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

^aHydrogen Isotope Research Center, University of Toyama, Gofuku3190,Toyama 930-8555,Japan
^bITER Organization, St.Paul-lez-Durance, France

Abstract

   Exposure of Type 316 stainless steel to tritium-containing hydrogen at an elevated temperature causes diffusion of the majority into the bulk and trapping of a small fraction in a thin oxide layer on the surface at concentrations far exceeding those in the bulk. The uptake by the bulk and surface layer is temperature and pressure dependent. After chemical erosion of the tritium-rich layer, the concentration of tritium on the topmost surface is slowly and asymptotically restored even at 298 K. Isothermal heating at 373 or 473 K until substantial release of the bulk tritium is associated with a comparatively much smaller liberation from the surface layer suggesting different retention and liberation mechanisms. The tritium inventory and profile evolution of homogeneously loaded Type 316 stainless steel caused by chronic release at the ambient temperature and radioactive decay were followed experimentally over several years and modeled successfully by a diffusion mechanism. The model has been adapted to specimens nonhomogeneously loaded with tritium only up to the subsurface. It simulates profile and inventory changes well even after prolonged aging. Chronic release constitutes an aging loss of tritium comparable to that of radioactive decay that should be taken into account for the assessment of tritium-contaminated stainless steel waste.

Keywords:tritium, stainless steel, aging

松山政夫¹⁾、山西敏彦²⁾

1）富山大学水素同位体科学研究センター

2)日本原子力研究開発枚構核融合研究開発部門トリチウム工学研究グループ

Abstract

   日本国内の大量トリチウム取り級いは,富山大学水素同位体科学研究センター、および日本原子力研究開発機構トリチウムプロセス研究棟(TPL)を中心に進められ、施設運転から約30年が経過した。その間、無事故での安全取り扱い実績を重ね、大量トリチウム取り扱いという課題が、ようやく日本でも特殊なものではなくなったと自負している。本講座では、これまでの上記施設における成果をレビューするとともに、施設建設準備の活動にも焦点をあて、今後の技術継承と、日本における大量トリチウム取り扱い施設の維持という観点から提言したい。

Keywords:Tritium, confinement, handling, fusion

波多野雄治¹⁾、山田正行²⁾、林　巧²⁾

¹⁾富山大学水素同位体科学研究センター

²⁾日本原子力研究開発機構核融合研究開発部門トリチウム工学研究グループ

Abstract

   　現在最も実現性の高い核融合炉は、重水素とトリチウム(三重水素、水素の放射性同位元素)を燃焼させる ものであり、百万kWの発電炉を想定すると約500g程度のトリチウムを毎日燃焼し生産することとなる。核融合 炉はもとより国有の安全性を有しているが、トリチウムの閉じ込め、安全取り扱いが施設全体の安全確保の要で ある。本章では、 1)国内の大学で最大級のトリチウム取り扱い施設である富山大学水素同位体科学研究セン ターと2) 国内唯一のグラムレベルトリチウム取り扱い施設である日本原子力研究開発機構トリチウムプロセス 研究棟におけるトリチウム安全取り扱い実続を整理するとともに、主要な関連研究開発の成果を纏めた。

Keywords: tritium, safety, confinement, contamination, decontamination, detritiation system, tritium transport, failure rate, tritium release behavior

松山政夫、林　巧¹⁾、山西敏彦¹⁾

富山大学水素同位体科学研究センター

1)日本原子力研究開発機構核融合研究開発部門トリチウム工学研究グループ

Abstract

   　富山大学水素同位体科学研究センターおよび原子力機構トリチウムプロセス研究棟(TPL)ともに、 ホット運転から25-30年が経過し、老朽化対策が必要となってきている。その中で、ホット施設の運転・保守の努力を 続けており、蓄積された貴重な経験を示した。一方で、両施設の更新、新たな施設も必要とされており、 富山大の施設の更新状況、六ヶ所に整備を開始したトリチウム取り扱い研究施設の概要、今後原型炉に向けた研究開発 において更に必要となる施設の考え方を示した。
　長期に亘る継続的な研究開発が必要な核融合炉研究において、人材の育成・確保に、技術の継承に、核融合炉の社会的受容性の熟成のために、トリチウム取り扱い施設が必要なことはいうまでもない。上記２つの大量トリチウム 取り扱い施設が必要なことはいうまでもない。上記2つの大量トリチウム取り扱い施設を核に、日本のトリチウム施設が連携して、国内外のトリチウム研究における協力研究体制、今後の展望を議論した。

