Y. Oya^a, S. Suzuki^a, M. Matsuyama^b, T. Hayashi^c, T. Yamanishi^c, Y. Asakura^d, K. Okuno^a
aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
^bITER Organization, St.Paul-lez-Durance, France
^cJapan Atomic Energy Agency, Ibaraki, Japan
^d National Institute for Fusion Science, Toki, Japan

Abstract

The ion fluence dependence of deuterium retention in SS-316 during oxidation at a temperature of 673 K was studied to evaluate the dynamics of deuterium retention in the oxide layer of SS-316. The correlation between the chemical state of stainless steel and deuterium retention was evaluated using XPS and TDS.
   It was found that the major deuterium desorption temperatures were located at around 660 K and 935 K, which correspond to the desorption of deuterium trapped as hydroxide. The deuterium retention increased with increasing deuterium ion fluence, since the deuterium retention as hydroxide increased significantly. However, retention saturated at an ion fluence of ~ 2.5 ×10²¹ D⁺+ m^-2. The XPS result showed that FeOOD was formed on the surface, although pure Fe also remained in the oxide layer. These facts indicate the nature of the oxide layer have a key role in deuterium trapping behavior.

J. Wu^a, Z. Yang^a, Q. Li^a, C.-Y. Xie^a, G.-N. Luo^a, M. Matsuyama^a

^a Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031, China
^b Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

Tritium exposed samples, including doped graphite, SiC coated graphite, tungsten coatings fabricated by vacuum plasma spraying, and polycrystalline tungsten, were examined by b-ray-induced X-ray spectrometry in an argon atmosphere. The changes in the X-ray spectra with time were followed for a maximum time of 100 h. The results indicated that the SiC coated graphite absorbed the most tritium and the polycrystalline tungsten the least. Preliminary computer simulations were performed to analyze the depth profiles of tritium in these materials.

T. Nozaki , H. Homma, Y. Hatano

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

Abstract

   Pressure-composition isotherms of TiC_1-x -H system were measured at 773,873 and 973 K over a pressure range from 10^-2 to 10⁵ Pa. The obtained isotherms could be divided in three concentration regions. In the low concentration region in which[H]/[TiC] ≦ 1.4 × 10^-3, Sieverts' law held, and the solubility of hydrogen clearly increased with decreasing temperature. The enthalpy of solutions was – 70 ± 5 kJ mol^-1. In the region in which 1.4 × 10 ^-3 < [H]/[TiC] ≦ 8.0 × 10^-3, the dependence of hydrogen concentration, C_H, on pressure and temperature weakened as pressure lincreased. Finally,C_H was independent of temperature when it reached 8.0 × 10^-3. This C_H was comparable with the concentration of carbon vacancies in TiC_1-x. In the high concentration region in which [H]/[TiC] ≧ 8.0 × 10^-3, Sieverts' law held agaln, but the solublilty of hydrogen was almost independent of temperature. These observations independent of temperature. These observations indeicated that hydrogen dissolved in carbon vacancies in the low and middle concentration regions with the above-mentioned value of enthalpy of solution, and also occupied much less stable sites in the high concentration region.

M.Matsuyama¹, Z.Chen², K.Nisimura³, S.Akamaru¹, Y.Torikai¹, Y.Hatano¹, N.Ashikawa³, Y.Oya⁴, K.Okuno⁴, T.Hino⁵

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

    Tritium amount retained in surface layers and release behavior from surface layers were examined using SS316L samples exposed to plasmas in the Large Helical Device and commercial Cu-Be alloy plate. BIXS analysis and observation by SEM indicate that carbon and titanium deposited on the plasma-facing surface of the SS316L samples. Larger amount of tritium was trapped in the plasma-faclng surface in comparison with the polished surface. Hligher enrichment of tritium in surface layers was similarly found in the pollShed surihce of both samples.The amount of surface trltlum in both samples was alrnost same, comparison with polished surface. Higher enrichment of tritium in surface layers was similarly found in the polished surface of both samples. The amount of surface tritium in both samples was almost same, while the bulk concentration of tritium in Cu-Be was much lower than that in SS316L. Tritium release from the SS316L and Cu-Be samples into water was examined by immersion experiments. Tritium elution was observed for both samples, but changes in the residual tritium amount in surface layers were different from each other.

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

M. Hara^a, K. Yudou^b, E. Kinoshita^b, K. Okazaki^b, K. Ichinose^b, K. Watanabe^a, M. Matsuyama^a

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^b Faculty of Human Development, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

   To study the alloying effects on ZrMn₂–H system, thermodynamic properties of Zr(Mn_1–xCo _x)₂ hydride were measured by volumetric method. ZrMn₂ gave a single plateau region in the pressure–composition isotherm. On the other hand, double plateaus were clearly observed in Zr(Mn_0.7Co_0.3) ₂ and Zr(Mn_0.6Co_0.4)₂–H systems. The appearance of the double plateau characteristics would be explained in view of the hydrogen binding in the tetrahedral occupation sites in Zr(Mn_1–xCo_x)₂. Since the hydrogen binding in the tetrahedral 2ZrMnCo site would be less stable than that in the 2Zr2Mn site, the equilibrium pressure increases with increasing cobalt content. The appearance of the first plateau was ascribed the increase in the bonding of Mn–H in 2Zr2Mn site adjoining the 2ZrMnCo site.

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

P. Calderonia^a, J. Sharpe^a, M. Shimada^a, B. Denny^a, B. Pawelko^a, S. Schuetz^a, G. Longhurst^a, Y. Hatano^b, M. Hara^b, Y. Oya^c, T. Otsuka^d, K. Katayama^d, S. Konishi^e, K. Noborio^e, Y. Yamamoto^e

^aFusion Safety Program, Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-7113, USA
^bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^cRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
^dInterdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581, Japan
^eInstitute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011, Japan

Abstract

   The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

Y.Oya^a, M.Kobayashi^a, R.Kurata^a, N. Yoshida^b, N. Ashikawa^c, A. Sagara^c, M.Hara^d, Y.Hatano^d, K.Okuno^a

^aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Oya, Suruga-ku, Shizuoka 422-8529, Japan
^bInstitute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^cNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi, Gifu 509-5292, Japan
^dHydrogen Isotope Research Center, University of Toyama, 3190, Gofuku, Toyama 930-8555, Japan

Abstract

    Behaviors of hydrogen isotope retention and damages in tungsten and SS-316 with simultaneous C⁺-D₂⁺implantation were compared to those with only D₂⁺ implantation using X-ray photoelectron spectroscopy (XPS), Thermal desorption spectroscopy (TDS), glow discharge-optical emission spectroscopy (GD-OES) and Transmission electron microscopy (TEM).
   The total D retention for SS-316 with only D₂⁺ implantation was about 45% as large as that for tungsten. The D retention for simultaneous C⁺-D₂⁺ implanted tungsten and SS-316 clearly increased as a factor of 1.7, which is almost the same among these samples. The density of dislocation loops was enhanced by the simultaneous C⁺-D₂⁺ implantation, indicating the D trapping site would be produced by C⁺ implantation. As for the D desorption temperature, small shift toward lower temperature side was found for SS-316 compared to tungsten, indicating the D trapping energy by dislocation loops and grain boundary for SS-316 is lower than that for tungsten.

