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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


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

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, 8 Pages, 2023/00

 Times Cited Count:0 Percentile:0.05(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, 26 Pages, 2023/00

 Times Cited Count:0 Percentile:0.05(Nuclear Science & Technology)

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

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:1 Percentile:62.84(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:2 Percentile:83.13(Nuclear Science & Technology)

Journal Articles

France-Japan collaboration on thermodynamic and kinetic studies of core material mixture in severe accidents of sodium-cooled fast reactors

Quaini, A.*; Goss$'e$, S.*; Payot, F.*; Suteau, C.*; Delacroix, J.*; Saas, L.*; Gubernatis, P.*; Martin-Lopez, E.*; Yamano, Hidemasa; Takai, Toshihide; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

CEA and JAEA defined new sub-tasks under the current implementing arrangement: Kinetics of interaction in core material mixtures- Physical properties of core material mixtures, High temperature thermodynamic data for the UO$$_{2}$$-Fe-B$$_{4}$$C system, Experimental studies on B$$_{4}$$C-SS kinetics and B$$_{4}$$C-SS eutectic material relocation (freezing), B$$_{4}$$C/SS eutectic and kinetics models for SIMMER code systems, Methodology for the modelling of mixtures liquefaction kinetics. The paper describes major R&D results obtained in the France-Japan collaboration under the previous implementing arrangement as well as experimental and analytical roadmaps under the current arrangement.

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.

Journal Articles

Modelling and simulation of source term for sodium-cooled fast reactor under hypothetical severe accident; Primary system/containment system interface source term estimation

Onoda, Yuichi; John Arul, A.*; Klimonov, I.*; Danting, S.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 13 Pages, 2022/04

Journal Articles

Cesium chemistry in the LWR severe accident and towards the decommissioning of Fukushima Daiichi Nuclear Power Station

Osaka, Masahiko; Gou$"e$llo, M.*; Nakajima, Kunihisa

Journal of Nuclear Science and Technology, 59(3), p.292 - 305, 2022/03

 Times Cited Count:1 Percentile:0.01(Nuclear Science & Technology)

Research on the fission product chemistry made after the severe accident of the Fukushima Daiichi Nuclear Power Station were reviewed with focus on the Cesium chemistry in terms of two regimes, namely the accidental source term and the long-term source term via aqueous phase towards the decommissioning. For the accidental source term, Cs chemical interaction with Mo, B and Si were reviewed. Regarding the unique issue of long-term source term via aqueous phase, Cs penetration into concrete and fuel debris leaching were mentioned as the main sources of FPs. Efforts on the preparation of thermodynamic data for the Cs complex oxides were described. All these Cs chemical behaviors should be modelled and validated/verified through the analysis and evaluation of the actual samples including fuel debris that would be taken from the Fukushima Daiichi Nuclear Power Station in near future.

Journal Articles

Release behaviors of elements from an Ag-In-Cd control rod alloy at temperatures up to 1673 K

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi*

Nuclear Technology, 208(3), p.484 - 493, 2022/03

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

An Ag-In-Cd control rod alloy was heated in argon or oxygen at 1073-1673 K for 60-3600 s and the release behavior of the elements was examined. Complete liquefaction of the alloy occurred between 1123 and 1173 K, and elemental release was quite limited below the liquefaction temperature. In argon, almost all of the Cd content was released within 3600 s at $$>$$ 1173 K and within 60 s at $$>$$ 1573 K, while the released fractions of Ag and In were $$<$$ 3% and $$<$$ 8%, respectively. In oxygen, the release of Cd, which was quite small at temperatures up to 1573 K, drastically increased to $$sim$$ 30-50% at 1673 K for short periods. Releases of Ag and In were also small in oxygen under the examined conditions. Comparison with the experimental data suggests that conventional empirical release models may underestimate the Cd release at lower temperatures just after control rod failure in severe accidents.

Journal Articles

Flame structures and ignition thresholds of hydrogen jets containing sodium mist under various gas concentrations

Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 59(2), p.198 - 206, 2022/02

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

JAEA Reports

Effect of nitrous acid on migration behavior of gaseous ruthenium tetroxide into liquid phase

Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi

JAEA-Research 2021-011, 12 Pages, 2022/01


In boiling and drying accidents involving high-level liquid waste in fuel reprocessing plants, emphasis is placed on the behavior of ruthenium (Ru). Ru would form volatile species, such as ruthenium tetroxide (RuO$$_{4}$$), and could be released to the environment with coexisting gases, including nitric acid, water, or nitrogen oxides. In this study, to contribute toward safety evaluations of these types of accidents, the migration behavior of gaseous Ru into the liquid phase has been experimentally measured by simulating the condensate during an accident. The gas absorption of RuO$$_{4}$$ was enhanced by increasing the nitrous acid (HNO$$_{2}$$) concentration in the liquid phase, indicating the occurrence of chemical absorption. In control experiments without HNO$$_{2}$$, the lower the temperature, the greater was the Ru recovery ratio in the liquid phase. Conversely, in experiments with HNO$$_{2}$$, the higher the temperature, the higher the recovery ratio, suggesting that the reaction involved in chemical absorption was activated at higher temperatures.

Journal Articles

Experiments of melt jet-breakup for agglomerated debris formation using a metallic melt

Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Abe, Yutaka*

Nuclear Engineering and Design, 386, p.111575_1 - 111575_17, 2022/01

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

262 (Records 1-20 displayed on this page)