Recently, the use of risk information in regulatory decision as well as operation and maintenance of nuclear power plant has been gradually increased. Application of risk information to optimize the inservice inspection of nuclear piping system has been one of such an attempt in Korea. An accurate evaluation of the risk, or failure probability for each pipe subcomponent is essential to the application of the risk-informed inservice inspection. For this purpose, a probabilistic fracture mechanics code based on Monte Carlo simulation techniques was developed to handle various flaw shapes and orientations in nuclear piping system. Then, the failure probabilities of various subcomponents of the main coolant loop in a pressurized water reactor were calculated, and the relative risk ranking was determined using the plant’s data. Also, the effects of in-service inspection and frequency on the risk were investigated. Finally, the application of the probabilistic fracture mechanics methods to the risk-informed inservice inspection of nuclear piping system was discussed.
Probabilistic failure analysis of nuclear piping components due to a combined degradation mechanisms is a challenging issue. At present the majority of analyses were done by assuming a single failure mechanism for a specific location of a piping system. But in reality, this might not be an accurate approach. A tiny crack might be present in a weld location due to fabrication defect or initiated due to fatigue after a short incubation time of plant’s start up. This pre-existing or initiated crack then may be further matured by the synergistic effect of different probable degradation mechanisms e.g. fatigue, stress corrosion cracking, etc. In this study the development process of an advanced probabilistic fracture mechanics code has been described which can handle this combined failure mechanisms. Numerical examples are also presented to rationalize the use of such methodology.
To design a Reactor Pressure Vessel (RPV), material property like crack must be considered as it is an unavoidable property of materials. Presence of crack in materials must be kept within limit to prevent material's failure. So, crack propagation must be analyzed and observed. In this paper, crack propagation due to stress and materials fracture toughness of reactor pressure vessel cladding has been observed to estimate cumulative probability of crack failure using Probabilistic Fracture Mechanics (PFM). Average crack size is guessed as 3 mm and geometry factor is considered as 1.12 to analyze edge crack. Final crack analysis range has been found to be 1.8 mm with crack propagation rate of 30% of its average size. Variation of critical crack size and crack initiation point for several design stresses and fracture toughness has been investigated with probabilistic fracture mechanics technique. The observed crack propagation by calculating final crack size and the cumulative crack failure probability of the reactor pressure vessel materials are presented in this work.
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