Determination of the initial oxidation behavior of Zircaloy-4 by in-situ TEM

Wayne Harlow, Hessam Ghassemi, Mitra L. Taheri, 2016


The corrosion behavior of Zircaloy-4 (Zry-4), specifically by oxidation, is a problem of great importance as this material is critical for current nuclear reactor cladding. The early formation behavior and structure of the oxide layer during oxidation was studied using in-situ TEM techniques that allowed for Zry-4 to be monitored during corrosion. These environmental exposure experiments were coupled with precession electron diffraction to identify and quantify the phases present in the samples before and after the oxidation. Following short-term, high temperature oxidation, the dominant phase was revealed to be monoclinic ZrO2 in a columnar structure. These samples oxidized in-situ contained structures that correlated well with bulk Zry-4 subjected to autoclave treatment, which were used for comparison and validation of this technique. By using in-situ TEM the effect of microstructure features, such as grain boundaries, on oxidation behavior of an alloy can be studied. The technique presented herein holds the potential to be applied any alloy system to study these effects.

Impact Statement

Zirconium-based alloys (Zircaloy) are important materials in nuclear fuel cladding, as they are corrosion resistant and exhibit a low neutron cross section. However, the oxidation behavior is not well understood. Researchers at Drexel University sought to better understand the oxidation of Zircaloys usingused the Atmosphere gas cell, focused ion beam (FIB) sample preparation and precession diffraction analysis. They found that in situ analysis compared well with ex situ analysis techniques, confirming the viability of in situ analysis for this material. Changes in grain structure and including texture and boundaries are reported.