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Due to the complex structure of the LFR fuel rod bundle, refined CFD simulation consumes significant computing resources and time, making it difficult to meet the immediate needs of engineering design and research. This paper introduces a method that integrates the sub-channel level thermal-hydraulic analysis code CorTAF-LBE with the BPNN algorithm to analyze the thermal-hydraulic phenomena within the LFR core. The method has been validated by refined simulation results of the THEADES test section, which indicate that the error is within an acceptable range. The feasibility of its application in LFR fuel assemblies was also evaluated. The flow characteristics and temperature distribution in unwrapped LFR rod bundles are investigated. Due to differences in fuel distribution and flow area, the temperature in the central channel is much higher than that in the side and trapezoidal channels. Three accident conditions, including inlet blockage, center blockage, and outlet blockage, were simulated. The corresponding peak temperatures of the fuel cladding were 670 K, 767 K, and 860 K, respectively, representing an increase of approximately 170 K compared to the nominal conditions. The analysis results effectively demonstrate the 3D temperature distribution of the fluid in the rod bundles, which is significant for reactor design and safety analysis of LFRs. [ABSTRACT FROM AUTHOR]

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]

Inter-wrapper flow (IWF) is a typical thermal hydraulic phenomenon that exists in narrow gaps among the adjacent assemblies of a liquid metal fast reactor (LMFR) that can dissipate heat from assemblies and transfer heat among adjacent assemblies. Moreover, when blocked or reduced flow accidents occur, the IWF has a significant effect on reactor safety, especially reducing the peak cladding temperature. In this study, a computational fluid dynamics (CFD) model of the Karlsruhe liquid metal laboratory IWF experimental setup (called MYRRHA design) is built. The model contains three fuel assemblies with seven rods using lead–bismuth eutectic as working fluid. The new mesh generation method, correlation of turbulence model, and fluid–solid coupling heat transfer were considered in the CFD model. The established model was initially validated, and the simulation results were found to be in good agreement with experimental data. The influence of IWF on the thermal hydraulic characteristics within the rod bundle was revealed. The results indicated that the IWF rate can affect the heat transfer capability of the bundle under. The amount of heat removed increased by up to 10% with the IWF rate in a single bundle. The occurrence of IWF reduces the wall surface temperature and appropriately balances the temperature of adjacent bundles; however, it cannot effectively reduce the peak temperature of the cladding. The complex heat transfer characteristics and flow mechanism under the influence of IWF and wire mixing were revealed. The work presented in this paper represents an important step toward the IWF research on blocked flow scenarios. • A CFD model of a typical three bundles with IWF in LFR is developed. • The three dimensional flow and heat transfer features of the IWF is investigated. • The IWF effects the distribution of coolant temperature and velocity in the bundles. • The heat dissipation capacity increases as the IWF rate increases in LFR. [ABSTRACT FROM AUTHOR]

• The three-dimensional CFD model of ELSY primary system lead pool is established. • The calculation process is simplified by multiple reasonable UDF settings, and the steady state operation condition and reactor shutdown transient process was simulated. • The thermal hydraulic phenomena under rated and natural circulation conditions are analyzed. • This work can provide an appropriate and effective reference for the design and optimization of ELSY in the future. • A complete set of thermal stratification calculation method of LFR is formed, which can provide safety assessment advice for the same type of fast reactor. Lead-cooled fast reactor (LFR) adopts the liquid metal lead as the coolant and is beneficial to the fuel sustainability and high safety. LFR is suffering from the serious thermal stratification due to the pool type design and high operation temperature, leading to the threatening thermal stress on the structure. In this paper, the three-dimensional CFD model of ELSY lead pool is established and the thermal stratification phenomena under the forced and natural circulations are investigated. The results show that under steady-state conditions, ELSY has temperature stratification along the wall in the upper lead pool where the steam generator is located, and a high temperature concentration zone and a speed stagnant zone in the lower plenum. Under natural circulation conditions, when the water decay heat removal system (W-DHR), isolation condenser (IC) and reactor vessel air cooling system (RVACS) are put into operations, there is also obvious temperature stratification in the steam generator area. This paper puts forward some corresponding optimization suggestions on the problems in the simulation. The above results can provide an appropriate and effective reference for the design of ELSY in the future. [ABSTRACT FROM AUTHOR]

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation. [ABSTRACT FROM AUTHOR]

• A new analysis method of liquid metal mixing in Gen-IV nuclear system is proposed. • The LES model for liquid metal mixing is developed and experimentally validated. • The effect of different jet structures on the mixing characteristics was analyzed. • The three-sided outlet head with offset hedge proved to be the optimal structure. Nuclear energy is essential in addressing the problem of climate change and energy crisis. Liquid metal mixing phenomenon at the core outlet of liquid metal fast reactors is an important issue in Gen-IV nuclear systems. In this paper, the experimental and numerical analysis of the liquid metal mixing process in complex jets were carried out based on the structure of fuel assembly heads in sodium-cooled fast reactors. The liquid metal mixing characteristics were obtained in the experiment. And a high fidelity LES model was developed and validated experimentally. The impact of different head structures on the mixing process was analyzed and an index of liquid metal temperature mixing degree was newly proposed and compared with previous formulas. Result shows that, the temperature mixing region can be divided into five zones along the flow direction. Compared to other three structures, the three-sided outlet head with offset hedge has the lowest fluctuations of velocity and temperature in the upstream of the mixing flow. Also, the three-sided outlet head with offset hedge has the best temperature mixing ability, and the index is two times larger than that of the prototype. The three-sided outlet head with offset hedge proved to be the optimal structure. [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]