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

T. Yamanishi^a, T. Hayashi^a, Y. Iwai^a, K. Isobe^a, M. Hara^b, T. Sugiyama^c, K. Okuno^d

^aJapan Atomic Energy Agency, Tokai, Ibaraki 319-1195 Japan
^bUniversity of Toyama, Gofuku, Toyama 930-8555, Japan
^cNagoya University, Chikusa-ku, Nagoya 464-8603, Japan
^dShizuoka University, Ooya, Shizuoka 422-8529, Japan

Abstract

    The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. It was found that the permeation rate of tritium was almost the same as that with no water vapor, and the major chemical form of tritium was HTO. It was also found that the permeation rate was decreased to ~1/10 of the above value by plated with gold, and the major chemical form was HT. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A set of data for an advanced chemical exchange column has been obtained as R&D on the tritiated water processing. The packing and catalyst were uniformly mixed into the advanced column. It has been proved that there was possibility that the advanced column has a large separation factor in comparison with the ordinary columns, where the packing and catalyst were alternately packed. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column. The ion conductivity of the main material of the cell (Nafion) was hardly decreased even in the case of 1600 kGy irradiation by gamma ray.

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

K. Kobayashi^a, T. Enokida^a, D. Iio^a, Y. Yamada^a, M. Hara^b, Y. Hatano^b

^aDepartment of Physics, University of Toyama, 3190 Gofuku, Toyama-city, Toyama, Japan, 930-8555
^bHydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama-city, Toyama, Japan, 930-8555

Abstract

    It is quite important to control and observe the concentration and total amount of tritium when nuclear fusion is utilized like ITER project. There are many kinds of molecular species, conditions, amount, and concentration in the environment and it is desirable to have multiple ways of observation. Tritium is often found as hydrogen, water and methane molecules. Their behavior differs by the molecular species and detection of molecular species is therefore important. Near-infrared spectroscopy can be a good molecular species sensitive method for this purpose. However, since basic spectroscopic information of tritiated water (HTO or T₂O) is unavailable, in this study we prepared tritiated water of high concentration, and carried out frequency modulation near-infrared spectroscopy.
   The tritiated water was synthesized by the oxidation reaction of 1 Ci of T₂. Near-infrared spectra at 1.3 micron were recorded. Many lines were observed which are not due to normal water. They are strong candidates of tritiated water spectral lines.

Y. Hatano¹, M. Hara,¹ H. Ohuchi,² H. Nakamura³, T. Hayashi,³ T. Yamanishi³

¹Hydrogen Isotope Research Center, University of Toyama: Gofuku 3190, Toyama 930-8555, Japan
²Graduate School of Pharmaceutical Sciences, Tohoku University, Aramaki-Aoba, Aoba-ku, Sendai 980-8578, Japan
³Tritium Technology Group, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195, Japan

Abstract

    Concentration of tritium in highly tritiated water was measured by exposing imaging plates (IPs) to water vapor. Tritium penetrated into photostimulated luminescence (PSL) phosphor through polyethylene terephthalate protection layer, and well detectable signal of PSL was induced at tritium concentration of 16 kBq cm^-3. In addition, tritium was reversibly desorbed by keeping IPs in air, and signal from IPs returned to background level. In other words, IPs exposed to tritiated water vapor were reusable; tritium concentration in water could be measured without any waste. In addition, no handling of tritiated water such as sampling and dilution was necessary.

Y. Kawamura^a, W. Shu^b, M. Matsuyama^C, T. Yamanishi^a

^aJapan Atomic Energy Agency: Shirakata Shirane 2-4, Tokai, Ibaraki 319-1195, Japan
^bITER organization, 13067 Saint-Paul-Lez-Durance Cedex, France
^cUniversity of Toyama, Gofuku 3190, Toyama-city, Toyama 930-8555, Japan

Abstract

    Beta ray induced X-ray spectrometry (BIXS) is one of the methods applicable to tritium gas monitor. It can measure tritium by counting the X-ray that is induced by interaction between the beta ray of tritium and the materials. Tritium gas monitor of BIXS use installed into Tritium Process Laboratory (TPL) in Japan Atomic Energy Agency (JAEA) uses NaI(Tl) as the scintillator. In this work, the NaI scintillator and the photo-multiplier that can work at 150℃ have been installed instead of the ordinary scintillator and photo-multiplier. And, the sample gas such as He, T₂, or T₂ (1%)/He mixture was introduced into the tritium gas monitor kept at 120℃. Then, the counting rate was observed. The counting rate at 120℃ was about a half of that at the room temperature. The counting rate after the heating was almost same with that before the heating. So, the deterioration of the scintillator by the heating has not been observed.

R.-D. Penzhorn¹, Y. Torikai¹, M. Saito¹, M. Hiro¹, A. Perevezentsev², M. Matsuyama¹

¹Hydrogen Isotope Research Center, University of Toyama, 930 8555 Toyama, Gofuku 3190, Japan
²ITER Organization, St. Paul-les-Durance, France

Abstract

    The uptake of tritium on the surface and in the bulk of copper upon exposure to a 50 ℃ T/H mixture at 300 or 473 K was investigated using a chemical etching technique. Concentrations of tritium approaching saturation are achieved fairly rapidly in Cu even at low temperatures because of comparatively high diffusivity and low solubility of hydrogen in this material. The results were interpreted by a diffusion model. Most notorious are the very high concentrations of tritium on the topmost surface and subsurface. They were quantified by etching and confirmed by BIXS. In addition, there is evidence for tritium trapping in the subsurface region.
   Tritium-loaded copper specimens release tritium chronically at ambient temperature. The egress of tritium manifests in the gas phase almost exclusively as tritiated water.

