Book Description

Nuclear power plants seeking to extend their operating license must first address the degradation of systems, structures, and components (SSCs) to ensure they can maintain a satisfactory level of reliability into the extended lifetime. Passive SSCs play an important role in determining the feasibility of life extension. Part of the feasibility analysis requires plants to demonstrate the viability and reliability of passive SSCs into the extended lifetime. The research carried out toward this thesis considers primary water stress corrosion cracking (PWSCC) of steam generator (SG) tubes as an example degradation mechanism. An empirical model for PWSCC crack growth is adopted to simulate crack growth over a 40-year operating lifetime. Surveillance and maintenance strategies similar to those performed by the industry are integrated with the PWSCC crack growth model to determine the effectiveness of surveillance strategies for detecting SG tube degradation. The results of this analysis were applied to a specific accident scenario in which steam generator tubes rupture following a depressurization of the secondary side due to the sudden rupture of a steam-line caused by flow-accelerated corrosion. Likelihood of a spontaneous steam generator tube rupture is also assessed. The analysis and application of the specific accident scenario indicates a maximum core damage frequency in the 16th year. Sensitivity analyses into the probability of detection (POD) and crack growth rates were also performed. As expected, the likelihood of the accident scenario occurring increased significantly as the maximum POD was decreased. When crack growth rates were slowed down, the overall likelihood of the accident scenario decreased and the expected occurrence of the accident scenario was delayed.