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MYRRHA is the Gen IV heavy liquid metal cooled fast reactor (LMFR), which is a pool-type reactor system developed and designed by the Belgian Nuclear Research Center SCK·CEN. The primary cooling system of the MYRRHA operates with molten lead-bismuth eutectic (LBE) as the working fluid, which offers several advantages. The equipment failures of fourth generation nuclear facilities will cause serious economic and environmental consequences, especially when flow blockage occurs, the threat to reactor safety is enormous. Rapidly detecting the occurrence of flow blockage and initiating shutdown procedures can effectively reduce the consequences of the accident. In this paper, a coupled sub-channel/CFD multi-scale platform is developed by implementing the sub-channel code SACOS-PB to the CFD code FLUENT based on the domine decomposition and explicit numerical scheme. The flow characteristics and temperature distribution in the upper plenum of a lead-cooled fast reactor(LFR) design based on MYRRHA reactor were investigated. The temperature at the core outlet increases 110K under the flow blockage condition, and the monitoring point for flow blockage is most suitable to be arranged in the center of the assembly and below the height of 0.6 m. The analysis results well demonstrated the 3D flow and local phenomena of the fluid in the upper plenum, which is significant to reactor design and reactor safety analysis of LFRs. [ABSTRACT FROM AUTHOR]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]