Keywords:tritium, core facility, broader approach, DEMO Reactor, thermonuclear fusion

Kazuki Hirakawa^a, Mitsuhiro Inoue^b and Takayuki Abe^a

^aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bDepartment of Materials Science and Technology, Faculty of Engineering, Nagaoka University of Technology, Kamitomika 1603-1, Nagaoka, Niigata 940-2188, Japan

Abstract

   Methanol oxidation performance of a carbon-supported Pt-Ru alloy catalyst used at the direct methanol fuel cell (DMFC) anode is improved by adding TiO₂. However, the methanol oxidation performance of the electrocatalyst described above must be enhanced further to realize practical application in DMFCs. In this study, we used our original surface-modifying technique termed the “polygonal barrel-sputtering method” to prepare a carbon-supported Pt-Ru and TiO₂ (Pt-Ru/TiO₂/C) electrocatalyst offering higher methanol oxidation performance. The obtained results show that the methanol oxidation performance of the prepared Pt-Ru/TiO₂/C is superior to that using wet process as the TiO₂ deposition method. Furthermore, for our sputtering method, the peak current of methanol oxidation on the Pt-Ru/TiO2/C is enhanced by increasing the TiO₂ deposited amount up to 2.8 wt.%. These results suggest that a Pt-Ru/TiO₂ interface area is increased using the polygonal barrel-sputtering method, providing the high methanol oxidation performance of Pt-Ru/TiO₂/C.

Keywords: Direct methanol fuel cells(DMFCs), Anode catalyst, Methanol oxidation, Polygonal barrel-sputtering method, Pt-Ru/Tio₂ interface

¹Department of Materials Science and Technology, Faculty of Engineering, Nagaoka University of Technology, Kamitomika 1603-1, Nagaoka, Niigata 940-2188, Japan
²Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

   In this study, we investigated the suppression of potential oscillations that occur while feeding H₂ gas including a large amount of CO to polymer electrolyte fuel cells. A carbon-supported Pt-Ru alloy (Pt-Ru/C) sample was prepared by the polygonal barrel-sputtering method. Electrochemical measurement was conducted in 1 N H₂SO₄electrolyte solution saturated with H₂ gas including CO of 1,000 ppm. From the measurement, it was found that the prepared Pt-Ru/C sample showed higher CO tolerance than a commercially available sample, and therefore, potential oscillations in the case of the prepared sample occurred for slightly longer intervals. Additional samples were prepared by sputtering different amounts of Ru on the prepared Pt-Ru/C sample, and it was observed that the average interval of potential oscillations increased with the amount of Ru. This result suggests that in the case of the polygonal barrel-sputtering method, Ru plays an important role in suppressing potential oscillations and influences the CO tolerance of Pt-Ru/C.

Keywords: Anode Catalyst, CO Oxidation, Polygonal Barrel-Sputtering Method, Polymer Electrolyte Fuel Cells, Potential Oscillations, Pt-Ru/C

Takahito Watakabe¹, Goroh Itoh², Yuji Hatano³

¹Graduate Student, Graduate School of Science and Engineering, Ibaraki University, 4-12-1 Nakanarusawa, Hitachi-city, Ibaraki 316-8511, Japan
²Department on Mechanical Engineering,Ibaraki University, 4-12-1 Nakanarusawa, Hitachi-city, Ibaraki 316-8511, Japan
³Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan

Abstract

   In recent years, the use of hydrogen as a clean energy has been paid attention to in terms of the prevention of global warming. Tanks composed solely of steel and cylinders consisting of aluminum liner reinforced with C-FRP in the surrounding are used to store high-pressure hydrogen gas in hydrogen stations and in fuel cell vehicles, respectively. On the other hand, hydrogen embrittlement has been known to occur in some metallic materials under several certain conditions. Also, it has been generally known that the environmental hydrogen that invades the material during services plays major role in hydrogen embrittlement. For this reason, investigations on the behavior of environmental hydrogen in metallic materials are needed. In this study, the behavior of environmental hydrogen in pure aluminum, 6061 aluminum alloy and 7075 aluminum alloy has been investigated by means of tritium autoradiography.

