Abstract
Irradiation-assisted stress corrosion cracking is of concern for the safe and economic operation of light water reactors. In this study, cracking susceptibility of austenitic stainless steels was investigated by using slow strain rate tensile (SSRT) tests in a simulated pressurized water reactor (PWR) environment. The specimens were irradiated to 5, 10, and 48 dpa in the BOR60 reactor at 320°C. The SSRT results showed that yield strength was increased significantly in irradiated specimens while ductility and strain hardening capability were decreased. Irradiation hardening was found to be saturated below 10 dpa. The irradiated yield strength of cold-worked specimens was higher than that of solution-annealed specimens. Fractographic examinations were also performed on the tested specimens, and the dominant fracture morphology was ductile dimples. Intergranular cracking was rarely seen on the fracture surface. Transgranular cleavage cracking, however, was found more frequently on the specimen tested in simulated PWR environment.
This is a preview of subscription content, log in via an institution to check access.
Access this chapter
Tax calculation will be finalised at checkout
Purchases are for personal use only
Preview
Unable to display preview. Download preview PDF.
Similar content being viewed by others
References
Craig F. Cheng, “Intergranular Stress-Assisted Corrosion Cracking of Austenitic Alloys in Water-Cooled Nuclear Reactors,” J. Nucl. Mater., 56 (1975) 11–33.
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.
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.
R. L. Jones, J. D. Gilman, and J. L. Nelson, “Controlling Stress Corrosion Cracking in Boiling Water Reactors,” J. Nucl. Mater., 143 (1993) 111.
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.
P. Scott, “A Review of Irradiation Assisted Stress Corrosion Cracking”, J. Nucl. Mater., 211, (1994) 101–122.
P. L. Andresen, F. P. Ford, S. M. Murphy, and J. M. Perks, “State of Knowledge of Radiation Effects on Environmental Cracking in Light Water Reactor Core Materials,” Proc. 4th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, NACE, Houston, TX, pp. 1.83–1.121, 1990.
G. S. Was, and P. L. Andresen, “Stress Corrosion Cracking Behavior of Alloys in Aggressive Nuclear Reactor Core Environments,” Corrosion, 63 (1), (2007), 19.
J. O. Stiegler and L. K. Mansur, “Radiation Effects in Structural Materials,” Ann. Rev. Mater. Sci., 9 (1979) 405.
S. J. Zinkle, P. J. Maziasz, and R. E. Stoller, “Dose Dependence of the Microstructural Evolution in Neutron-Irradiated Austenitic Stainless Steel,” J. Nucl. Mater., 206 (1993) 266.
P. J. Maziasz, “Overview of Microstructural Evolution in Neutron-Irradiated Austenitic Stainless Steels,” J. Nucl. Mater., 205 (1993) 118.
G. S. Was, and S. M. Bruemmer, “Effect of Irradiation on Intergranular Stress Corrosion Cracking,” J. Nucl. Mater., 216 (1994) 326.
S. M. Bruemmer, and G. S. Was, “Microstructural and Microchemical Mechanisms Controlling Intergranular stress corrosion cracking in light-water-reactor systems,” J. Nucl. Mater., 216 (1994) 348.
Electric Power Research Institute, “CIR II Program: Description of the Boris 6 and 7 Experiment in the BOR-60 Fast Breeder Reactor,” Report No. 1011787, Palo Alto, CA, 2005.
Y. Chen, O. K. Chopra, W. K. Soppet, W. J. Shack, Y. Yang, and T. Allen, “Cracking Behavior and Microstructure of Austenitic Stainless Steels and Alloy 690 Irradiated in BOR-60 Reactor, Phase I,” Report ANL/09–32, Argonne National Laboratory, 2010.
R. N. Parkins, “Development of Strain-Rate Testing and Its Implications,” ASTM-STP 665, ASTM 1977.
A. Ikeda, M. Ueda, and H. Okamoto, “The Role of Slow Strain Rate Testing on Evaluation of Corrosion Resistant Alloys for Hostile Hot Sour Gas Production,” ASTM STP 1210, ASTM 1993.
Greenwood, L. R., “Neutron Interactions and Atomic Recoil Spectra,” J. Nucl. Mater. 216, 29–44, 1994
D. Edwards, E. Simonen, S. Bruemmer, “Radiation-Induced Segregation Behavior in Austenitic Stainless Steels: Fast Reactor versus Light Water Reactor Irradiations,” Proc. 13th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 2007.
Author information
Authors and Affiliations
Editor information
Editors and Affiliations
Rights and permissions
Copyright information
© 2011 TMS (The Minerals, Metals & Materials Society)
About this paper
Cite this paper
Chen, Y., Alexandreanu, B., Soppet, W.K., Shack, W.J., Natesan, K., Rao, A.S. (2011). Slow Strain Rate Tensile Tests of Irradiated Austenitic Stainless Steels in Simulated PWR Environment. In: Busby, J.T., Ilevbare, G., Andresen, P.L. (eds) Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors. Springer, Cham. https://doi.org/10.1007/978-3-319-48760-1_78
Download citation
DOI: https://doi.org/10.1007/978-3-319-48760-1_78
Publisher Name: Springer, Cham
Online ISBN: 978-3-319-48760-1
eBook Packages: Chemistry and Materials ScienceChemistry and Material Science (R0)