A conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

Abstract

The authors are developing the design concept of the pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDG) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident. The design concepts of a reactor vessel (RV) and its internal structures have been investigated whether they could withstand severe seismic conditions in Japan and thermal loads. The design adopted enhanced RV support structure, enhanced conical-shaped core support structure, a thickened knuckle part of the RV, and a flat plenum separator with ribs. A three-dimensional steady-state thermal hydraulic analysis of the RV revealed that the temperature difference of the upper and lower surfaces of the flat plenum separator could be effectively reduced by installing layers of thermal insulation plates. The authors have also conducted a transient analysis of loss of flow type anticipated transients without scram events to evaluate the feasibility of a self-actuated shutdown system. Moreover, the configuration of the decay heat removal system has been investigated considering sufficient utilization of natural circulation capability of sodium coolant, heat removal capacity of each cooling system, conformance with design requirements, and recommendations of SDC and SDG such as diversity and redundancy of components.

Study on post-buckling crack propagation in thin-walled cylinders under dynamic cyclic load

“…Under seismic load, buckling is regarded as a critical failure mode of the Fast Reactor Main Vessel (FRMV) which is a thin-walled large-diameter cylindrical structure, as shown in Fig. 1 (Kubo et al, 2020). The FRMV is suspended from the pedestal of the reactor building such that the weight of core and internal structures is concentrated at the lower part of the vessel.…”

“…Thus, a stable crack propagation is Fig. 1 Schematic view of fast reactor structure (Kubo et al, 2020).…”

Development of core design optimization process by integrated analysis with neutronics and plant dynamics

“…The appropriate representative problem can efficiently progress the development of the core design optimization process. To apply the developed process to the practical core design of advanced SFRs, we investigated experiences in the wide range of core designs of fast reactors from 100 to 1500 MWe (Fabbris et al, 2016;Hourcade et al, 2011Hourcade et al, , 2013Ingremeau et al, 2011;JAEA, 2006JAEA, , 2023aJAEA, , 2023bKubo et al, 2020;Onoda et al, 2016;Power Reactor and Nuclear Fuel Development Corporation, 1980;Yamano et al, 2012a;Zeng et al, 2020) from the viewpoint of design objectives for optimization, design constraints, including trade-off boundary conditions to be satisfied, core design parameters, and safety evaluation events.…”

“…Table 2 lists the transient events regarding the safety evaluation of the core design studied in JOYO (JAEA, 2023a), MONJU (Power Reactor and Nuclear Fuel Development Corporation, 1980;Onoda et al, 2016), and the design investigations of the Advanced SFR (Kubo et al, 2020;Yamano et al, 2012a). Events are categorized as design basis events (DBE) comprising an anticipated operational occurrence (AOO), a design basis accident (DBA), and the beyond DBE.…”

Time-dependent change in occurrence rate of steam generator tube leak in sodium-cooled fast reactors—Phenix and BN-600

“…If the DHRS shares a part of the secondary heat transport system, which uses SG heat transfer tubes, the reliability of the DHRS depends on the occurrence rate of leak from the SG tubes. This type of DHRS is considered as a candidate in the conceptual design study of next-generation SFRs, which have been developed in Japan (Kubo et al, 2020). From this safety point of view, it is necessary to estimate the leak occurrence rate for the next-generation SFRs.…”