青木　純

名古屋工業大学大学院工学研究科物質工学専攻
〒455-8555 名古屋市昭和区御器所町つくり領域

Abstractなし

田口　明

富山大学水素同位体科学研究センター
〒930-8555　富山市五福3190

A.Taguchi

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

Abstract

Zeolite, crystalline microporous alminosilicates, is an attractive material owing to its sorption capacity, molecular sieve or ion exchange properties. Thus, there are many publications of zeolite applying to an hydrogen isotope separation. In this review, in the initial part, structural properties and some example of applications of zeolite have briefly described. Then, recent publications of hydrogen isotope separation, including a gas phase and a liquid phase of tritium, by zeolite have been reviewed.

K.Watanabe, Y.Torikai, M.Hara, N.Nunomura^*, R-D. Penzhorn

Hydrogen Isotope Research Center, Universyt of Toyama
^*Information Technology Center
Gofuku 3190, Toyama 930-8555, Japan

Abstract

  Adsorption/desorption processes of water and hydrogen on the Cr₂O₃ (0001) surface were studied by ab initio calculations with Gaussian 03 for small clusters of Cr3O3.
  It was found that water and hydrogen can be adsorbed molecularly with practically no activation energy.　Subsequently the adsorbed molecules are dissociated to form (Cr)OH(a) + (O)H(a) from water and (O)H(a) + (O)H(a) from hydrogen. The process of water adsorption can be described as H₂O(g) + Cr3O3 → (Cr3O3)H₂O(a) =(Cr3O3)H₂O(a)* → (Cr)OH(a)-(O)H(a). The process of hydrogen adsorption is H₂(g) + Cr3O3 → (Cr3O3)H₂(a)=(Cr3)H₂(a)*→(Cr)H-(O)H(a)=(Cr)H-(O)H(a)*= - 2 (O)H(a). With respect to water adsorption, the energies of the adsorbed species were calculated to be -103, -94.8 and -164 kJ/mol for the molecularly adsorbed species, the activated complex and the dissociatively adsorbed species, respectively. Regarding hydrogen adsorption, the energies were -24.7, 38.2, -45.1, 48.0 and -111 kJ/mol for H₂(a), H₂(a)*, (Cr)H-(O)H(a), (Cr)H-(O)H(a)* and (O)H-(O)H(a), respectively. The present results indicate that both water and hydrogen can be dissociatively adsorbed on the Cr₂O₃ surface. However, the dissociative adsorption of water takes place much more easily than that of hydrogen.

赤丸悟士、原　正憲、加藤剣一^A)、野口恒行^A)、中村 和^A)、波多野雄治、松山政夫
富山大学水素同位体科学研究センター
〒930-8555　富山市五福3190
^A)株式会社化研
〒310-0903 茨城県水戸市堀町 1044

S.Akamaru, M.Hara, K.Kato^A), T.Noguchi^A), K.Nakamura^A), Y.Hatano, M.Matsuyama
Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^A)Kaken co. ltd., Horimati 1044, Mito, Ibaraki 310-0903, Japan

Abstract

  The Glove box detritiation system (GBDS) was replaced to new one at October 2009. The oxidation-adsorption method to remove tritium is used for the new system as well as the case for the old system. Several components in the system were modified in order to improve tritium confinement and durability. Performance tests of the new GBDS were carried out, and it was proved that the new system has the ability to reduce tritium concentration in the glove box from 3.7 MBq/cm³to 0.5 Bq/cm³ within 5 hours.

Keyword: tritium detritiation system.

M. Oyaidzu, T. Yamanishi

Tritium Process Laboratory, Japan Atomic Energy Agency, Ibaraki 319-2295, Japan

Abstract

  A study on the corrosion behavior of F82H steel and the effects of tritium on it using the electrochemical techniques of Tafel extrapolation and anodic polarization has just been started and the preliminary results without the use of tritiated water were introduced in the present report. Using Tafel extrapolation, the corrosion rate of F82H steel in 5w% H₂SO₄/H₂O could be estimated as approximately 2 cm year^-1, which was approximately 2 orders of magnitude larger than that of SUS 304. In anodic polarization curve, the regions of passivation, passive layer sustention and transpassivation could be observed under the same circumstance as Tafel extrapolation experiment and the quantum of electricity for F82H steel to passivate was approximately 100 C cm^-2, which was approximately 3 orders of magnitude larger than that of SUS304. Therefore it was expected that the corrosion rate of F82H steel could far higher than that of SUS304, even though only a preliminary experiments under the highly corrosive circumstance has been performed.

