Coupling of neutronics and thermal-hydraulics codes for the simulation of reactivity insertion accident for LFR

Abstract

As one of the advanced reactors proposed by the Generation IV International Forum (GIF), Lead-bismuth eutectic (LBE) coolant cooled fast reactor has shown great advantages over the other reactors. To analysis this reactor, several steps are carried out covering the code development and the coupled analysis. Firstly, the physical property of lead-bismuth, heat transfer correlation, turbulent mixing and pressure drop models are implemented into the thermal-hydraulics code COBRA-YT, where the inlet flow rate boundary condition is calculated by Computational Fluid Dynamics (CFD). The neutronics model has seven neutron energy groups and six groups of delayed neutron precursors. The neutronics parameters such as cross-section and feedback coefficient are calculated by the DRAGON code with endfb7 library. Secondly, based on sub-channel code COBRA-YT and neutron code SKETCH-N, the coupled of neutronics and thermal-hydraulics code has been developed via Parallel Virtual Machine (PVM) software platform. COBRA-YT code performs the thermal-hydraulics calculation and transfers its results such as coolant density and fuel temperature to the neutronics code SKETCH-N to update the cross-section; then SKETCH-N carries out the neutron-physical simulation of the reactor and provides the power density to the thermal-hydraulics code COBRA-YT as boundary conditions. Finally, this coupled code platform is exploited to the lead-bismuth fast reactor design to simulate some transient and control rod withdrawal accidents. The results achieved so far indicates that an increase of the reactor power and cladding temperature occurs after the control rod withdrawal, but the thermal-hydraulics results exceed their design limits. Time step size effect and withdrawal velocity analysis are also discussed.