Analysis of Radiation-Induced Embrittlement Gradients on Fracture Characteristics of Thick-Walled Pressure Vessel Steels

Abstract

The fracture behavior of thick-walled nuclear vessels is considered for the case of a radiation-induced toughness gradient through the wall which characteristically results from neutron attenuation by the wall material itself. Fracture-safe design analyses based on linear elastic formulations or extrapolations of these formulations to the elastic-plastic regime are not sufficiently developed to characterize the integrated behavior of a wall whose toughness can range from brittle at the inner surface to highly ductile at the outer surface. Solutions to the problem in the foreseeable future will be obtained only by experimental means. The present approach uses the Fracture Analysis Diagram (FAD) together with a new interpretative method for fracture extension resistance based on modified dynamic tear specimens as the tools for gradient assessments. With these techniques the significance of the toughness gradient through the wall is assessed in terms of thick section mechanical constraint, and fracture characteristics of the complete wall are predicted. Characterization of a hypothetical 8-in. vessel wall is based on measured through-thickness fluence levels; this behavior is correlated with fracture toughness degradation for steels of varying sensitivity to irradiation using the FAD. This analysis indicates that major portions of the vessel wall remain above the FTE temperature, which dictates yield stress loading requirements for fracture, when the wall temperature is maintained at the limiting value, NDT + 60 deg F, for the inside surface as suggested by current AEC criteria. Fracture extension resistance measurements based on data from a 3-in. thick plate having a metallurgically induced toughness gradient suggest that nuclear vessels with analogous gradients will not fracture in an unstable fashion and will not generate missiles capable of breaching the containment system. Additional research is necessary to fully develop the approach for application to the individual reactor vessel.