An intergranular stress corrosion cracking (IGSCC) growth model for unirradiated, thermally sensitized stainless steels and a methodology for assessment of through-wall crack growth in horizontal weld heat affected zones (HAZs) of boiling water reactor (BWR) core shrouds has been developed. This empirical model accounts for the variability of important IGSCC parameters such as coolant conductivity, stress intensity factor, K, temperature and electrochemical corrosion potential (ECP) in providing a conservative, yet realistic assessment of the crack growth rate. Data from various sources were used to derive the empirical crack growth correlation, including work from Electric Power Research Institute (EPRI)-sponsored research, work sponsored by the U. S. Nuclear Regulatory Commission (NRC) and in-plant crack arrest verification system (CAVS) data as well as laboratory data developed by the General Electric Nuclear Energy (GENE). The combined database from all the sources was evaluated to ensure that only relevant data was used in the model development. This refined database was used to derive the crack growth correlation using pattern recognition and multivariate modeling tools. For practical application to crack growth evaluation of stainless steel components, three approaches were developed for dispositions of flaws, i.e., a K-independent approach, a conservative 95th percentile K-dependent approach and a plant specific approach using actual BWR water chemistry data. An example problem representing actual BWR shroud conditions is presented that demonstrates how the current crack growth model, the weld residual stress and K can be used to perform a plant-specific evaluation of flawed shrouds. The results of this example demonstrated that significant operating periods are likely for most flawed conditions in BWR core shroud welds before ASME Code Section XI core shroud safety margins are challenged.

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