Over the service life of a nuclear power plant, the Boiling Water Reactor (BWR) may undergo many cool-down and heat-up thermal-hydraulic transients associated with, for example, scheduled refueling outages and other normal operational transients, or even possible overcooling transients. These thermal-hydraulic events can act on postulated surface flaws in BWRs and therefore impose potential risk on the structure integrity of Reactor Pressure Vessels (RPVs). Internal surface flaws are of interest for the BWRs under overcooling accidental scenarios, while external surface flaws are more vulnerable when the BWRs are subjected to heat-up transients. Stress Intensity Factor Influence Coefficient (SIFIC) databases for application to linear elastic fracture mechanics analyses of BWR pressure vessels which typically have an internal radius to wall thickness ratio (Ri/t) between 15 and 20 were developed for external surface breaking flaws. This paper presents three types of surface flaws necessary in fracture analyses of BWRs: (1) finite-length external surface flaws with aspect ratio of 2, 6, and 10. (2) infinite-length axial external surface flaws; and (3) 360° circumferential external surface flaws. These influence coefficients have been implemented and validated in the FAVOR fracture mechanics code developed at Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission (NRC). Although these SIFIC databases were developed in application to RPVs subjected to thermal-hydraulic transients, they could also be applied to RPVs under other general loading conditions.

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