Slow Strain Rate Tensile Tests of Irradiated Austenitic Stainless Steels in Simulated PWR Environment

Author:

Chen Y.,Alexandreanu B.,Soppet W. K.,Shack W. J.,Natesan K.,Rao A. S.

Publisher

Springer International Publishing

Reference19 articles.

1. Craig F. Cheng, “Intergranular Stress-Assisted Corrosion Cracking of Austenitic Alloys in Water-Cooled Nuclear Reactors,” J. Nucl. Mater., 56 (1975) 11–33.

2. F. Garzarolli, H. Rubel, and E. Steinberg, “Behavior of Water Reactor Core Materials with Respect to Corrosion Attack,” Proc. Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, NACE, Houston, TX, pp. 1–24, 1984.

3. F. Garzarolli, D. Alter, and P. Dewes, “Deformability of Austenitic Stainless Steels and Nickel-Base Alloys in the Core of a Boiling and a Pressurized Water Reactor,” Proc. Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, American Nuclear Society, La Grange Park, IL, pp. 131–138, 1986.

4. R. L. Jones, J. D. Gilman, and J. L. Nelson, “Controlling Stress Corrosion Cracking in Boiling Water Reactors,” J. Nucl. Mater., 143 (1993) 111.

5. Y. Chen, O. K. Chopra, W. K. Soppet, N. L. Dietz Rago, and W. J. Shack, “IASCC Susceptibility of Austenitic Stainless Steels and Alloy 690 in High Dissolved Oxygen Water Environment,” Proc. 13th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 2007.

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