Fatigue crack growth rates of irradiated pressure vessel steels in simulated nuclear coolant environment

Fatigue crack growth rates, in a simulated pressurized water reactor primary loop environment, are presented for A508-2, A533-B steels and submerged arc weld metal (Linde 0091 flux) in both the irradiated and unirradiated conditions. The initial results, for specific experimental conditions, constant load amplitude waveforms, and load ratio = 0.2, indicate that irradiation does not enhance fatigue crack growth rates more than the enhancement due to the environment alone. Fractographic examination of the fatigue fracture surfaces showed that the crack propagated by a ductile striation formation mechanism. This article outlines the experimental techniques involved in this testing and describes the results and their possible significance with respect to the existing safety analysis codes