Neutronics and sub-channel thermal-hydraulics analysis of the Iranian VVER-1000 fuel bundle

In the present research, sub-channel thermal-hydraulic analysis of the Bushehr Nuclear Power Plant (BNPP) core, a Russian VVER-1000 type, at hot full power (HFP) and steady state BOC conditions is presented herein using modified COBRA-EN code. Required power distribution and hot channel factors were computed by our past neutronics calculations using MCNP-5 code. Due to increasing temperature, modification on the continuous-energy cross sections in the MCNP libraries are carried out using NJOY code. Maximum and average fuel temperature, enthalpy, void fraction, coolant temperature and density, coolant mass flow rate and pressure drop are calculated using different models. Thermal-hydraulics calculations of the most rated channel (hottest sub-channel), which is determined based on coupled neutronics-thermal hydraulics calculations are investigated and results in temperature, enthalpy, critical heat flux and MDNBR of the hottest sub-channel are found. Finally, our results are compared with analytical approaches and the reactor FSAR.