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June 13 @ 9:00 am - 10:00 am MDT
Title: Oxidation Behavior of Zirconium Alloys in Transient Conditions
Program: Master of Science in Materials Science and Engineering
Advisor: Dr. Brian Jaques, Materials Science and Engineering
Committee Members: Dr. Michael Hurley, Materials Science and Engineering, and Dr. Clemente Parga, Materials Science and Engineering
The Transient Reactor Test (TREAT) facility, located at the Idaho National Laboratory, is a necessary platform for the advancement of nuclear power across the world. TREAT’s fuel is being converted from highly enriched uranium to low-enriched uranium. With this conversion, a new cladding material must be identified that performs as well as, or improves the performance, of the legacy Zircaloy-3 cladding. TREAT’s operating conditions are unique compared to light water reactors due to its design to simulate transient conditions. TREAT is an air-cooled test reactor designed to replicate rapid transients to 400-600 °C at rates up to 700 °C/s. The goal of this work is to directly compare the oxidation behavior of candidate zirconium alloy cladding materials to understand the effects of plastic deformation (chamfering), welding, and nitrogen in the oxidizing atmosphere. Through these studies, it is determined which zirconium alloy would be best suited as nuclear fuel cladding material in the TREAT facility.
Isothermal and non-isothermal oxidation studies were completed on pure Zr, Zircaloy-3 (Zry-3), Zircaloy-4 (Zry-4), Zr-1Nb, and Zr-2.5Nb plate specimens in both Ar+20%O2 and N2+20%O2 to study the effect of nitrogen on the oxidation behavior. Electron beam welded (EBW) and tungsten inert gas (TIG) welded Zry-3, Zry-4, and Zr 1Nb tube samples were oxidized under rapid heating isothermal conditions in dry and humid N2+20%O2. Macroscopic images of the samples were taken after oxidation, the oxide thickness was measured, and mass gain data was used to determine the oxidation rate constants and activation energies. It was found that Zry-3, Zry-4, and Zr-1Nb experience faster oxidation in N2+20%O2 than Ar+20%O2 at 800 °C, while Zr and Zr-2.5Nb were relatively unaffected. Zry-3, Zry-4, and Zr-1Nb were found to experience accelerated oxidation in the weld region. Additionally, Zry-3 and Zry-4 experienced accelerated oxidation at the chamfers, while the chamfered region of Zr-1Nb experienced less oxidation. In all oxidation experiments, Zr 1Nb had the most favorable oxidation behavior.