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Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding SCIE SCOPUS

Title
Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding
Authors
Dabney TylerJohnson GregYeom HwasungMaier BenWalters JorieSridharan Kumar
Date Issued
2019-12
Publisher
Elsevier Limited
Abstract
Two FeCrAl alloy coatings with different Cr and Al contents were deposited on Zr-alloy substrates via the cold spray deposition method to study the efficacy of these coatings to improve the accident tolerance of Zr-alloy fuel cladding in light water reactors (LWRs). First, the coatings were tested in a 400 degrees C steam autoclave for 72 h as an accelerated method to simulate the normal LWR operating conditions. In these tests, both coatings resulted in notable improvements in oxidation resistance compared to the Zr-alloy, with the higher Cr containing coating exhibiting a thinner oxide layer. Both coatings provided good oxidation resistance when tested in 1200 degrees C ambient air environment; however, the formation of a low melting point eutectic (-928 degrees C) between Fe and Zr resulted in significant melting associated with inter-diffusion between Fe in the coating and the Zr in the substrate. Therefore, to take advantage of the superior oxidation resistance that FeCrAl coatings can provide, a Mo interlayer was deposited, also by the cold spray process, in-between the protective FeCrAl coating and the Zr-alloy substrate, thus creating a dual cold spray layer accident tolerant coated cladding concept. The wear resistance of the FeCrAl coating was superior to the Zr-alloy substrate based on pin-on-disk wear tests, providing an indication of the benefits of such coatings to reduce grit-to-rod fretting (GTRF) damage. The feedstock powders, microstructure, phases, hardness of coatings, as well as oxide and inter-diffusion layers of oxidized coatings were examined using SEM, XRD, and XPS characterization techniques.
URI
https://oasis.postech.ac.kr/handle/2014.oak/118881
DOI
10.1016/j.nme.2019.100715
ISSN
2352-1791
Article Type
Article
Citation
Nuclear Materials and Energy, vol. 21, 2019-12
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염화성YEOM, HWASUNG
Div. of Advanced Nuclear Enginrg
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