under solid-liquid sorption was investigated. The adsorption abilities and separation factor of organic functionalized SBAs were comparable to those of bare SBA. The desorption of water from bare SBA and -COOH functionalized SBA were studied by Fourier transform infra-red spectroscopy using D2O as a probe molecule. An interaction was observed for D2O with -COOH group where the hydrogen bonds became weaker than D2O with bare SBA.

[2015_14]

Tritium Release from SS316 Under Vacuum Condition

Y. Torikai, R.-D. Penzhorn

Abstract

Type 316 stainless steel specimens loaded with tritium either by exposure to 1.2 kPa HT at 573 K or submersion into liquid HTO at 298 K showed characteristic thin surface layers trapping tritium in concentrations far higher than those determined in the bulk. The evolution of the tritium depth profile in the bulk during heating under vacuum was non-discernable from that of tritium liberated into a stream of argon. Only the relative amount of the two released tritium-species, i.e. HT or HTO, was different. Temperature-dependent depth profiles could be predicted with a one-dimensional diffusion model. Diffusion coefficients derived from fitting of the tritium release into an evacuated vessel or a stream of argon were found to be (1.4 × 1.0)×10-7 and (1.3 × 0.9) × 10-9 cm2/s at 573 and 423 K, respectively. Polished surfaces on type SS316 stainless steel inhibit considerably the thermal release rate of tritium.

[2015_15]

Tritium Trapping on the Plasma Irradiated Tungsten Surface

Y. Torikai, V. Kh. Alimov, K. Isobe, M. Oyaidzu, T. Yamanishi, R.-D. Penzhorn, Y. Ueda, H. Kurishita, V. Philipps, A. Kreter, M. Zlobinski, TEXTOR Team

Abstract

Tungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.

[2015_16]

Influence of He implantation on deuterium trapping at defects induced in W by irradiation with MeV-range W ions

V. Kh. Alimov1,2, Y. Hatano1, K. Sugiyama3, B. Tyburska-Püschel3, M. Oya4, Y. Ueda4, K. Isobe5, A. Hasegawa6

1 Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama-shi, Toyama 930-8555, Japan
2 National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow 115409, Russia
3 Max-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany
4Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871, Japan
5Tritium Technology Group, Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195, Japan
6 Department of Quantum Science and Energy Engineering, Tohoku University, 6-6-01-2, Aramaki-aza-Aoba, Aoba-ku, Sendai 980-8579, Japan
Abstract

The effect of helium (He) ion irradiation on the concentration of the ion-induced defects preliminary generated in tungsten (W) was investigated. Polycrystalline mechanically-deformed W samples were damaged by irradiation with 20 MeV W ions at room temperature to 0.5 displacements per atom (dpa) at the damage peak situated at a depth of 1.35 μm, followed by 1 keV He ion irradiation at 473, 623, and 823 K to fluences in the ranges of (3.7-5)×1022 He/m2. The samples were then loaded with deuterium (D) by D2 gas exposure at a pressure of 100 kPa at 673 K for 10 h. Concentration of deuterium trapped in the W-ion induced defects was determined by nuclear reaction analysis using the D(3He,p)4He reaction. It has been found that the effect of following He irradiation on the concentration of the W-ion-induced traps responsible for D trapping is negligible.

[2015_17]

Formation of α-alumina scales in the Fe-Al(Cr) diffusion coating on China low activation martensitic steel

Q. Zhana, W. Zhaoa, H. Yanga, Y. Hatanob, X. Yuana, T. Nozakib, X. Zhua
a China Institute of Atomic Energy, P.O. Box 275-55, Beijing 102413, PR China
b Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

