Hoop Tensile Test of Zirconium Alloy Cladding of Spent Fuel Rod

Zirconium alloys used as cladding in pressurized water reactor (PWR) fuel rods undergo significant mechanical properties changes due to prolonged neutron irradiation and high-temperature coolant corrosion during in-reactor service. This study examined the impacts of irradiation and hydrogen absorption on the mechanical properties of zirconium alloy cladding. Zirconium alloy cladding from spent fuel rods of a commercial pressurized water reactor served as the research subject. Hoop tensile tests were conducted in a hot cell facility to compare the mechanical properties of in-reactor irradiated samples with those of out-of-reactor unirradiated but hydrogen-charged samples. The results demonstrate that irradiation substantially increases strength while reducing ductility, although elevated temperatures restore plasticity and promote ductile fracture modes. Hydrogen absorption exerts a comparatively limited influence: even at high hydrogen concentrations, specimens retain ductility elongations above 40% at elevated temperature and about 20% at room temperature, without evident brittle fracture. Fractographic observations confirm irradiation-induced brittle fracture at room temperature and ductile cup-cone fracture at elevated temperature, whereas hydrogen-charged specimens show no distinct brittle features. Under the present conditions, irradiation governs the mechanical property evolution of in-reactor zirconium alloys, with its strengthening and embrittlement effects exceeding those associated with hydride formation.