Radiation response of the overlay cladding from the decommissioned WWER-440 Greifswald unit 4 reactor pressure vessel

Results of tensile and crack extension testing conducted on irradiated austenitic overlay cladding material of WWER 440 reactor pressure vessels are presented. The specimens were sampled from three trepans originating from the decommissioned WWER-440 reactor pressure vessel of the nuclear power plant Greifswald Unit 4. Crack extension curves were measured with Charpy size SE(B) specimens using the unloading compliance technique according to ASTM E1820-11 at different temperatures. Crack initiation fracture toughness values JQ and KJQ are determined with the crack extension curves. The highest KJQ values were found in the temperature range from 20 to 75 °C. A significant scatter was observed in the initiation values. The reasons are seen in the scatter of the estimation of very low crack extensions and crack jumps caused by regions of low tearing strength in the overlay cladding. The comparison of the measured KJQ values and conservatively estimated stress intensity factors at an assumed surface crack shows that the cladding would remain intact during pressurized thermal shock transient.