To study the formation mechanism of stable α-Al2O3 scales, the oxidation behavior of Fe-Al(Cr) diffusion coating on China low activation martensitic steel has been investigated under the oxygen partial pressure ranging from 1 to 20,000 Pa at 1253 K. A single, continuous Al2O3 scale with the maximum thickness of about 2000 nm was formed on the Fe-Al(Cr) diffusion layer. The phase transformation of alumina scales on the surface of Fe-Al(Cr) layer was studied at different oxidation times ranging from 3 to 180 min. With the increase in oxygen partial pressure, the phase transformation time of α-Al2O3 is decreased. The metastable γ-Al2O3 and transition a-(Al0.948Cr0.052)2O3 phases were formed in the earlier oxidation process and finally transformed to the stable α-Al2O3 phase, which were detected by grazing incidence angle X-ray diffraction and confirmed by transmission electron microscopy. This implies that Cr shows the third element effect and serves as a template for the nucleation of the stable α-Al2O3.

[2015_18]

Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

M. Shimadaa, M. Harab, T. Otsukac, Y. Oyad, Y. Hatanob
aFusion Safety Program, Idaho National Laboratory, Idaho Falls, ID, USA
bHydrogen Isotope Research Center, University of Toyama, Toyama, Japan
cKyushu University, Interdisciplinary Graduate School of Engineering Science, Higashi-ku, Fukuoka, Japan
dRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, Japan

Abstract

Three tungsten samples irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to deuterium plasma (ion fluence of 1 × 1026 m-2) at three different temperatures (100, 200, and 500℃, in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy was performed with a ramp rate of 10 ℃ min-1 up to 900 ℃, and the samples were annealed at 900 ℃ for 0.5 h. These procedures were repeated three times to uncover defect-annealing effects on deuterium retention. The results show that deuterium retention decreases approximately 70% for at 500℃ after each annealing, and radiation damages were not annealed out completely even after the 3rd annealing. TMAP modeling revealed the trap concentration decreases approximately 80% after each annealing at 900 ℃ for 0.5 h.

[2015_19]

Development of positron annihilation spectroscopy for investigating deuterium decorated voids in neutron-irradiated tungsten

C.N. Taylora*, M. Shimadaa, B.J. Merrilla, D.W. Akersb, Y. Hatanoc
aFusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
bExperimental Programs, Idaho National Laboratory, Idaho Falls, ID 83415, USA
cHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan

Abstract

The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900℃ , and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium- free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined.

[2015_20]

In-situ observation of sputtered particles for carbon implanted tungsten during energetic hydrogen isotope ion implantation

Y. Oyaa, M. Satoa, H. Uchimuraa, N. Ashikawab, A. Sagarab, N .Yoshidac, Y. Hatanod, K. Okunoa
aGraduate School of Science, Shizuoka University, Shizuoka, Japan
bNational Institute for Fusion Science, Gifu, Japan
cInstitute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka, Japan
dHydrogen Isotope Research Center, University of Toyama, Toyama, Japan


Abstract

The effect of carbon implantation for the dynamic recycling of deuterium, which demonstrates tritium recycling, including retention and sputtering, was investigated using in-situ sputtered particle measurements. The C+ implanted W, WC and HOPG were prepared and dynamic sputtered particles were measured during H2+ irradiation. It was found that the major hydrocarbon species for C+ implanted tungsten was found to be CH3, although those for WC and HOPG were CH4. The chemical state of hydrocarbon is controlled by the H concentration in a W-C mixed layer. The amount of C-H bond and the retention of H trapped by carbon atom should control the chemical form of hydrocarbon sputtered by H2+ irradiation and the desorption of CH3 and CH2 was due to chemical sputtering, although that for CH was physical sputtering. The activation energy for CH3 desorption was estimated to be 0.4 eV, corresponding to the trapping process of hydrogen by carbon through the diffusion in W. It was concluded that the chemical states of hydrocarbon sputtered by H2+ irradiation for W was determined by the amount of C-H bond on the W surface.

