Hydrogen Isotope Research Center -University of Toyama
Sep. 14-15, 2009, Multi-Purpose Hall, Faculty of Science, University of Toyama, Toyama, Japan
[workshop 2009_01]
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
[workshop 2009_02]
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
[workshop 2009_03]
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
[workshop 2009_04]
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
[workshop 2009_05]
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
[workshop 2009_06]
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
[workshop 2009_07]
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
[workshop 2009_08]
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