Modeling of control rod ejection transient for WWER-1000-model 446 using RELAP5m3.3/PARCSv2.6 coupled codes

Abstract

By using the Best Estimate (BE) method instead of conservative assumptions for the evaluation of reactor safety, significant economic considerations with optimal fuel burn-up could be obtained in addition to reactor safety.In this method, due to the detailed simulation and feedback considerations, special attention has been paid to the coupling of neutronic and thermo-hydraulic codes to achieve more reliable results. In this study, the Control Rod Ejection (CRE) transient has been simulated for Bushehr Nuclear Power Plant (BNPP) as a WWER-1000 power plant model 446 according to Final Safety Analysis Report (FSAR). CRE is a transient of Reactivity Initiated Accidents (RIA) category. In this study, the reactor thermo-hydraulic system has been simulated by RELAP5/mod3.3, while the neutron kinetic system of the reactor core has been simulated by the PARCSv2.6 code. These codes have been coupled utilizing Parallel Virtual Machine (PVM) interface software to consider the effects of thermal hydraulic and neutronic feedbacks. Thus, the power calculated by the PARCS code is used by the RELAP5 code and the obtained thermal hydraulic parameters are inserted to the PARCS code for macroscopic cross-section calculations.A computer program written by C++ has been used for the cycle execution of the WIMS code to produce the macroscopic cross-section library with the format required by the PARCS code. After the three-dimensional (3D) thermo-neutronic modeling of the reactor core, the Hot Zero Power (HZP) and Hot Full Power (HFP) versions of CRE transients, which have been considered in the plant’s FSAR, have been simulated. Some parameters such as power, inserted reactivity, primary side pressure, fuel temperature, and break mass flow have been investigated.An acceptable agreement could be observed between the calculated result of the coupled codes and FSAR data, confirming the validity of the proposed model. To evaluate the reactor safety more accurately, the effects of these events have locally been studied by investigating the spatial radial distribution of power in the two versions of the transients.