[2015_21]

Preheating temperature effect on tritium retention in VPS-W

M. Satoa, H. Uchimuraa, K. Todaa, T. Tokunagab, H. Watanabeb, N. Yoshidab, Y. Hatanoc, R. Kasadad, T. Nagasakad, A. Kimurad, Y. Oyaa, K. Okunoa
aGraduate School of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, JAPAN
bResearch Institute for Applied Mechanics, Kyushu University, Kasuga-koen, Kasuga-city, Fukuoka 816-8580, JAPAN
cHydrogen Isotope Research Center, University of Toyama, 3190 Gohuku, Toyama-city, Toyama 930-8555, JAPAN
dInstitute of Advanced Energy, Kyoto University,Gokasho, Uji, Kyoto 611-0011, JAPAN
eNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki-city, Gifu 509-5292, JAPAN

Abstract

The deuterium retention behavior for the Vacuum Plasma Spraying (VPS) tungsten (W) coating was studied to demonstrate the tritium retention as a function of heating temperature. It was found that two major deuterium desorption stages were observed at the temperature regions of 400 - 700 K (Stage 1) and 900 - 1100 K (Stage 2), considering that Stage 1 was linked to the desorption of deuterium trapped by near surface and intrinsic defects, and Stage 2 was related to the desorption of deuterium bound to impurities as C-D bonds. By heating the sample above 673 K, the major peak of C-1s was shifted from C-O bond to C-C bond, where the retention of deuterium as Stage 2 was increased. Therefore it was indicating that the hydrogen isotope retention was controlled by the amount of C-C bond in VPS, most of which was contaminated during the VPS coating process. The comparison of several samples (VPS-W with shading, VPS-W without shading and Polycrystalline W (PCW)) shows that the carbon impurity has a large affinity with deuterium and make stable trapping states compared to that with intrinsic defects and grain boundaries. However, most of them was reduced by heating at 1173 K. Therefore, heating treatment is quite important to get rid of carbon impurities and refrain higher tritium retention in VPS.

[2015_22]

Dispersion and distribution of bimetallic oxides in SBA-15, and their enhanced activity for reverse water gas shift reaction

B. Lua, Y. Jub,T. Abea, K. Kawamotoc

Abstract

We used the direct hydrothermal synthesis method to obtain various well-dispersed bimetallic oxides/ SBA-15 for the first time. It is possible that well-dispersed relatively large bimetallic sulfates are formed during the hydrothermal synthesis process and then re-dispersed with difficulty during the heat treatment process resulting in the formation of well-dispersed oxide particles in SBA-15. TEM elemental maps of CuO-NiO/SBA-15 clearly illustrated that CuO and NiO particles were monodispersed in SBA-15. TEM- EDX line analysis revealed that NiO particles were well distributed on the SBA-15 surface, and then covered by CuO particles. TEM elemental maps of CuO-CeO2/SBA-15 clearly showed that CuO and CeO2 particles aggregated slightly in SBA-15. TEM-EDX line analysis showed that CeO2 particles were well distributed on the SBA-15 surface, and then covered by CuO particles. TEM elemental maps of NiO- CeO2/SBA-15 clearly illustrated that NiO and CeO2 particles aggregated slightly in SBA-15. TEM-EDX line analysis revealed that NiO particles were largely mixed with CeO2 on the SBA-15 surface. Therefore, TEM elemental maps can be used to study the dispersion of bimetallic oxides, and TEM-EDX line analysis is very effective for investigating their distribution in SBA-15. Compared with monometallic oxides/SBA-15, the obtained bimetallic oxides/SBA-15 catalysts exhibited excellent efficiency as regards reducing CO2 to CO by the reverse water-gas shift (RWGS) reaction. In particular, the bimetallic oxides/SBA-15 catalysts could result in the high CO2 conversion to CO at low temperature.

