Oxidation of Zirconium Alloy Fuel Cladding Material

Jordan Vandegrift; Clemente J. Parga; Ben Coryell; Brian J. Jaques

Oxidation of Zirconium Alloy Fuel Cladding Material

Jordan Vandegrift, Clemente J. Parga, Ben Coryell, Brian J. Jaques

Micron School of Materials Science and Engineering

Research output: Contribution to conference › Presentation

Abstract

Zirconium alloys are often used as nuclear fuel cladding. During transients the cladding may be exposed to air at high temperatures, which can cause accelerated oxidation that compromises the structural integrity of the cladding. There exist discrepancies in literature regarding the behavior of zirconium alloys under these conditions. The research presented compares the behavior of Zircaloy-3, Zircaloy-4, and Zr-1Nb during isothermal oxidation (500-820 °C) as well as rapid transient oxidation (≈ 700 ^°C / _s ) scenarios. The effect of alloying elements, plastic deformation, and welding are analyzed. The oxidation behavior is compared using TGA data, oxide thickness measurements, and post-oxidation imaging.

Original language	American English
State	Published - 12 Apr 2019

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@conference{f51f7cf4acf24f3fa3bf0ffdf81dc252,

title = "Oxidation of Zirconium Alloy Fuel Cladding Material",

abstract = " Zirconium alloys are often used as nuclear fuel cladding. During transients the cladding may be exposed to air at high temperatures, which can cause accelerated oxidation that compromises the structural integrity of the cladding. There exist discrepancies in literature regarding the behavior of zirconium alloys under these conditions. The research presented compares the behavior of Zircaloy-3, Zircaloy-4, and Zr-1Nb during isothermal oxidation (500-820 °C) as well as rapid transient oxidation (≈ 700 °C / s ) scenarios. The effect of alloying elements, plastic deformation, and welding are analyzed. The oxidation behavior is compared using TGA data, oxide thickness measurements, and post-oxidation imaging.",

author = "Jordan Vandegrift and Parga, {Clemente J.} and Ben Coryell and Jaques, {Brian J.}",

year = "2019",

month = apr,

day = "12",

language = "American English",

}

TY - CONF

T1 - Oxidation of Zirconium Alloy Fuel Cladding Material

AU - Vandegrift, Jordan

AU - Parga, Clemente J.

AU - Coryell, Ben

AU - Jaques, Brian J.

PY - 2019/4/12

Y1 - 2019/4/12

N2 - Zirconium alloys are often used as nuclear fuel cladding. During transients the cladding may be exposed to air at high temperatures, which can cause accelerated oxidation that compromises the structural integrity of the cladding. There exist discrepancies in literature regarding the behavior of zirconium alloys under these conditions. The research presented compares the behavior of Zircaloy-3, Zircaloy-4, and Zr-1Nb during isothermal oxidation (500-820 °C) as well as rapid transient oxidation (≈ 700 °C / s ) scenarios. The effect of alloying elements, plastic deformation, and welding are analyzed. The oxidation behavior is compared using TGA data, oxide thickness measurements, and post-oxidation imaging.

AB - Zirconium alloys are often used as nuclear fuel cladding. During transients the cladding may be exposed to air at high temperatures, which can cause accelerated oxidation that compromises the structural integrity of the cladding. There exist discrepancies in literature regarding the behavior of zirconium alloys under these conditions. The research presented compares the behavior of Zircaloy-3, Zircaloy-4, and Zr-1Nb during isothermal oxidation (500-820 °C) as well as rapid transient oxidation (≈ 700 °C / s ) scenarios. The effect of alloying elements, plastic deformation, and welding are analyzed. The oxidation behavior is compared using TGA data, oxide thickness measurements, and post-oxidation imaging.

UR - https://scholarworks.boisestate.edu/gss_2019/34

M3 - Presentation

ER -

Oxidation of Zirconium Alloy Fuel Cladding Material

Abstract

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