Application of Corrosion Fatigue Crack Growth Rate Data to Integrity Analyses of Nuclear Reactor Vessels

Abstract

The methodology of fatigue crack growth analysis in evaluating structural integrity of nuclear components has been well established over the years, even to the point where a recommended practice has been incorporated in Appendix A to Section XI of the ASME Code. The present reference curve for crack growth rates of pressure vessel steels in reactor water environment was developed in 1973, and since that time a great deal of data have become available. The original curve was meant to be a bounding curve, and some recent data have exceeded it, so an important question to address is which reference curve to use for these calculations. The important features of fatigue crack growth behavior in a reactor water environment are reviewed, along with some suggested explanation for the observed environmental enhancement and overall trends. The variables which must be accounted for in any reference crack growth rate curve are delineated and various methods for accomplishing this task are discussed. Improvements to the present reference curve are suggested, and evaluated as to their accuracy relative to the present curve. The impact of the alternative curves is also evaluated through solution of an example problem. A discussion of the conservatisms included in the alternative reference curves as compared with the present reference curve is included. Also research work is identified which could lead to further improvement in the reference curves.