Development of a droplet breakup model for dry spacer grid in the dispersed flow film boiling regime during reflood transients

Abstract

The liquid droplet breakup process at the leading edge of a dry spacer grid during reflood transients is studied theoretically based on the experimental observations in the present work. Various rod bundle quenching experiments were carried out at the Rod Bundle Heat Transfer (RBHT) test facility to simulate the reflood transients of a Loss of Coolant Accident (LOCA) for Light Water Reactors. In the experiment, variation of the droplet size distributions across a spacer grid are measured and analyzed. The size of droplet generated downstream of a spacer grid is found to be affected by the condition of the spacer grid as well as the quench front propagation. In addition, the substantially reduced droplet size at downstream of the spacer grid indicates that droplet breakup takes place. Based on available experimental data obtained in the Rod Bundle Heat Transfer tests and in literature, a theoretical model incorporating two droplet groups (large and small droplet groups) is developed for the dry spacer grid droplet breakup by considering the conservation of liquid mass, energy transport, and including various important flow and heat transfer parameters including the Nusselt number, Reynolds number, Jakob number, Weber number, dimensionless radiation number, distance from the initial breakup point, incoming droplet size, and spacer grid blockage ratio. Moreover, a liquid droplet mass loss coefficient was proposed for small droplet group to specially account for the liquid mass evaporation due to interfacial heat transfer and radiation. The model so developed is found to be able to predict the RBHT data within 17% error with RMSE of 0.1060 and standard deviation of 0.0990 for the validation data set, showing significantly increased accuracy. Comparison made with other experimental results indicates that the current model developed correctly captures the mass and energy transport process during breakup. In addition, sensitivity analyses are performed to investigate the overall performance and relative importance of various parameters that are related to the breakup process in the current study. The current model developed can be incorporated into nuclear reactor thermal-hydraulic and safety analysis codes to enhance their capabilities of predicting the droplet behaviors and the fuel rod peak cladding temperature in the dispersed flow film boiling regime during reflood transients of postulated accident scenarios such as LOCA for light water reactors.