Keywords: Hydrogen embrittlement, environmental hydrogen, pure aluminum, 6061 aluminum alloy, 7075 aluminum alloy, tritium autoradiography

Masanori Hara^a, Yuka Togashi^a, Masao Matsuyama^a, Yasuhisa Oya^b,Kenji Okuno^b

^a Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
^bRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Ohya 836, Suruga-ku Shizuoka 422-8529, Japan

Abstract

   To investigate the aging behavior of Li_2+xTiO_3+y, the materials Li_2.0TiO_3+y, Li_2.2TiO_3+y and Li_2.4TiO_3+y were stored under moist air at ambient temperature. Under these conditions the weight of Li_2.0TiO_3+y increased by only 1% in 100 days, which provides evidence for the high chemical stability of Li₂TiO₃. In comparison, under the same storage conditions the weights of Li_2.2TiO_3+y and Li_2.4TiO_3+yincreased by 14% and 18%, respectively. The observed weight gain is attributed to the uptake of water, and to the reaction of a by-product of Li_2+xTiO_3+y with water and carbon dioxide. The weight gain curves were evaluated by the Jander equation. From the analysis, it was possible to obtain the diffusion coefficient of water through Li₂TiO₃ particles at ambient temperature. The determined value was 2 × 10^-17 m²/s.

Sanjeev K. SHARMA¹, Hideki ZUSHI², Ikuji TAKAGI³, Yuki HISANO¹, Mizuki SAKAMOTO², Yuta HIGASHIZONO², Taichi SHIKAMA³, Sigeru MORITA⁴, Tetsuo TANABE¹, Naoaki YOSHIDA², Kazuaki HANADA², Makoto HASEGAWA², Osamu MITARAI⁵, Kazuo NAKAMURA², Hiroshi IDEI², Kohnosuke N. SATO², Shoji KAWASAKI², Hisatoshi NAKASHIMA², Aki HIGASHIJIMA², Yousuke NAKASHIMA⁶, Nobuyuki NISHINO⁷, Yuji HATANO⁸, Akira SAGARA⁴, Yukio NAKAMURA⁴, Naoko ASHIKAWA⁴, Takashi MAEKAWA³, Yasuaki KISHIMOTO³, Yuichi TAKASE⁹ and QUEST Group

¹IGSES,Kyushu University
²RIAM, Kyushu University
³Kyoto University
⁴NIFS
⁶Kyushu Tokai University
⁷Hiroshima University
⁸Toyama University
⁹University of Tokyo

Abstract

   Particle retention and recycling are still major issues for the long pulse fusion devices. Plasma driven permeation(PDP) is one of the main reason behind the particle retention/recycling in these devices. To enhance the understanding of the particle retention under various discharge parameters, the PDP of hydrogen through a 30 μm thick Ni membrance heated at 523 K has been studied in RF and ohmic plasmas in the spherical tokamak QUEST. The permeation has been allowed by reflected neutral atoms by using a baffle before the Ni membrance to avoid any impurity deposition. The baffle before the Ni membrance also provides a way to simulate the particles retension behind plasma facing components(PFC) inside the tokamak chamber. It has been observed that the PDP fluence (Q_PDP)increases linearly with RF power (P_RF) and the chamber pressure. A linear relationship with scattered date between Q_PDP and Hα fluence (Qα) is found in all dischaeges, which suggests the fundamental proportionality of Q_PDP to the incident atomic fluence. For the longer discharges pulse, the PDP flux(Γ_H2) tends to saturate at τ ≈ 30 s, though the incidene ion flux to the chamber wall is ~5× 10²¹ m^-2 s^-1. Q_PDP is also found proportional to the plasma discharge widths(τ ›5s).