Y. Torikai^a, M. Saito^a, A. Taguchi^a, R.-D. Penzhorn^a, K. Akaishi^a, K. Tatenuma^b, K. Isobe^c, T. Hayashi^c, T. Yamanishi^c

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bKAKEN, Hori 1044, Mito, Ibaraki 310-0903, Japan
^cTritium Technology group, Japan Atomic Energy Agency, Tokai-mura Naka-gun, Ibaraki, 319-1195, Japan

Abstract

    A batch process concept for the decontamination from tritium of fusion reactor materials based on a hydrothermal treatment is under development at HRC. Essentially, tritium-loaded material is heated in a tightly closed vessel containing a defined amount of water. The objective of the water is to "capture" the released tritium in a small volume of liquid. For the detritiation, stainless steel temperatures in the range 393-473 K over a period of several days were found to be adequate. From the results it appears that by and large the released tritium accumulates in the purposely introduced water. The achieved degree of decontamination was estimated from the tritium concentration in the water and the tritium that remained in the decontaminated material. Tritium trapped in the surface layer of stainless steel was not reduced by the isochoric hydrothermal treatment in the same proportion as that in the bulk.

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

^aDept. Advanced Energy Engineering Science, Kyushu University, Fukuoka 812-8581, Japan
^b Tritium 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 transfer in porous concrete materials for the tertiary tritium safety containment is performed to investigate; (i) how fast tritium is transferred through porous concrete walls coated with or without a hydrophobic paint, and (ii) how well the hydrophobic paint coating works as a film protecting against tritium migrating through concrete. The experiment is comparatively carried out using two types of cement-paste and mortar disks with or without two kinds of paints. The results obtained here are summarized as follows: (1) Tritium transfer can be correlated in terms of the effective tritium diffusivity of DT=1.2x10^-11 m²/s in porous cement. (2) Adsorbed or condensed liquid HTO itself is transferred only through pores in cement, and no tritium transfer path is present in non-porous sand. (3) Rates of tritium sorption and dissolution in cement and mortar coated with an epoxy-resin paint is correlated in terms of the diffusivity through the paint film of DT=1.0x10^-16 m²/s. (4) The epoxy paint works more effectively as an anti-tritium diffusion coating than the acrylic-silicon resin paint. (5) The hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H₂O.

N. Nunomura¹, S. Sunada², K. Watanabe³

¹Information Technology Center Universiy of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
²Graduate School of Science and Engineering for Research, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
³Hydogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan

Abstract

    Adsorption of H₂O on the α-Al₂O₃ (0001) surface was studied by means of a first-principles calculation based on density functional theory (DFT). We also investigated the behavior of the isotope exchange by substituting a protium atom with deuterium or tritium. The oxygen atom of H₂O adsorbs on the Al atom of the outermost surface layer, the entire water molecule is slanted at the direction of a hollow site, and a molecular plane is nearly parallel to the surface. The adsorbed states are mostly due to coupling of lone-pair electrons of H₂O with the empty p orbitals of the Al atom of surface. The behavior of dissociation for H₂O is clarified from molecular dynamics simulations, indicating that the second neighbor oxygen atom is more preferable adsorption site for dissociation than the nearest neighbor oxygen atom on the surface.

A. Taguchi^a, R. Akai^a, M, Saito^a, Y. Torikai^a, M. Matsuyama^a, M. Ogura^b, S. Uchida^c

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bInstitute of Industrial Science, The Universiy of Tokyo, Komaba 4-6-1, Meguro-ku, Tokyo 153-8505, Japan
^cGraduate School of Art and Science, The University of Tokyo, Komaba 3-811, Meguro-ku, Tokyo 153-8902, Japan

Abstract

    The ability of various solid adsorbents to adsorb tritium from tritiated water was studied. The tritium removability and adsorption ability of mesoporous silica (MCM-41) were found to be larger than those of conventional microporous zeolites such as mordenite (MOR) and Linde-type A (LTA). The different adsorbents can be arranged in order of tritium removability and tritium adsorption ability as follows: MCM-41 > LTA(5A) > high-silica MOR [approximately equal] low-silica MOR [approximately equal] LTA(4A). The adsorbents can also be arranged in decreasing order of the separation factor (α) as follows: MCM-41 > LTA(5A) > low-silica MOR [approximately equal] LTA(4A) > high-silica MOR.

T. Oda^a, M. Shimada^b, K. Zhang^c, P. Calderoni^b, Y. Oya^d, M .Sokolov^e, R. Kolasinski^f, J. P. Sharpe^b,Y. Hatano^c

^aDepartment of Nuclear Engineering and Management, The University of Tokyo, Tokyo, 113-8656, JAPAN
^bFusion Safety Program, Idaho National Laboratory, Idaho Falls, ID, 83415, USA
^cHydrogen Isotope Research Center, University of Toyama, Toyama, 930-8555, Japan
^dRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, 422-8529, JAPAN
^eOak Ridge National Laboratories, Oak Ridge, TN, 37831, USA
^fHydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, C4, 94551, USA

Abstract

    The behavior of hydrogen isotopes implanted into tungsten containing vacancies was simulated using a Monte Carlo technique. The correlations between the distribution of implanted deuterium and fluence, trap density and trap distribution were evaluated. Throughout the present study, qualitatively understandable results were obtained. In order to improve the precision of the model and obtain quantitatively reliable results, it is necessary to deal with the following subjects: (1) how to balance long-time irradiation processes with a rapid diffusion process, (2) how to prevent unrealistic accumulation of hydrogen, and (3) how to model the release of hydrogen forcibly loaded into a region where hydrogen densely exist already.

M. Saito, Y. Torikai, R.-D. Penzhorn, K. Akaishi, M. Matsuyama

Hydogen Isotope Research Center, University of Toyama, 930 8555 Toyama, Gofuku 3190, Japan

Abstract

    Uptake, distribution, and release behavior of tritium in Ni was investigated by chemical etching and thermal release rate measurements. Liberated tritium was found to consist almost exclusively of tritiated water. The chronic release rate of tritium from Ni was significantly larger than that from type 316 stainless steel. Depth profiles in specimens that partially lost tritium due to its chronic release into vacuum, air or a stream of argon could be reproduced by a one-dimensional diffusion model using best fit diffusion coefficients. Values of the best-fit diffusion coefficients at 298 K were found to be independent from the ambient into which tritium was released. The average diffusion coefficient from all measurements at 298 K, i.e. (2.7 ± 1.3) × 10^-10 [cm²/s] was in line with diffusion coefficients calculated from literature data at the same temperature. Hence, the diffusion model constitutes a useful tool for the prediction of tritium bulk depth profiles in Ni during chronic release (CR).

