Interim activity status report of "the group for investigation of reasonable safety assurance based on graded approach" (from September, 2019 to September, 2020)

Yonomoto, Taisuke; Nakashima, Hiroshi*; Sono, Hiroki; Kishimoto, Katsumi; Izawa, Kazuhiko; Kinase, Masami; Osa, Akihiko; Ogawa, Kazuhiko; Horiguchi, Hironori; Inoi, Hiroyuki; et al.

JAEA-Review 2020-056, 51 Pages, 2021/03

JAEA-Review-2020-056.pdf:3.26MB

A group named as "The group for investigation of reasonable safety assurance based on graded approach", which consists of about 10 staffs from Sector of Nuclear Science Research, Safety and Nuclear Security Administration Department, departments for management of nuclear facility, Sector of Nuclear Safety Research and Emergency Preparedness, aims to realize effective graded approach (GA) about management of facilities and regulatory compliance of JAEA. The group started its activities in September, 2019 and has had discussions through 10 meetings and email communications. In the meetings, basic ideas of GA, status of compliance with new regulatory standards at each facility, new inspection system, etc were discussed, while individual investigation at each facility were shared among the members. This report is compiled with expectation that it will help promote rational and effective safety management based on GA by sharing contents of the activity widely inside and outside JAEA.

New PbTiO-type giant tetragonal compound BiZnVO and its stability under pressure

Yu, R.*; Hojo, Hajime*; Oka, Kengo*; Watanuki, Tetsu; Machida, Akihiko; Shimizu, Keisuke*; Nakano, Kiho*; Azuma, Masaki*

Chemistry of Materials, 27(6), p.2012 - 2017, 2015/03

https://doi.org/10.1021/cm504133e

no abstracts in English

Corrosion modeling for carbon steel under oxygen depleted underground environment

Shibata, Toshio*; Watanabe, Masatoshi; Taniguchi, Naoki; Shimizu, Akihiko*

Zairyo To Kankyo, 62(2), p.70 - 77, 2013/02

https://doi.org/10.3323/jcorr.62.70

In the oxygen depleted underground environment, carbon steel reacts with HO, producing H gas and forming corrosion film on the steel surface. Corrosion rate is controlled by diffusion of reaction species through corrosion film. Diffusion constants of some species working in the corrosion process were obtained from literatures. However, no data were found on the diffusion constant of HO in iron oxides based on an appropriate assumption. Mass transfer model for the corrosion rate was used to simulate the corrosion rate of carbon steel. Liquid phase diffusion model of Fe or HO through pores in the corrosion film and solid phase diffusion model of HO through corrosion film itself were examined by Excel simulation. Change in corrosion current density and corrosion loss with time and pH and temperature dependence of corrosion current density were examined. By comparing the results, it is suggested that the solid phase diffusion of HO in the corrosion film controls the corrosion rate of carbon steel in oxygen depleted environment.

Electrical resistance measurement techniques for metal hydrides under high-pressure H conditions & electrical transport and structural properties of FeH

Matsuoka, Takehiro*; Hirao, Naohisa*; Oishi, Yasuo*; Shimizu, Katsuya*; Machida, Akihiko; Aoki, Katsutoshi

Koatsuryoku No Kagaku To Gijutsu, 21(3), p.190 - 196, 2011/09

https://doi.org/10.4131/jshpreview.21.190

We present the newly developed electrical resistance measurement technique for metal hydrides compressed in high-pressure H and the first successful simultaneous measurements of electrical resistance and X-ray diffraction of FeH at high pressures and low temperatures. The electrical resistivity showed a sharp increase with the formation of iron-hydride FeH () at 3.5 GPa. The '-phase of FeH was found to be metallic up to 25.5 GPa. The vs. curves up to 16.5 GPa approximately follow Fermi-liquid law below 25 K. However, was found to be better fitting at 25.5 GPa. This change can be related to the previously reported ferromagnetism collapse at corresponding pressures.

Structural and valence changes of europium hydride induced by application of high-pressure H

Matsuoka, Takahiro*; Fujihisa, Hiroshi*; Hirao, Naohisa*; Oishi, Yasuo*; Mitsui, Takaya; Masuda, Ryo; Seto, Makoto*; Yoda, Yoshitaka*; Shimizu, Katsuya*; Machida, Akihiko; et al.

