Modified COBRA-EN code to investigate thermal-hydraulic analysis of the Iranian VVER-1000 core

Thermal-hydraulic analysis of a typical VVER-1000 core at steady-state condition, using COBRA-EN code, is presented herein. Required power distribution was computed by the WIMS-D4 and CITATION codes based on the neutronic calculations. Maximum and average fuel temperature, enthalpy, void fraction, coolant temperature and density, coolant mass flow rate and pressure drop are calculated using EPRI model. Thermal-hydraulic calculations of the most rated channel which is determined based on neutronic calculations results in temperature, enthalpy, critical heat flux and minimum DNBR (MDNBR) of the core hottest channel are investigated. The COBRA-EN code is modified in order to make a thermal-hydraulic analysis for the VVER reactor, and this is the main objective of the present article. Our results are compared with analytical approaches and the reactor FSAR.