Zirconium-based alloys are one of the most significant materials in thermal-neutron reactor systems. With very low neutron capture cross section, good corrosion resistance, mechanical strength and resistance to neutron radiation damage, zirconium alloys are used for fuel claddings. Cladding materials are still improved and tested in normal as well as critical reactor conditions. Zircaloy-4 (Zr-1.5Sn-0.2Fe-0.1Cr) is used for west types of light-water reactors, Pressurized Water Reactors (PWR). In our study, Zircaloy-4 cladding tubes were high-temperature oxidized in steam at the series of temperatures from 950 up to 1425 °C to simulate PWR reaching severe accident conditions. To observe the influence of hydrogen (H) diffusing from the coolant water on oxidation process, the specimens with ∼1000 ppm H were compared to the specimens with almost no hydrogen content. Wave Dispersive Spectroscopy (WDS) and nanoindentation were performed in line profiles across the cladding wall. Both methods contributed to verify the pseudobinary Zircaloy-4/oxygen phase diagram with focus on determination of phase boundaries. The increase of oxygen concentration with increasing temperature was observed. Moreover, oxygen concentration profiles and related change in nanohardness and Young's modulus showed the effect of hydrogen on the cladding microstructure. Hydrogen dissolved in metallic matrix increases the oxygen solubility in prior β-phase, the specimens with 1000 ppm H showed the higher oxygen content at almost all temperatures. As well, material hardening was observed on specimens with 1000 ppm H with significant difference in β-phase, measured on specimens exposed to lowest and highest oxidation temperature. Thus, with increasing temperature and hydrogen content, increased oxygen solubility affects the cladding ductility.