Full scope thermal-neutronic analysis of LOFA in a WWER-1000 reactor core by coupling PARCS v2.7 and COBRA-EN

Highlights: • Developing a package for thermal-neutronic simulation of NPP core. • External explicit parallel Coupling of PARCS, COBRA-EN and WIMS codes. • Steady-state convergence calculations. • Simulation of LOFA in WWER-1000 reactor core. • Validation of results with Bushehr nuclear power plant's FSAR. - Abstract: A full scope thermal-neutronic couple has been developed to analyze WWER-1000 nuclear reactor core by using PARCS v2.7, COBRA-EN and WIMSD-5B. First, WIMSD-5B has been used to calculate macroscopic cross sections for each fuel assembly. Neutronic calculations have been done by PARCS v2.7 code to obtain multiplication factor and PPF distributions of all fuel assemblies in the core. COBRA-EN calculates thermal-hydraulic parameters for all the fuel assemblies. This couple can be used for steady-state and transient calculations. Neutronic accidents such as Rod Ejection Accident that its initial event is a neutronic event (control rod movement that yields to reactivity changes) and thermal-hydraulic accidents such as LOFA that its initial event is a thermal-hydraulic event (coolant flow rate decrease) can be analyzed by this couple. In this study, steady-state calculations and LOFA have been simulated by this couple in the reactor core. A new "steady-state convergence" method is introduced. Results have been compared to Bushehr Nuclear Power Plant FSAR. The results show great similarities with FSAR results; and confirm the ability of this package to simulate accident in the core; especially thermal-hydraulic accident that is out of the ability of PARCS code individually.