Transient Reactor Core Behavior by Multi-Physics Approach with Neutron Kinetics, Fuel Performance and Sub-channel Thermal-Hydraulics Coupling

Abstract

Steam line break accident and reactivity-initiated accident in the nuclear reactor are representative Non-LOCA accidents in which thermal-hydraulics and fuel behavior are strongly coupled each other. Moreover, those accidents might yield a spatially asymmetric core status that cannot be calculated by conventional system-scale reactor safety analysis. Thus, the multi-scale and multi-physics approach based on sub-channel scaled resolution is applied in reactor core region in order to examine a realistic safety margin. In this study, full core pin-wise non-LOCA safety analysis have been performed. An activation of transient behavior is triggered by the system thermal-hydraulics code, MARS-KS, and core behavior is calculated by sub-channel thermal hydraulics code, CUPID-RV and neutron kinetics code, MASTER. Fuel behavior is calculated by the fuel performance code, FRATPRAN. For the realistic core simulation, the initial condition as the fuel burnup information was obtained by pin-wise steady thermal-mechanical code, GIFT. KAERI developed the MARU platform for easy assessment to simulate the MSMP approach in high-performance computing environment. Two accident scenarios for End-Of-Cycle burnup status are simulated. In SLB simulation, the detailed thermal behavior inside the reactor pressure vessel during the transient such asymmetric power distribution and occurrence of the void fraction at upper plenum. Furthermore, the enhancement of safety margin was quantitatively investigated by obtaining a realistic DNBR distribution. Other accident scenario is control rod ejection accident. By the coupled simulation, the safety margin is quantitatively investigated by comparing the stored energy during the transient.