Oxidation Behavior of Welded Zry-3, Zry-4, and Zr–1Nb Tubes

The Transient Reactor Test (TREAT) facility is a research reactor designed to simulate rapid transients to test new fuel designs. TREAT\u27s cladding is exposed to unique conditions compared to normal water reactors. These conditions include: exposure to air at high temperatures (≥600 °C), rapid heating (≈700 °C/s), and cladding geometry that includes chamfers and welds. This work investigates the effects of chamfering and welding on the oxidation behavior of zirconium alloys (Zircaloy-3, Zircaloy-4, and Zr–1Nb). Tube specimens were examined under isothermal and transient conditions in dry and humid air. The effect of weld type (tungsten inert gas or electron beam), the number of welds, and alloying elements are compared. Thermogravimetric analysis was used to collect mass gain data during isothermal oxidation and the data was used to quantify the oxidation rate constant and the activation energy of oxidation. Oxide behavior in the weld region, chamfered region, and bulk tube was measured and compared. The microstructure and secondary phase precipitates in EBW tubes before and after breakaway were characterized. The electron beam welded Zr–1Nb specimen was found to have the most favorable oxidation behavior under both isothermal and transient conditions. Zry-4 oxidized the most readily and was the most affected by mechanical deformation

Oxidation of Zirconium Alloy Fuel Cladding Material

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

Oxidation Behavior of Zirconium, Zircaloy-3, Zircaloy-4, Zr-1Nb, and Zr- 2.5Nb in Air and Oxygen

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%O2 and N2+20%O2 atmospheres at temperatures ranging from 400–800 °C to determine which alloy should be implemented as TREAT\u27s 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