High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

Yeom, Hwasung; Maier, Benjamin; Johnson, Greg; Dabney, Tyler; Lenling, Mia; Sridharan, Kumar

doi:10.1016/j.jnucmat.2019.151737

Title: High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

Journal Article · Wed Jul 31 00:00:00 EDT 2019 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2019.151737· OSTI ID:1801185

Yeom, Hwasung ^[1]; Maier, Benjamin ^[1]; Johnson, Greg ^[1]; Dabney, Tyler ^[1]; Lenling, Mia ^[1]; Sridharan, Kumar ^[1]

Univ. of Wisconsin, Madison, WI (United States)

Oxidation kinetics and microstructural evolution of cold sprayed Cr coatings on Zircaloy-4 at 1130–1310 °C in flowing steam at atmospheric pressure have been studied. The study is aimed at understanding of the response of Cr coated Zr-alloy under a steam environment and high temperatures pertinent to design basis accidents (DBAs) and beyond design basis accidents (BDBAs) in light water reactors (LWRs). Surface morphology, microstructure, and phases of post-oxidation test samples were characterized using Scanning Electron Microscopy (SEM), x-ray diffraction (XRD), and Scanning Transmission Electron Microscopy (STEM). Growth kinetics of the Cr-oxide scale and interdiffusion layers between the Cr coating and the Zr-alloy substrate were quantified from cross-sectional SEM images. Cross-sectional analysis showed that the Cr coatings offered a 50 times reduction in oxidation rate over bare Zircaloy-4 in a 1310 °C steam environment. Oxidation kinetics at 1130 °C followed parabolic law (i.e., n ~ 0.5 in power law kinetics) but at 1230 °C and 1310 °C the value of n was suppressed to below 0.5 possibly due to the volatilization of Cr species at the two highest temperatures. Interdiffusion at coating/substrate interface resulted in formation of a brittle Cr₂Zr or Zr(Fe, Cr)₂ intermetallic compound layer on the order of micrometer in thickness and scattered Cr-rich precipitates were observed well below the interface within the Zr-alloy substrate after cooling. These experimental results could provide data to LWR system simulation codes for better estimation of coping time in the event of accidents.

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Research Organization:: Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: NE0008222

OSTI ID:: 1801185

Alternate ID(s):: OSTI ID: 1564592

Journal Information:: Journal of Nuclear Materials, Vol. 526; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

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Cited by: 68 works

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