Activities of the GIF safety and operation project of the sodium-cooled fast reactors

Yamano, Hidemasa; Tsige-Tamirat, H.*; Kang, S. H.*; Summer, T.*; Chenaud, M. S.*; Rozhikhin, E.*; Wang, J.*

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

The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 "Methods, Models and Codes", WP-SO-2 "Experimental Programs and Operational Experience", and WP-SO-3 "Studies of Innovative Design and Safety Systems". This paper reports various activities in 2019 within the SO project.

Development of density control technologies for MOX pellet using dry recycled powders

Nishina, Masahiro; Takeuchi, Kentaro; Murakami, Shinichi

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

no abstracts in English

Material data acquisition activities to develop the material strength standard for sodium-cooled fast reactors

Toyota, Kodai; Onizawa, Takashi; Wakai, Takashi; Hashidate, Ryuta; Kato, Shoichi

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

Investigation on natural circulation for decay heat removal in reactor vessel of sodium-cooled fast reactor

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*; Nakane, Shigeru*; Ishida, Katsuji*

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

In sodium-cooled fast reactors (SFRs), optimizing the design and operate decay heat removal systems (DHRSs) is important for safety enhancement against severe accidents that could lead to core melting. The natural circulation phenomena in a reactor vessel during operating a DHRS were clarified by conducting water experiments using a 1:10 scale experimental facility (PHEASANT) simulating the reactor vessel of loop-type SFRs. In this study, we investigated the natural circulation phenomena under conditions of operating the dipped-type DHX and RVACS using the results of temperature and particle image velocimetry (PIV) measurements, respectively. Furthermore, the effects of temperature fluctuation on the PIV measurement were quantitatively evaluated.

Progress in conceptual design of a pool-type sodium-cooled fast reactor in Japan

Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Miyagawa, Takayuki*; Uchita, Masato*; Suzuno, Tetsuji*; Endo, Junji*; Kubo, Koji*; Murakami, Hisatomo*; Uzawa, Masayuki*; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 11 Pages, 2022/04

The authors are carrying out conceptual design studies for a pool-type sodium-cooled fast reactor. There are main challenges such as measures against severe earthquake in Japan, thermal hydraulic in a reactor vessel (RV), a decay heat removal system design. When the JP-pool SFR of 650 MWe is installed in Japan, it shall be designed against the severe seismic conditions. Additionally, a newly three-dimensional seismic isolation system is under development.

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.

Development of integrated severe accident analysis code, SPECTRA for sodium-cooled fast reactor

Uchibori, Akihiro; Sonehara, Masateru; Aoyagi, Mitsuhiro; Takata, Takashi*; Ohshima, Hiroyuki

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

A new computational code, SPECTRA, has been developed for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors. The in-vessel thermal hydraulics module includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.

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.