Keywords: tritium, radiochemistry, electrochemistry, corrosion

K.Isobe, T.Yamanishi

Tritium Technology group, Japan Atomic Energy Agency, Tokai,Ibaraki,319-1195,Japan

Abstract

　The highly tritiated water is expected to be produced in the vacuum chamber and in a bleeding blanket of fusion reactor system. A ceramic electrolysis would be only a method to recover tritium from such highly tritiated water. Aiming at improving the water decomposition efficiency of this process, we have developed effective electrode containing cerium oxide (ceria). In this study, the effect of ceria concentration in the electrode on the efficiency has been studied. The current density increased with ceria concentration in the electrode and reached the value of high density 100mA/cm² for 30% ceria-containing electrode at 1.5V. This current density is one order of magnitude higher than that of usual (Pt-YSZ) electrode.

Keywords: tritium, highly tritiated water, water detritiation system

T.Nozaki¹⁾, E.Yamakawa²⁾, A.Hachikawa²⁾, K.Ichinose²⁾, M.Hara¹⁾, Y.Hatano¹⁾

1) Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
2) Faculty of Human development, University of Toyama, Toyama 930-8555, Japan

Abstract

　Permeation of hydrogen through Pd-Ta-Pd composite membrane was measured at 500 and 600°C to examine the feasibility of the concept to use this type of membrane as permeation window to recover tritium from liquid Pb-Li alloy. At 500°C, the permeation rate gradually decreased with increasing operation time due to surface contamination by carbon. These observations indicated that partial pressures of residual hydrocarbons in vacuum permeator have to be strictly controlled to avoid carbon deposition. Nevertheless, permeability could be restored by removal of carbon through reaction with O₂ gas. At 600°C, however, permanent degradation took place due to development of open porosity in Pd layer and interdiffusion between Pd and Ta. Preparation of HfN intermediate layer between Pd and Ta hindered the porosity development and interdiffusion.

Keywords: Tritium Recovery, Permeation, Tantalum, Palladium, Fusion Blanket

M.Hara, Y.Togashi

1)Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
2)Faculty of Science, University of Toyama, Toyama 930-8555, Japan

Abstract

   　The permeability of water through the polypropylene film was studied using tritiated water. Experiments were carried out in a flow type vessel designed for the determination of the permeability of water. All permeated water was captured by a water bubbler. The permeability was evaluated from the amount of tritiated water permeated through the polymer. The obtained apparent permeability increased with increasing temperature. The temperature dependence was governed by the water vapor pressure. The permeability of water through polypropylene normalized by water vapor pressure was found to be 1.9×10^-13 g cm cm^-2 s^-1 Torr^-1 at 296 K.

Keywords: tritiated water, permeation, polypropylene

Yasuhisa Oya¹, Sachiko Suzuki¹, Makoto Kobayashi¹, Rie Kurata¹, Shoji Miyamoto², Naoaki Yoshida², Naoko Ashikawa³, Akio Sagara³, Yuji Hatano⁴,Kenji Okuno¹

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

Abstract

　  Deuterium retention in tungsten irradiated simultaneously with deuterium, helium and carbon ions was investigated by TDS. Effect of irradiation pre-damage on the deuterium retention was examined by microstructure observation with taking account of the TDS results. It was found that deuterium was preferentially trapped by intrinsic defects for simultaneous C⁺and D₂⁺implanted tungsten. The deuterium trapping by ion-induced defects was enhanced by the ion implantation.
　In the case of simultaneous implantation with He+, the deuterium retention has largely changed. Especially, in the case of He⁺and D₂⁺simultaneous implantation, the deuterium retention increased compared to the sequential implantation. However, the triple ions (He⁺-D₂⁺-C⁺) implantation, the deuterium retention was almost the same as that for the only D₂⁺implanted tungsten, indicating the dynamic desorption would be enhanced.

