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Yamano, Hidemasa; Kubo, Shigenobu; Tokizaki, Minako*; Nakamura, Hironori*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 12 Pages, 2022/10
Specific design features of advanced sodium-cooled fast reactors (SFRs) designed in Japan are a passive reactor shutdown system, a passive decay heat removal system (DHRS), and an in-vessel retention (IVR) concept against an anticipated transients without scram (ATWS) in design extension condition (DECs). The present paper describes numerical analysis methodologies for event sequences studied in Japan and some numerical analyses of DECs to show the effectiveness of the passive shutdown system against a typical ATWS and severe accident mitigation measures for the IVR of molten core. For the passive shutdown capability, the numerical analysis has demonstrated the effectiveness of a self-actuated shutdown system against a severe ATWS event, for which the temperature response time was separately evaluated by a computational fluid dynamics (CFD) code. A recently developed debris-bed cooling analysis methodology coupled with a CFD code and a debris-bed module has successfully simulated a three-dimensional coolant flow field near the debris bed with the passive DHRS capability in order to demonstrate the debris-bed coolability on a core catcher.
Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called "pump/diagrid link". Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Yamada, Yumi*; Tokizaki, Minako*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kubo, Shigenobu; Sakaba, Hiroshi*
no journal, ,
JSFR demonstration reactor employes SASS: Self Actuated Shutdown System as passive reactor shutdown system, probability of the loss of function is demanded through reliability evaluation to demoonstrate level 1PSA for this plant. Unreliability of SASS in representative abnormal transient without scram (ATWS) event was evaluated using developed method for reliability evaluation.
Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
no journal, ,
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture, which is a candidate of the design basis event in pool type SFR. Sensitivity has been studied for the parameters which are thought significantly to affect the detection of the pump/diagrid link rupture, such as initial reactor power (at rated and 30% power operation), reactivity feedbacks and so forth. The sufficiency of the function was confirmed by the analysis results that at least two signals are transmitted for the detection of the pump/diagrid link rupture, which is the development target of the plant protection system in pool type SFR.
Hasegawa, Takashi*; Kan, Taro*; Moriwaki, Hiroyuki*; Tokizaki, Minako*; Yamano, Hidemasa; Takano, Kazuya
no journal, ,
no abstracts in English
Tokizaki, Minako*; Tani, Akihiro*; Ando, Masato*; Onoda, Yuichi
no journal, ,
Decay heat removal characteristics were evaluated for the Pool-Type Sodium-Cooled Fast Reactor (600 MWe class power) under the categories of Anticipated Operational Occurrences (AOO) and Design Basis Accidents (DBA), and it is expected that the integrity of the core fuel and coolant boundary can be ensured in light of the safety criteria. In addition, from the viewpoint of robustness confirmation of decay heat removal function, decay heat removal characteristics of immersion type DRACS during independent operation were evaluated, and the prospect of core cooling by natural circulation was obtained.
