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Journal Articles

Sodium fire collaborative study progress; CNWG fiscal year 2020

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.

Journal Articles

Sodium fire collaborative study progress; CNWG fiscal year 2019

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.

JAEA Reports

Development of correlation of gaseous ruthenium transfer rate to condensed water in accident of evaporation to dryness by boiling of reprocessed high level liquid waste in Fuel Reprocessing Facilities

Yoshida, Kazuo; Tamaki, Hitoshi; Yoshida, Naoki; Amano, Yuki; Abe, Hitoshi

JAEA-Research 2017-015, 18 Pages, 2018/01

JAEA-Research-2017-015.pdf:3.08MB

An accident of evaporation to dryness by boiling of high level liquid waste is postulated as one of the severe accidents at a fuel reprocessing facility. It was observed at the experiments that a large amount of ruthenium (Ru) is volatilized and transfer to the vapor phase in the tank. The nitric acid and water mixed vapor released from the tank is condensed. Volatilized Ru is expected to transfer into the condensed water at the compartments in the building. Quantitative estimation of the amount of Ru transferred condensed water is key issues to evaluate the reduction the amount of Ru through leak path in the facility building. This report presents that a correlation has been developed for Ru transfer rate to condensed water with vapor condensing rate based on the experimental results and additional thermal-hydraulic simulation of the experiments. Applicability of the correlation has been also demonstrated with the accident simulation of typical facilities in full-scale.

Journal Articles

Examination of $$^{131}$$I and $$^{137}$$Cs releases during late phase of Fukushima Daiichi NPP accident by using $$^{131}$$I/$$^{137}$$Cs ratio of source terms evaluated reversely by WSPEEDI code with environmental monitoring data

Hidaka, Akihide; Yokoyama, Hiroya

Journal of Nuclear Science and Technology, 54(8), p.819 - 829, 2017/08

AA2016-0500.pdf:0.44MB

 Times Cited Count:9 Percentile:75.2(Nuclear Science & Technology)

To clarify what happened during the Fukushima accident, the phenomena within RPV and the discussion of ties with the environmental monitoring are very important. However, the previous study has not necessarily advanced until the present that passed almost six years from the accident. The present study investigated $$^{131}$$I and $$^{137}$$Cs release behaviors during the late phase of the accident based on $$^{131}$$I/$$^{137}$$Cs ratio of the source terms that were recently evaluated backward by WSPEEDI code based on environmental monitoring data. The $$^{131}$$I release from the contaminated water in the basement of 1F2 and 1F3 reactor buildings was evaluated to be about 10% of $$^{131}$$I source term. The increase in $$^{137}$$Cs release from March 21 to 23 and from March 30 to 31 could be explained by the release of CsBO$$_{2}$$ which is formed as a result of chemical reactions of Cs with B$$_{4}$$C due to re-ascension of the core temperature caused by slight shortage of the core cooling water.

Journal Articles

Characteristics of severe accidents of Reduced-Moderation Water Reactor (RMWR)

Yonomoto, Taisuke; Akie, Hiroshi; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao; Iwamura, Takamichi

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10

Reduced-Moderation Water Reactor (RMWR) is a light-water cooled high-conversion reactor that is being developed by JAERI with collaboration from the Japanese industries. Since RMWR utilizes the highly enriched plutonium, the safety concern for RMWR includes the possibility of recriticality during severe accidents as is the case with the liquid metal cooled fast breeder reactor. In order to clarify this concern, characteristics of severe accidents of RMWR are analyzed in this study. The results obtained so far indicate that (1) the mechanical impact of recriticality in the core, if occurs, is supposed to be insignificant due to the absence of water, (2) the mixture of the fuel and cladding debris in the lower plenum does not cause recriticality when they are well mixed and distributed flatly, and (3) if requires, the installation of neutron-absorption material with realistic geometry can effectively prevent recriticality in the lower plenum even for the conservatively-assumed spherical accumulation of core debris.

Journal Articles

Analysis of pressure- and temperature- induced steam generator tube rupture during PWR severe accident initiated from station blackout

Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00

Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.

Journal Articles

Numerical analysis on ingress of coolant event in vacuum vessel using modified TRAC-BF1

Kurihara, Ryoichi; Ajima, Toshio*; Ueda, Shuzo; Seki, Yasushi

Journal of Nuclear Science and Technology, 38(7), p.571 - 576, 2001/07

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Depressurization analyses of PWR station blackout with MELCOR 1.8.4

Antariksawan, A. R.*; Hidaka, Akihide; Moriyama, Kiyofumi; Hashimoto, Kazuichiro*

JAERI-Tech 2001-011, 116 Pages, 2001/03

JAERI-Tech-2001-011.pdf:4.02MB

no abstracts in English

JAEA Reports

Vectorization, parallelization and porting of nuclear codes (Porting); Progress report fiscal 1999

Kawasaki, Nobuo*; Nemoto, Toshiyuki*; Kawai, Wataru*; Ogasawara, Shinobu*; Ishizuki, Shigeru*; Kume, Etsuo; Yatake, Yoichi*; Adachi, Masaaki*

JAERI-Data/Code 2000-039, 134 Pages, 2001/01

JAERI-Data-Code-2000-039.pdf:4.32MB

no abstracts in English

JAEA Reports

Analysis of steam generator tube rupture as a severe accident using MELCOR1.8.4

H.Yang*; Hidaka, Akihide; Sugimoto, Jun

JAERI-Tech 99-013, 97 Pages, 1999/03

JAERI-Tech-99-013.pdf:3.24MB

no abstracts in English

Journal Articles

Comparative study of source terms of a BWR severe accident by THALES-2, STCP and MELCOR

Hidaka, Akihide; *; Soda, Kunihisa; Muramatsu, Ken; Sakamoto, Toru*

ANS Proc. of the 1992 National Heat Transfer Conf., p.408 - 416, 1993/00

no abstracts in English

Oral presentation

Influence of iodine chemistry in accumulated water at reactor buildings on late phase source term at Fukushima NPP accident

Hidaka, Akihide

no journal, , 

During core cooling at Fukushima Daiichi NPP accident, large amount of contaminated water was accumulated in the basements of reactor buildings at Units 1 to 4. The estimated ratios of $$^{131}$$I and $$^{137}$$Cs quantities in water to the core inventories are 0.51%, 0.85% at Unit 1, 74%, 38% at Unit 2 and 26%, 18% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere. Many evaluations for I-131 release have been performed so far by MELCOR or the reverse estimation with SPEEDI. The SPEEDI reverse predicted significant release until March 26 while no prediction in MELCOR after March 17. The present study showed that iodine release from accumulated water due to radiolytic conversion from I$$^{-}$$ to I$$_{2}$$ and gas-liquid partition of I$$_{2}$$ may explain the release between March 17 and 26. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks.

Oral presentation

Development of dynamic PRA using multi-fidelity models

Zheng, X.; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu

no journal, , 

no abstracts in English

13 (Records 1-13 displayed on this page)
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