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Journal Articles

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05

 Times Cited Count:1 Percentile:63.33(Nuclear Science & Technology)

Journal Articles

Cohesive/Adhesive strengths of CsOH-chemisorbed SS304 surfaces

Li, N.*; Sun, Y.*; Nakajima, Kunihisa; Kurosaki, Ken*

Journal of Nuclear Science and Technology, 61(3), p.343 - 353, 2024/03

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

During the Fukushima Daiichi nuclear power plant (1F) accident, an overwhelming amount of the cesium remaining in the pressure vessel could have been deposited onto 304 stainless steel (SS304) steam separators and dryers, both with large surface areas. During 1F's decommissioning, the deposited cesium is a safety hazard as it can generate radioactive dust. However, the cohesive and adhesive strengths of CsOH-chemisorbed oxide scales are yet to be defined. In this study, we investigated how CsOH-chemisorption affects the cohesive and adhesive strengths between oxide scales and SS304 substrates with a scratch tester. The scratch test results revealed that the cohesive strengths of the oxide scales decreased after CsOH-chemisorption, while adhesive failure could not be reached.

Journal Articles

MAAP code analysis focusing on the fuel debris conditions in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 3

Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Shimomura, Kenta; Cibula, M.*; Mizokami, Shinya*

Nuclear Engineering and Design, 414, p.112574_1 - 112574_20, 2023/12

Journal Articles

Main outputs from the OECD/NEA ARC-F Project

Maruyama, Yu; Sugiyama, Tomoyuki*; Shimada, Asako; Lind, T.*; Bentaib, A.*; Sogalla, M.*; Pellegrini, M.*; Albright, L.*; Clayton, D.*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4782 - 4795, 2023/08

Journal Articles

Estimation for mass transfer coefficient under two-phase flow conditions using two gas components

Nanjo, Kotaro; Shiotsu, Hiroyuki; Maruyama, Yu; Sugiyama, Tomoyuki; Okamoto, Koji*

Journal of Nuclear Science and Technology, 60(7), p.816 - 823, 2023/07

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

Journal Articles

Comprehensive analysis and evaluation of Fukushima Daiichi Nuclear Power Station Unit 3

Yamashita, Takuya; Honda, Takeshi*; Mizokami, Masato*; Nozaki, Kenichiro*; Suzuki, Hiroyuki*; Pellegrini, M.*; Sakai, Takeshi*; Sato, Ikken; Mizokami, Shinya*

Nuclear Technology, 209(6), p.902 - 927, 2023/06

 Times Cited Count:2 Percentile:84.55(Nuclear Science & Technology)

Journal Articles

Journal Articles

Double diffusive dissolution model of UO$$_{2}$$ pellet in molten Zr cladding

Ito, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*

Journal of Nuclear Science and Technology, 60(4), p.450 - 459, 2023/04

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

Journal Articles

Study on safety characteristics of a sodium-cooled fast reactor with negative void reactivity during initiating phase in severe accident

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04

JAEA Reports

Improvement of model for cesium chemisorption onto stainless steel in severe accident analysis code SAMPSON (Joint research)

Miwa, Shuhei; Karasawa, Hidetoshi; Nakajima, Kunihisa; Kino, Chiaki*; Suzuki, Eriko; Imoto, Jumpei

JAEA-Data/Code 2021-022, 32 Pages, 2023/01

JAEA-Data-Code-2021-022.pdf:1.41MB
JAEA-Data-Code-2021-022(errata).pdf:0.17MB

The improved model for cesium (Cs) chemisorption onto stainless steel (SS) in the fission product (FP) chemistry database named ECUME was incorporated into the severe accident (SA) analysis code SAMPSON for the more accurate estimation of Cs distribution within nuclear reactor vessels in the TEPCO's Fukushima Daiichi Nuclear Power Station (1F). The SAMPSON with the improved model was verified based on the analysis results reproducing the experimental results which were subjected to the modeling of Cs chemisorption behavior. Then, the experiment in the facility with the temperature gradient tube to simulate SA conditions such as temperature decrease and aerosol formation was analyzed to confirm availability of the improved model to the analysis of Cs chemisorption onto SS. The SAMPSON with the improved model successfully reproduced the experimental results, which indicates that the improved model and the analytical method such as setting a method of node-junction, models of aerosol formation and the calculation method of saturated CsOH vapor pressure can be applicable to the analysis of Cs chemisorption behavior. As the information on water-solubility of Cs deposits was also prerequisite to estimate the Cs distribution in the 1F because Cs can be transported through aqueous phase after the SA, the water-solubility of chemisorbed Cs compounds was investigated. The chemisorbed compounds on SS304 have been identified to CsFeO$$_{2}$$ at 873 K to 973 K with higher water-solubility, CsFeSiO$$_{4}$$ at 973 K to 1273 K and Cs$$_{2}$$Si$$_{4}$$O$$_{9}$$ at 1073 K to 1273 K with lower water-solubility. From these results, the water-solubility of chemisorbed Cs compounds can be estimated according to the SA analysis conditions such as temperature in the reactor and the CsOH concentration affecting the amount of chemisorbed Cs.

