Abstract
This paper aims to review new results regarding Irradiation Assisted Stress Corrosion Cracking (IASCC) of neutron irradiated Ti-stabilized austenitic stainless steel 08Ch18N10T (chemically similar to AISI 321) from WWER 440 reactor’s core internals of NPP Greifswald decommissioned after 15 years in service. Two components (core barrel and core shroud basket) irradiated in the LWR conditions (5×10−9 — 4×10−8 dpa/s, 260–330°C) to doses about 2–5 dpa were used for the testing.
IASCC was investigated by Slow Strain Rate Tensile (SSRT) and Crack Growth Rate (CGR) tests in simulated WWER water environment at 320°C. The IASCC presence been demonstrated if detected the presence of areas of mixed intergranular (IG) and transgranular (TG) fracture on fracture surface. The two tests represent different stress strain conditions for IASCC development, namely for the crack initiation. The test results showed that plane stress condition facilitates IASCC initiation in the thick components. The fracture surface observations indicate that IASCC crack grows based on strain-controlled fracture mechanism.
The results are compared with other data obtained by SSRT tests on the steel irradiated in fast reactor. Relation between the results on fast and in-service irradiated materials is mentioned, but disparateness in data not allowed any conclusions.
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References
V. S. Neustroev, V. N. Golovanov, V. K. Shamardin, “Mechanical Properties of Fe-18Cr-10Ni-Ti Austenitic Steel Irradiated in the WER-1000 and VVER-440 Reactors”, Proceedings of IAEA TM on Reactor Core Internals Behavior, Erlangen, Germany, 2004.
J. Kocik, M. Postler, M. Zamboch, E. Keilova, J. Burda, “Effect of Neutron Irradiation on Microstructure and Mechanical Properties of VVER-type Reactor Vessel Internals”, Fontevraud5, International Symposium on PWR, France, 2002.
A. Hojna, M. Falcnik, O. Hietanen, L. Hulinova, R. Korhonen, F. Oszvald, “Behavior of 08Ch18N10T Steel After 15 Years of Operation as Core Shroud of WWER 440 Plant”, 11th Int. Conf. on Material Issues in Design, Manufacturing and Operation of NPP Equipment, Saint Petersburg, Russia, 2010.
B. Z. Margolin, I. P. Kursevich, A. J. Minkin, V. I. Smirnov, V. A. Fedorova, A. N. Lapin, V. I. Kokhonov, V. S. Neustroev, S. V. Belozerov, V. K. Shamardin, V. G. Vasilev, R. Ahlstrand, J. Sainero, M. Navas, J. Lapena, “Influence of Neutron Irradiation and Environment on Materials Properties of Internals of WWER Reactors”, 11th Int. Conf. on Material Issues in Design, Manufacturing and Operation of NPP Equipment, Saint Petersburg, Russia, 2010.
J. Kocik, E. Keilova, J. Michalicka, “Microstructure Characterization of Reactor Vessel Internals of the Decommissioned VVER-440 Reactor”, to be submitted to JNM.
P. Andresen, “SCC Testing and Data Quality Considerations”, 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, Newport Beach, California, pp. 411–422, 1999.
R. Pathania, R. Carter, A. Demma, “Overview of EPRI Research on Effect of Irradiation on LWR Internals”, Proceedings of Fontenvraud 7, Avignon, 2010.
B. K. Neale, “The analytical fracture behaviour of a bar in tension containing a circumferential edge crack”, Int. J. Pres. Ves. & Piping 173, pp. 191–198, 1997.
F. P. Ford, “Quantitative Prediction of Environmentally Assisted Cracking”, Corrosion 52 (5), pp. 375–395, 1996.
K. Fujimoto, T. Yonezawa, E. Wachi, Y. Yamaguchi, M. Nakano, R. P. Shogan, J. P. Massoud, T. R. Mager, “Effect of the Accelerated Irradiation and Hydrogen/Helium Gas on IASCC Characteristics for Highly Irradiated Austenitic Stainless Steels”, 12th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, ed. T.R. Allen, P.J. King and L. Nelson, TMS, Warrendale, PA, pp. 299–310, 2005.
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© 2011 TMS (The Minerals, Metals & Materials Society)
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Hojná, A., Ernestová, M., Hietanen, O., Korhonen, R., Hulinová, L., Oszvald, F. (2011). Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel WWER Reactor Core Internals. 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_77
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DOI: https://doi.org/10.1007/978-3-319-48760-1_77
Publisher Name: Springer, Cham
Online ISBN: 978-3-319-48760-1
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