Abstract:Loss of coolant accident (LOCA) is an important type of nuclear power plant design basis accident.Large break accidents in the cold pipe section and the hot pipe section may lead to the occurrence of counter-current flow limitation (CCFL).During the refilling phase,the lower part of the reactor with high upward flow rate causes the injection water to be unable to be injected downwards into the core in time,resulting in the phenomenon of CCFL Commonly used thermal-hydraulic system programs such as RELAP 5 use special models to describe this phenomenon,but for different geometric structures,the program requires users to select different models and model parameters.A model study of the interaction mechanism of the phenomenon is carried out,and the Dukler test bench with straight pipe section as the structural feature and the Karei test bench with the rod bundle channel as the test section are modeled.Through comparative analysis with experimental data,the applicability of different CCFL models is studied.

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