The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0.

• A neutron diffusion model is developed based on OpenFOAM. • The model is validated by the benchmark. • The nuclear reactor neutronic thermal–hydraulic coupling code CorTAF-2.0 is developed. • The neutronic thermal–hydraulic coupling phenomena in the full core were carried out using CorTAF-2.0. Numerical reactor technology is the development direction of advanced nuclear energy system simulation, which has the ability of high-fidelity multi-physical field coupling calculation. The thermal–hydraulic and neutron physics fields in reactor core exhibit strong feedback relationships. Establishing the neutronic thermal–hydraulic coupling calculation model of the whole core "pin-by-pin" is the key technology for the development of numerical reactors. In this study, the neutron diffusion analysis model and cross-section database have been developed based on the open source CFD platform OpenFOAM. The three-dimensional neutron diffusion analysis model based on the finite volume method has been validated by benchmark problems. By integrating the neutron diffusion analysis model and cross-section database into nuclear reactor core three-dimensional thermal–hydraulic characteristics analysis code CorTAF, the strongly coupled neutronic thermal–hydraulic coupling analysis code CorTAF-2.0 has been developed. The neutronic thermal–hydraulic coupling phenomena in the full core of AP1000 have been simulated using CorTAF-2.0, yielding the sub-channel level parameters distribution of the whole core. This work offers valuable references for reactor full-core neutronic thermal–hydraulic coupling analysis. [ABSTRACT FROM AUTHOR]