Compact Crack Arrest Tests for the Validation of a Finite Element Material Model of the Reactor Pressure Vessel Steel of the Nuclear Power Plant KKG

Abstract

The integrity of a reactor pressure vessel (RPV) has to be given for its operating time in accordance with the regulations. An assessment of the RPV against brittle failure needs to be conducted specially for one of the most severe loading cases. This is the loss-of-coolant accident (LOCA). Cold water is injected into the RPV at operating conditions. This thermal shock of the ferritic pressure vessel wall leads to loading conditions at the beltline area known as Pressurized Thermal Shock (PTS). The assessment against brittle failure is based on a deterministic fracture analysis. Common parameters like stress intensity factors are employed to calculate the PTS event for an assumed (postulated) flaw. Subsequently the results of the fracture mechanics analysis are compared with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle fracture initiation. The validation considers the material data and the velocity of the crack growing into the specimen until it stops. The measured crack propagation velocity for tests performed according to ASTM E1221 [1] is compared to the result of the Finite Element (FE) simulation of a Compact Crack Arrest (CCA) test. For five of eight tests performed at −60 °C crack propagation velocity values were determined ranging from 509 m/s to 694 m/s with an average value of 618 m/s.