Physical Review Letters, 107(2), p.025501_1 - 025501_4, 2011/07

https://doi.org/10.1103/PhysRevLett.107.025501

Europium-hydride EuH exposed to high-pressure H conditions has been found to exhibit structural and valence changes, (=2, divalent) 63/ (=2, 7.2-8.7 GPa) 4/ (2, 8.7-9.7 GPa) 4/ (2, 9.7 GPa-, trivalent). Having trivalent character and distorted cubic fcc structure, the 4/ structure is the -phase which has been commonly observed for other rare-earth metal hydrides. Our study clearly demonstrates that EuH is no longer an "irregular" member of the rare earth metal hydrides.

Investigation of the long term corrosion resistance of the overpack FY2006 (Contract research)

Tachikawa, Hirokazu*; Kawakubo, Fumie*; Shimizu, Akihiko*; Shibata, Toshio*; Azumi, Kazuhisa*; Inoue, Hiroyuki*; Sugimoto, Katsuhisa*; Tsuru, Toru*; Fujimoto, Shinji*

JAEA-Research 2007-086, 74 Pages, 2008/02

JAEA-Research-2007-086.pdf:5.96MB

The corrosion life time of the overpack has been investigated on the basis of experimental data and past research, assuming the ranging geological environment of Japan. However, some subject for the realization of the overpack design, such as the behavior of the overpack under high pH conditions, the behavior of the overpack with change of near-field environmental condition and the corrosion behavior of the welds have still been left. To take into account these conditions, expert committee composed of metal corrosion science experts were established in the Nuclear Safety Research Association and past research outcomes and the theory of safety assessment for long term corrosion resistance were investigated from the view points of metal corrosion science.

Overview of national centralized tokamak program; Mission, design and strategy to contribute ITER and DEMO

Ninomiya, Hiromasa; Akiba, Masato; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hayashi, Nobuhiko; Hosogane, Nobuyuki; Ikeda, Yoshitaka; Inoue, Nobuyuki; et al.

Journal of the Korean Physical Society, 49, p.S428 - S432, 2006/12

To contribute DEMO and ITER, the design to modify the present JT-60U into superconducting coil machine, named National Centralized Tokamak (NCT), is being progressed under nationwide collaborations in Japan. Mission, design and strategy of this NCT program is summarized.

Overview of the national centralized tokamak programme

Kikuchi, Mitsuru; Tamai, Hiroshi; Matsukawa, Makoto; Fujita, Takaaki; Takase, Yuichi*; Sakurai, Shinji; Kizu, Kaname; Tsuchiya, Katsuhiko; Kurita, Genichi; Morioka, Atsuhiko; et al.

Nuclear Fusion, 46(3), p.S29 - S38, 2006/03

https://doi.org/10.1088/0029-5515/46/3/S05

The National Centralized Tokamak (NCT) facility program is a domestic research program for advanced tokamak research to succeed JT-60U incorporating Japanese university accomplishments. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility and mobility is pursued in aspect ratio and shape controllability, feedback control of resistive wall modes, wide current and pressure profile control capability for the demonstration of the high-b steady state.

Overview of design and R&D of test blankets in Japan

Enoeda, Mikio; Akiba, Masato; Tanaka, Satoru*; Shimizu, Akihiko*; Hasegawa, Akira*; Konishi, Satoshi*; Kimura, Akihiko*; Koyama, Akira*; Sagara, Akio*; Muroga, Takeo*

Fusion Engineering and Design, 81(1-7), p.415 - 424, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.08.097

no abstracts in English

Plan and strategy for ITER blanket testing in Japan

Enoeda, Mikio; Akiba, Masato; Tanaka, Satoru*; Shimizu, Akihiko*; Hasegawa, Akira*; Konishi, Satoshi*; Kimura, Akihiko*; Koyama, Akira*; Sagara, Akio*; Muroga, Takeo*

Fusion Science and Technology, 47(4), p.1023 - 1030, 2005/05

https://doi.org/10.13182/FST05-A822

The Fusion Council of Japan has established the long-term research and development program of the blanket in 1999. In the program, the solid breeder blanket was selected as the primary candidate blanket of the fusion power demonstration plant in Japan. In the program, Japan Atomic Energy Research Institute (JAERI) has been nominated as a leading institute of the development of solid breeder blankets, in collaboration with universities, for the near term power demonstration plant, while, universities including National Institute for Fusion Science (NIFS) are assigned mainly to develop advanced blankets for longer term power plant development. In the long term research and development program, ITER blanket module testing is identified as the most important milestone, by which integrity of candidate blanket concepts and structures are evaluated. In Japan, universities, NIFS and JAERI cover a variety of types of blanket development. This paper presents a plan and strategy for ITER blanket module testing in Japan.