为了研究锆合金在离子辐照下的显微组织演变行为，采用1.8 MeV Ar+在300 ℃下对N36（Zr-1.0Sn-1.0Nb-0.3Fe，质量分数，%）锆合金进行辐照，制备辐照损伤剂量分别为2、5、10和17 dpa的样品。利用扫描电子显微镜和透射电子显微镜表征分析了合金辐照前后的显微组织。结果表明，辐照前合金为完全再结晶组织，第二相主要为hcp-Zr(Nb,Fe)2。随辐照剂量的增加，损伤峰附近的Ar泡密度先增大后趋于稳定，但Ar泡尺寸逐渐增大；合金基体辐照损伤区内均形成了<a>型位错环，位错环密度随辐照剂量的增加而增大，后趋于饱和，但未检测到<c>型位错环；辐照损伤区中的Zr(Nb,Fe)2第二相均发生了非晶化，非晶化的阈值低于0.4 dpa，未发现明显的元素扩散现象。

Zirconium alloys have been used as fuel cladding materials in water-cooled nuclear reactors for decades due to their low thermal neutron absorption cross-section, reasonable mechanical properties, good compatibility with fuels, and good corrosion resistance. During the operation, zirconium alloy cladding materials are subjected not only to the erosion and corrosion from high-temperature and high-pressure cooling water but also to intense neutron irradiation. When zirconium alloys are subjected to irradiation, structure defects including <a> and <c> type dislocation loops and voids or bubbles are often formed in matrix because of the displacement cascades. Neutron irradiation also causes second phase particles (SPPs) amorphization and elemental diffusion. Due to the long period and high cost of neutron irradiation experiments, the radioactivity of irradiated samples, and the effect of ion irradiation on the microstructure of zirconium alloys is similar to that of neutron irradiation, so ion irradiation is often used to simulate the damage introduced by neutron irradiation under laboratory conditions. In order to study the microstructure evolution behavior of zirconium alloys under ion irradiation, N36 (Zr-1.0Sn-1.0Nb-0.3Fe, mass fraction, %) alloy is irradiated with 1.8 MeV Ar+ at 300 ℃ to prepare samples with irradiation damage doses of 2, 5, 10, and 17 displacement per atom (dpa) respectively. Microstructures of the specimens before and after irradiation are characterized and analyzed by SEM and TEM. The results show that the alloy is completely recrystallized before irradiation, and SPPs are mainly hcp-Zr(Nb,Fe)2. With the increase of irradiation dose, the density of Ar bubbles near the damage peak increases first and then tends to be stable, but the size of Ar bubbles increases gradually. <a> type dislocation loops are formed in the irradiation damage area of the alloy matrix. The density of <a> type dislocation loops increases with the dose, and then tends to be stable. However, no <c> type dislocation loops are detected. Zr(Nb,Fe)2 SPPs in the irradiation damage region are all amorphized, and the threshold of amorphization is lower than 0.4 dpa, and no obvious element diffusion phenomenon is found.