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Title: Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.
Authors:
 [1] ;  [1] ;  [1] ;  [1] ;  [1] ;  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
Grant/Contract Number:
AC05-00OR22725
Type:
Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 467; Journal Issue: P2; Journal ID: ISSN 0022-3115
Publisher:
Elsevier
Research Org:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Center for Nanophase Materials Sciences (CNMS)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
OSTI Identifier:
1324043
Alternate Identifier(s):
OSTI ID: 1459743

Yamamoto, Yukinori, Pint, Bruce A., Terrani, Kurt A., Field, Kevin G., Yang, Ying, and Snead, Lance Lewis. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors. United States: N. p., Web. doi:10.1016/j.jnucmat.2015.10.019.
Yamamoto, Yukinori, Pint, Bruce A., Terrani, Kurt A., Field, Kevin G., Yang, Ying, & Snead, Lance Lewis. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors. United States. doi:10.1016/j.jnucmat.2015.10.019.
Yamamoto, Yukinori, Pint, Bruce A., Terrani, Kurt A., Field, Kevin G., Yang, Ying, and Snead, Lance Lewis. 2015. "Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors". United States. doi:10.1016/j.jnucmat.2015.10.019. https://www.osti.gov/servlets/purl/1324043.
@article{osti_1324043,
title = {Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors},
author = {Yamamoto, Yukinori and Pint, Bruce A. and Terrani, Kurt A. and Field, Kevin G. and Yang, Ying and Snead, Lance Lewis},
abstractNote = {Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.},
doi = {10.1016/j.jnucmat.2015.10.019},
journal = {Journal of Nuclear Materials},
number = P2,
volume = 467,
place = {United States},
year = {2015},
month = {10}
}