Y. Togashi, M. Hara

Hydogen Isotope Research Center, Universy of Toyama: Gofuku 3190, Toyama, 930-8555, Japan

Abstract

    To understand the water vapor transport through a polypropylene film at near ambient temperature, water vapor permeation and sorption measurements were carried out using tritiated water as tracer. The activation energy and frequency factor of the permeability were found to be 11 kJ/mol and 1.5 x 10^-10 cm³(STP) cm cm^-2 s^-1 Pa^-1, respectively. The corresponding values of the solubility were determined to be -30 kJ/mol and 2.9 x 10^-10 cm³(STP) cm^-3 Pa^-1. Because the permeation can be described by a one-dimensional diffusion model, the diffusion coefficient was evaluated from the quotient of permeability and solubility. The activation energy of water diffusion through polypropylene was calculated to be 41 kJ/mol.

M. Matsuyama^a, K. Shinmura^a, Z. Chen^b, Y. Torikai^a

^aHydogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bSchool of Nuclear Science and Technology, University of Science and Technology of China Hefei, Anhui Province 230026, P. R. China

Abstract

    Solubility of tritium in Cu-Be(2 mass%) alloy was determined by means of measurement of a tritium depth profile in the alloy. Tritium exposure to the samples was conducted under the following conditions: pressure, 0.4 to 2.6 kPa; temperature, 350 to 450℃ exposure time, 4 to 11 hours. Tritium depth profiles were obtained by chemical etching after the exposure. Remarkably high tritium concentration appeared in surface layers within 0.5 m, whereas almost constant concentration was observed from 10 m to the bulk. It was found, therefore, that surface tritium should be omitted in evaluation of the solubility of tritium. In addition, it was seen that dissolution of tritium into Cu-Be alloy obeys the Sieverts' law from the pressure dependence, and the solubility of tritium in Cu-Be alloy was lower than that in pure copper. From the temperature dependence of solubility, the heat of solution of tritium was determined as 17 kJ/mol.

Y. Ozeki, Y. Hatano, H. Taniguchi, M. Matsuyama

Hydogen Isotope Research Center, University of Toyama: Gofuku 3190, Toyama 930-8555, Japan

Abstract

K. Nishimura^a, K. Takatsuka^b, M. Matsuyama^b, N. Noda^b, M. Tanaka^a

^aNational Institute for Fusion Science, 322-6 0roshi-cho, Toki, 509-5292, Japan
^bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama, 930-8555, Japan

Abstract

    Hydrogen isotope retention and its removal in/from a plasma-facing wall and structural materials have been recognized as key issues for fusion reactor from the viewpoints of operational safety and environmental preservation. Baking and/or a glow discharge are possible effective methods to remove hydrogen isotopes from materials. To investigate the absorption and desorption characteristics of hydrogen isotopes in structural materials, a glow discharge apparatus with twin chambers (Glow-1, 2) made with stainless steel was provided. The sample tip can be moved between the Glow-1 chamber and Glow-2 chamber without exposing it to the air. The controlled infrared (IR) heating device was installed in the Glow-1 chamber to heat the sample tip. Results of absorption and desorption by glow discharge, and the thermal desorption by IR heating are analyzed. No oxides of hydrogen isotopes were observed in these experiments. Most gases are retained near the solid surface and are desorbed at a low temperature below 150℃ The activation energy was estimated to be about 50 kJ/mol.

S. K. Sharma¹, H. Zushi², I. Takagi³, Y. Hisano¹, T. Shikama³, S. Morita⁴, T. Tanabe¹, N. Yoshida², M. Sakamoto², Y. Higashizono², K. Hanada², M. Hasegawa², O. Mitarai⁵, K. Nakamura², H. Idei², K. N. Sato², S. Kawasaki², H. Nakashima², A. Higashijima², Y. Nakashima⁶, N. Nishino³, Y. Hatano⁸, A. Sagara⁴, Y. Nakamura⁴, N. Ashikawa⁴, T. Maekawa³, Y. Kishimoto³, Y.Takase⁹, QUEST²

¹IGSES, Kyushu University, Kasuga, Fukuoka, 816-8580, Japan
²RIAM, Kyushu University, Kasuga, Fukuoka, 816-8580, Japan
　　　　　 ³DNE, Graduate School of Engineering, Kyoto University, Japan
⁴National lnstitute for Fusion Science, Toki, Japan
⁵Kyushu Tokai University, 9-1-1 Toroku, Kumamoto 862-8652, Japan
⁶Plasma Research Center, University of Tsukuba, Japan
⁷DMSE, Graduate School of Engineering, Hiroshima University, Japan
⁸Hydrogen Isotope Research Center, Toyama University, Toyama 930-8555, Japan
⁹Graduate School of Frontier Science, University of Tokyo, Ibaragi, Japan

Abstract

    A permeation measuring system with a nickel membrane of 30 m thickness was installed near the mid plane of the spherical tokamak, QUEST. Hydrogen permeation through the membrane heated at fix temperatures (422 - 506 K) was measured during short pulse (< 1 s) and long pulse (1 hour) plasma discharges. After the membrane was heated to a required temperature, hydrogen plasma was discharged using a 2.45 GHz or 8.2 GHz RF system. Significant plasma-driven permeation was observed even for very short plasma discharges (e.g. 0.1s). Numerical calculations with the use of diffusion equation under recombination boundary conditions were conducted to simulate the transient permeation behavior. The numerical calculations were also used to estimate diffusion coefficient and recombination coefficients of membrane material. Temperature dependence of both the coefficients was explained by the Arrhenius law. A one hour long permeation curve was also numerically reproduced using the same set of parameters except an increasing recombination coefficient on the plasma side of the membrane.

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

^aLaboratoty 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 Mukouyama, Naka-shi, Ibaraki, 311-0193 Japan

Abstract

    Tritium retention in plasma facing materials is a primary issue for ITER and next step fusion devices, since it greatly affects its safety and operational schedule. In the ITER, carbon and tungsten are used as divertor materials. In the present study, co-deposited carbon film, tungsten and isotropic graphite were exposed to tritium gas, and then the amount of absorbed tritium was investigated. During the tritium exposure, the partial pressure of tritium gas was kept at 10 Pa. The sample temperature was kept a constant in the range from RT to 573 K. The amounts of absorbed tritium were evaluated by -ray-induced X-ray spectrometry (BIXS). The amounts of absorbed tritium in co-deposited carbon films were one or two orders of magnitude larger than that of polycrystalline tungsten and isotropic graphite. The amount of absorbed tritium for co-deposited carbon film with a high volume density (1.53 g/cm³) was several times larger than that of the film with a low volume density (1.13 g/cm³). The amount of absorbed tritium increased with the temperature. These results indicate that co-deposited carbon films can absorb much larger amount of tritium than tungsten and graphite, and carbon film density affects the amount of absorbed tritium.

