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Journal Articles

Numerical simulation technologies for safety evaluation in plant lifecycle optimization method, ARKADIA for advanced reactors

Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.

Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11

The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.

Journal Articles

Development of plant lifecycle optimization method, ARKADIA for advanced reactors

Uchibori, Akihiro; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Doda, Norihiro; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai; Ohshima, Hiroyuki

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 10 Pages, 2022/09

The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies of the ARKADIA-Design. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event.

Journal Articles

Development of methodology to evaluate mechanical consequences of vapor expansion in SFR severe accident transients; Lessons learned from previous France-Japan collaboration and future objectives and milestones

Bachrata, A.*; Gentet, D.*; Bertrand, F.*; Marie, N.*; Kubota, Ryuzaburo*; Sogabe, Joji; Sasaki, Keisuke; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

In the frame of France-Japan collaboration, one of the objectives is to define and assess the calculation methodologies, and to investigate the phenomenology and the consequences of severe accident scenarios in sodium fast reactors (SFRs). A methodology whose purpose is to assess the loadings of the structures induced by a Fuel Coolant Interaction (FCI) taking place in the sodium plenum of SFR has been defined in the frame of the collaboration between France and Japan during 2014-2019. The work progress will be spread over the period 2020-2024 and the main objectives and milestones will be introduced in the paper. The objective of studies is to comprehensively address the margin between the limit of integrity of the main vessel structures and the loadings resulting from severe accidents. For this purpose, the SIMMER mechanistic calculation code simulates core disruptive accident sequences in SFRs. A fluid structure dynamics tool evaluates this interaction i.e. EUROPLEXUS is used in CEA studies and AUTODYN tool is used in JAEA studies. In the paper, a benchmark study is described in order to illustrate the evaluation of vapour expansion phase in the hot plenum. To do that, joint input data are used on the basis of an ASTRID 1500 MWth core degraded state after the power excursion which leads to vapour expansion. The most penalizing case was evidenced in this study by suppressing the action of transfer tube in-core mitigation devices in SIMMER input deck and thus privileging the upward molten core ejection. Even if the most penalizing case was evidenced in this paper, no significant RV deformation was observed in both EUROPLEXUS and AUTODYN calculation results. The assumed mechanical energy was small for the core expansion phase.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 2; Assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00393_1 - 16-00393_10, 2017/04

The achievement of In-Vessel Retention (IVR) of the accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into the low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the present assessments, it should be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

no abstracts in English

Journal Articles

In-vessel retention of unprotected accident in a fast reactor; Assessment of material-relocation and heat-removal behavior in ULOF

Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 2; Assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06

no abstracts in English

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

Journal Articles

A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 Times Cited Count:25 Percentile:90.75(Nuclear Science & Technology)

Journal Articles

Safety evaluation of prototype fast-breeder reactor; Analysis of ULOF accident to demonstrate in-vessel retention

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Ito, Kei

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07

Journal Articles

Numerical study on sodium-water reaction mechanism in the gas phase using counter-flow reaction region

Sogabe, Joji; Takata, Takashi*; Yamaguchi, Akira*; Kikuchi, Shin; Ohshima, Hiroyuki

Journal of Nuclear Science and Technology, 49(11), p.1067 - 1077, 2012/11

 Times Cited Count:1 Percentile:10.14(Nuclear Science & Technology)

Sodium water reaction (SWR) is a design basis accident of a sodium-cooled fast reactor (SFR). In a steam generator (SG) of the SFR, when a heat transfer tube fails, highly pressurized water or water vapor will leak into liquid sodium resulting in a chemical reaction between sodium and water or water vapor. The mechanisms of the SWR are complicated and have not been fully elucidated. The authors have developed a numerical code and have carried out SWR experiments of a counter-flow diffusion flame (in gas phase). In this paper, the authors perform numerical simulations based on the experimental conditions to validate two chemical reaction models. In addition, sensitivity analyses are performed for various hydration numbers and water vapor flow velocities. It is founded that hydration reaction occurs somewhat in the gas-phase reaction and that influences of the water vapor flow velocity are not negligible mainly from the viewpoint of the reaction surface location.

