Abstract
Thermal hydraulic analysis of the steam generators as one of the main components of the power cycle in pressurized water reactor (PWR) is crucial in the design and safety of the nuclear power plants. Two phase flow field simulation near the tube bundles is important in obtaining logical numerical results however the complexity of the tube bundles due to geometry and arrangement makes the numerical analysis complicated. In this research tube bundle has been assumed as the porous media and the outlet boundary condition as the one of the main challenge in these kind of simulations has been optimized according to similar researches. In order to adjust and tune the available computational fluid dynamic (CFD) code, pressure drop of the typical kettle reboiler tube bundle in two various heat fluxes and vapor volume fraction distribution in VVER 1000 steam generator in normal operation have been investigated. The typical transient mode of the VVER 440 steam generator, loss of feed water accident, has been studied eventually. It was observed that obtained vapor volume fraction can predict experimental data with more accuracy than the similar researches and would be increased with the elevation during the accident. On the other hand, pressure drop and level of the feed water value reduces through time and show good adoption with the measurements.
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