Zircaloy fuel element cladding changes dimensions during service in a pressurized water reactor (PWR) as a result of stress-free irradiation-induced growth, creep-driven by fuel pellet expansion and hydriding. The application of a tensile load in a high-temperature autoclave environments has previously been reported to increase the corrosion rate of Zircaloy, and heat treatments (beta quenching) that reduce the irradiation-induced stress-free growth of Zircaloy have previously been reported to reduce Zircaloy corrosion in-reactor. However, the effect of in-situ straining on Zircaloy corrosion in a PWR environment has not been systematically studied and reported in the literature. This paper presents experimental results regarding the effect of in-situ straining on Zircaloy corrosion in a PWR environment, both in-reactor and in an autoclave. In-situ electrochemical data and post-test metallographic data are presented. In-situ straining is seen to increase the corrosion rate of the Zircaloy, likely by breaking the passivating layer that is forming on the surface. However, the effect is a function of the applied strain rate.