Keywords: plasma driven permeation, PDP, recycling, retention, tritium

Kun Zhang, Yuji Hatano

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

Abstract

   Layers of Mg or Al phosphate were prepared on type 430 ferritic stainless steel (SS430) by wet-chemical methods as tritium permeation barrier. Disk-type specimens of SS430 were first coated with ZrO₂/ZrO₂ or ZrO₂/Al₂O₃ layers (100 nm) by sol-gel and electrolytic deposition techniques. Then, the phosphate layers were prepared by dip-coating method; the total thickness of coating was 200 nm. The permeation rate of hydrogen was measured at 300-600 ℃ under driving pressure of 0.1 MPa. The Mg or Al phosphate layers provided significant barrier effects, and the permeation reduction factor observed under the present conditions was 200-3000 against that of Pd coated 430 specimen. Further investigation, however, was required to improve the stability at high temperatures.

Keywords: Tritium permeation barrier, Wet-chemical method, Sol-gel method, Electrolytic deposition, Phosphate

松山政夫

富山大学水素同位体科学研究センター

Abstract

   From viewpoint of establishment of a standard system to determine the absolute amount and/or concentration of high-level tritium which is used as fuel of a future fusion reactor and one of radioactive materials, the appllicability of a specially designed twin type conduction calorimeter has been examined at the tritium handling facility in Hydrogen Isotope Research Center, University of Toyama. Two examples of the examinations were shown: one is the measurement of a given amount of tritium stored in ZrNi alloy powder, and the other is the measurement of tritiated water. Both examinations gave enough results as a standard device. lt was seen, therefore, that the present calorimeter has high potential for measurement of the absolute amount of tritium. Furthermore, an example that the present calorimeter was applied to nondestructive measurement of the amount of tritium in a glass ampoule was described.

Keywords: tritium, calorimeter, absolute measurement, non-destructive measurement, fusion reactor

Masao Matsuyama, Kenichi Takatsuka, Masanori Hara

Hydrogen Isotope Research Center, University of Toyama, Gofuku3190,Toyama 930-8555,Japan

Abstract

   Applicability of a specially designed high-sensitivity conduction calorimeter has been examined to determine the absolute amount of tritium in tritiated water. A given concentration of tritiated water was prepared for examinations. Prior to measurements by the calorimeter, the concentration of tritium was measured by a liquid scintillation counter, and it was determined as 46.0±0.5MBq/cm³. This tritiated water, 20cm³, was loaded into a sample cell of the calorimeter. The concentration of tritium determined by the calorimeter was 45.2±2.0MBq/cm³, and it agreed with that by the liquid scintillation counter within a few percent. In addition, detection limit of the present calorimeter was evaluated to be 40 MBq, considering the standard deviation of background level measured without tritiated water. It was concluded, therefore, that the present calorimeter can be utilized as a standard device for determination of the absolute amount of tritium in tritiated water.

Keywords: Calorimeter, Tritiated water, Measurement, Liquid scintillation counter

Teo Nozaki, Yuji Hatano, Eriko Yamakawa, Asuka Hachikawa, Kazuyoshi Ichinose

Hydrogen Isotope Research Center, University of Toyama, Gofuku3190,Toyama 930-8555,Japan

Abstract

   Hafnium nitride (HfN) was chosen as a material for non-porous intermediate layer to improve the high temperature stability of Pd-Ta composite membranes for hydrogen separation. A layer of dense HfN (70 nm) was prepared between Ta substrate and thin Pd film (300 nm), and the high temperature stability of Pd coating was examined by hydrogen absorption experiments at 573 K after the heat treatments at 873 and 973 K. The HfN layer showed obvious hydrogen permeability, though the permeation rate in HfN appeared to be smaller than that in Pd and Ta. In addition, the degradation in coating effects of Pd at elevated temperatures was substantially retarded by HfN layer. Such improved stability was ascribed to retardation of open porosity development in Pd films and interdiffusion between Pd and Ta. It was concluded that HfN is a potential candidate material for intermediate layer to improve high temperature stability of Pd-group 5 metal composite hydrogen separation membranes.

