High Temperature Behavior of Zirconium Alloys
Zirconium alloys are routinely used as cladding material in the nuclear industry due to the favorable mechanical properties and reasonably good corrosion resistance in normal water reactor operating conditions. The effect of transient conditions on zirconium alloys is not well understood. The Idaho National Laboratory has an on-site Transient Reactor Test Facility (TREAT), designed to test the behavior of nuclear fuel in transient conditions. TREAT currently utilizes zircaloy-3 cladding, however other zirconium alloy cladding materials are being considered. The effect of TREAT’s conditions (mixed O2+N2 atmosphere and temperatures up to 820 °C) on zircaloy-3 and zircaloy-4 are compared in this work. It has previously been shown that a mixed O2+N2 atmosphere increases the corrosion rate of these alloys. Additionally, zirconium undergoes a structural phase transformation from HCP→BCC at 863°C, which results in a volume contraction and can compromise the mechanical stability of the material. The high temperature behavior of the zirconium alloys in various atmospheres were investigated using thermogravimetric analysis, differential scanning calorimetry, and dilatometry. The chemical stability and structural phase transformation was investigated with differential scanning calorimetry. Post-test characterization included optical and scanning electron microscopy, and energy dispersive x-ray spectroscopy (EDS). The behavior of these alloys are presented here.
Vandegrift, Jordan, "High Temperature Behavior of Zirconium Alloys" (2017). 2017 Undergraduate Research and Scholarship Conference.