Dynamic strain aging behavior of accident tolerance fuel cladding FeCrAl-based alloy for advanced nuclear energy


The Fe-13Cr-4Al alloy has become a promising candidate material for accident tolerance fuel (ATF) cladding of light-water reactors (LWRs) due to its excellent oxidation resistance to high-temperature water vapor. However, the tensile deformation behavior of the Fe-13Cr-4Al alloy under different strain rates at different temperatures is still unclear. In the present study, the tensile behavior and deformed microstructure of the Fe-13Cr-4Al alloy were investigated at strain rates from 5 × 10–4 to 1 × 10–2 s−1 in the temperature range from RT to 600 °C. Serrations were observed in the tensile engineering stress–strain curves of the intermediate temperature ranging from 300 to 450 °C at all the three strain rates, indicating the occurrence of dynamic strain aging (DSA). As the strain rate increased, the temperature range where serrated plastic flow occurred shifted to the high temperature regime. Serrated plastic flow occurred after a certain critical plastic strain, and the critical plastic strain decreased with the increase in temperature and increased with the increase in strain rate. In the DSA regime of the Fe-13Cr-4Al alloy, the plateau in the yield strength, ductility minima, negative strain rate sensitivity, the peaks of strain hardening exponent and work hardening rate were observed, which were the typical manifestations of the DSA. The activation energies of serrated plastic flow evaluated by three different methods were 157 ± 35, 92 ± 10 and 93 ± 10 kJ/mol, respectively. According to the values of activation energy, the controlling mechanism responsible for the DSA of the Fe-13Cr-4Al alloy was found to be the interaction between substitutional aluminum atoms and dislocations. The fracture surfaces of the specimens tested at 400 °C under three strain rates showed the mixed fracture mode, which contained dimples and cleavage facets. The tensile samples at 600 °C showed a completely ductile fracture mode with large and deep dimples. A large number of dislocation tangles in the microstructure of the specimens tested at 400 °C under all the strain rates were observed, with a large amount of Fe2Nb Laves particles along the grain boundaries and in the matrix. Dislocation pile-up was also observed around the grain boundaries and second-phase particles. With the increase in the strain rate, the number of subgrains increased, and a clear dislocation cell structure was also observed.

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This work was supported by the National Natural Science Foundation of China [grant number: 51971207,11805293, 51801194, U1904194].

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Correspondence to Hui Wang.

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Zhang, Y., Wang, H., An, X. et al. Dynamic strain aging behavior of accident tolerance fuel cladding FeCrAl-based alloy for advanced nuclear energy. J Mater Sci 56, 8815–8834 (2021). https://doi.org/10.1007/s10853-021-05820-6

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