The Multiple Roles of Small-Angle Tilt Grain Boundaries in Annihilating Radiation Damage in SiC

Interface-Driven Strain in Heavy Ion-Irradiated Zr/Nb Nanoscale Metallic Multilayers: Validation of Distortion Modeling via Local Strain Mapping

Nanolayered metallic alloys are promising materials for nuclear applications thanks to their resistance to radiation damage. Here, we investigate the effect of ion (C, Si, and Cu) irradiation at room temperature with different fluences into sputtered Zr/Nb metallic multilayer films with periods 27 nm (thin) and 96 nm (thick). After irradiation, while a high strain in the entire thin nanoscale metallic multilayer (NMM) is observed, a quite small strain in the entire thick NMM is established. This difference is further analyzed by a semianalytical model, and the reasons behind it are revealed, which are also validated by local strain mapping. Both methods show that within a thick layer, two opposite distortions occur, making the overall strain small, whereas in a thin layer, all the atomic planes are affected by the interface and are subjected to only a single type of distortion (Nb─tension and Zr─compression). In both thin and thick NMMs, with increasing damage, the strain around the interface increases, resulting in a release of the elastic energy at the interface (decrease in the lattice mismatch), and the radiation-induced transition of the Zr/Nb interfaces from incoherent to partially coherent occurs. Density functional theory simulations decipher that the inequality of point defect diffusion flux from the inner to the interface-affected region is responsible for the presence of opposite distortions within a layer. Technologically, based on this work, we estimated that Zr/Nb55 with thicknesses around Zr = 24 nm and Nb = 31 nm is the most promising multilayer system with the high radiation damage resistance and minimum swelling for nuclear applications.

Small-Angle Twist Grain Boundaries as Sinks for Point Defects

It is known that grain boundaries (GBs) provide sinks for defects induced into a solid by irradiation. At the same time radiation can change the atomic structure and chemistry of GBs, which in turn impacts the ability of GBs to continue absorbing defects. Although a number of studies have been reported for tilt GBs acting as defect sinks, the questions of how twist GBs evolve to absorb non-equilibrium concentrations of defects and whether GBs saturate as defect sinks for typical irradiation conditions have remained largely unanswered. Here, we used a combination of molecular dynamics and grand canonical Monte Carlo simulations to determine how twist GBs accommodate point defects. We used SiC and {001} and {111} twist GBs as model systems. We found that diffusion of defects along GBs in this material is slow and for most experimentally relevant conditions point defects will accumulate at twist GBs, driving structural and chemical evolution of these interfaces. During irradiation, screw dislocations within GB planes absorb interstitials by developing mixed dislocation segments that climb. Formation of mixed dislocations occurs either by nucleation of interstitial loops or by faulting/unfaulting of stacking faults. Both types of twist GBs can accommodate a high density of interstitials without losing the crystalline structure, irrespectively of the interstitial flux.