Investigating the time-dependent behavior of nuclear reactors during loss of flow accidents is necessary for safety assessment.Coupled neutronic/thermal-hydraulic codes are used to simulate a full three-dimensional mo...Investigating the time-dependent behavior of nuclear reactors during loss of flow accidents is necessary for safety assessment.Coupled neutronic/thermal-hydraulic codes are used to simulate a full three-dimensional model and predict the essential safety parameters.MCNP6/ANSYS-FLUENT17.2 coupled scheme is used in the present study to simulate a three-dimensional model for VVER-1000 assembly and analyze its behavior during a LOFA(loss of flow accident).Three LOFA scenarios are proposed to represent the failure of one,two or three of the coolant pumps.The influence of the accident on the reactivity and axial power distribution of the assembly is determined considering thermal-hydraulic feedbacks.Then the data obtained are provided to the thermal-hydraulic code to calculate the actual temperature values.The results of the study showed that the developed coupling scheme granted an actual and precise description of the axial behavior of the assembly during LOFA.The output data obtained from both neutronic and thermal-hydraulic calculations have a strong feedback effect;this demonstrated the effect of data exchange between codes to predict accurate values for the main safety parameters.Moreover,it revealed the importance of studying the detailed axial distribution of the safety parameters for the reactor assessment during accidents rather than taking average values in calculations.展开更多
文摘Investigating the time-dependent behavior of nuclear reactors during loss of flow accidents is necessary for safety assessment.Coupled neutronic/thermal-hydraulic codes are used to simulate a full three-dimensional model and predict the essential safety parameters.MCNP6/ANSYS-FLUENT17.2 coupled scheme is used in the present study to simulate a three-dimensional model for VVER-1000 assembly and analyze its behavior during a LOFA(loss of flow accident).Three LOFA scenarios are proposed to represent the failure of one,two or three of the coolant pumps.The influence of the accident on the reactivity and axial power distribution of the assembly is determined considering thermal-hydraulic feedbacks.Then the data obtained are provided to the thermal-hydraulic code to calculate the actual temperature values.The results of the study showed that the developed coupling scheme granted an actual and precise description of the axial behavior of the assembly during LOFA.The output data obtained from both neutronic and thermal-hydraulic calculations have a strong feedback effect;this demonstrated the effect of data exchange between codes to predict accurate values for the main safety parameters.Moreover,it revealed the importance of studying the detailed axial distribution of the safety parameters for the reactor assessment during accidents rather than taking average values in calculations.
