Lifetime prediction of components in Alloy 600 is a major concern for nuclear power plants. Alloy 600 components have been shown to be susceptible to stress corrosion cracking (SCC). In the 1990’s, an engineering model was developed in order to predict the life time as a function of the main macroscopic parameters (stress, environment, material), based on laboratory results. This model has since been used to predict the ranking of various Alloy 600 components, using the knowledges of the manufacturing and service conditions for each component. It was applied successfully in the case of forged control rod drive mechanism (CRDM) nozzles. However, it was found necessary to improve this model to account for the strain history of the different components. Predictions using the model, investigated from an array of test results on Alloy 600 in laboratory primary water, have demonstrated that the time for initiation differed significantly according to the strain path applied to the specimen. The present work is dedicated to assess SCC results from samples with different strain paths and different level of cold work in order to better understand the manufacturing conditions on SCC. The samples are machined in three different directions and tested at different durations in order to model the time for transition (transition between slow and fast propagation) as a function of cold work, strain path and stress. Thermomechanical treatments are also applied on two different heats of Alloy 600: forged WF675 (very susceptible to SCC in as received conditions) and rolled 78456/337 (non susceptible to SCC in as-received conditions) in order to transform the forged microstructure into a microstructure close to the rolled microstructure and vice-versa. These microstructures are then tested in primary conditions and the results are compared to the results obtained on as-received material in order to get a better understanding of manufacturing process and microstructure parameters regarding SCC behaviour.

Stress corrosion cracking (SCC) of alloy 600 is regarded as one of the most important challenges to nuclear power plant operation worldwide. This study investigates two heats of alloy 600 (forged control rod drive mechanismnozzle and rolled divider plate) in order to obtain a better understanding of the effects of the material parameter on the SCC phenomenon. The experimental approach was designed to determine the effect of the manufacturing process (forged vs. rolled), the cold-work (as-received vs. cold-worked) and the strain path (monotonic vs. complex) on SCC of alloy 600. Specimens with different strain paths have been produced from two materials representative of plant components and tested in high temperature (360°C) high pressure primary water environment. The manufacturing process has been proven to have a great effect on the stress corrosion cracking behaviour of alloy 600. Indeed, the SCC susceptibility assessment has demonstrated that the rolled materialis resistant to SCC even after cold work, whereas the forged material is susceptible in the as-received state. Microstructural characterisations have been undertaken to explain these differences in SCC behaviour. The carbide distribution is the main microstructural parameter influencing SCC but the misorientation, in synergy with the carbide distribution, has been proven to give a better representation of the materials SCC susceptibilities.