| The single-phase and two-phase heat transfer enhancement associated with flow through a nuclear fuel assembly grid spacer, which is of important concern on the cooling of fuel rods in nuclear power plants under abnormal conditions, is investigated. Grid spacers are present within reactor fuel assemblies to maintain the structural spacing of the fuel rods. These grid spacers also provide a flow blockage such that the flow within the assembly is disrupted, which causes the flow to redevelop and provides an enhancement in the local heat transfer. However, as the flow continues to evolve downstream of the grid, the heat transfer enhancement will decrease as the flow becomes fully developed. In the first part of the present work, single-phase steam cooling experiments conducted at the Rod Bundle Heat Transfer (RBHT) Test Facility located at The Pennsylvania State University are utilized to develop an upgraded correlation which models the single-phase local heat transfer enhancement. The improvement of the upgraded correlation over the models currently in use is then demonstrated.;The Pennsylvania State University and the United State Nuclear Regulatory Commission joined in creating a Rod Bundle Heat Transfer Test Facility to obtain experimental data that can be used to assess and improve the computer code model predictions. The facility is well suited for various experiments, including reflood heattransfer, two-phase level swell, single-phase steam cooling, and steam cooling experiments with droplet injection. The RBHT facility is well instrumented with sensitive differential pressure cells, traversing steam probe rakes, and over 500 different channels of instrumentation for monitoring fuel rod temperatures, spacer grid temperatures, housing temperatures, and fluid temperatures and pressures. The rod bundle was constructed to model a portion of a pressurized water reactor fuel assembly, with similar pitch, rod diameter, mixing vane grids, and full assembly length.;Data from the steam cooling experiments with droplet injection conducted in the RBHT facility provide very detailed insight to the temperature distribution and heat transfer characteristics of dispersed flow film boiling regime during the reflood phase of the emergency response to a large break loss of coolant accident. The information and data required to effectively analyze this regime include: the entrained liquid droplet sizes and velocity, vapor temperature, steam flow rate, and the interfacial heat and mass transfer. The droplet sizes are measured with the aid of a high speed camera system using an infrared laser as the lighting source and specially designed computer software, known as VisiSizer, which is capable of analyzing droplet data in real time. Droplet diameter and droplet distribution data were measured at a series of elevations in order to obtain detailed information on the evaporation and breakup of the droplets as they travel upwards through the bundle and pass through the spacer grids.;The experimental data are compared with the predictions of Best-Estimate computer codes, such as TRACE and COBRA-TF. The droplet diameters and distribution provide a good data set for future model development. However, the data can be made more complete by determining the droplet velocity and Weber number, which are missing in the measured data but are key parameters for modeling both droplet breakup and spacer grid effects. The droplet velocities can be predicted with the aid of the current computer codes to obtain an approximate value that can be used as a basis for modeling. The detailed data sets obtained from the RBHT experiments gathered in this study along with the droplet velocities calculated from COBRA-TF in the present work provide useful insight into the two-phase heat transfer phenomena that occur during the reflood period of loss of coolant accident. (Abstract shortened by UMI.). |
