Development and steady-state validation of the fine sub-channel code CUNLUN

<bold>Background</bold>The safety assessment of reactor cores and their components relies on thermal-hydraulic analysis. System codes are suitable for system-level analysis, sub-channel codes provide more detailed thermal-hydraulic parameter distributions but struggle to capture complex three-dimensional flows, and Computational Fluid Dynamics (CFD) codes offer high accuracy but are limited by computational cost and modeling complexity. Existing thermal-hydraulic analysis methods, due to insufficient accuracy and applicability, fail to meet the requirements of modern reactor design and safety analysis.<bold>Purpose</bold>This study aims to develop a sub-channel analysis method based on fine-grid control volumes, which incorporates a more complete mixing grid model and a fully three-dimensional numerical solution approach for steady-state thermal-hydraulic analysis of pressurized water reactor (PWR) cores.<bold>Methods</bold>Firstly, the traditional sub-channel control volumes were further subdivided into more than 16 control volumes that were suited to the flow characteristics, increasing resolution to capture fine flow patterns inside coolant channels without significantly adding computational burden. Then, the fuel rods were divided into spacer and bare rod regions, with distributed resistance methods separately modeling the spacer and bare rod regions, forming the computational source terms. Subsequently, three-dimensional fluid flow and heat transfer differential equations were established based on the refined control volumes. The differential equations were discretized using the finite volume method, and numerical solutions were obtained using the SIMPLE algorithm. The code was developed using C++ and named as CUNLUN. Finally, the accuracy and effectiveness of the CUNLUN code in predicting flow and heat transfer characteristics were validated using the OECD/NEA recommended MATiS-H and NESTOR international benchmark tests.<bold>Results</bold>The results indicate that: 1) Compared to international benchmark data, the code demonstrates high predictive accuracy, validating its precision and reliability in thermal-hydraulic analysis. 2) In the MATiS-H benchmark, the maximum average relative error of axial velocity at four different positions is 9.07%. The verification results for the NESTOR benchmark also exhibit good consistency, indicating that the CUNLUN code can accurately capture and represent the variations in flow characteristics both in the near-field (close to the mixing grid region) and far-field (away from the mixing grid region). 3) In the NESTOR benchmark, the maximum deviation of sub-channel center temperatures is 1.982 5 ℃, with a maximum relative error of 0.64%. 4) In the NESTOR benchmark, among 32 experimental measurement points, 24 points show a 69.08% improvement in prediction accuracy using the distributed resistance momentum source model, significantly outperforming cases without the model.<bold>Conclusions</bold>The developed code in this study provides effective design and research tools for the thermal-hydraulic analysis of PWR fuel assemblies, aiding in the analysis of coolant flow and heat transfer characteristics. It also offers valuable references and insights for the development of reactor core thermal-hydraulic analysis programs.