Advanced assessment of the integrity of ductile components

Nuclear Reactor Pressure Vessels (RPV) are manufactured from medium strength low alloy ferritic steel, specifically selected for its high toughness and good weldability. The ability of the pressure vessel to resist crack growth is crucial given that it is one of the fundamental containment safety systems of the reactor. For most of their lifetime, the pressure vessel operates at sufficiently elevated temperatures to ensure the material is ductile. However, the development of ductile damage, in the form of voids, and the ability to predict the ductile crack growth in RPV materials requires further work. The Gurson-Tvergaard-Needleman (GTN) model of void nucleation, growth and coalescence provides one tool for predicting ductile damage development. The model is normally calibrated against fracture toughness test data. However, recent work [1] has demonstrated the benefit of refining calibrations against measured void volume fractions generated from notched and pre-cracked specimen tests. This paper described the measurement of void distributions below the fracture surface of a range of notched and precracked specimens. The void distribution below the fracture surface is shown to be dependent upon the local stress triaxiality and plastic strain distribution. As a result, precracked specimens show a greater concentration of voids close to the fracture surface, whilst notched tensile specimens show a lower volume fraction of voids close to the crack surface. In both specimen types, voids are observed to extend between 2.5 and 3.5 mm below the fracture surface.