M. Shimada^a, Y. Hatano^b, P. Calderoni^a, T. Oda^c, Y. Oya^d, M. Sokolov^e, K. Zhang^b, G. Cao^f, R. Kolasinski^g, J.P. Sharpe^a

^aFusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
^bHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^cDepartment of Nuclear Engineering and Management, The University of Tokyo, Tokyo 113-8656, Japan
^dRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
^eOak Ridge National Laboratory, Oak Ridge, TN 37831, USA
^fDepartment of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706, USA
^gHydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, CA 94551, USA

Abstract

    With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 10²¹ m^-2 s^-1, ion fluence: 4 × 10²⁵ m^-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

N. Nunomura¹, S. Sunada², K. Watanabe³

I. Takagi^a, S.K. Sharma^b, H. Zushi^c, R. Imade^a, T. Komura^a, Y. Hisano^b, Y. Hatano^d, Y. Nakamura^e, A. Sagara^e, N. Ashikawa^e n

Abstract

    A permeation measuring system with a nickel membrane of 30 μm thickness was installed in a spherical tokamak QUEST. The membrane was located near the mid plane of the tokamak so that one side of the membrane was faced to the plasma. After the membrane was heated to 502 K, hydrogen plasma was discharged using 2.45 GHz RF system. The permeation flux through the membrane increased with a proper time-lag, that is, a significant plasma-driven permeation was observed. A numerical calculation of the diffusion equation under recombination boundary conditions was successfully conducted to simulate the transient permeation behavior. The recombination coefficients, estimated in a temperature range of 412- 575 K, can be explained by a model of thermally activated processes. Stable operation of the permeation system indicates the suitability of this system for the measurements of atomic flux with known diffusivity and recombination coefficients.

Y. Oya^a, M. Kobayashi^a, R. Kurata^a, W. Wang^a, N. Ashikawa^b, A. Sagara^b, N. Yoshida^c, Y. Hatano^d, K. Okuno^a

^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
^dHydrogen Isotope Research Center, University of Toyama, 3190, Gofuku, Toyama 930-8555, Japan

Abstract

    To elucidate the simultaneous implantation effects on deuterium retention in polycrystalline tungsten, simultaneous 10 keV C⁺, 3 keV D₂⁺ and 3 keV He⁺ implantation was performed at room temperature. He⁺/D⁺ flux ratio dependence of deuterium retention indicated that He⁺ implantation induced high deuterium retention even if the He₂⁺ flux ratio was low. D₂ TDS spectrum for C⁺-D₂⁺ implanted tungsten was clearly different from the samples with only D₂⁺, D₂⁺-He⁺ and C⁺-D₂⁺- He⁺ implantation, indicating that D was trapped by C as C-D bond. For D₂⁺ -He⁺ implantation sample, He⁺ implantation would introduce irradiation damages and D retention increased. However, C⁺- D₂⁺-He⁺ implantation, the accumulation of C on tungsten was suppressed and the retention of D trapped by C clearly decreased. These facts indicated the D retention for C⁺-D₂⁺-He⁺ implanted tungsten was limited by He⁺ implantation and most of D would be retained in the interstitial sites or irradiation defects under simultaneous implantation with He.

T. Otsuka^a, M. Shimada^b, R. Kolasinski^c, P. Calderoni^b, J.P. Sharpe^b, Y. Ueda^d, Y. Hatano^e, T. Tanabe^a

^aInterdisciplinary Graduate School of Engineering and Science, Kyushu University, Japan
^bFusion Safety Program, Idaho National Laboratory, USA
^cHydrogen and Metallurgical Science Department, Sandia National Laboratories, USA
^dGraduate School of Engineering, Osaka University, Japan
^eHydrogen Isotope Research Center, University of Toyama, Japan

Abstract

    We have applied a tritium imaging plate technique to measure the tritium distribution profile on surface and in bulk of various metal materials after exposure to a deuterium-tritium plasma in a linear plasma experimental apparatus. The experimental tritium concentration profiles in mm range are interpreted according to a simple hydrogen diffusion model in each metal. We found that a significant amount of tritium is localized in near surface regions and is clearly distinguishable from tritium diffused in the bulk. The amount of surface tritium is not likely correlated to bulk properties (diffusivity and solubility), but is related to trapping in surface defects or metal impurities such as oxide and carbide.

W. Wang,^a,e M. Kobayashi^a, R. Kurata^a, S. Suzuki^a, N. Ashikawa^b, A. Sagara^b N. Yoshida^c, Y. Hatano^d, G.N. Luo^e, 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, Japan
^cResearch Institute for Applied Mechanics, Kyushu University, Kyushu, Japan
^dHydrogen Isotope Research Center, University of Toyama, Toyama, Japan
^eInstitute of Plasma Physics, Chinese Academy of Sciences, Hefei, China

Abstract

    The simultaneous implantation of 10 keV C⁺ and 3 keV D into tungsten was carried out at elevated temperatures to establish the retention mechanism of energetic deuterium and carbon. Thermal desorption spectroscopy showed that the deuterium retention obviously decreased as the implantation temperature increased. All the desorption stages disappeared when the implantation was performed at 673 K, indicating no deuterium trapping as C-D bonds was processed in this temperature. The results of X-ray photoelectron spectroscopy and glow discharge-optical emission spectroscopy showed that a mixed carbon layer had been formed during the implantation, resulting from the carbon deposition and accumulation near the surface of tungsten during the implantation. The carbon layer would enhance the chemical sputtering with deuterium and reduce the deuterium retention, as the implantation temperature increases. During discussion, a simple retention mechanism has been proposed, which shows the importance of implantation temperature.

K. Zhang, Y. Hatano

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

Abstract

>    An electrolytic deposition technique was applied to seal open pores in sol-gel derived ZrO₂ coating and to improve barrier effects against permeation of hydrogen isotopes. Disk-type specimens of type 430 ferritic stainless steel were first covered by thin ZrO₂ films (50 nm) with a conventional sol-gel technique. Then, pores in the ZrO₂ film was sealed with ZrO₂ or Al₂O₃ by cathodic processes in ethanol solution of Zr or Al nitrate and subsequent heat treatments in air. The permeation rate of hydrogen was measured at 300-600 ℃. The sol-gel derived ZrO₂ coatings showed only limited barrier effects; the permeation reduction factor (PRF) was about 6-800. Nevertheless, the treatments by electrolytic deposition technique resulted in considerable improvement in the barrier effects, especially at high temperature region (›500℃), and the PRF increased to 100-1000.

