Abstract
The work presented here is an experimental investigation of the two-phase choked flow of initially subcooled water through a stress corrosion crack on the steam generator tube. The study was designed to provide an extensive experimental data set for predicting leak flow rates of nuclear reactor coolant using actual tube steam generator tubes sample. Five samples of steam generator tubes with stress corrosion axial cracks taken from steam generators of the US pressurized water reactor and Canadian pressurized heavy water reactor (CANDU) were used in this study. Wall thicknesses of the samples vary from 1.1 mm to 1.3 mm with the crack channel length to hydraulic dimeter ratios from 3 to 15. The experiment was carried out for various stagnation pressures up to 6.89 MPa and liquid subcoolings ranging from 5 °C to 70 °C. Parametric dependence of choked flow rate on stagnation pressure, crack channel length to hydraulic diameter ratio, and degree of liquid subcooling was demonstrated for different crack geometries. All the parameters are shown to have significant effects on the choking mass flux. The experimental results indicate that the channel length strongly affects the choking mass flux as well as the correlation between this parameter and the stagnation pressure. These data are unique as they were obtained with actual steam generator tube cracks and hence are advantageous in understanding the choked flow in the very short channel length channel and similar flow geometries.
Published Version
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