The Effects of Irradiation on the Improvement in Oxidation Behavior of MX-ODS Steel in Liquid Pb.

Lead-cooled fast reactors exhibit strong inherent safety performance and good economic features, while material degradation due to corrosion and irradiation is still challenging. Oxide dispersion-strengthened steels are one of the promising candidates for fuel cladding materials. The effects of both irradiation and corrosion on ODS steel need to be further studied. In this work, MX-ODS steel was irradiated by Fe ions at 500 °C up to 46 dpa. Later, the as-received specimen and the irradiated specimen were used to conduct corrosion tests in oxygen-saturated Pb at 550 °C for 1 h. In the as-received specimen, discontinuous oxides penetrated by Pb and Pb in contact with steel matrix were observed, demonstrating unsatisfactory corrosion resistance of the material. However, in the irradiated specimen after corrosion experiment, a protective oxide layer formed and prevented Pb attack. The oxidation behavior differences between the two specimens can be attributed to the defects produced by irradiation and the structural discrepancy in oxides caused by the formation process. A possible mechanism of irradiation on the corrosion is discussed. In the as-received specimen, Fe atoms loss led to voids in the oxides, and lead penetrated the oxides through these voids. In the irradiated specimen, defects left by previous irradiation helped to form a more uniform oxide layer. The adhesive outer magnetite oxide and the Fe ions generated from where grain boundary oxidation developed retarded the presence of voids and made the oxide layer protective.

Investigation of the dissolution and diffusion properties of interstitial oxygen at grain boundaries in body-centered-cubic iron by the first-principles study.

Oxidation corrosion of steel is a universal problem in various industries and severely accelerated in nuclear reactors. First-principles calculations are performed to explore the dissolution and diffusion properties of interstitial oxygen in the body-centered-cubic iron grain boundaries Σ3〈110〉(111) and Σ5〈001〉(310). Solution energies indicate that interstitial oxygen atoms prefer to dissolve in body-centered-cubic iron, and energetically segregate to grain boundaries. Energy barriers show that oxygen atoms would segregate towards Σ3〈110〉(111) with a low energy barrier. However, they concentrate to the transition region of Σ5〈001〉(310) due to the high-energy barrier in the transition zone. When O atoms arrive at grain boundaries, they would stay there due to the larger solution energy and diffusion energy barrier in grain boundaries compared to that in the defect-free Fe bulk. These results indicate that O atoms would prefer to diffuse through the bulk, and oxidize grain boundaries. This study provides insight into oxidation phenomena in experiments and necessary parameters for future studies on the oxidation of steel under irradiation in nuclear reactors.