ALLOY 690 SURFACE NANOSTRUCTURES DURING EXPOSURE TO PWR PRIMARY WATER AND POTENTIAL INFLUENCE ON STRESS CORROSION CRACKING INITIATION
Alloy 690 (30 wt% Cr) is currently being used to replace lower Cr alloy 600 (16 wt% Cr) in critical pressurized water reactor (PWR) components due to its superior protection against stress corrosion cracking (SCC). The increased Cr content is believed to assist in forming a protective chromia film on the surface that inhibits corrosion and intergranular attack (IGA) in particular. A key objective for this work is to understand how surface microstructural features act as precursor sites for SCC nucleation. Unstressed alloy 690 specimens have been exposed to simulated PWR water conditions at 360°C including cold-worked materials shown to susceptible to SCC in crack-growth tests. Detailed surface and near-surface characterizations were performed on cross-section samples starting with low kV backscatter scanning electron microscopy (SEM) imaging to document the extent of general and localized corrosion. Selected surface and local corrosion regions were then prepared by focused ion beam milling for transmission electron microscopy (TEM) and atom probe tomography (APT) analyses. TEM is used to determine the oxidation microstructures as well as compositional gradients associated with the degradation, while APT is conducted to investigate the atomistic redistribution of alloying and impurity elements in the oxides and surrounding metal.
- Research Organization:
- Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- AC05-76RL01830
- OSTI ID:
- 1991314
- Report Number(s):
- PNNL-SA-96869
- Resource Relation:
- Conference: Proceedings of the 16th International Conference on Environmental Degradation of Nuclear Power Systems - Water Reactors, August 11-15, 2013, Asheville, NC
- Country of Publication:
- United States
- Language:
- English
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