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Hydrogen embrittlement induced by hydrogen absorption on the inner surface of the cladding is a key failure mode for zirconium alloy fuel claddings in nuclear reactors, with particularly significant risks under conditions such as LOCA (Loss of Coolant Accident). I. Core mechanism of occurrence: Source and diffusion of hydrogen: When the cladding is damaged, steam enters the fuel rod, where high-temperature oxidation reactions occur on the inner surface, generating large amounts of hydrogen; these hydrogen atoms then rapidly penetrate into the zirconium alloy matrix. Cracking due to hydride precipitation: At room temperature, the solubility of hydrogen in zirconium is extremely low; the absorbed hydrogen precipitates as fine, brittle zirconium hydrides (ZrH₁.₅) near the original β-grain boundaries, significantly reducing the material’s ductility. Recent research by Xi’an Jiaotong University has also found that vacancy clusters are the sites where hydrides preferentially nucleate, which explains the \"memory effect\" of hydride precipitation – wherein hydrides repeatedly precipitate at the same locations during thermal cycling, accelerating embrittlement. II. Key influencing factors: Hydrogen concentration threshold – When the oxidation temperature is above 1000°C and the hydrogen absorption level exceeds 200–300 wt.ppm, the cladding becomes brittle ； In the range of 932–972°C, the embrittlement threshold increases to 500–750 wt.ppm. Inner surface atmosphere: The volume ratio of hydrogen to steam, V(H₂)/V(H₂O), has a direct and significant impact on the hydrogen absorption behavior of the cladding; the peak in hydrogen absorption usually occurs in a region 15–45 mm away from the crack. Stress and temperature: Circumferential tensile stress induces the reorientation of hydrides, resulting in a radial distribution of hydrides along the thickness direction of the cladding, which significantly increases the risk of cracking. Reactor design criteria typically limit the hydrogen uptake during operation of zirconium alloy claddings to 250–500 wt.ppm. III. Characteristics for failure determination: The fracture surfaces of failed cladding elements typically exhibit features of intergranular brittle fracture, with almost no visible plastic deformation. Cracks propagate along the precipitated hydrides, and such elements are highly prone to fragmentation under conditions such as quenching, which directly threatens the safety and integrity after a reactor accident.