Alloy 690 is widely used as steam generator heat transfer tube in PWR nuclear power plants.During the process of heat exchange,the thin tubes are subjected to vibration,i.e.,flow induced vibration.Many operating experiences and experiments showed that corrosion fatigue(CF)is one of potential failure modes for nuclear-grade structural materials.The testing results based on round bar materials documented in the open literature is not representative of actual behavior of steam generator tubing due to the mutually interactive influences of microstructural features,section thickness and processing variables.The CF behavior of Alloy 690 used in actual steam generator tubing in low dissolved oxygen PWR environment(<5 ppb)was investigated using a “boat-shaped” specimens.The laboratory test data is compared to those published in the open literature and the fatigue lives predicted by ANL model for Ni-alloy.The present testing results reveal the tube material has longer fatigue life when compared to the round bar materials,especially in low stain amplitude.This may suggest that the standard specimen geometry,such as round bar specimen,may be conservative to characterize the response of the tube in PWR environments.