首页|First-principles computational tensile tests of grain boundaries of zirconia with Cu segregation

First-principles computational tensile tests of grain boundaries of zirconia with Cu segregation

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© 2021 Elsevier B.V.Zirconium alloys are widely used as nuclear cladding in pressured water reactor of nuclear power plant, and its corrosion resistance determines the service life of Zr alloys. Alloying can improve the corrosion resistance of Zr alloys, however, the mechanism of alloying element changing the corrosion resistance is still unclear. Since the properties of oxide of Zr alloys are critical to corrosion resistance, we use first-principles calculation to investigate the role of a typical alloying element Cu on the tensile strength of monoclinic zirconia grain boundary (GB) in oxide of Zr alloys. It is found that Cu atom can segregate at GB, and enhance the tensile strength of GB, leading to the improvement of corrosion resistance of Zr alloys. Moreover, we also check the results from calculations with various exchange-correlation functionals, we confirm that it is necessary to use generalized gradient approximation (GGA) plus a Hubbard U correction or hybrid functional instead of standard GGA to calculate the tensile strength of GB with Cu.

Corrosion resistanceFirst-principles computational tensile testsGrain boundaryZirconium alloys

Xu C.-H.、Xie Y.-P.、Yao M.-Y.、Hu L.-J.、Lin X.-D.

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Institute of Materials School of Materials Science and Engineering Shanghai University

2022

10.1016/j.jnucmat.2021.153498

Journal of Nuclear Materials

Journal of Nuclear Materials

EISCI
ISSN：0022-3115
年,卷(期)：2022.560
  • 61