Abstract: 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.