Abstract
Ferritic/martensitic (F/M) steels whose matrix is Fe-Cr are important candidate materials for fuel cladding of fast reactors, and they have excellent irradiation-swelling resistance. However, the mechanism of irradiation-swelling of F/M steels is still unclear. We use a first-principles method to reveal the influence of irradiation defects, i.e., Frenkel pair including atomic vacancy and self-interstitial atom, on the change of lattice volume of Fe-13Cr lattice. It is found that vacancy causes lattice contraction, while a self-interstitial atom causes lattice expansion. The overall effect of a Frenkel pair on the change of lattice volume is lattice expansion, leading to swelling of the alloy. Furthermore, the diffusion properties of point defects in Fe-13Cr are investigated. Based on the diffusion barriers of the vacancies and interstitial atoms, we find that the defects in Fe-13Cr drain out to surfaces/grain boundaries more efficiently than those in pure α-Fe do. Therefore, the faster diffusion of defects in Fe-13Cr is one of important factors for good swelling resistance of Fe-13Cr compared to pure α-Fe.
Highlights
Ferritic/martensitic (F/M) steels have excellent high-temperature strength, creeprupture strength, corrosion resistance, neutron irradiation resistance, and low price, which are promising candidates for the new generation of cladding materials of fast reactors [1,2]
The volume change of Fe-13Cr induced by the Frenkel pair has a wide range, while the volume change of pure α-Fe induced by the Frenkel pair is close to the maximum of these for Fe-13Cr. This indicates that volume expansion induced by the Frenkel pair has a chance of being smaller than that of pure α-Fe, which is a conducive factor contributing to the swelling resistance of Fe-13Cr being better than that of pure α-Fe
The thermodynamic and kinetic properties of irradiation defects that are related to swelling of Fe-Cr alloys are investigated
Summary
Ferritic/martensitic (F/M) steels have excellent high-temperature strength, creeprupture strength, corrosion resistance, neutron irradiation resistance, and low price, which are promising candidates for the new generation of cladding materials of fast reactors [1,2]. Zheng et al, used in situ TEM to study the commercial F/M steels HT9, and they found that the size and density of dislocation loops and dislocation lines depends on dose and temperature [7] They performed irradiation on HT9 in the BOR-60 reactor, Academic Editors: Meng Wu and Anderson Janotti. Since first-principles calculation can provide reliable structural and thermodynamic properties of defects in materials, we perform first-principles calculation to investigate the thermodynamic and kinetic properties of Frenkel pairs in F/M steels, including the stability of the defects, the influence of defects on the lattice volume and the diffusion barriers of defects, and analyze their effects on irradiation swelling
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