Keywords: tritium retention, plasma-surface interactions, tungsten, simultaneous implantation

V.Kh. Alimov¹, S. Lindig², M. Balden², K. Isobe¹, W.M. Shu³, J. Roth², T. Yamanishi¹

¹ Tritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki, 319-1195, Japan
² Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
³ ITER International Organization, CEA Cadarache, 13108 Saint Paul-lez-Durance, France

Abstract

   　Three-dimensional morphology of blister-like structures and deuterium retention in recrystallized tungsten have been examined after exposure to a low-energy (38 eV/D), high-flux (10²² D/m²s) deuterium plasma at ion fluences in the range from 10²⁶ to 10²⁷ D/m² and various temperatures. The methods used were scanning electron microscopy equipped with focused ion beam, thermal desorption spectroscopy, and the D(³He,p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV. Blister-like structures with various shapes and sizes depending on the exposure fluence and temperature are formed on the W surface. In doing so, cracks with distorted areas (<600 K) and large cavities (100-1000 μm³) at grain boundaries (≧500 K) were observed beneath the surface. The surface blister-like structures and the defects underneath are correlated along crystallographic orientation of the W grains in accordance to the low-indexed sliding systems. The defects are mobile and accumulate under D plasma exposure. Samples exposed near room temperature do not form such large cavities even by subsequent heating up to 1300 K. The D plasma exposure at temperatures above 700 K does not lead to formation of blister-like structures. At ion fluences of 10²⁶-10²⁷ D/m², the D retention increases with the exposure temperature, reaching its maximum value of about 10²² D/m² at 480-530 K, and then decreases to about 10¹⁹ D/m² at 800 K.

Keywords: blistering; deuterium; deuterium retention; high ion flux; tungsten

T. Chikada ¹, A. Suzuki ¹, T. Tanaka ², T. Kobayashi ³, H. Maier ⁴, T. Terai ¹, T. Muroga ²

¹ School of Engineering, University of Tokyo, Tokyo 113-8656, Japan
² National Institute for Fusion Science, Gifu 509-5292, Japan
³ Institute of Physical and Chemical Research (RIKEN), Saitama 351-0198, Japan
⁴ Max-Planck-Institut fur Plasmaphysik, D-85748 Garching, Germany

Abstract

   　Development of tritium permeation barrier is considered one of principal investigations to build up a fuel cycle in a fusion power plant. For the practical application as tritium permeation barrier, fabrication of erbium oxide coatings by metal-organic decomposition method has been carried out on reduced activation ferritic/martensitic steels. The coated samples showed various surface states after the heat-treatment process under different conditions. An oxide layer was generated between the erbium oxide coating and the substrate when heat-treated in high purity argon. It is indicated that the oxide layer has caused defects in the coating and resulted in the degradation of the samples during deuterium permeation measurements. The sample heat-treated in high purity hydrogen with moisture shows the thinner oxide layer and stable permeation fluxes during the measurements. A coating of 0.3 μm in thickness indicates a permeation reduction factor of 700-1000 at 773-973 K, which is a comparable level to the coating deposited by physical vapor deposition technique performed in the previous study.

Keywords: tritium, hydrogen, permeation barrier, ceramic coating, erbium oxide

Masao Matsuyama

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

Abstract

A tritium measuring technique is one of key technologies for the precise control and safe handling of fuel particles in the fusion systems. A variety of techniques have been already established so far for measurements of low-level tritium such as a tracer of chemical reactions and environmental tritium. However, high-level and huge amount of tritium is applied to the fusion reactors. Some techniques developed so far are applicable for fusion environment, but new techniques will be also required for establishment of a tritium plant in future fusion reactors. From this viewpoint, application of β-ray-induced X-ray spectrometer (BIXS) for highly tritiated waters is described with the conventional devices, where the BIXS was newly developed by author to measure high-level tritium in-situ. Present state and future subjects of tritium measuring techniques will be discussed in this paper.

Key words: tritium, high concentration, measurement technique, BIXS, fusion reactor