The U.S. Nuclear Regulatory Commission (NRC) promulgated Part 50.69 to Title 10 of the Code of Federal Regulations (CFR), “Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors,” in November 2004 (hereafter referred to as 10 CFR 50.69). The rule provides a voluntary alternative to compliance with many regulations which require “special treatment,” or regulatory requirements which go beyond industrial controls, including: specific inspection, testing, qualification, and reporting requirements. The voluntary alternative includes a process for categorization of structures, systems, and components (SSCs) as having either low safety significance (LSS) or high safety significance (HSS). The categorization process can result in increased requirements for HSS SSCs which were previously treated as non-safety-related, and reduced requirements for LSS SSCs which were previously treated as safety-related.
The categorization process includes plant-specific risk analyses which are used in combination with an integrated decision-making panel (IDP) to determine whether the SSC has a low or high safety significance. Seismic probabilistic risk assessment (SPRA) is one of the risk analyses options to account for the seismic risk contribution. Because the 10 CFR 50.69 rule has currently not been implemented widely, the significance of various SPRA assumptions and sources of uncertainty to the categorization process has had limited evaluation for a broad spectrum of U.S. nuclear power plants.
This paper will assess the importance of certain aspects of the seismic risk contribution to the categorization process. NRC Standardized Plant Analysis Risk (SPAR) models will be used to perform sensitivity studies to quantify the impact of various assumptions and sources of uncertainty on the outcome of the categorization process.
Impact of structural aging on seismic risk assessment of reinforced concrete structures in nuclear power plants
