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Yamashita, Kiyoto; Maki, Shota; Yokosuka, Kazuhiro; Fukui, Masahiro; Iemura, Keisuke
JAEA-Technology 2023-023, 97 Pages, 2024/03
JAEA-Technology-2023-023.pdf:8.21MB
The incinerator adopted to incineration room, Plutonium Waste Treatment Facility had been demonstrated since 2002 for developing technologies to reduce the volume of fire-resistant wastes such as vinyl chloride (represented by Polyvinyl chloride bags) and rubber gloves for Radio Isotope among radioactive solid wastes generated by the production of mixed oxide fuels. The incinerator, cooling tower, and processing pipes were replaced with a suspension period from 2018 to 2022, which fireproof materials on the inner wall of the incinerator was cracked and grown caused by hydrogen chloride generated when disposing of fire-resistant wastes. This facility consists of the waste feed process, the incineration process, the waste gas treatment process, and the ash removal process. We replaced the cooling tower in the waste gas treatment process from March 2020 to March 2021, and the incinerator in the incineration process from January 2021 to February 2022. In addition, samples were collected from the incinerator and the cooling tower during the removing and dismantling of the replaced devices, observed by Scanning Electron Microscope and X-ray microanalyzer, and analyzed by X-ray diffraction to investigate the corrosion and deterioration of them. This report describes the method of setting up the green house, the procedure for replacing them, and the results from analysis in corrosion and deterioration of the cooling tower and incinerator.
Tashiro, Shinsuke; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Watanabe, Koji*; Abe, Hitoshi
Nuclear Technology, 208(10), p.1553 - 1561, 2022/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)To contribute to the confinement safety evaluation of the radioactive materials in the Glove box (GB) fire accident, combustion tests with the Polymethyl methacrylate (PMMA) and the Polycarbonate (PC) as typical panel materials for the GB have been conducted with a relatively large scale apparatus. As the important data for evaluating confinement safety, the release ratio and the particle size distribution of the soot generated from the burned materials were obtained. Furthermore, the rise of the differential pressure (
P) of the high efficiency particle air (HEPA) filter by the soot loading was also investigated. As results, the release ratio of the soot from the PC was about seven times as large as the PMMA. In addition, it was found that the behavior of the rise of the P with soot loading could be represented uniformly regardless of kinds of combustion materials by considering effect of the loading volume of the soot particle in the relatively low loading region.
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
SAND2022-14235 (Internet), 29 Pages, 2022/10
This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2022. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, JAEA F7-1 and F7-2 sodium pool fire experiments are used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2023.
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
Proceedings of International Topical Meetings on Advances in Thermal Hydraulics (ATH 2022) (Internet), p.316 - 329, 2022/06
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
SAND2021-15469 (Internet), 45 Pages, 2021/12
This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2021. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, JAEA F7-1 and F7-2 sodium pool fire experiments are used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2021.
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
SAND2021-0136 (Internet), 53 Pages, 2021/01
This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2020. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. These input requirements are flexible enough to permit further model development via control functions to enhance the current model without modifying the source code. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, a JAEA F7-1 sodium pool fire experiment is used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2021.
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
JAEA-Research-2020-007.pdf:15.83MB
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.
Louie, D. L. Y.*; Uchibori, Akihiro
SAND2019-15043 (Internet), 35 Pages, 2019/12
This report describes the progress on the sodium fire research in fiscal year 2019 in the Civil Nuclear Energy Research and Development Working Group (CNWG). In this study, the validation study of the sodium pool fire model incorporated into the MELCOR code, which was originally developed for accident analysis in light water reactors, was carried out through the numerical analysis on the sodium pool fire experiment named F7-1. In this preliminary analysis, pool and atmosphere temperature went up to the same level with the measured results, while the unnatural behavior appeared in the latter half of the analysis. Based on this result, recommendations for improvement were made for a new analysis in next fiscal year, 2020.
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.
Okano, Yasushi; Yamano, Hidemasa
Mechanical Engineering Journal (Internet), 4(3), p.16-00517_1 - 16-00517_10, 2017/06
A sensitivity study on forest fire hazard curves was performed. The probability fluctuation on forest fire breakout time affects the reaction intensity and the fireline intensity around 4% and 14% respectively. The probability fluctuation on forest fire breakout points affects the hazard curve frequency around +70% to -40%. The probability fluctuation due to forest firefighting operation only affects the frequency of the hazard curves, but not the intensity. The hazard curves without the effect of firefighting remarkably increase around 40 to 80 times in frequency in comparison with those with considering the forest firefighting operation effect outside the plant. This study indicated that the most significant factor in the forest fire hazard risk is whether the forest firefighting operation outside the plant is expected before the forest fire arrival at the plant.
