Flaws detected in nuclear power plant components during in-service inspections are typically evaluated based on stress intensity factor influence coefficient databases and solutions from industry standards and public literature (e.g. API-579, ASME Section XI Code, WRC-175 Bulletin, Raju-Newman, etc). For certain components in the Pressurized Water Reactor (PWR) nuclear power plants, such as Bottom Mounted Instrumentation (BMI) nozzles, the cylindrical component geometry may fall outside the applicability limits of stress intensity factor influence coefficient databases. This situation occurs where the thickness to inner radius ratios of the cylindrical geometry is greater than 1.0. Accurate stress intensity factor (SIF) solutions are essential to flaw evaluation since the SIFs are used in the determination of both the allowable flaw size and crack growth in order to determine acceptability of the detected flaw.

In this paper, stress intensity factor influence coefficients are generated based on a three-dimensional finite element analysis for axial flaws located on the inside surface and outside surface of a cylindrical component with thickness to inner radius ratios (t/Ri) of 1, 2, 4, & 6. Non-dimensional influence coefficients are determined at the deepest point of the crack front and the surface point of the flaw based on a 4th order polynomial fit for a through-wall stress profile. The influence coefficients are generated for semi-elliptical flaws with a/c ratios = 0.125 through 2; where a is depth of the elliptical flaw, and c is the half-length of the elliptical flaw. The influence coefficients developed are suitable for calculating stress intensity factors for cylindrical components with high thickness to inner radius ratios.

This content is only available via PDF.
You do not currently have access to this content.