Y. Ueda^a, K. Miyata^a, Y. Ohtsuka^a, H.T. Lee^a, M. Fukumoto^b, S. Brezinsek^c, J.W. Coenen^c, A. Kreter^c, A. Litnovsky^c, V. Philipps^c, B. Schweer^c, G. Sergienk^c, T. Hirai^d, A. Taguchi^e, Y. Torikai^e, K. Sugiyama^f, T. Tanabe^g, S. Kajita^h, N. Ohnoⁱ, The TEXTOR team

^aGraduate School of Engineering, Osaka University, Suita, Osaka 565-0871, Japan
^bJapan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193, Japan
^cInstitute für Energieforschung, Forschungszentrum Julich, Association EURATOM-FZJ, Germany
^dITER Organization, 13067 St. Paul-lez-Durance, France
^eHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^fMax–Planck–Institut für Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
^gInterdisciplinary Graduate School of Engineering Science, Kyushu University, Fukuoka, Japan
^hEcoTopia Science Institute, Nagoya University, Nagoya 464-8603, Japan
ⁱGraduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan

Abstract

   W nano–structures (fuzz), produced in the linear high plasma device, NAGDIS, were exposed to TEXTOR edge plasmas (ohmic He/D mixed plasma and pure D plasma) to study formation, erosion and C deposition on W fuzz in tokamak plasmas for the first time. Fuzz layers were either completely eroded or covered by C deposit. There was no clear indication of W fuzz growth under the present conditions. There was no significant difference of C deposition between 'thick' fuzz (500–600 nm in thickness) and 'thin' fuzz (300–400 nm) in the He/D plasma. On the W fuzz surface, C deposition was enhanced probably due to reduction of effective sputtering yield and effective reflection coefficient of carbon ions, similar to roughness effects. Formation and erosion of W fuzz in tokamak devices and role of impurities are discussed.

V. N. Alimov^a, Y. Hatano^b, A. O. Busnyuk^a, D. A. Livshits^c, M.E. Notkin^a, A. I. Livshits^a

^aBonch-Bruevich University, 61 Moika, St. Petersburg 191186, Russia
^bHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
^cInnolume GmbH, Konrad-Adenauer-Allee 11, 44263 Dortmund, Germany

Abstract

    The composite membranes based on Group 5 metals are capable of H₂ separation with the high speed and infinite selectivity. The chemical and thermal stability are critical issues for the application of such membranes in the field of hydrogen energy. In order to understand the degradation mechanisms, the H₂ permeation through composite Pd_{2 μm}-Nb_100μm-Pd_2μm membranes was investigated in a very wide pressure range: (10^-5-10⁴) Pa. At higher pressures the surface contaminations only moderately decreased the permeation. However the permeation experiments at lower pressures demonstrated that an orders of magnitude change in the probability of H2 molecule dissociative sticking is actually hidden behind this relatively moderate effect. The membranes poisoned by the surface contaminations could be recovered by their exposure to O₂ at (300-400) ℃. Heating at temperature higher than 500 ℃ resulted in the irreversible decrease of permeation and in the pronounced change of permeation behavior at the variation of H₂ pressure. An extremely high permeation was observed at lower pressures at the clean surface of Pd coating. That allows developing an effective membrane pump for hydrogen isotopes.

Keywords: Hydrogen separation, Composite membranes, Niobium, Palladium

H. Katsui¹, A. Hasegawa², Y. Katoh³, Y. Hatano,⁴ T. Tanaka⁵, S. Nogami², T. Hinoki², T. Shikama¹

¹Institute for Materials Research, Tohoku University, Sendai, Japan
²Department of Quantum Science and Energy Engineering, Tohoku University Sendai, Japan
³Malerials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN
⁴Hydrogen lsolope Research Center, University of Toyama, Toyama, Japan
⁵Department of Helical Plasma Research, National Institute for Fusion Science, Toki, Japan
⁶Institute of Advanced Energy, Kyoto University, Kyoto, Japan

Abstract

    Compatibility of monolithic silicon carbide(SiC) with ternary lithium ceramics (Li_1-x AlO_2-y, Li_2-xTiO_3-y, Li_2-xZrO_3-y and Li_4-xSiO_4-y) under irradiation of neutros at high temperatures was studied. Disk samples of SiC in contact with sintered ternary lithium ceramics were irradiated in High Flux Isotope Reactor (HFIR) at 800 ℃ to 5.9 displacements per atom(dpa). Chemical reactions of SiC as determined by appearance of the surface were relatively less significant for the systems of SiC/Li_1-xAlO_2-y and SiC/Li_2-x TiO_3-y, whereas some bonding likely due to chemical reaction between SiC and the lithium ceramics and broken samples were observed in the systems of SiC/Li_2-xZrO_3-y and SiC/Li_4-xSiO_4-y. The effect of lithium burnup due to the (n, α) nuclear reaction was also examined by using samples of lithium ceramics whose lithium ratio was hypo-stoichiometric in the fabrication process. More reaction products were observed on the surface of β-SiC in contact with Li_1-xAlO_2-y having the lower lithium ratio (Li/Al). It was considered that the formation of LiAl₅Ｏ₈ phase due to lithim loss could deteriorate the compatibility of the SiC -Li_1-x AlO_2-y system.

阿部孝之、井上光浩

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 なし

Keywords : Polygonal Barrel-Sputtering Method, Dry Process, Surface Modification, Zero-, One-, Three-Dimensional Materials, Functional Materials

Y. Ueda¹, H. T. Lee¹, N. Ohno², S. Kajita³, A. Kimura⁴, R. Kasada⁴, T. Nagasaka⁵, Y. Hatano⁶, A. Hasegawa⁷, H. Kurishita⁸, Y.Oya⁹

¹Graduate School of Engineering, Osaka University, Osaka 565-0871, Japan
²Graduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan
³EcoTopia Science Institute, Nagoya University, Nagoya 464-8603, Japan
⁴Institute of Advanced Energy, Kyoto University, Kyoto 611-0011, Japan
⁵National Institute for Fusion Science, Gifu 509-5292, Japan
⁶Hydrogen Isotope Research Center, Toyama University, Toyama 930-8555, Japan
⁷Department of Quantum Science and Energy Engineering, Tohoku University, Sendai 980-8579, Japan
⁸Institute for Materials Research (IMR), Tohoku University, Ibaraki 311-1313, Japan
⁹Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan

Abstract

    The status of ongoing research projects of tungsten R&D in Japan is summarized in this paper. For tungsten material development, a new improved fabrication technique, the so-called superplasticity-based microstructural modification, is described. This technique successfully improved fracture strength and ductility at room temperature. Recent results on vacuum plasma spray W coating and W brazing on ferritic steels and vanadium alloys are explained. Feasibility of these techniques for the manufacture of the blanket is successfully demonstrated. The latest findings on the effect of neutron damage in tungsten on T retention and on the change in mechanical and electrical properties are described. Retention characteristics for neutron-damaged W were different compared to those for ion-damaged W. Upon neutron irradiation, tungsten alloys containing transmutation elements of W (Re and Os) show changes in properties that are different compared with those shown by pure W. The effects of mixed plasma exposure (D/He/C) are described. Both D/He and D/C mixed ion irradiations significantly affect ion-driven permeation in W. He bubble dynamics play a key role in nano-structure formation on the W surface.