Journal Articles

Sodium-water reaction elucidation with counter-flow diffusion flame experiment and its numerical simulation

Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Sogabe, Joji*; Deguchi, Yoshihiro*; Kikuchi, Shin

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05

Sodium-water reaction (SWR) is a design basis accident of a sodium fast reactor (SFR). A breach of the heat transfer tube in a steam generator (SG) results in contact of liquid sodium with water. The purpose of the present paper is to delineate the mechanism and process of the SWR by a counter-flow diffusion flame experiment and a numerical simulation.

Oral presentation

Investigation of counterflow diffusion reaction of sodium and water vapor in low-pressure condition

Sogabe, Joji*; Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Kikuchi, Shin; Deguchi, Yoshihiro*

no journal, , 

no abstracts in English

Oral presentation

Numerical study on sodium dynamics near reaction surface in counterflow of sodium and water vapor

Sogabe, Joji*; Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Kikuchi, Shin

no journal, , 

Evaluation of sodium-water reaction phenomena is one of important safety issues in sodium-cooled fast reactors. In this study, we have considered sodium dynamics near reaction surface based on the experimental results, including concentration distributions of sodium, water vapor, and aerosol measured in counter-flow diffusion flame experiment of sodium and water.

Oral presentation

Validation analyses of a debris bed heat removal model using D series experiments

Sogabe, Joji; Koyama, Kazuya*; Tobita, Yoshiharu

no journal, , 

no abstracts in English

Oral presentation

Oral presentation

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 6; Assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Wada, Yusaku; Suzuki, Toru; Tobita, Yoshiharu

no journal, , 

no abstracts in English

Oral presentation

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 11; Best estimate and uncertainty assessment of PAMR/PAHR phase in ULOF

Sogabe, Joji; Wada, Yusaku*; Suzuki, Toru*; Tobita, Yoshiharu

no journal, , 

no abstracts in English

Oral presentation

Development of advanced reactor knowledge- and AI-aided design integration approach through the whole plant lifecycle, ARKADIA, 6; Development of ex-vessel analysis model in ARKADIA-Safety for safety evaluation

Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*

no journal, , 

The ARKADIA-Safety for safety evaluation and design optimization considering severe accident has been developed. This paper presents validation of sodium fire analysis model and integration of ex-vessel analysis model for improvement of the evaluation method.

Oral presentation

Development of advanced reactor knowledge- and AI-aided design integration approach through the whole plant lifecycle, ARKADIA, 10; Development of safety design optimization methods in ARKADIA-Safety

Okano, Yasushi; Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Ishida, Shinya; Sogabe, Joji; Takata, Takashi*

no journal, , 

no abstracts in English

Oral presentation

Model development for blockage of disrupted core materials in flow path

Sogabe, Joji; Kamiyama, Kenji; Tobita, Yoshiharu; Okano, Yasushi

no journal, , 

During severe accidents by an anticipated transient without scram, it is important to evaluate multiphase multi-component flow behavior, when a part of the disrupted core material is discharged outside the disrupted core region through control rod guide tubes. In particular, the blockage behavior of the disrupted core material in a flow path is an important phenomenon that affects the amount of relocated fuels (the fuel discharged outside the disrupted core region and the fuel remaining in the disrupted core region). A fast reactor safety analysis code, SIMMER, is currently being developed for application to the post-accident material relocation (PAMR) phase. In the paper, aiming at actual reactor analyses for the PAMR phase of the SIMMER code, a model for the blockage in the flow path for possible phenomena in the PAMR phase. The model improves the applicability of the SIMMER code to the PAMR phase on the actual reactors.

21 (Records 1-20 displayed on this page)