Effect of Strain Rate and High Temperature Water on Deformation Structure of VVER Neutron Irradiated Core Internals Steel
Components of reactor core internals suffer from Irradiation Assisted Stress Corrosion Cracking. Here is studied 08Ch18N10T austenitic stainless steel acquired from decommissioned NPP Nord Unit 1, VVER 440-230 type in Greifswald, where had been irradiated to 5.2 dpa. The material was tensile tested at 20 °C in air and slow strain rate tested at 320 °C in air and in water. SEM observations of the fracture surface found ductile fracture for the air tests, but areas of intergranular fracture typical of IASCC in the water. This paper emphasizes the microscopic examination from three samples to determine the underlying physical damage processes. TEM observations close to the fractured surface focused to the interaction of dislocations with local radiation damage defects and grain boundaries owing to different test conditions. Determination of local chemical composition around the grain boundaries indicated radiation induced segregation; as well presence of helium gas in voids. The observation of tensile tests found the presence of twinning and regions of strained martensite transformation. The nano features of tests at elevated temperature were tangled dislocations, similar in air and water. No effect of the water environment on the deformation structures was observed.
KeywordsAustenitic stainless steel Neutron irradiation Radiation damage Slow strain rate High temperature water
The authors would like to express their sincere thanks to Edita Lecianova and Vit Rosnecky for their support in handling and active sample preparation, especially in the hot-cells.
The presented work was financially supported by the SOTERIA project (EC: No. 661913) and by the Ministry of Education, Youth and Sport Czech Republic project LQ1603 (Research for SUSEN). The work has been realized within the SUSEN project (established in the framework of the European Regional Development Fund (ERDF) in project CZ.1.05/2.1.00/03.0108).
- 2.G.S. Was, K.J. Stephenson, Y. Ashida, Insights into the IASCC mechanism in neutron irradiated austenitic alloys with varying microstructure and microchemistry, in Proceedings 16th Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (2013)Google Scholar
- 6.K. Fukuya, K. Fujii, Deformation structure in 316 stainless steel irradiated in a PWR, in 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (2005), pp. 389–393Google Scholar
- 8.M.N. Gussev, J.T. Busby, Deformation localization in highly irradiated austenitic stainless steels, in 16th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (2013)Google Scholar
- 12.W. Karlsen, J. Pakarinen, A. Toivonen, U. Ehrnstén, Deformation microstructures of 30 dpa AISI 304 stainless steel after monotonic tensile and constant load autoclave testing, in 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (2011), pp. 1429–1443Google Scholar
- 13.J.L. Nelson et al., Critical issues report and roadmap for the advanced radiation-resistant materials program, in TR-1026482 (EPRI, Palo Alto, 2012)Google Scholar
- 14.M. Zamboch et al., Influence of radiolysis and hydrogen embrittlement on the in-service cracking of pwr internal structures, in TR-112593 (EPRI, Palo Alto, 1999)Google Scholar
- 17.A. Hojná, M. Ernestová, O. Hietanen, R. Korhonen, L. Hulinová, F. Oszvald, Irradiation assisted stress corrosion cracking of austenitic stainless steel WWER reactor core internals,in 15th Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors (2012), pp. 1257–1272Google Scholar
- 18.H.K. Namburi, P. Bublíkova, V. Rosnecky, J. Michalicka, TEM foil preparation from irradiated metallic materials: a practical approach. J. Environ. Sci. Eng. 4B, 432–444 (2015)Google Scholar