V. Kh. Alimov^1,2, B. Tyburska-Püschel³, M. H. J't.Hoen⁴, J. Roth³, Y. Hatano¹, K. Isobe², M. Matsuyama¹, T. Yamanishi²

¹Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
²Tritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan
³Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
⁴Institute for Plasma Physics Rijnhuizen, EURATOM-FOM, NL-3439 MN Nieuwegein, The Netherlands

Abstract

    Hydrogen isotope exchange in tungsten was investigated at various temperatures both after sequential exposure to low-energy deuterium (D) and protium (H) plasmas and after sequential irradiation with low-energy D and H ions. The methods used were thermal desorption spectroscopy, and the D (³He, p) ⁴He nuclear reaction at ³ He energies varied from 0.69 to 4.0MeV, allowing the determination of the D concentration at depths of up to 6 μm. It was found that a major portion of the deuterium initially accumulated in the D-implanted W is released on subsequent exposure to H plasma or irradiation with H ions. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface near room temperatures, but extending to all analyzable depths at elevated temperatures.

Y Hatano¹, K Zhang¹, K Hashizume²

¹Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
²Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Fukuoka 812-8581, Japan

Abstract

    Layers of tetragonal ZrO₂ (180 nm) were prepared on the surfaces of ferritic steel specimens by dip coating and electrolytic deposition techniques. These methods were selected because of their applicability to large and complicated structures. Hydrogen permeation tests were carried out at 300-550 ℃ and the results were compared with those obtained in the previous study for thinner ZrO₂ coating (100 nm). Although the thickness of the present coatings was less than twice that of the previous coatings, the permeation reduction factor for the former was larger by an order of magnitude than that for the latter at/below 400 ℃. Tritium retention in the coatings was also measured after exposure to a deuterium-tritium mixture at 300 ℃. The obtained results suggest that the permeation rate was determined by transportation through defects in the coatings, and the density of defects penetrating into the coating/substrate interface was significantly reduced by an increase in coating thickness.

Yasuhisa Oya¹, Masashi Shimada², Makoto Kobayashi¹, Takuji Oda³, Masanori Hara⁴, Hideo Watanabe⁵, Yuji Hatano⁴, Pattrick Calderoni², Kenji Okuno¹

¹Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
²Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
³Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, Tokyo 113-8656, Japan
⁴Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
⁵Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan

Abstract

    The behavior of D retention for Fe²⁺-irradiated tungsten with a damage of 0.025-3 dpa was compared with that for neutron-irradiated tungsten with 0.025 dpa. The D2 thermal desorption spectroscopy (TDS) spectra for Fe²⁺-irradiated tungsten consisted of two desorption stages at 450 and 550 K, while that for neutron-irradiated tungsten was composed of three stages and an addition desorption stage was found at 750 K. The desorption rate of the major desorption stage at 550K increased as the displacement damage increased due to Fe²⁺ irradiation increasing. In addition, the first desorption stage at 450K was found only for damaged samples. Therefore, the second stage would be based on intrinsic defects or vacancy produced by Fe²⁺irradiation, and the first stage should be the accumulation of D in mono-vacancy and the activation energy would be relatively reduced, where the dislocation loop and vacancy is produced. The third one was found only for neutron irradiation, showing the D trapping by a void or vacancy cluster, and the diffusion effect is also contributed to by the high full-width at half-maximum of the TDS spectrum. Therefore, it can be said that the D₂ TDS spectra for Fe²⁺-irradiated tungsten cannot represent that for the neutron-irradiated one, indicating that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten is different from that for the ion-irradiated one.

M. Shimada¹, G. Cao², Y. Hatano³, T. Oda⁴, Y. Oya⁴, M. Hara³, P. Calderoni¹

¹ Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
²Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706, USA
³ Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
⁴Department of Nuclear Engineering and Management, The University of Tokyo,
Tokyo 113-8656, Japan

Abstract

    Tungsten samples (99.99% purity from A.L.M.T. Corp., 6mm in diameter, 0.2mm in thickness) were irradiated by high-flux neutrons at 50 ℃ C to 0.025 dpa in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Subsequently, the neutron-irradiated tungsten samples were exposed to high-flux deuterium plasmas (ion flux: 10²¹-10 ²²m^-2 s^-1, ion fluence: 10²⁵-10²⁶ m^-2) in the Tritium Plasma Experiment at Idaho National Laboratory. This paper reports the results of deuterium depth profiling in neutron-irradiated tungsten exposed to plasmas at 100, 200 and 500 % via nuclear reaction analysis (NRA). The NRA measurements show that a significant amount of deuterium (>0.1 at.% D/W) remains trapped in the bulk material (up to 5μm) at 500 %. Tritium Migration Analysis Program simulation results using the NRA profiles indicate that different trapping mechanisms exist for neutron-irradiated and unirradiated tungsten.