Takata, Takashi; Ohno, Shuji; Tajima, Yuji*
Mechanical Engineering Journal (Internet), 4(3), p.16-00577_1 - 16-00577_11, 2017/06
Evaluation of accidental sodium leak, combustion, and its thermal consequence is one of the important issues to be assessed in the field of sodium-cooled fast reactor (SFR). The present paper deals with the sodium pool fire and subsequent heat transfer behavior in air atmosphere two-cell geometry both experimentally and analytically because such two-cell configuration is considered as a typical one to possess important characteristic of multi-compartment system seen in an actual plant. As a result of the numerical analysis using a lumped-parameter based zonal model safety analysis code SPHINCS, the applicability of the ventilation model implemented in SPHINCS has been demonstrated. It is also investigated that the buoyancy- driven ventilation is dominant in the experiment.
Okano, Yasushi; Yamano, Hidemasa
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 9 Pages, 2016/11
This paper deals with an event sequence by forest fire heat effect on a decay heat removal function of a sodium-cooled fast reactor. Related to the potential vulnerability, an event scenario was developed using conservative assumptions. An event tree was developed with an initiating event of the loss of off-site power, and the headings are related to "external diesel fuel tanks", "emergency diesel generator and its auxiliary system", "alternative cooling system and its power source", and "decay heat air cooler". A failure probability on each heading was given from a fragility curve as a function of reaction intensity or by assumptions based on conservative models. A core damage frequency, under the conditional of the loss of off-site power, was conservatively evaluated around 10
/year. A key heading in the event tree with large effect on the frequency is the intactness of the external diesel fuel tanks.
Okano, Yasushi; Yamano, Hidemasa
Journal of Nuclear Science and Technology, 53(8), p.1224 - 1234, 2016/08
Times Cited Count：4 Percentile：36.53(Nuclear Science & Technology)A method to obtain a hazard curve of a forest fire was developed. The method has four steps: a logic tree formulation, a response surface evaluation, a Monte Carlo simulation, and an annual exceedance frequency calculation. The logic tree consists domains of forest fire breakout and spread conditions, weather conditions, vegetation conditions, and forest fire simulation conditions. The new method was applied to evaluate hazard curves of a reaction intensity and a fireline intensity for a typical location around a sodium-cooled fast reactor in Japan.
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.
Ohno, Shuji; Takata, Takashi; Tajima, Yuji*
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
From various kinds of sodium fire situations postulated in SFR plants, the present paper treats the sodium pool fire and subsequent heat transfer behavior in an important air atmosphere two-cell geometry as one of the important cell configuration conditions. The detailed analysis and investigation of sodium fire and thermal-hydraulics in horizontally arranged two cells with an opening between them are made both from experimental measurement and from numerical simulation with a multi-cell sodium fire analysis code SPHINCS.
Shimomura, Yusuke; Hanari, Akira*; Sato, Isamu*; Kitamura, Ryoichi
JAEA-Technology 2015-062, 47 Pages, 2016/03
JAEA-Technology-2015-062.pdf:1.85MB
In response to new standards for regulating waste management facilities, it was carried out impact assessment of forest fires on the waste management facilities existed in Oarai Research and Development Center of Japan Atomic Energy Agency. At first, a fire spread scenario of forest fires was assumed. The intensity of forest fires was evaluated from field surveys, forest fire evaluation models and so on. As models of forest fire intensity evaluation, Rothermel Model and Canadian Forest Fire Behavior Prediction (FBP) System were used. Impact assessment of radiant heat to the facilities was carried out, and temperature change of outer walls for the assumed forest fires was estimated. The outer wall temperature of facilities was estimated around 160
C at the maximum, it was revealed that it doesn't reach allowable temperature limit. Consequently, it doesn't influence the strength of concrete. In addition, a probability of fire breach was estimated to be about 20%. This report illustrates an example of evaluation of forest fires for the new regulatory standards through impact assessment of the forest fires on the waste management facilities.
Okano, Yasushi; Yamano, Hidemasa
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05
Numerical simulations of forest fire propagation and smoke transport were performed with sensibility studies to weather conditions, and the effect by the smoke on the air filter was quantitatively evaluated. Forest fire propagation simulations were performed using FARSITE code. A temporal increase of a forest fire spread area, a position of the frontal fireline, "reaction intensity" and "frontal fireline intensity" are obtained and used for the smoke transport simulations by ALOFT-FT where spatial distribution of PM2.5 and PM10 are evaluated. The total amount of particle matter at the air filter at the nuclear power plant is around several hundred grams per m
which is well below the operational limit of the air filter of 15 kg/m.
Ohno, Shuji
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 9 Pages, 2014/12
This paper presents the typical characteristics of sodium combustion and subsequent reaction heat transfer behaviors observed and investigated in sodium columnar leak and fire experiment which was conducted in an enclosed steel vessel with large inner volume of about 100 m
. Especially the experiment was carried out with the main focus on the burning phenomenon within a limited spatial area in the case of large sodium leak rate as well as on the multi-dimensional thermal-hydraulics both near a sodium burning zone and in a whole region in the vessel. The investigated experimental results show us that the sodium combustion of columnar leak and its splashed droplets would lead to important oxygen deficiency behavior near the burning region, and be followed by the limitation or saturation of maximum sodium burning rate.
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.
