Investigation of Water-Vapor Two-Phase Flow Characteristics in a Tight-Lattice Core by Large-Scale Numerical Simulation, (I)

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Although subchannel codes are used for the thermal-hydraulic analysis of fuel bundles in nuclear reactors from the former, many composition and empirical equations based on experimental results are needed to predict the two-phase flow behavior. When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.
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