Kondratieva O. Heat transfer in coolant flow in channels of pressurized water nuclear reactors under sub- and supercritical pressures

The thesis is devoted to a study of fluid flow and heat transfer in the channels, modeling the channels of the nuclear reactor core of a rod type. Specifics of perspective reactors of the IV generation with supercritical water used as a coolant was analyzed. Effect of the inter-channel mixing on the flow regime in the fuel assembly was considered. Enthalpy of cross-flow between the cells of the rod bundle under subcritical pressure was studied experimentally. Regime parameters ranged from a single-phase flow up to the subcritical heat transfer in two-phase flows. Results for the mixing flow enthalpy in central and corner rods were analyzed. Influence of flow regime changeover on enthalpy variation was revealed. Computer simulations were performed based on the developed model, whose results include distributions of fluid flow (velocity, kinetic energy, energy dissipation rate) and heat transfer (temperature, heat transfer coefficient) parameters for supercritical water flow in a seven-rod fuel assembly.