Thermal-Hydraulic modelling and analysis of VVER-1200 reactor core

Abstract

A mathematical model has been developed to simulate the thermal–hydraulic performance of the VVER-1200 pressurized water reactor under normal operation. The energy equation and the heat conduction equation are solved analytically in order to predict the coolant, clad and fuel temperature distributions. The core active length is divided into axial regions and the fuel rod is divided into radial zones, nodal calculation is performed for three types of cooling channels; the core mean channel with mean core power, and both the mean and hot channels of the most intensive energy fuel assembly. Through this model, the heat flux leading to the Departure from Nucleate Boiling (DNB) as well as the Departure from Nucleate Boiling Ratio (DNBR) predicted at each axial node for each channel using EPRI correlation. High accuracy correlations and/or models valid under the reactor operating conditions are selected to estimate the heat transfer coefficients under single-phase forced convection, subcooled boiling and bulk boiling regimes. The vapor quality and void fraction are estimated for boiling regimes as well. The model is then used to simulate the VVER-1200 reactor performance under both the rated and conservative operational parameters where boiling is predicted in the hot channel under both cases. The results are verified against the available calculations in the literature for coolant temperature, clad temperature and void fraction, where a very good agreement is achieved. The developed model is considered an appropriate tool for independent thermal–hydraulic safety assessment of VVER-1200 reactor core under steady-state operation.