Weld Material Investigations of a WWER-440 Reactor Pressure Vessel: Results From the First Trepan Taken From the Former Greifswald NPP

Between 1973 and 1990 four units of the Russian nuclear power plants type WWER-440/230 were operated in Greifswald (former East Germany). Material probes from the pressure vessels were gained in the frame of the ongoing decommissioning procedure. The investigations of this material started with material from the circumferential core weld of unit 1. First, this paper presents results of the reactor pressure vessel (RPV) fluence calculations depending on different loading schemes and on the axial weld position based on the Monte Carlo code TRAMO. The results show that the use of the dummy assemblies reduces the flux by a factor of 2–5 depending on the azimuthal position. The circumferential core weld (SN0.1.4) received a fluence of $2.4×1019neutrons/cm2$ at the inner surface; it decreases to $0.8×1019neutrons/cm2$ at the outer surface. The material investigations were done using a trepan from the circumferential core weld. The reference temperature $T0$ was calculated with the measured fracture toughness values, $KJc$⁠, at brittle failure of the specimen. The $KJc$ values show a remarkable scatter. The highest $T0$ was about $50°C$ at a distance of 22 mm from the inner surface of the weld. The Charpy transition temperature $TT41J$ estimated with results of subsized specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The VERLIFE procedure prepared for the integrity assessment of WWER RPV was applied on the measured results. The VERLIFE lower bound curve indexed with the Structural Integrity Assessment Procedures for European Industry (SINTAP) reference temperature, $RTT0SINTAP$⁠, envelops the $KJc$ values. Therefore for a conservative integrity assessment the fracture toughness curve indexed with a RT representing the brittle fraction of a data set of measured $KJc$ values has to be applied.