Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

Kevin G. Field; Xunxiang Hu; Kenneth C. Littrell; Yukinori Yamamoto; Lance L. Snead

doi:10.1016/j.jnucmat.2015.06.023

Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

Kevin G. Field, Xunxiang Hu, Kenneth C. Littrell, Yukinori Yamamoto, Lance L. Snead

Oak Ridge National Laboratory

Research output: Contribution to journal › Article › peer-review

237 Scopus citations

Abstract

Abstract The Fe-Cr-Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe-Cr-Al alloys has not been fully established. In this study, a series of Fe-Cr-Al alloys with 10-18 wt % Cr and 2.9-4.9 wt % Al were neutron irradiated at 382°C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2<111> and a<100> were detected and quantified. Results indicate precipitation of Cr-rich α' is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure-property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

Original language	English
Article number	49161
Pages (from-to)	746-755
Number of pages	10
Journal	Journal of Nuclear Materials
Volume	465
DOIs	https://doi.org/10.1016/j.jnucmat.2015.06.023
State	Published - Jul 20 2015

Funding

Research was sponsored by DOE Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program . Research on the CG-2 General Purpose SANS and neutron irradiation of Fe–Cr–Al alloys at ORNL's HFIR user facility was sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, DOE . Transmission electron microscopy was partially sponsored by ORNL's Center for Nanophase Materials Sciences , which is sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, DOE .

Funders	Funder number
ORNL's Center for Nanophase Materials Sciences
Scientific User Facilities Division
U.S. Department of Energy
Office of Nuclear Energy
Basic Energy Sciences

Keywords

Accident tolerant
Dislocation
FeCrAl
Hardening
Phase stability

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10.1016/j.jnucmat.2015.06.023

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title = "Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys",

abstract = "Abstract The Fe-Cr-Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe-Cr-Al alloys has not been fully established. In this study, a series of Fe-Cr-Al alloys with 10-18 wt % Cr and 2.9-4.9 wt % Al were neutron irradiated at 382°C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2<111> and a<100> were detected and quantified. Results indicate precipitation of Cr-rich α' is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure-property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.",

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AU - Field, Kevin G.

AU - Hu, Xunxiang

AU - Littrell, Kenneth C.

AU - Yamamoto, Yukinori

AU - Snead, Lance L.

PY - 2015/7/20

Y1 - 2015/7/20

N2 - Abstract The Fe-Cr-Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe-Cr-Al alloys has not been fully established. In this study, a series of Fe-Cr-Al alloys with 10-18 wt % Cr and 2.9-4.9 wt % Al were neutron irradiated at 382°C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2<111> and a<100> were detected and quantified. Results indicate precipitation of Cr-rich α' is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure-property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

AB - Abstract The Fe-Cr-Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe-Cr-Al alloys has not been fully established. In this study, a series of Fe-Cr-Al alloys with 10-18 wt % Cr and 2.9-4.9 wt % Al were neutron irradiated at 382°C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2<111> and a<100> were detected and quantified. Results indicate precipitation of Cr-rich α' is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. A structure-property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

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KW - Hardening

KW - Phase stability

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Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

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