A Transient Neutronic-Thermalhydraulic Coupled Modelling Approach

Abstract

Simulating two-phase flow and forced-convection boiling heat transfer constitute important aspects in performing nuclear power-reactor safety analysis. In this work, a transient thermalhydraulic code named ARTHUR (Advanced Routines of Thermal-Hydraulics for Unsteady-states Reactors) was developed and coupled to an existing neutronic code, DONJON-3. To this aim, the flow equations based on the drift-flux model were discretised using a second-order finite-difference for the space domain and a first-order fully implicit method for the time. The flow model has been validated by comparing the simulations with experimental boiling two-phase flow data obtained in uniformly heated tubes. Also, the heat transfer equations for the central pin were discretised using the same discretisation scheme and the model was validated by comparing the simulations with analytical solutions. Finally, coupled neutronic-thermalhydraulic simulations of a simplified CANDU four-channel reactor have been done for several transient conditions.