Journal Articles

Improvement of JASMINE code for ex-vessel molten core coolability in BWR

Matsumoto, Toshinori; Kawabe, Ryuhei*; Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Maruyama, Yu

Annals of Nuclear Energy, 178, p.109348_1 - 109348_13, 2022/12

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

The Japan Atomic Energy Agency extended the applicability of their fuel-coolant interaction analysis code JASMINE to simulate the relevant phenomena of molten core in a severe accident. In order to evaluate the total coolability, it is necessary to know the mass fraction of particle, agglomerated and cake debris and the final geometry at the cavity bottom. An agglomeration model that considers the fusion of hot particles on the cavity floor was implemented in the JASMINE code. Another improvement is introduction of the melt spreading model based on the shallow water equation with consideration of crust formation at the melt surface. For optimization of adjusting parameters, we referred data from the agglomeration experiment DEFOR-A and the under-water spreading experiment PULiMS conducted by KTH in Sweden. The JASMINE analyses reproduced the most of the experimental results well with the common parameter set, suggesting that the primary phenomena are appropriately modelled.

Journal Articles

Revolatilization of iodine by bubbly flow in the suppression pool during an accident

Nanjo, Kotaro; Ishikawa, Jun; Sugiyama, Tomoyuki; Pellegrini, M.*; Okamoto, Koji*

Journal of Nuclear Science and Technology, 59(11), p.1407 - 1416, 2022/11

 Times Cited Count:7 Percentile:89.36(Nuclear Science & Technology)

Journal Articles

The OECD/NEA Working Group on the Analysis and Management of Accidents (WGAMA); Advances in codes and analyses to support safety demonstration of nuclear technology innovations

Nakamura, Hideo; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*

Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10

Journal Articles

BWR lower head penetration failure test focusing on eutectic melting

Yamashita, Takuya; Sato, Takumi; Madokoro, Hiroshi; Nagae, Yuji

Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08

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

Journal Articles

Development of dynamic PRA methodology for external hazards (Application of CMMC method to severe accident analysis code)

Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07

Identifying accident scenarios that could lead to severe accidents and evaluating their frequency of occurrence are essential issues. This study aims to establish the methodology of the dynamic Probabilistic Risk Assessment (PRA) for sodium-cooled fast reactors that can consider the time dependency and the interdependence of each event. Specifically, the Continuous Markov chain Monte Carlo (CMMC) method is newly applied to the SPECTRA code, which analyzes the severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. Currently, a fault-tree model of air coolers of decay heat removal system is implemented as the CMMC method, and a series of preliminary analysis of the plant's transient characteristics under the scenario of volcanic ashfall has been conducted.

Journal Articles

Study on initiating phase of core disruptive accident (Validation study of SAS4A code for the unprotected transient overpower accident)

Ishida, Shinya; Fukano, Yoshitaka

Nihon Kikai Gakkai Rombunshu (Internet), 88(911), p.21-00304_1 - 21-00304_11, 2022/07

In previous studies, the reliability and validity of the SAS4A code was enhanced by applying Phenomena Identification and Ranking Table (PIRT) approach to the Unprotected Loss of Flow (ULOF). SAS4A code has been developed to analyze the early stage of Core Disruptive Accident (CDA), which is named Initiating Phase (IP). In this study, PIRT approach was applied to Unprotected Transient over Power (UTOP), which was one of the most important and typical events in CDA as well as ULOF. The phenomena were identified by the investigation of UTOP event progression and physical phenomena relating to UTOP were ranked. 8 key phenomena were identified and the differences in ranking between UTOP and ULOF were clarified. The code validation matrix was completed and an SAS4A model, which was not validated in ULOF, was identified and validated. SAS4A code became applicable to various scenarios by using PIRT approach to UTOP and the reliability and validity of SAS4A code were significantly enhanced.

Journal Articles

Status of the uncertainty quantification for severe accident sequences of different NPP-designs in the frame of the H-2020 project MUSA

Brumm, S.*; Gabrielli, F.*; Sanchez-Espinoza, V.*; Groudev, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; Bocanegra, R.*; Herranz, L. E.*; Berda$"i$, M.*; et al.

Proceedings of 10th European Review Meeting on Severe Accident Research (ERMSAR 2022) (Internet), 13 Pages, 2022/05

Journal Articles

Post-test analyses of the CMMR-4 test

Yamashita, Takuya; Madokoro, Hiroshi; Sato, Ikken

Journal of Nuclear Engineering and Radiation Science, 8(2), p.021701_1 - 021701_13, 2022/04

Journal Articles

Time-resolved 3D visualization of liquid jet breakup and impingement behavior in a shallow liquid pool

Kimura, Fumihito*; Yamamura, Sota*; Fujiwara, Kota*; Yoshida, Hiroyuki; Saito, Shimpei*; Kaneko, Akiko*; Abe, Yutaka*

Nuclear Engineering and Design, 389, p.111660_1 - 111660_11, 2022/04

 Times Cited Count:3 Percentile:63.91(Nuclear Science & Technology)

Journal Articles

French-Japanese experimental collaboration on fuel-coolant interactions in sodium-cooled fast reactors

Johnson, M.*; Delacroix, J.*; Journeau, C.*; Brayer, C.*; Clavier, R.*; Montazel, A.*; Pluyette, E.*; Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04

Fuel-coolant interactions in the event of molten fuel discharge to the lower plenum of a sodium cooled fast reactor is under investigation as part of a French-Japanese experimental collaboration on severe accidents. The MELT facility enables the X-ray visualisation of the quenching of molten core material jets in sodium at kilogram-scale. The SERUA facility, currently under preparation, is presented for the investigation of boiling heat transfer at elevated melt-coolant interface temperatures. In this article, the status of the collaboration using these facilities is presented.

275 (Records 1-20 displayed on this page)