The accurate assessment of critical crack size plays a vital role in demonstrating the Leak-Before-Break (LBB) criterion for the safety demonstration of a sodium-cooled Fast Breeder Reactor (FBR) piping system. The advancement of the crack size will increase the stress intensity factor and reduce the load-carrying capacity of the piping system. The prototype-sized pipe bend test revealed that even under a large-size crack growth situation, the ductile pipe bend fails by collapse rather than tearing instability. The critical crack size was realistically estimated based on a prototype-sized pipe bend cyclic test and compared with elastoplastic numerical analysis.
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Themes: Mechanical Behavior in Nuclear Materials
CORROSION FATIGUE OF HOLLOW SPECIMENS IN SIMULATED LWR WATER ENVIRONMENT
Pipe systems in nuclear power plants are subjected to cyclic thermal and mechanical loads from mixing points between hot and cool water and mechanically induced vibrations. The light water reactor (LWR) water environment flowing through the pipes at 300 °C with an internal pressure of 120 bars has a detrimental effect on the fatigue lives of the pipe systems. In this study, an experimental setup has been developed and designed to assess the corrosion fatigue lives of hollow specimens subjected to an alternating uniaxial cyclic load and simulated LWR water environment simultaneously. The corrosion fatigue tests have been conducted for both boiling water reactor (BWR) and pressurized water reactor (PWR) water environments.
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