[2015_23]

Sensing hydrogen in the gas phase using ferromagnetic Pd-Co films

S. Akamarua, T. Matsumotob, M. Muraia, K. Nishimurab, M. Haraa, M. Matsuyamaa

aHydrogen Isotope Research Center, Organization for Promotion of research, University of Toyama, 3190 Gofuku, Toyama, 930-8555, Japan
b Faculty of Engineering, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

It is demonstrated that Pd-Co films can be used to sense the hydrogen concentration in the gas phase using the changes in the magnetic susceptibility, caused by the hydrogen uptake at ambient temperature. The induced voltage, which is proportional to the magnetic susceptibility of the Pd-Co films, was changed by the hydrogen concentration under the flow of the hydrogen/nitrogen mixture, and saturated under a constant flow. The saturated value of the induced voltage decreased with the hydrogen concentration in the gas phase and the change in the induced voltage was detected at a hydrogen concentration of 0.2 vol%. Under the flow of dry air, the induced voltage also reacted to hydrogen. However, the induced voltage did not have a clear change under a hydrogen concentration of 0.2 vol% in dry air. Moreover, the response time it took to reach saturation, from the injection of hydrogen, was slower than that under a hydrogen/nitrogen mixture. This behavior was likely caused by the water-forming reactions, hindering the absorption of hydrogen in the Pd-Co films in the presence of oxygen.
Keywords:Pd-Co film, Hydrogen sensor, Magnetic susceptibility

[2015_24]

Grafting Ni particles onto SBA-15, and their enhanced performance for CO2 methanation

B. Lua, Y. Jub, T. Abea, K. Kawamotoc

aHydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama, 930-8555, Japan
bLaboratory of Computational Geodynamics College of Earth Science, University of Chinese Academy of Sciences, Beijing 100049, China
cGraduate School of Environmental and Life Science, Okayama University, 3-1-1 Tsushima-naka, Kita-ku, Okayama-shi, Okayama, 700-8530, Japan

Abstract

By a post synthesis method, nickel (Ni) particles could be grafted onto SBA-15 for the first time through chemical bond (-O-Ni-O-Si-O-) formation between silicon (Si) and Ni via oxygen (O) using Ni ammonia (NH3) complex ions (Ni(NH3)x)2+ with an NH3/Ni mole ratio of 1-5, which existed as Ni phyllosilicate on the SBA-15 surface, while Ni particles could not be grafted onto SBA-15 in the absence of NH4OH (NH3/Ni mole ratio of 0). An NH3/Ni mole ratio of 2-4 was suitable for grafting conditions, which could give a product with the closest Ni amount to that of raw Ni complex ion solution. The product obtained was named as the Ni-grafted SBA-15 sample. XPS, UV-vis and H2-TPR analyses demonstrated that a chemical bond was formed between Ni and silicon (Si) via oxygen (O), and no bulk nickel oxides existed in the Ni-grafted SBA-15 sample. The formation of -O-Ni-O-Si-O- was completed via the reaction between hydrolyzate (Ni(OH)(NH3)x-1)+ from (Ni(NH3)x)2+ and = Si-OH (silanol sites) on the SBA-15 surface. The Ni-grafted SBA-15 catalyst suited CO2 methanation, resulting in higher CO2 conversion and methane selectivity than a NiO dispersed SBA-15 catalyst obtained by the conventional post synthesis method. The activation energy for CO2 methanation increased with a decreasing initial Ni amount used. The rate equation for CO2 methanation could be expressed as: r = kCCO20.64CH2 4.05, where C is the concentration. The Ni-grafted SBA-15 catalyst had high thermal stability for CO2 methanation.

[2015_25]

Analysis of ion recombination in ionization chambers for tritiummeasurements

Z. Chena,b, S. Pengaa,S. Chengaa, Z. Tanaa, H. Wangaa, M. Matsuyamaa

aInstitute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang, Sichuan 621900, China
bUniversity of Toyama, Gofuku 3190, Toyama-city, Toyama 930-8555, Japan

Abstract

Ion recombination in ionization chambers for tritium measurements has been studied theoretically and experimentally. A two-dimensional model for ionization chamber of cylindrical form was introduced to derive theoretical formulas for evaluating ion recombination loss quantitatively, and it can be extended to three dimensions easily. Experiments were carried out with an gold plated ionization chamber of 1.0 L at tritium concentration 1.10 × 1011 Bq/m3 and 1.98 × 1010 Bq/m3. With the established theory, recombination coefficient was determined to be around 6.9 × 10−6 cm3 /ion s in dry air, which agreed with value reported recently in the literature. In addition, ion densities in ionization chamber have been investigated, which shows good accordance with experiments. Furthermore, this proposed method for analyzing recombination coefficient is also suitable for other kinds of working gas besides dry air and for ionization chambers of different sizes.

