Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger; Project overview

Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

Emura, Yuki; Takai, Toshihide; Kikuchi, Shin; Kamiyama, Kenji; Yamano, Hidemasa; Yokoyama, Hiroki*; Sakamoto, Kan*

Journal of Nuclear Science and Technology, 10 Pages, 2023/00

https://doi.org/10.1080/00223131.2023.2287109

Thermophysical properties of molten FeO, (FeO)-(ZrO) and (FeO)-(UO)

Kondo, Toshiki; Toda, Taro*; Takeuchi, Junichi*; Kargl, F.*; Kikuchi, Shin; Muta, Hiroaki*; Oishi, Yuji*

Journal of Nuclear Science and Technology, 59(9), p.1139 - 1148, 2022/09

https://doi.org/10.1080/00223131.2022.2032445

no abstracts in English

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.

Prediction of RPV lower structure failure and core material relocation behavior with MPS method (Contract research)

Yoshikawa, Shinji; Yamaji, Akifumi*

JAEA-Research 2021-006, 52 Pages, 2021/09

JAEA-Research-2021-006.pdf:3.89MB

In Fukushima Daiichi Nuclear Power Station (referred to as "FDNPS" hereafter) unit2 and unit3, failure of the reactor pressure vessel (RPV) and relocation of some core materials (CRD piping elements and upper tie plate, etc.) to the pedestal region have been confirmed. In boiling water reactors (BWRs), complicated core support structures and control rod drive mechanisms are installed in the RPV lower head and its upper and lower regions, so that the relocation behavior of core materials to pedestal region is expected to be also complicated. The Moving Particle Semi-implicit (MPS) method is expected to be effective in overviewing the relocation behavior of core materials in complicated RPV lower structure of BWRs, because of its Lagrangian nature in tracking complex interfaces. In this study, for the purpose of RPV ablation analysis of FDNPS unit2 and unit3, rigid body model, parallelization method and improved calculation time step control method were developed in FY 2019 and improvement of pressure boundary condition treatment, stabilization of rigid body model, and calculation cost reduction of debris bed melting simulation were achieved in FY2020. These improvements enabled sensitivity analyses of melting, relocation and re-distribution behavior of deposited solid debris in RPV lower head on various cases, within practical calculation cost. As a result of the analyses of FDNPS unit2 and unit3, it was revealed that aspect (particles/ingots) and distribution (degree of stratification) of solidified debris in lower plenum have a great impact on the elapsed time of the following debris reheat and partial melting and on molten pool formation process, further influencing RPV lower head failure behavior and fuel debris discharging behavior.

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 2; Kinetic study on eutectic reaction process between stainless steel with low boron carbide concentration and stainless steel

Kikuchi, Shin; Takai, Toshihide; Yamano, Hidemasa; Sakamoto, Kan*

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 9 Pages, 2021/08

https://doi.org/10.1115/ICONE28-62252

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (BC) and stainless steel (SS) may occur. Thus, behavior of BC-SS eutectic melting is one of the phenomena to evaluate the core disruptive accidents in SFR. In this study, the reaction experiments using SS crucibles and the pellets of SS with low BC concentration as samples were performed to simulate the state of the reaction interface in which the eutectic reaction and interdiffusion of BC-SS have progressed to a certain extent. It was revealed that the rate constants of eutectic reaction between SS and SS with low BC concentration are smaller than that of BC-SS eutectic reaction at high temperatures.

Experiments of self-wastage phenomena elucidation in steam generator tube of sodium-cooled fast reactor

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12

https://doi.org/10.3327/taesj.J19.012

Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.

Now is the time of fast reactor

Negishi, Hitoshi; Kamide, Hideki; Maeda, Seiichiro; Nakamura, Hirofumi; Abe, Tomoyuki

Nihon Genshiryoku Gakkai-Shi ATOMO, 62(8), p.438 - 441, 2020/08

https://doi.org/10.3327/jaesjb.62.8_438

Prototype Fast Breeder Reactor, Monju, was under decommission since April, 2018. It is the first time for Japan to make a sodium cooled reactor into decommission. It is significant work and will take 30 years. The Monju has provided wide spectrum and huge amount of findings and knowledge, e.g., design, R&D, manufacturing, construction, and operation up to 40% of full power over 50 years of development history. It is significant to utilize such findings and knowledge for the development and commercialization of a fast rector in Japan.

Phase 1 code assessment of SIMMER-III; A Computer program for LMFR core disruptive accident analysis

Kondo, Satoru; Tobita, Yoshiharu

JAEA-Research 2019-009, 382 Pages, 2020/03

JAEA-Research-2019-009.pdf:22.82MB

The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.

Development of a fast reactor and related thermal hydraulics studies in Japan

Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.2095 - 2107, 2019/08

Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures. This paper describes the progress of research and development related to safety enhancement and the severe accident countermeasures. A volcanic PRA methodology was developed for the proper consideration of external hazards. Water and sodium experiments were carried out for the decay heat removal in a core disruptive accident (CDA), and also thermal hydraulic interactions between the core and upper plenum where dipped heat exchanger was operated. In order to elucidate the behavior of molten fuel during CDA, basic experiments of core melt fragmentation in deep and shallow sodium pools were carried out. X-ray visualization showed the liquid column of molten steel was intensively fragmented nearly simultaneously with a rapid expansion of sodium vapor.