In order to conduct effective and rational maintenance activity of components in nuclear power plants, it is proposed to manage fatigue degradation based on crack size corresponding to an extent of cumulative fatigue effect. This study is aimed at developing a prediction model for fatigue crack growth in simulated reactor coolant environment. In order to investigate influence of reactor coolant environment on crack initiation and crack growth, two-step replica observations were conducted for environmental fatigue test specimens (type 316 stainless steel) subjected to three kinds of strain range. Crack initiation, growth, and coalescence were observed in the experiments. It is clarified that crack coalescence is one of the dominant factors causing fatigue life reduction, and fatigue life reduction depends on crack size and distance of two coalescing cracks. Then, a model was developed for predicting statistical crack initiation and growth behavior. The relationship between dispersion of crack initiation life and strain range was approximated by the Weibull model to predict crack initiation. Then, the statistical crack growth was modeled using the relation of crack growth rate and strain intensity factor. Furthermore, the crack coalescence was taken into account to the crack growth prediction considering the distance between two cracks. Finally, the crack growth curve, which is the relationship between crack size and operation period, was derived through Monte Carlo simulation with the developed model. The crack growth behavior and residual life in the simulated reactor coolant environment can be reviewed by the crack growth curve obtained with crack initiation, and the growth model developed was compared with the fatigue test results.
Development of Fatigue Crack Growth Prediction Model in Reactor Coolant Environment
Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received March 25, 2017; final manuscript received April 19, 2018; published online May 21, 2018. Assoc. Editor: David L. Rudland.
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Ishizawa, T., Takeda, S., Kitada, T., Nakamura, T., and Kamaya, M. (May 21, 2018). "Development of Fatigue Crack Growth Prediction Model in Reactor Coolant Environment." ASME. J. Pressure Vessel Technol. August 2018; 140(4): 041402. https://doi.org/10.1115/1.4040095
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