Microstructural Analysis of Zirconia at the Fuel-Cladding Interface in Medium and High Burnup Irradiated Fuel Rods.

Autor: Schneider, C., Fayette, L., Zacharie-Aubrun, I., Blay, T., Sercombe, J., Favergeon, J., Chevalier, S.
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Zdroj: Oxidation of Metals; Oct2021, Vol. 96 Issue 3/4, p295-306, 12p
Abstrakt: During irradiation in a nuclear reactor, the UO2 fuel and the Zr cladding come into contact due to fuel thermal expansion and cladding creep. The inner surface of the Zr cladding in contact with the UO2 fuel therefore oxidizes. This paper analyzes the microstructure of zirconia formed at the Zr cladding/UO2 interface in two samples with different burnups. It seems that the zirconia develops 'normally' until it comes into contact with the fuel. The development of internal zirconia is, however, constrained mechanically by the contact of the UO2 fuel. Four different zones can be distinguished in the zirconia layer of each sample. In the medium-burnup sample (37 GWd/tU), the zirconia in the first zone in contact with the cladding exhibits the same kind of microstructure as 'ordinary' zirconium oxide formed on the cladding outer surface, which is characterized by columnar and equiaxed grains. Thereafter, two intermediate zones can be observed: a first composed of very small equiaxed grains and a second with a mixture of intermediate and large grains. In the high-burnup sample (61 GWd/tU), zirconia presents a wavy interface with some circumvolutions. The ZrO2 in contact with the fuel is more developed in this case than in the medium-burnup sample, but it shows the same microstructure. [ABSTRACT FROM AUTHOR]
Databáze: Complementary Index
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