The Response of Alloy 690 Tubing in a Pressurized Water Reactor Environment
In this paper, the corrosion-fatigue data on Alloy 690 plate documented in the open literature is compared with test data obtained using steam generator tubing. The results on plate material 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. This paper also examines if response of steam generator tubing in a pressurized water reactor environment can be modeled using test data obtained for the material on plate stock. The fatigue crack growth rate data for alloy IN 690 used in actual steam generator tubing in low dissolved oxygen PWR environment was generated using a circumferentially through-wall cracked tube specimen. The laboratory test data is compared with the data published in the open literature using a modified corrosion-fatigue model. Test results reveal the tube material to have near similar fatigue crack growth rate behavior when compared to the plate material. This provides conclusion that the standard specimen geometry, such as compact tension specimen, can be used to characterize the response of the tube in adverse environments.
Materials and Design
Young, B. A.; Gao, Xiaosheng; Srivatsan, Tirumalai S.; and King, Peter J., "The Response of Alloy 690 Tubing in a Pressurized Water Reactor Environment" (2007). Mechanical Engineering Faculty Research. 650.