During the loss-of-water accident in the nuclear reactor, the high-temperature steam can react with the Zr alloy cladding tube and lead to oxidation of the Zr alloys, which lead to the embrittlement and fracture of the cladding tube. The process is closely related to the microstructural evolution of the Zr alloys. In order to reveal the embirttlement mechanisms of the Zr-Sn-Nb cladding tube, the steam oxidation experiments of Zr-Sn-Nb alloy at 1000~1250 ℃ were performed. The microstructure was investigated by an optical microscope, a scanning electron microscope and a transmission electron microscope. The hydrogen content was identified by a oxygen, nitrogen and hydrogen analyzer. The results show that the Zr-Sn-Nb alloy can be oxidized into three layers, including ZrO2, α-Zr(O) and Prior-β. The thickness of ZrO2 layer and α-Zr(O) layer increase with increasing oxidation time, meanwhile, the number of cracks increase in α-Zr(O) layer. The lath β-Zr phase of the Prior-β layer is transformed into sheet-like α-Zr phase and grain width of α-Zr phase become larger when the alloy is exposed to longer steam oxidation time. Loose ZrO2 layer with a large number of transverse cracks can be formed in 1000℃ steam, while dense ZrO2 layer can be observed at higher temperatures. H content of the Zr-Sn-Nb alloy matrix increases with the oxidation time. The hydrogen pickup of the Zr-Sn-Nb alloy matrix is much higher at 1000°C steam than those at other temperatures. The orientation relationship between the α-Zr matrix and the hydride is (0002)α-Zr//(-20-2)δ-ZrH1.66 and [2-1-10]α-Zr//[011]δ-ZrH1.66 in 1000℃ steam, and (-2110)α-Zr//(20-2)δ-ZrH1.66 and [01-10]α-Zr//[111]δ-ZrH1.66 in 1200℃ steam.