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JAEA Reports

Status of study of long-term assessment of transport of radioactive contaminants in the environment of Fukushima (FY2018) (Translated document)

Nagao, Fumiya; Niizato, Tadafumi; Sasaki, Yoshito; Ito, Satomi; Watanabe, Takayoshi; Dohi, Terumi; Nakanishi, Takahiro; Sakuma, Kazuyuki; Hagiwara, Hiroki; Funaki, Hironori; et al.

JAEA-Research 2020-007, 249 Pages, 2020/10


The accident of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. occurred due to the Great East Japan Earthquake, Sanriku offshore earthquake, of 9.0 magnitude and the accompanying tsunami. As a result, large amount of radioactive materials was released into the environment. Under these circumstances, Japan Atomic Energy Agency (JAEA) has been conducting "Long-term Assessment of Transport of Radioactive Contaminants in the Environment of Fukushima" concerning radioactive materials released in environment, especially migration behavior of radioactive cesium since November 2012. This report is a summary of the research results that have been obtained in environmental dynamics research conducted by JAEA in Fukushima Prefecture.

Journal Articles

HEPA filter clogging and volatile material release under solvent fire accident in fuel reprocessing facility

Ono, Takuya; Watanabe, Koji; Tashiro, Shinsuke; Amano, Yuki; Abe, Hitoshi

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07

After the Fukushima-Daiichi accident, countermeasures against the severe accident are newly required as regulatory items for nuclear facilities. Organic solvent fire in cell was defined as one of the accidents in the fuel reprocessing plant. When the solvent burns, aerosols including soot are released. The substances clog HEPA filters in the ventilation system and their breakthrough may happen because of differential pressure rising. Moreover, the fire can also release volatile radioactive gaseous species, which can pass through HEPA filters. These phenomena are important for evaluation of confinement capability of the facility and public exposure. We have investigated, in relating to the clogging behavior, release behavior of aerosols as well as of volatile materials from burnt solvent. In the presentation, we will report experimental data and evaluation results obtained from recent research.

JAEA Reports

Solvent extraction and release behavior of ruthenium and europium in fire accident conditions in reprocessing plants (Contract research)

Amano, Yuki; Watanabe, Koji; Masaki, Tomoo; Tashiro, Shinsuke; Abe, Hitoshi

JAEA-Technology 2016-012, 21 Pages, 2016/06


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.

Journal Articles

Study on safety evaluation for nuclear fuel cycle facility under accident conditions

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.

Journal Articles

Study on safety evaluation for nuclear fuel cycle facility under fire accident conditions

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.

Journal Articles

Analytical study on fire and explosion accidents assumed in HTGR hydrogen production system

Inaba, Yoshitomo; Nishihara, Tetsuo; Nitta, Yoshikazu*

Nuclear Technology, 146(1), p.49 - 57, 2004/04

 Times Cited Count:3 Percentile:24.74(Nuclear Science & Technology)

One of the most important safety design issues for a hydrogen production system coupling with a High Temperature Gas-cooled Reactor (HTGR) is to ensure reactor safety against fire and explosion accidents because a large amount of combustible fluid is dealt with in the system. The Japan Atomic Energy Research Institute (JAERI) has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High Temperature engineering Test Reactor (HTTR). In the plan, we developed the P2A code system to analyze event sequences and consequences in detail on the fire and explosion accidents assumed in the HTGR or HTTR hydrogen production system. This paper described the three accident scenarios assumed in the system, the structure of P2A, the analysis procedure with P2A and the results of the numerical analyses based on the accident scenarios, and it was showed that P2A was a useful tool for the accident analysis in the system.

JAEA Reports

Differential pressure analysis for ventilation filter by smoke under fire accident with CELVA-1D

Watanabe, Koji; Tashiro, Shinsuke; Abe, Hitoshi; Takada, Junichi; Morita, Yasuji

JAERI-Tech 2004-029, 48 Pages, 2004/03


In a part of building ventilating installation of a nuclear fuel facility, a reprocessing plant for example, the pre-filters are adopted as one of the ventilation filters. In a fire accident, it is supposed that, because of the pre-filter clogging by large smoke, its differential pressure ($$Delta$$p) is evolved up to the value at its breakage. Therefore, in regard to maintaining the confinement of radioactive materials, it is important to predict the time course of $$Delta$$p of the pre-filter accurately. In the current study, it was assumed that the empirical equation for the DF of the pre-filter as function of smoke particle diameter (SPD), between 0.7-2 $$mu$$m, could be applied to estimating its DF for SPD above 2 $$mu$$m. Under this assumption, the time corresponding to its $$Delta$$p at its breakage, calculated by CELVA-1D, was agreed well with the experimental result.

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