Abstract
Abstract In Pressurized Water Reactors and in CANDU reactors steam generator (SG) tubes represent a major fraction of the reactor primary coolant pressure boundary. The ability to estimate the leak rates from the through wall cracks in the steam generator tube is important in terms of radiological source terms and overall operational management of steam generators as well as demonstration of the leak-before- break condition. In this study an experimental program and analysis methods were developed to measure and assess the choking flow rate of subcooled water through simulated steam generator tube crack geometries. Experiments were conducted on choking flow for various simulated crack geometries for vessel pressures up to 7 MPa with various subcoolings. Measurements were done on subcooled flashing flow rate through well defined simulated crack geometries with L/D
Published Version
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