Zirconium alloys used as nuclear fuel cladding materials are subject to neutron irradiation inside the reactor, which affects their corrosion resistance. Ion irradiation can be used to simulate neutron irradiation to study the effect of irradiation on corrosion behavior. In this study, the Zircaloy-4 plate was irradiated with Ar_⁺ at 360°C with an electrostatic accelerator. The unirradiated and 5 dpa irradiated samples were corroded in 360°C/18.6 MPa/3.5 ppm Li+1000 ppm B aqueous solution and 400°C/10.3 MPa super-heated steam for 300 days, respectively. The microstructure of the samples was characterized by scanning electron microscopy and transmission electron microscopy. It was found that the Zr(Fe,Cr)₂ second phase particles in the unirradiated samples were hexagonal close-packed structures, and the Fe/Cr atomic ratio was in the range of 1.8 ~ 2.0, and the second phase particles were amorphized after irradiation. Under the two corrosion conditions, during the corrosion process of the irradiated damage zone in the alloy matrix, the oxide film thickness of the irradiated samples was smaller than that of the unirradiated samples, indicating that Ar+ irradiation can enhance the corrosion resistance of Zircaloy-4 to a certain extent. However, once the irradiated damage zone was fully oxidized, as corrosion progressed, the oxide film thickness of the irradiated samples became greater than that of the unirradiated samples, suggesting that the oxide film formed in the irradiated damage zone facilitated the diffusion of oxidative ions, accelerating the corrosion of Zircaloy-4. The influence of irradiation on the corrosion resistance of Zircaloy-4 at different stages is discussed from the perspectives of microstructure evolution and stress accumulation in the oxide film induced by irradiation-induced defects.