Validation of the subchannel code LINDEN for void fraction simulation based on the NUPEC PWR tests (PSBT)

Subchannel analysis code is one of the most important thermal-hydraulic analysis tools for nuclear reactor cores. A new subchannel analysis code for PWRs named LINDEN code has been developed by China Nuclear Power Technology Research Institute (CNPRI), and it is also as a part of the self-reliant software package of CNPRI for reactor core design and safety analysis. The flow model of the LINDEN code has four conservation equations, and the four-equation flow model is actually a non-equilibrium drift flux model, which takes into account thermal disequilibrium of liquid phase and slip velocity between liquid and vapor phases under sub-cooled boiling condition. In addition, other auxiliary correlations are used for closing these four conservation equations, such as the heat transfer model between fuel rod and coolant, the turbulent mixing model between adjacent subchannels, etc. In this paper, the typical NUPEC PWR tests (PSBT) are used to evaluate the void fraction predicable ability of LINDEN. The LINDEN code is used to model and calculate three PSBT void fraction tests, including steady single-channel steady tests, steady 5×5 rod bundle tests and transient 5×5 rod bundle tests. The calculated void fraction results are compared with the measured results to validate that the LINDEN code has the capability to simulate the void fraction distribution under steady and transient test conditions with acceptable accuracy.