Atom-probe tomography of surface oxides and oxidized grain boundaries in alloys from nuclear reactors
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Atom-probe tomography of surface oxides and oxidized grain boundaries in alloys from nuclear reactors Karen Kruska1, David W Saxey12, Takumi Terachi3, Takuyo Yamada3, Peter Chou4, Olivier Calonne5, Lionel Fournier5, George D W Smith1, Sergio Lozano-Perez1 1
University of Oxford, Department of Materials, Oxford, United Kingdom. University of Western Australia, School of Physics, Perth, WA, Australia 3 Institute of Nuclear Safety Systems Inc., Tsuruga, Fukui, Japan. 4 EPRI, Palo Alto, CA, United States. 5 Areva NP, Paris, France. 2
ABSTRACT The preparation of site-specific atom-probe tomography (APT) samples containing localized features has become possible with the use of focused ion beams (FIBs). This technique was used to achieve the analysis of surface oxides and oxidized grain boundaries in this paper. Transmission electron microscopy (TEM), providing microstructural and chemical characterization of the same features, has also been used, revealing crucial additional information. The study of grain boundary oxidation in stainless steels and nickel-based alloys is required in order to understand the mechanisms controlling stress corrosion cracking in nuclear reactors. Samples oxidized under simulated pressurized water reactor primary water conditions were used, and FIB lift-out TEM and APT specimens containing the same oxidized grain boundary were prepared and fully characterized. The results from both techniques were found fully consistent and complementary. Chromium-rich spinel oxides grew at the surface and into the bulk material, along grain boundaries. Nickel was rejected from the oxides and accumulated ahead of the oxidation front. Lithium, which was present in small quantities in the aqueous environment during oxidation, was incorporated in the oxide. All phases were accurately quantified and the effect of different experimental parameters were analysed. INTRODUCTION Because of their excellent mechanical properties and corrosion resistance at operating temperatures, stainless steels (SSs) and Ni-based alloys are frequently used in pressurised water reactors (PWRs) [1]. With increasing years of operation, more incidences of stress corrosion cracking (SCC) were identified [2] (although most failures in stainless steels have been attributed to non-specification conditions). A major advance in the combat against SCC was the replacement of Alloy 600 with the more Cr-rich Alloy 690 in steam generator tubes [3]. Cracking in the commonly used stainless steels and Ni-based alloys is predominantly intergranular [4, 5], which suggests that preferential oxidation of grain boundaries (GBs) is an important part (not the only part) of the SCC mechanisms in these alloys. Although the crystal structure of these alloys is identical and they have similar mechanical properties, the ratios of the main alloying elements are different. This leads to different corrosion potentials at the surfaces in
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the same environment and potentially to the formation of different oxides at the surfaces and at GBs. Previous studies on the
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