Development and verification of a general nuclear-thermal coupling code based on preCICE and OpenFOAM adapter

BackgroundThe safe operation of nuclear reactors relies on precise simulation of multi-physics couplings, including neutron transport, conjugate heat transfer, and fluid dynamics. Traditional customized coupling programs often suffer from low efficiency and insufficient accuracy, struggling to meet the demands of complex scenarios in advanced reactor designs, such as high-resolution digital twin simulations.PurposeThis study aims to develop a general-purpose 3D neutronics-thermal hydraulics coupling code with high precision, leveraging the open-source multi-physics coupling library preCICE and its adapter OpenFOAM-adapter. To enable accurate prediction of safety parameters and in-depth analysis of multi-physics interactions within reactor cores.MethodsA neutron transport module and a thermal-hydraulic module were integrated in this proposed framework. Firstly, a self-developed neutron transport solver based on the finite volume method was employed in the neutron physics module to validate against benchmark problems, and a 3D solid heat conduction model (laplacianFoam) together with a buoyant turbulent flow model (buoyantPimpleFoam) for detailed analysis of temperature and velocity fields were combined in the thermal-hydraulics module. Then, the conjugate heat transfer (CHT) module was enhanced through secondary development of the preCICE adapter to support volume coupling and bidirectional data exchange between neutron and thermal-hydraulics solvers. Finally, a typical pressurized water reactor (PWR) single-rod benchmark was used for validation, involving grid independence analysis and comparison of three data mapping methods: nearest-neighbor, nearest-projection, and radial basis function (RBF) interpolation.ResultsThe validation results demonstrate that the above-mentioned neutronics-thermal hydraulics coupling code accurately outputs key parameters, including 3D power distribution, neutron flux density, velocity fields, and temperature fields of PWR. Quantitative verification shows a relative error of less than 0.1% for the coolant outlet temperature and 0.14% for the maximum cladding temperature, satisfying the precision requirements of reactor design codes. Grid independence analysis confirms that a medium-fidelity thermal-hydraulic grid combined with a coarse neutron grid balances accuracy and computational efficiency. Among data mapping methods, nearest-neighbor mapping provides acceptable precision with the lowest computational cost, while RBF interpolation, though more accurate, incurs higher computational overhead.ConclusionsThe developed coupling code in this study breaks through the limitations of traditional customized coupling approaches, supporting heterogeneous grid configurations and large-scale parallel computing. It offers a robust tool for high-resolution neutronics-thermal-hydraulics coupling simulations, facilitating safety analysis, design optimization, and digital twin modeling of reactor cores.