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Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu*
JAEA-Review 2017-010, 93 Pages, 2017/06
JAEA-Review-2017-010.pdf:2.47MB
There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee.
Amano, Yuki; Watanabe, Koji; Masaki, Tomoo; Tashiro, Shinsuke; Abe, Hitoshi
JAEA-Technology 2016-012, 21 Pages, 2016/06
JAEA-Technology-2016-012.pdf:1.81MB
To contribute to safety evaluation of fire accident in fuel reprocessing plants, solvent extraction behavior of ruthenium, which could form volatile species, was investigated. Distribution ratios of ruthenium at fire accident conditions were obtained by extraction experiments with several solvent composition at different temperature as parameters. In order to investigate release behavior of ruthenium and europium at fire accident, release ratios of ruthenium and europium were also obtained by solvent combustion experiments.
Abe, Hitoshi; Masaki, Tomoo; Amano, Yuki; Uchiyama, Gunzo
JAEA-Research 2014-022, 12 Pages, 2014/11
JAEA-Research-2014-022.pdf:1.03MB
To contribute safety evaluation of boiling and drying accident of high active liquid waste (HALW) in fuel reprocessing plant, release behavior of Ru, which was considered as an important nuclide for evaluating public dose from the volatile viewpoint, has been investigated. It has been reported that release of Ru becomes conspicuously after HALW is dried up. In this work, to grasp the release behavior of Ru, release ratio of Ru with thermal decomposition of Ru nitrate, which would be in the dried HALW, was measured and release rate constant of Ru from the nitrate was estimated. It was found that the calculation result of release rate of Ru from the nitrate with rise of temperature by using the constant could well simulate the result acquired from the beaker-scale experiment.
Abe, Hitoshi; Tashiro, Shinsuke; Morita, Yasuji
JAERI-Conf 2005-007, p.199 - 204, 2005/08
Source term data for estimating release behavior of radioactive nuclides is necessary to evaluate synthetic safety of nuclear fuel cycle facility under accident conditions, such as fire and criticality. In JAERI, the data has been obtained by performing some demonstration tests. In this paper, the data for the criticality accident in fuel solution obtained from the TRACY experiment, will be mainly reviewed. At 4.5 h after the transient criticality, the release ratio of the iodine were about 0.2% for re-insertion of transient rod at just after transient criticality and about 0.9% for not re-insertion. Similarly the release coefficient and release ratio for Xe were estimated. It was proved that the release ratio of Xe-141 from the solution was over 90% in case that the inverse period was over about 100 (1/s). Furthermore, outline of the study on the fire accident as future plan will be also mentioned.
Tashiro, Shinsuke; Abe, Hitoshi; Morita, Yasuji
JAERI-Conf 2005-007, p.348 - 350, 2005/08
Hot test at Rokkasho Reprocessing plant has been started since last year. In addition, construction of the MOX fuel fabrication facility at Rokkasho site is planning. So, the importance of safety evaluation of the nuclear fuel cycle facility is increasing. Under the fire accident, one of the serious postulated accidents in the nuclear fuel cycle facility, the equipments (glove-box, ventilation system, ventilation filters etc.) for the confinement of the radioactive materials within the facility could be damaged by a large amount of heat and smoke released from the combustion source. Therefore, the fundamental data and models calculating for the amount of heat and smoke released from the combustion source under such accident are important for the safety evaluation of the facility. In JAERI, the study focused on the evaluation of amount of heat and smoke released from the combustion source is planning. In this paper, the outline of experimental apparatus, measurement items and evaluation terms are described.
Nuclear Fuel Cycle Facility Safety Research Committee
JAERI-Tech 2001-055, 92 Pages, 2001/07
JAERI-Tech-2001-055.pdf:5.39MB
no abstracts in English
Fujiki, Kazuo
Decommissioning Policies for Nuclear Facilities, p.111 - 120, 1992/00
no abstracts in English
Yanagisawa, Hiroshi; Takeshita, Isao; Miyoshi, Yoshinori; Sugikawa, Susumu; Suzaki, Takenori; Tachimori, Shoichi
Proc. of the 91 Int. Conf. on Nuclear Criticality Safety,Vol. 2, p.V-65 - V-72, 1991/00
no abstracts in English
Yanagisawa, Hiroshi; Takeshita, Isao; Nomura, Masayuki; ; Tsujino, Takeshi
Proc. of the CSNI Specialist Meeting on Safety and Risk Assessment in Fuel Cycle Facilities, p.461 - 470, 1991/00
no abstracts in English
Kato, Shohei; *; Yamamoto, Hideaki
EPA-520/1-90-013, p.44 - 56, 1990/00
no abstracts in English
