Abstract
The effect of environment and prestrain on IASCC of austenitic stainless steels was investigated in post-irradiation CERT tests in Ar at 288°C and simulated BWR NWC. Two alloys susceptible to IASCC were selected for this study, which are commercial alloy 304 and high purity Fe-15Crl2Ni. Samples were irradiated to 5 dpa with 2 MeV protons at 360°C and strained to about 2% in Ar at 288°C. The results showed that the water environment is the key to inducing IASCC at 288°C. No intergranular cracking was observed in either alloy following straining in Ar up to ~3.5%. However, cracking occurred once the samples were subsequently strained to an additional 1% in simulated BWR NWC. Cracks tended to be long and extend over many grain boundaries rather than single grain facets, perhaps due to the prestrain in Ar. Cracking induced by 0.2% plastic strain in simulated BWR NWC in CP304 with 3.4% prestrain in Ar was also observed. Normal stress is critical in determining the crack initiation location in CP304 and Fe-15Cr12Ni when strained in simulated BWR NWC because the cracked grain boundaries were preferentially aligned perpendicular to the tensile direction.
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
Reference
Busby, J.T., G.S. Was, and E.A. Kenik, Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels. Journal of Nuclear Materials, 2002. 302(1): p. 20–40.
Fukuya, K., et al., A prediction model of IASCC initiation stress for bolts in PWR core internals. Nuclear Engineering and Design, 2010. 240(3): p. 473–481.
Busby, J. T., in Nuclear Engineering and Radiological Sciences Department. 2001, University of Michigan: Ann Arbor.
Jiao, Z. and G.S. Was, Localized deformation and IASCC initiation in austenitic stainless steels. Journal of Nuclear Materials, 2008. 382(2–3): p. 203–209.
M.C. Hash, J.T.B., G.S. Was. in Effects of Radiation on Materials: 21st International Symposium. 2002. West Conshohocken, PA, USA: American Society for Testing of Materials.
Jiao, Z. and G.S. Was, Impact of localized deformation on IASCC in austenitic stainless steels. Journal of Nuclear Materials, 2011. 408(3): p. 246–256.
Nishioka, H., et al., Deformation structure in highly irradiated stainless steels. Journal of Nuclear Science and Technology, 2008. 45(4): p. 274–287.
Onchi, T., et al., Mechanism of irradiation assisted stress corrosion crack initiation in thermally sensitized 304 stainless steel. Journal of Nuclear Materials, 2005. 340(2–3): p. 219–236.
E. A. West, G.S.W., Strain Incompatibilities and their Role in Intergranular Cracking of Irradiated 316L Stainless Steels.
Jiao, Z., J.T. Busby, and G.S. Was, Deformation microstructure of proton-irradiated stainless steels. Journal of Nuclear Materials, 2007. 361(2–3): p. 218–227.
Hiromasa Nishioka, K.F., Katsuhiko Fujii and Yuji Kitrunai, Deformation Structure in Highly Irradiated Stainless Steels. Journal of Nuclear Science and Technology, 2008. 45: p. 274–287.
J.F. Ziegler, J.P.B., SRIM2008. 2008, IBM Corp.: Yorktown, NY.
Gupta, G., et al., Microstructural evolution of proton irradiated T91. Journal of Nuclear Materials, 2006. 351(1–3): p. 162–173.
T. Onchi, K.H., M.L. Castaño, M. Navas. A role of localized plastic deformation on intergranular crack initiation of irradiated, thermally sensitized type 304 stainless steels in inert gas and in water. in Proceedings of 10th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. 2001. Lake Tahoe, Nevada.
A. Toivonen, U.E., W. Karlsen, P. Aaltonen, J-P. Massoud, J-M Boursier. Fractographic observations on a highly irradiated AISI 304 steel after constant load tests in simulated PWR water and argon and after supplementary tensile and impact tests. in Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power System — Water Reactors. 2005.
Elaine, A., West, Influence of Local Stress and Strain on Intergranular Cracking of 316L Stainless Steel in Supercritical Water, in Nuclear Engineering and Radiological Sciences Department. 2010, University of Michigan: Ann Arbor.
Furutani, G., et al., Stress corrosion cracking on irradiated 316 stainless steel. Journal of Nuclear Materials, 2001. 288(2–3): p. 179–186.
Manahan, M.P. and R. Kohli, A Phenomenological Investigation of In-Reactor Cracking of Type 304 Stainless Steel Control Rod Cladding. Nuclear Engineering and Design, 1989. 113: p. 297–321
McMurtrey, M.D., et al., Relationship between localized strain and irradiation assisted stress corrosion cracking in an austenitic alloy. Materials Science and Engineering a-Structural Materials Properties Microstructure and Processing, 2011. 528(10–11): p. 3730–3740.
Cookson, J.M., et al., Irradiation Assisted Stress-Corrosion Cracking of Controlled Purity 304l Stainless-Steels. Journal of Nuclear Materials, 1993. 202(1–2): p. 104–121.
Morisawa, J., et al., Hydrogen analysis and slow strain rate test in Ar gas for irradiated austenitic stainless steel. Journal of Nuclear Materials, 2001. 294(3): p. 241–249.
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
Lai, W., Jiao, Z., Was, G.S. (2011). Effect of Environment and Prestrain on IASCC of Austenitic Stainless Steels. 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_86
Download citation
DOI: https://doi.org/10.1007/978-3-319-48760-1_86
Publisher Name: Springer, Cham
Online ISBN: 978-3-319-48760-1
eBook Packages: Chemistry and Materials ScienceChemistry and Material Science (R0)