Modeling and tests on the damage and fracture behaviors of non-irradiated zirconium alloys with different hydrogen concentrations at RT and 350°C

Abstract

Zirconium alloys are widely used as the materials of nuclear cladding tubes or some other structural components. The chemical reaction of zirconium with coolant water leads to hydrogen pickup during reactor exposure, degrading the mechanical properties of zirconium alloy tubes. In this study, the load-displacement curves and fracture morphologies of ring-typed un-irradiated zirconium alloy specimens with various hydrogen contents are obtained at RT and 350 °C by circumferential tensile tests. An efficient modeling strategy for the damage and fracture behaviors of hydrided zirconium alloy specimens is proposed and numerically implemented, reproducing the local necking, damage initiation and evolution, crack propagation and ultimate fracture of the specimens. The predictions of load-displacement curves for different hydrided specimens agree well with the experimental data, and the fracture morphologies are also captured. The combinations of experimental tests and numerical simulation herein successfully identify the mechanical constitutive models of hydrided zirconium alloy materials, well describing their hydrogen-concentration related characteristics of strain hardening, softening and collapsing rupture. In the considered mean hydrogen concentrations less than 800 ppm, the hydride-induced hardening and embrittlement are predicted to appear dominantly at room temperature (RT); these effects are negligible at 350 °C, resulting from the greatly improved ductility of hydrides and zirconium alloy matrix at high temperatures. This work lays a foundation for simulation of the failure behaviors of hydrided zirconium alloy tubes.