Case Study of the Propagation of a Small Flaw Under PWR (Pressurized Water Reactor) Loading Conditions and Comparison with the ASME Code Design Life - Comparison of ASME Code Sections III and XI.

摘要

A cooperative study was performed to investigate the degree of conservatism and consistency in the ASME Boiler and Pressure Vessel Code Sect. III fatigue evaluation procedure and Sect. XI flaw acceptance standards. A single, realistic, sample problem was analyzed to determine the significance of certain points of criticism of an earlier parametric study by staff members of the Division of Engineering Standards of the Nuclear Regulatory Commission. The problem was based on a semielliptical flaw located on the inside surface of the hot-leg piping at the reactor vessel safe-end weld for the Zion 1 pressurized-water reactor (PWR). The current study shows that the criticism of the earlier report should not detract from the conclusion that if a component experiences a high level of cyclic stress corresponding to a fatigue usage factor near 1.0, very small cracks can propagate to unacceptable sizes.