Abstract: During the in-reactor operation of pressurized water reactor (PWR) fuel rods, hydrogen from the coolant environment diffuses into the zirconium alloy cladding, leading to the precipitation of hydrides within the zirconium matrix. This phenomenon represents a critical issue in nuclear materials because hydride formation embrittles the cladding material, which can initiate crack propagation and ultimately result in material failure. Conventional methods for quantifying the hydride content in damaged fuel rods rely mainly on destructive analysis techniques, such as post-sectioning metallographic examination or high temperature hydrogen extraction. However, these approaches suffer from significant limitations, including the complexity of preparing irradiated samples and the inability to examine multiple locations on a single specimen. Such constraints hinder a comprehensive assessment of cladding integrity under operating conditions. To address the technical challenges in hydrogen analysis, this study employed neutron imaging as an advanced non-destructive testing method. The fundamental principle of neutron imaging is based on the difference in neutron attenuation between hydrogen and zirconium atoms. Through dedicated data-processing algorithms, the neutron attenuation data were converted into hydrogen distribution maps, with corrections applied for factors such as material thickness and the neutron energy spectrum. The results clearly demonstrate that neutron imaging can effectively visualize the distribution patterns of hydrides within the cladding, revealing pronounced hydride enrichment adjacent to cracks and other structural defects. Semi-quantitative analysis indicates that the average hydrogen concentration on the surface of the irradiated fuel rod is about 100 ppm, while localized hydrogen concentrations in hydride enriched regions near cracks reach approximately 5 000-10 000 ppm. This work validates the feasibility of neutron imaging to characterize the hydrogen content in damaged fuel rods. The findings contribute to optimizing the in-service performance of zirconium alloy cladding materials and provide a basis for future design improvements aimed at enhancing reactor safety and longevity.