Zircaloy-2 and its almost identical sister alloy Zr-4 are used extensively as fuel cladding, as well as other components in nuclear power reactors where it operates at 250-350C and also as the calandria tubes in CANDU reactors which operate at 80C. The high cycle fatigue properties are important in the calandria tube application, where the potential effects of intergranular constraints from manufacturing are unknown. In fuel cladding applications, the low cycle fatigue properties of these alloys are extremely important, especially when they contain hydrogen picked up from the corrosion reaction with the heavy water coolant. This proposal is a first step to investigate the internal strain response of these materials during in situ fatigue at realistic temperatures.