M. Muroga¹, D. K. Sze², K. Okuno³, T. Terai⁴, A. Kimura⁵, R. J. Kurtz⁶, A. Sagara¹, R. Nygren⁷, Y. Ued⁸, R. P. Doerner², J. P. Sharpe⁹, T. Kunugi¹⁰, N. B. Morley¹¹, Y. Hatano¹², M. A. Sokolov¹³, T. Yamamoto¹⁴, A. Hasegawa¹⁵, Y. Katoh¹³, N. Ohno¹⁶, K. Tokunaga¹⁷, S. Konishi⁵, S. Fukada¹⁷, P. Calderoni⁹, T. Yokomine¹⁰, K. Messadek¹¹, Y. Oya³, N. Hashimoto¹⁸, T. Hinoki⁵, H. Hashizume¹⁵, T. Norimatsu⁸, T. Shikama¹⁵, R. E. Stoller¹³, K .A. Tanaka⁸, M. S. Tillack²

¹NIFS, Toki, Japan, ²UCSD, San-Diego, CA, USA, ³Shizuoka University, Shizuoka, Japan, ⁴University of Tokyo, Tokyo, Japan, ⁵Kyoto University, Uji, Japan, ⁶PNNL, Richland, WA, USA, ⁷SNL, Albuquerque, NM, USA, ⁸Osaka University, Suita, Japan, ⁹INL, Idaho Falls, ID, USA, ¹⁰Kyoto University, Kyoto, Japan, ¹¹UCLA, Los Angeles, CA, USA, ¹²Toyama University, Toyama, Japan, ¹³ORNL, Oak Ridge, TN, USA, ¹⁴UCSB Santa-Barbara, CA, USA, ¹⁵Tohoku University, Sendai, Japan, ¹⁶Nagoya University, Nagoya, Japan, ¹⁷Kyushu University, Kasuga, Japan, ¹⁸Hokkaido University, Sapporo, Japan

Abstract

    Japan-US cooperation program TITAN (Tritium, Irradiation and Thermofluid for America and Nippon) started in April 2007 as 6-year project. This is the summary report at the midterm of the project. Historical overview of the Japan-US cooperation programs and direction of the TITAN project in its second half are presented in addition to the technical highlights.

¹Institute for Energy and Climate Research, Forschungszentrum Julich, EURATOM Association,Trilateral Euregio Cluster, Julich, Germany, ²SCK-CEN, Belgian Nuclear Research Centre, Association EURATOM, Partner In the Trilateral Euregio Cluster, Mol, Belgium, ³Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Garching, Germany ⁴Hydrogen Isotope Research Center, University of Toyama, Japan, ⁵International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University, Oarai, Ibaraki, Japan, ⁶Graduate School of Engineering, Osaka University, Japan

Abstract

    The behaviour of tungsten (W) plasma-facing components (PFCs) has been investigated in the plasma edge of the TEXTOR tokamak to study melt-layer ejection, macroscopic tungsten erosion from the melt layer as well as the changes of material properties such as grain-size and abundance of voids or bubbles. The parallel heat flux at the radial position of the exposed tungsten tile in the plasma ranges around q|| ~ 45 MW m^-2 causing samples to be exposed at an impact angle of 35° to 20-30 MW m^-2. Locally the temperature reached up to 6000 K, high levels of evaporation and boiling are causing significant erosion in the form of continuous fine spray or droplet ejection. The amount of fine-spray tungsten emission depends strongly on the material properties: in the case of the tungsten-tantalum alloy the effect of spraying and droplet emission is significantly higher at even low temperatures when compared with regular tungsten or even ultra-high purity tungsten which shows almost no spraying at all. Differences in the material composition, grain structure and size may be related to the different evolution of macroscopic erosion. In addition the re-solidified material is studied and strong differences in terms of re-crystallized grain size and evolution of the grain structure and grain orientation are observed. The build up of large voids has been observed.

H. Ohuchi¹, Y. Hatano²

¹Graduate School of Pharmaceutical Sciences, Tohoku University, 6-3 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi, 980-8578 Japan
²Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, Toyama, 930-8555 Japan

Abstract

    Tritium measurement is indispensable for the fuel-processing systems of deuterium-tritium (DT)-fusion facilities. A new approach to detect tritium in regions deeper than the escape depth of beta-rays from tritium is being developed using an imaging plate (IP). The measurement principle of this approach is to observe bremsstrahlung X-rays induced by the tritium beta-rays. An IP made of europium-doped BaFBr(I), a photostimulated luminescence (PSL) material, is a two-dimensional radiation sensor. In the present study, the characteristics of this IP for measuring tritium by detecting bremsstrahlung X-rays, in particular a fading effect and the energy dependence of PSL sensitivities, are examined.

C. Hiromi, M. Inoue, A. Taguchi, T. Abe

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

Abstract

    The optimum Pt and Ru atomic composition of a carbon-supported Pt-Ru alloy (Pt-Ru/C) used in a practical direct methanol fuel cell (DMFC) anode was investigated. The samples were prepared by the polygonal barrel-sputtering method. Based on the physical properties of the prepared Pt-Ru/C samples, the Pt-Ru alloy was found to be deposited on a carbon support. The microscopic characterization showed that the deposited alloy forms nanoparticles, of which the atomic ratios of Pt and Ru (Pt:Ru ratios) are uniform and are in accordance with the overall Pt:Ru ratios of the samples. The formation of the Pt-Ru alloy is also supported by the electrochemical characterization. Based on these results, methanol oxidation on the Pt-Ru/C samples was measured by cyclic voltammetry and chronoamperometry. The results indicated that the methanol oxidation activities of the prepared samples depended on the Pt:Ru ratios, of which the optimum Pt:Ru ratio is 58:42 at.% at 25 ℃ and 50:50 at.% at 40 and 60 ℃. This temperature dependence of the optimum Pt:Ru ratio is well explained by the relationship between the methanol oxidation reaction process and the temperature, which is reflected in the rate-determining steps considered from the activation energies. It should be noted that at 25-60 ℃, the Pt-Ru/C with Pt:Ru = 50:50 at.% prepared by our sputtering method has the higher methanol oxidation activity than that of a commercially available sample with the identical overall Pt:Ru ratio. Consequently, the polygonal barrel-sputtering method is useful to prepare the practical DMFC anode catalysts with the high methanol oxidation activity.

Key words: Direct methanol fuel cells (DMFCs), Pt-Ru/C, Pt and Ru atomic composition, Methanol oxidation, Polygonal barrel-sputtering method

A. Aoki¹, M. Naruse¹, T. Abe²

¹Materials Science & Engineering, Graduate School of Engineering, Nagoya Institute of Technology, Gokiso, Showa-ku, Nagoya 466-8555, Japan
²Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, Toyama 930-8555, Japan

Abstract

    Series circuit of six organic thin-film solar cells (OSCs) consisting of poly(3-hexylthiophene) and PCBM has been fabricated in order to electrolyze water into hydrogen and oxygen gases. The open circuit voltage increases linearly with the number of unit cells and becomes 2.9 V at the six unit cells. On the other hand, the short circuit current, the fill factor and the energy conversion efficiency are almost constant, independent of the number of unit cells. The series circuit of six OSCs was combined with the water electrolysis cell with two platinum electrodes. Hydrogen and oxygen gases were generated at each platinum electrodes of the electrolysis cell under illumination. The operating current and voltage was determined to be 1.3 mA/cm2 and 2.6 V, respectively.

Key words: Organic solar cells, Hydrogen gas, Energy conversion, Water electrolysis