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The Lead-bismuth cooled Fast Reactors (LFR) utilizes wrapped hexagonal fuel assemblies arranged in a honeycomb pattern. The inter-wrapper flow (IWF) between adjacent assemblies has a significant influence on the core thermal-hydraulics. In this paper, the LFR core assembly analysis model, inter-wrapper flow and heat transfer model, and multi-scale coupling analysis model were developed and were implemented into the previously self-developed subchannel-level three-dimensional thermal hydraulic analysis code CorTAF-LBE for whole core of LFR, which is based on the OpenFOAM computational fluid dynamics simulation platform. The inter-wrapper flow and heat transfer model was validated against the KALLA-IWF experiment. Whole core steady state operating condition and low flow rate condition simulations were performed with reference to the 19-rod bundle geometry from KALLA laboratory and typical LFR core assembly arrangement. The whole core steady state distribution laws of key thermal-hydraulic parameters were obtained, and the influence of low flow rate operation on safety parameters such as cladding and pellet temperatures was analyzed. Under low flow rate operating conditions, the IWF heat transfer was accounted for 4.87% of the core thermal power, which was 315.6 kW higher than that under normal operating conditions, playing an important role in core heat removal and safe operation. This work provides important insights into LFR core design and the assessment of the role of IWF under low flow rate conditions. • The Multi-scale coupling analysis model is established and implanted into the CorTAF-LBE code. • The steady-state thermal-hydraulic parameter distribution at whole core scales is analyzed. • The effect of inter-wrapper flow on core heat transfer at low flow rate condition is evaluated. [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]

• This study utilizes the increasing flow regime experimental data to extend the prediction range and improve the accuracy. • Integrated flow regime prediction models considering vertical and horizontal flow direction were developed and evaluated. • The MLP, RF, and SVM algorithms were utilized to develop the flow regime prediction model respectively. • The flow regime prediction model shows better performance than existing models in three directions respectively. The nuclear reactor core is the pivotal component in a nuclear power plant to generate and transfer heat, so accurate prediction of reactor core flow and heat transfer characteristics is a crucial problem for the reactor system design. Flow regime is closely related to the thermal–hydraulic characteristics of two-phase fluid. Still, flow regime models used in thermal–hydraulic calculation codes rely heavily on earlier experimental data, featuring a narrow application range and unsatisfactory accuracy. To extend the prediction range and improve the accuracy of flow regime prediction by making full use of the increasing experimental data, integrated flow regime prediction models considering vertical flow direction for light water reactor (LWR) and horizontal flow direction for Canadian Natural Deuterium Uranium (CANDU) were developed and evaluated based on three standard machine learning algorithms in this study. Firstly, the experimental data of horizontal and vertical flow collected from the literature was modified and preprocessed to be training data. Then, the multi-layer perceptron (MLP) algorithm, random forest (RF) algorithm, and support vector machine (SVM) algorithm were utilized to develop the flow regime prediction model respectively. The performance of the three models was evaluated and compared and the results showed that the flow regime prediction model based on RF was the optimal model with higher prediction and more efficiency than the other two models. Finally, the flow regime prediction model was compared with existing models in three directions respectively, which indicated that the range and accuracy of the model based on RF were superior to the existing models. This study provides an implantable and scalable approach for flow regime prediction, and the application range and accuracy of flow regime prediction can be continuously improved with updating training data. [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]