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Sukegawa, Takenori; Shimada, Taro; Katsurai, Kiyomichi; Tanaka, Tadao; Nakayama, Shinichi
JAEA-Review 2009-075, 86 Pages, 2010/03
JAEA-Review-2009-075.pdf:8.13MB
In the field of safety regulation systems for nuclear facilities after completion of operation, criteria of residual radioactivities and confirmation and verification procedures for termination of decommissioning are important problems that should concretely be made a study. Safety standards and criteria in IAEA, USA, etc., and practical examples of site release of power reactors in USA were studied, therefore, problems for introducing the regulation system in Japan were discussed. In this report, final status survey of Trojan power plant were investigated as a particular case of site release, and, concept of specifying survey areas to be measured radioactivities and demonstrated to compliance with release criteria was discussed. In addition, the idea of confirmation and verification procedure for termination of decommissioning in Japan was proposed referring to the US guidance (MARSSIM).
Taniguchi, Naoki; Suzuki, Hiroyuki*; Nakanishi, Tomoaki*; Nakayama, Takenori*; Masugata, Tsuyoshi*; Tateishi, Tsuyoshi*
Zairyo To Kankyo, 56(12), p.576 - 584, 2007/12
The long term hydrogen absorption behavior and the possibility of hydrogen embrittlement were studied for titanium overpack for high level radioactive waste disposal. The results of galvanostatic cathodic polarization tests showed that as the cathodic current density is lowered, the amount of absorbed hydrogen for a constant cathodic charge was increased as well as hydrogen permeated into inside of titanium. The hydrogen absorption ratio for a cathodic current density equivalent to the corrosion rate under anaerobic condition was estimated to nearly 100 percent, and the amount of absorbed hydrogen for 1000 years was evaluated to be 400 ppm. The mechanical property of titanium containing hydrogen depended on not only hydrogen concentration but also hydrogen distribution type. The more hydrogen distribution is uniform, the degree of embrittlement was larger. It was expected that the rupture of titanium overpack with 6 mm thickness would be initiated if the crack size in titanium is over about 2-3 mm under the stress corresponds to yield strength.
Wada, Ryutaro*; Nishimura, Tsutomu*; Nakanishi, Tomoaki*; Nakayama, Takenori*; Sakashita, Shinji*; Fujiwara, Kazuo*; Tateishi, Tsuyoshi*
JNC TJ8400 2005-001, 224 Pages, 2004/02
JNC-TJ8400-2005-001.pdf:16.23MB
For repository container material of high-level radioactive waste, titanium and nickel-base alloys have been investigated as high corrosion resistance metal. In this study, the effects of environmental and material factors on hydrogen absorption of titanium were investigated experimentally. As for nickel-base allys, previous studies on corrosion behavior were serched.
Taniguchi, Naoki; Suzuki, Hiroyuki*; Yui, Mikazu; Nakanishi, Tomoaki*; Nakayama, Takenori*; Masugata, Tsuyoshi*; Tateishi, Tsuyoshi*
no journal, ,
Titanium (including titanium alloy) is one of the candidate materials of overpacks for geological disposal of high-level radioactive waste, and required long term integrity against the groundwater for more than 1000 years. As the corrosion of titanium occurs, hydrogen is generated since the deep underground environment is originally low oxygen concentration condition. There is a possibility that the titanium overpack will be attackd by the hydrogen embrittlement due to long term hydrogen absorption. In this study, the amount of hydrogen and the possibility of embrittlement were investigated based on the experimental data on the corrosion rate, hydrogen absorption behavior, mechanical proparty of titanium containing hydrogen.
Mizukoshi, Seiji; Sukegawa, Takenori; Nakayama, Shinichi
no journal, ,
Technical information including available technologies and residual radioactive contamination is required to establish safety standards for regulatory review for the future decommissioning of nuclear fuel facilities. We have acquired the technical information on the past and on-going decommissioning projects of domestic and foreign nuclear fuel facilities. This report presents the information and the study on the safety issues concerning the decommissioning of fuel reprocessing facilities.
Shimada, Taro; Oshima, Soichiro; Sukegawa, Takenori; Nakayama, Shinichi
no journal, ,
no abstracts in English
Takamura, Atsushi; Shimada, Taro; Oshima, Soichiro; Uno, Yuichi; Gunji, Soichi; Ito, Takeshi; Sukegawa, Takenori; Nakayama, Shinichi
no journal, ,
no abstracts in English
Takamura, Atsushi; Shimada, Taro; Oshima, Soichiro; Uno, Yuichi; Gunji, Soichi*; Ito, Takeshi; Sukegawa, Takenori; Tanaka, Tadao; Nakayama, Shinichi
no journal, ,
no abstracts in English
Shimada, Taro; Takamura, Atsushi; Oshima, Soichiro; Uno, Yuichi; Gunji, Soichi*; Ito, Takeshi; Sukegawa, Takenori; Tanaka, Tadao; Nakayama, Shinichi
no journal, ,
no abstracts in English
Sukegawa, Takenori; Shimada, Taro; Uno, Yuichi; Oshima, Soichiro; Ito, Takeshi; Takamura, Atsushi; Tanaka, Tadao; Nakayama, Shinichi
no journal, ,
no abstracts in English
Hayakawa, Takehito; Katakura, Junichi; Takeda, Seiji; Sukegawa, Takenori; Nakayama, Shinichi
no journal, ,
no abstracts in English
Mukai, Masayuki; Shimada, Taro; Tanaka, Tadao; Sukegawa, Takenori; Nakayama, Shinichi
no journal, ,
Nuclear Safety Research Center in JAEA has been conducting R&D activities to support national safety regulatory systems. (1) Research on safety assessment for decommissioning of nuclear fuel cycle facilities: A prototype computer code for fuel cycle facilities was developed by improving the code previously developed for nuclear reactor facilities. (2) Research on dose evaluation during dismantling: Values on evaluation parameters were obtained through the cutting test using actual contaminated pipes. Observed values for dispersion ratio of Co-60 and filtration efficiency indicated that existing recommendation values were conservative for safety assessment. (3) Research on site release on termination of decommissioning: A computer code was developed to calculate remaining radionuclide concentrations as site release criteria, and improved by adding practical settings for external and internal exposed doses. Preliminary calculations were conducted on typical land reuse scenario.
Mukai, Masayuki; Shimada, Taro; Tanaka, Tadao; Sukegawa, Takenori; Nakayama, Shinichi
no journal, ,
A code for evaluation of exposed doses for public and workers by decommissioning of nuclear cycle facilities was developed based on DecDose code which was developed by JAEA and intended to assess safety of decommissioning for atomic power reactors. We discussed on characteristic aspects particular to the nuclear cycle facilities and improved DecDose on the subject of operation process, target nuclides, and exposure pathways. Evaluation result by the developed code applied on a trial basis to uranium-enrichment factory was reasonably reflected the particular aspects of the nuclear cycle facilities in exposure doses.