[2015_A]

製氷化を利用した放射能汚染水の減容化技術

松山 政夫
富山大学 水素同位体科学研究センター

[2015_26]

Sputtered nano-cobalt on H-USY zeolite for selectively converting syngas to gasoline

PengLua,b,JianSunb,c,∗,PengfeiZhub,TakayukiAbed,RuiqinYanga,AkiraTaguchid,TharapongVitidsante,∗,NoritatsuTsubakib,∗
aZhejiang Provincial Key Lab for Chem. & Bio.Processing Technology of Farm Product, School of Biological and Chemical Engineering, Zhe jiang University of Science and Technology, Hangzhou 310023, Zhejiang,China
bDepartment of Applied Chemistry, School of Engineering, University of Toyama, Gofuku, Toyama930-8555, Japan
cDalian National Laboratory for Clean Energy(DNL), Dalian Institute of Chemical Physics, Chinese Academy of Sciences, Dalian 116023, Liaoning,China
dHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
eDepartment of Chemical Technology, Faculty of Science, Chulalongkorn University, Bangkok 10330, Thailand


abstract

Selectively converting CO and H2 to gasoline product (isoparaffin and olefin) in one steps till remains a great challenge. We demonstrate effective H-USY zeolite supported nano-cobalt bifunctional catalysts for this cat-alytic reaction, which a reprepared by the novel physical sputtering process.  Particles of the sputtered cobalt exist in nano-level and are well-dispersed on acid USY zeolite. Easy activation of the loaded nano-cobalt is also achieved in alow-temperature hydrogen reduction atmosphere. In the tandem catalytic reaction, the sputtered bifunctional Co/USY catalyst exhibits a much higher CO conversion and higher isoparaffins electiv-ity than the conventional impregnated one. Compared with H-Mor, H-Beta and other zeolites supported cata-lysts, H-USY zeolite supported cobalt catalyst shows the clearest promotional effect on the activity of Fischer-Tropsch synthesis. The described synthesis herein provides a new pathway to solve the problem caused by the strong metal-support interaction (MSI) in heterogeneous catalysis.


Keywords:Cobalt, USY,Zeolite,Fischer-Tropsch synthesis, Hydrocarbon

 

[2015_27]

Post-annealing Effects on Reaction Selectivity of Methanol Oxidation at Carbon-based Platinum Co-sputtered Electrocatalyst

Mitsuhiro INOUE,a Kazuhiro NAGAI,bSayoko SHIRONITA,b Takayuki ABE,a and Minoru UMEDAb,*
a Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
b Department of Materials Science and Technology, Faculty of Engineering, Nagaoka University of Technology, 1603-1 Kamitomioka, Nagaoka, Niigata 940-2188, Japan


abstract
A carbon with Pt (C–Pt) catalyst was prepared by a co-sputtering technique. Pt4f peaks of the prepared sample were observed at higher binding energies than those of a Pt sputtered sample. These Pt4f peaks shifted to the low binding energies with an increase in a post-annealing temperature, whereas the C1s peak position was unchanged. For the C–Pt catalyst as sputter-deposited, the methanol oxidation performance was enhanced by the presence of O2. This O2-enhnaced methanol oxidation was suppressed by the post-annealing at 160°C, while the reaction selectivity of the methanol oxidation was retained.


Keywords : Carbon-platinum Co-sputtered Anode Catalyst, Direct Methanol Fuel Cells, Methanol Oxidation Reaction Selectivity, Post-annealing Effect