Abstract: The damage of fuel elements in pressurized water reactor (PWR) will cause changes in the performance and behavior of the core blocks inside the elements, such as microstructure evolution and fission gas release, which directly affects the operation of the reactor. The microstructural evolution and fission gas release behavior in UO₂ fuel pellets following cladding failure in PWR were systematically investigated to enhance nuclear safety assessment. Fuel samples, including both intact and failed rods from a commercial nuclear power plant with 4.45% enriched UO₂ pellets, were carefully selected for analysis. Specimens were precisely extracted from three critical positions relative to the breach: the breach center, breach edge, and approximately 7 mm away from the breach. Metallographic analysis, electron probe microanalysis (EPMA), and scanning electron microscopy (SEM) were employed to characterize microstructural features and fission gas distribution. The radial porosity distribution and residual xenon (Xe) content, with neodymium (Nd) used as a quantitative reference due to similar fission yields, were meticulously measured across different radial zones. Results demonstrate that failed fuel pellets develop distinct annular, circumferentially stratified porosity bands along the radial direction. The edge region retains the high burnup structure (HBS), while the central region exhibits significantly increased pore density. In the intermediate zone near 0.6r/r₀, pore migration occurs with partial formation of columnar grains. The Xe release rate decreases with increasing distance from the breach. At the breach center, approximately 46% of Xe is released from the central region and 31% from the edge region of the pellet. Fracture surface analysis reveals numerous isolated gas bubbles at grain boundaries without interconnected channels, indicating that fission gas release occurs primarily through atomic diffusion rather than via percolation pathways. The microstructural evolution and enhanced fission gas release are attributed to the synergistic effects of elevated temperature (estimated at 1 400-1 600 ℃ in the center), steep thermal gradients, and oxidation-induced hyper-stoichiometry (UO_2+x) following cladding failure. These experimental findings provide critical data for improving mechanistic models of defective fuel behavior and enhancing safety protocols for managing failed fuel rods during reactor operation, particularly regarding the prediction of fission product release into the coolant system.