TEM characterization of irradiation-induced dislocation loops and voids in ion-irradiated pure chromium

Chromium (Cr) based coatings have been proposed for Zircaloy fuel rods to enhance the accident tolerance in water-cooled reactors. However, the response of Cr to irradiation has received limited attention in prior studies. To get a better understanding of radiation damage in Cr for future nuclear applications, in the present study, the irradiation-induced microstructure evolution of dislocation loops and voids was studied in pure Cr subjected to 2.8 MeV iron ions irradiation at 550 °C (belonging to stage III, intermediate-temperature). TEM observation results show that the distribution of dislocation loops extends the SRIM simulated irradiation depth to 3.5 µm, while voids were only formed in the irradiated region up to a depth of 1.3 µm. A rigorous method with combined consideration of edge-on view and inside-outside analysis was proposed to determine the habit plane. With this method, what has normally considered as an unsafe zone becomes a favored zone for determining the nature of dislocation loops. The nature of dislocation loops was determined including both interstitial and vacancy types with ½<111> Burgers vector in all regions. Within the irradiation region, the loops are approaching half interstitial and half vacancy, while beyond the irradiated region interstitial loops became dominant. The irradiation-induced swelling was quantified by the void size and density. The swellings are mostly corresponding well with the irradiation dose distribution which is simulated by SRIM. The swelling rate close to the peak region is around 0.3%/dpa. Results demonstrate that Cr tends to experience some similarities in irradiation loop evolution to W-based, Mo-based materials, as they are in the same group in the periodic table, despite the discrepancy in the characteristic of loops such as the irradiation temperature-dependent dislocation loop formation was observed as well. The present study is significant not only for supplementing the fundamental knowledge of irradiation damage in Cr but also for the application of Cr coating to the accident tolerant fuel claddings or other nuclear structure materials.