The FeCrAl alloy system is recognized as one of the candidate materials for accident-tolerant fuel (ATF) cladding in the nuclear power industry due to its high oxidation resistance under irradiation and high-temperature environments. The concentrations of Cr and Al have a significant effect on elastic properties of the FeCrAl alloy. In this work, elastic constants C₁₁, C₁₂, C₄₄, bulk modulus and shear modulus of FeCrAl alloy were calculated with molecular dynamics methods. We explored compositions with 1–15 wt.% Cr and 1–5 wt.% Al at temperatures from 0 K to 750 K. The results show that the concentrations of Al and Cr have different effects on the elastic constants. When the concentration of Al was fixed, a decrease in bulk modulus and shear modulus with increasing Cr content was observed, consistent with previous experimental results. The dependence of elastic constants on temperature was also the same as in the experiments. Investigations into elastic properties of defect-containing alloys have shown that vacancies, voids, interstitials and Cr-rich precipitations have different effects on elastic properties of FeCrAl alloys. Investigations of elastic properties of defect-containing alloys have shown that vacancies, void, interstitials and Cr-rich precipitations have different effects on elastic properties of FeCrAl alloys. Therefore, the present results indicate that both the Cr and Al concentrations and radiation defects should be considered to develop and apply the FeCrAl alloy in ATF design.

In the present work, the evolution of atomic structures and related changes in energy state, atomic displacement and free volume of symmetrical grain boundaries (GB) under the effects of external strain in body-centered cubic (bcc) iron are investigated by the molecular dynamics (MD) method. The results indicate that without external strain, full MD relaxations at high temperatures are necessary to obtain the lower energy states of GBs, especially for GBs that have lost the symmetrical feature near GB planes following MD relaxations. Under external strain, two mechanisms are explored for the failure of these GBs, including slip system activation, dislocation nucleation and dislocation network formation induced directly by either the external strain field or by phase transformation from the initial bcc to fcc structure under the effects of external strain. Detailed analysis shows that the change in free volume is related to local structure changes in these two mechanisms, and can also lead to increases in local stress concentration. These findings provide a new explanation for the failure of GBs in BCC iron systems.

FeCrAl is regarded as one of the most promising cladding materials for accident-tolerant fuel at nuclear fission reactors due to its comprehensive properties of inherent corrosion resistance, excellent irradiation resistance, high-temperature oxidation resistance, and stress corrosion cracking resistance. In this work, the irradiation response of FeCrAl irradiated by 2.4 MeV He²⁺ ions with a fluence of 1.1 × 10¹⁶ cm⁻² at room temperature was studied using X-ray diffraction, transmission electron microscopy, and nanoindentation. The characterization results of structural and mechanical properties showed that only black-dot defects exist in irradiated FeCrAl samples, and that the hardness of the irradiated samples was 11.5% higher than that of the unirradiated samples. Similar to other types of radiation defects, black-dot defects acted as fixed defect obstacles and hindered the movement of slip dislocations moving under the applied load, resulting in a significant increase in the hardness of FeCrAl. Importantly, this work points out that irradiation-induced black-dot defects can significantly affect the mechanical properties of materials, and that their contribution to radiation hardening cannot be ignored.

Materials typically experience serious microstructure and performance degradations under irradiation in nuclear reactors. To explore radiation-resistant metals with high design flexibility is urgently requested for the safe application of nuclear energy. In this work, we discover an anti-radiation mechanism for this purpose in a gradient nanostructured nuclear grade austenitic stainless steel prepared by a flexible surface nano-crystallization approach. A special 3-dimensional microstructure network, consisting of low-energy grain boundaries, stacking faults, and dislocation networks, is introduced in the nanostructure, so that a large-scale adaptive martensitic transformation mechanism is activated under irradiation even at extremely high radiation doses and high temperatures. Consequently, the radiation resistance is significantly enhanced, while a superior mechanical property is retained, in nanostructured samples compared to coarse-grained counterparts. Results presented in this work thus explore a strategy to prepare radiation-resistant metals in future.