TY - JOUR
T1 - Oxidation Behavior of Zirconium, Zircaloy-3, Zircaloy-4, Zr-1Nb, and Zr- 2.5Nb in Air and Oxygen
AU - Vandegrift, Jordan L.
AU - Price, Patrick M.
AU - Stroud, John-Paul
AU - Parga, Clemente J.
AU - Van Rooyen, Isabella J.
AU - Jaques, Brian J.
AU - Butt, Darryl P.
N1 - Vandegrift, Jordan L.; Price, Patrick M.; Stroud, John-Paul; Parga, Clemente J.; Van Rooyen, Isabella J.; Jaques, Brian J.; and Butt, Darryl P. (2019). "Oxidation Behavior of Zirconium, Zircaloy-3, Zircaloy-4, Zr-1Nb, and Zr- 2.5Nb in Air and Oxygen". Nuclear Materials and Energy, 20, 100692-1 - 100692-9. https://doi.org/10.1016/j.nme.2019.100692
PY - 2019/8/1
Y1 - 2019/8/1
N2 - The Transient Reactor Test (TREAT) facility at the Idaho National Laboratory currently utilizes a legacy Zircaloy- 3 cladding, which is no longer commercially available. TREAT is air cooled and routinely operates at temperatures well above that of traditional reactor designs. This study investigates the oxidation behavior of pure zirconium and its alloys (Zircaloy-3, Zircaloy-4, Zr-1Nb, Zr-2.5Nb) in Ar+20%O 2 and N 2 +20%O 2 atmospheres at temperatures ranging from 400–800 °C to determine which alloy should be implemented as TREAT's cladding. While the oxidation behavior of zirconium based cladding materials has been extensively documented, this study focuses on direct comparison between legacy Zircaloy-3 and contemporary alloys using a flat plate geometry and similar conditions seen at the TREAT facility. In this work, thermogravimetric analysis was used to measure both steady state and breakaway oxidation, which was then used to calculate oxidation rate constants and activation energies of each material. Oxide thickness was evaluated through microscopy of oxidized specimen cross sections. The Zircaloy-3 and Zr-1Nb alloys were found to be the most resistant to oxidation under the conditions of this study, whereas the Zr-2.5Nb alloy was found to be the most susceptible.
AB - The Transient Reactor Test (TREAT) facility at the Idaho National Laboratory currently utilizes a legacy Zircaloy- 3 cladding, which is no longer commercially available. TREAT is air cooled and routinely operates at temperatures well above that of traditional reactor designs. This study investigates the oxidation behavior of pure zirconium and its alloys (Zircaloy-3, Zircaloy-4, Zr-1Nb, Zr-2.5Nb) in Ar+20%O 2 and N 2 +20%O 2 atmospheres at temperatures ranging from 400–800 °C to determine which alloy should be implemented as TREAT's cladding. While the oxidation behavior of zirconium based cladding materials has been extensively documented, this study focuses on direct comparison between legacy Zircaloy-3 and contemporary alloys using a flat plate geometry and similar conditions seen at the TREAT facility. In this work, thermogravimetric analysis was used to measure both steady state and breakaway oxidation, which was then used to calculate oxidation rate constants and activation energies of each material. Oxide thickness was evaluated through microscopy of oxidized specimen cross sections. The Zircaloy-3 and Zr-1Nb alloys were found to be the most resistant to oxidation under the conditions of this study, whereas the Zr-2.5Nb alloy was found to be the most susceptible.
KW - breakaway
KW - oxidation
KW - thermogravimetric analysis
KW - zirconium alloys
UR - https://scholarworks.boisestate.edu/mse_facpubs/395
UR - http://www.scopus.com/inward/record.url?scp=85070370173&partnerID=8YFLogxK
U2 - 10.1016/j.nme.2019.100692
DO - 10.1016/j.nme.2019.100692
M3 - Article
SN - 2352-1791
VL - 20
JO - Nuclear Materials and Energy
JF - Nuclear Materials and Energy
M1 - 100692
ER -