Zircaloy-2 and Zircaloy-4 were widely used as cladding materials in commercial light water reactors due to their corrosion resistance, low neutron absorption, and mechanical durability. After reactor discharge, the cladding of the spent nuclear fuel can experience elevated temperatures for extended periods during dry storage, potentially altering its mechanical properties. While prior research often focuses on irradiation damage, less attention has been given to isolating the thermal response of unirradiated, as-manufactured cladding. This study examines how heat treatments ranging from 300°C to 570°C and lasting 6.8 to 2064 hours affect the mechanical behavior of unirradiated Zircaloy-2 and Zircaloy-4. Vickers microhardness testing at room temperature was used to assess changes in hardness, while axial tensile testing at both room temperature and 200°C, supported by a digital image correlation extensometer, was used to evaluate mechanical performance. Scanning electron microscopy micrographs were also collected on select samples to qualitatively examine changes in microstructure across heat treatment conditions. Heat treatment generally led to reductions in yield strength, ultimate tensile strength, and hardness compared to as-manufactured baselines, with more significant degradation at higher temperatures and longer durations. Hardness trends were compared with tensile data to assess whether microhardness can serve as a proxy for overall mechanical performance. These findings establish a baseline for strength evolution in thermally treated, unirradiated zirconium cladding and may inform future material assessments and modeling efforts related to dry storage.
Maddison Ann Rex (Wed,) studied this question.