Keywords: Hydrogen separation, Permeation membranes, Composite materials, Palladium, Tantalum, Nitrides

K. Okuno^a, S. Suzuki^a, M. Kobayashi^a, R. Kurata^a, M. Matsuyama^b, N. Ashikawa^c, A. Sagara^c, Y. Oya^a

^aRadioscience Research Laboratory, Faculty of Science, Shizuoka ersity, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan

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

^cNational Institute for Fusion Science, Gifu, Japan

Abstract

    The chemical states and the deuterium retention of boron films deposited on type 316 stainless steel used as first wall of LHD, which contained 52% oxygen, were studied by means of XPS and TDS. The high oxygen concentration in the boron film could be attributed to oxygen in the surface oxide layer of the stainless steel. The major chemical state of boron with 52% oxygen concentration was B₂O₃. The B₂O₃ structure was decomposed to B–B and B–O bonds by the energetic deuterium irradiation. It was found that the deuterium could remain in boron films under operation temperature in the first wall of LHD. The deuterium retention in the boron film deposited on SS was smaller compared to that in the boron film deposited on silicon substrate with 36% oxygen concentration. As the oxygen concentration in boron films increased, the amount of sputtered oxygen was clearly increased, indicating that more heavy water was formed by reacting with implanted deuterium and released, resulting in the smaller deuterium retention in the boron films. The knowledge for the tritium retention in boron films which cannot remove in the operation temperature and the release of tritiated water should be required for the D–D discharge in LHD.

Keywords: Boronization, LHD, SS316, XPS, TDS

Sanjeev K. Sharma^a, Hideki Zushi^b, Ikuji Takagi^c, Yuki Hisano^a, Mizuki Sakamoto^b, Yuta Higashizono^b, Taichi Shikama^c, Sigeru Morita^d, Tetsuo Tanabe^a, Naoaki Yoshida^b, Kazuaki Hanada^b, Makoto Hasegawa^b, Osamu Mitarai^e, Kazuo Nakamura^b, Hiroshi Idei^b, Kohnosuke N. Sato^b, Shoji Kawasaki^b, Hisatoshi Nakashima^b, Aki Higashijima^b, Yousuke Nakashima^f, Nobuyuki Nishino^g, Yuji Hatano^h, Akira Sagara^d, Yukio Nakamura^d, Naoko Ashikawa^d, Takashi Maekawa^c, Yasuaki Kishimoto^c, Yuichi Takaseⁱ and QUEST Group

^aIGSES, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^bRIAM, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
^cDNE, Graduate School of Engineering, Kyoto University, Japan
^dNational Institute for Fusion Science, Toki, Japan
^eKyushu Tokai University, 9-1-1 Toroku, Kumamoto 862-8652, Japan
^fPlasma Research Center, University of Tsukuba, Japan
^gDMSE, Graduate School of Engineering, Hiroshima University, Japan
^hHydrogen Isotope Research Center, Toyama University, Toyama 930-8555, Japan
ⁱGraduate School of Frontier Science, University of Tokyo, Ibaragi, Japan

Abstract

    Particle retention and recycling in plasma fusion devices are generally associated with the diffusion of atomic hydrogen into the materials. The resulted permeation of atomic hydrogen is known as plasma driven permeation (PDP). This permeation may also be significant, even in the walls, which are not directly exposed to the plasma. Under similar conditions, the permeation flux (Γ_perm) of hydrogen through a 30 μm thick Ni membrane heated at 412-575 K has been measured in the spherical tokamak QUEST. Γ_perm is being measured during the scans of different operating parameters like RF power (P_RF), chamber pressure (P_chamber), discharge widths (τ_dis) and vertical magnetic field (B_Z). Simultaneously edge plasma density and spectral intensities of atomic (Balmer) lines and molecular (Fulcher) bands have been compared with the permeation measurements. A linear relationship has been established between the time integrated Γ_perm i.e. permeation fluence (Q_perm) and the time integrated H_α intensity i.e. H_α fluence (Q_α). Q_perm also shows a strong relationship with the edge plasma density and various spectral fluences. The obtained results are discussed for exploring the applicability of the permeation probes in measuring the atomic flux near the first walls.

Keywords: Plasma driven permeation, Atomic diffusion, Recycling; Retention, PWI