Zirconium alloys are widely used as the fuel cladding material in pressurized water reactors, accumulating a significant population of defects and dislocations from exposure to neutrons. We present and interpret synchrotron microbeam X-ray diffraction measurements of proton-irradiated Zircaloy-4, where we identify a transient peak and the subsequent saturation of dislocation density as a function of exposure. This is explained by direct atomistic simulations showing that the observed variation of dislocation density as a function of dose is a natural result of the evolution of the dense defect and dislocation microstructure driven by the concurrent generation of defects and their subsequent stress-driven relaxation. In the dynamic equilibrium state of the material developing in the high dose limit, the defect content distribution of the population of dislocation loops, coexisting with the dislocation network, follows a power law with exponent $\alpha \approx 2.2$. This corresponds to the power law exponent of $\beta \approx 3.4$ for the distribution of loops as a function of their diameter that compares favourably with the experimentally measured values of $\beta$ in the range $3\le \beta \le 4$.

锆合金被广泛用作压水反应堆中的燃料包壳材料，由于暴露于中子而积累了大量的缺陷和错位。我们提出并解释了质子辐照的 Zircaloy-4 的同步加速器微束 X 射线衍射测量值，其中我们确定了一个瞬态峰值和随后的位错密度饱和度作为曝光的函数。这可以通过直接原子模拟来解释，表明观察到的位错密度随剂量的变化是致密缺陷和位错微观结构演变的自然结果，这些缺陷和位错微观结构是由同时产生的缺陷及其随后的应力驱动松弛驱动的。在高剂量极限发展的材料动态平衡状态下，$\alpha \approx \mathrm{2个}.\mathrm{2个}$. 这对应于幂律指数$\beta \approx \mathrm{3个}.\mathrm{4个}$用于环的分布作为其直径的函数，与实验测量值相比具有优势$\beta$在范围中$\mathrm{3个}\le \beta \le \mathrm{4个}$.