It is essential to understand basic deformation mechanism(s) of conventional alloys in order to develop improved or novel alloys for their applications in much more challenging conditions. Zircaloy-4 is extensively used in pressurized water reactor for nuclear fuel cladding application. It operates at very high temperature in the presence of mechanical loads, corrosive atmosphere, and neutron irradiation environment. Present work explores the fundamental plastic deformation mechanism(s) of Zircaloy-4 in the temperature range 20 to 600 °C by subjecting tensile samples to uniaxial tensile loads under quasi-static deformation conditions. Based on the results of uniaxial tensile testing as a function of temperature, repeated stress-relaxation experiments were carried out to determine the activation volume of the alloy at 20 and 500 °C. The results from uniaxial tensile and stress-relaxation testing were used to gain insight into potential deformation mechanism(s) in Zircaloy-4.

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