Numerical simulation of transient boiling and void cross flow in non-uniformly heated 5 × 5 rod bundle

Abstract

A multi-dimensional numerical simulation has been conducted for a transient boiling experiment by applying rapid and non-uniform power to fluid in a 5×5 bundle channel simulated a BWR fuel assembly in atmospheric pressure. The experiment is to investigate thermal hydraulic behavior in applying transient power to subcooled reactor coolant in a stand-by mode of a nuclear power plant such as reactivity initiated accident (RIA). The numerical simulation has been conducted by TRACE (version 5.0 / patch 5) which is a nuclear system analysis code to solve thermal hydraulics of boiling two phase flow by a two fluid model. The bundle channel in the upward vertical direction has been simulated by a three dimensional component with the Cartesian coordinate system attaching heat structures to simulate heating rods for thermal conduction and heat transfer calculation. It has been revealed that the numerical result is possible to simulate the void cross flow through the sub-channel in the bundle channel qualitatively as observed in the experiment. However, there was room for improvement for a wall heat transfer model in TRACE because the numerical result is larger than the experimental result in terms of the rising rate of the rod temperature during the rapid initial increasing of applied power, and it has been inferred that the evaluation model for an onset of nucleate boiling in TRACE also has to be improved because the void initiation time in the numerical result was earlier than the experimental result. Therefore, TRACE has been optimized by modifying the two physical models described above. The numerical results with the modified TRACE have been in better agreement with the experimental results than those with the original TRACE. Using the modified TRACE, a numerical experiment has been conducted putting an inlet velocity and subcooling of fluid and applied power as parameters in addition to the experimental conditions. The results have shown that there is a square law correlation between a void cross flow and applied power to the fluid. The experiments and analyses in this study have elucidated the behavior of transient boiling two phase flow in a fuel assembly of a nuclear power plant at the transient event such as RIA.