Mechanisms of Uniform Corrosion of Zirconium Alloys in Water and Steam
Doctoral thesis, 1993
Zirconium alloys are used in key components in nuclear reactors, e.g. fuel cladding tubes in pressurised water reactors. Current trends towards extended burnup of the nuclear fuel in these reactors have accentuated the demand for Zr alloys with higher uniform corrosion resistance under irradiation and lower hydrogen absorption.
The first part of this work was performed to investigate the matrix composition, that was proposed to influence the uniform corrosion resistance of Zircaloy, the traditional fuel cladding material. The first direct measurements were made of the concentrations of the alloying elements Fe, Cr and Ni in the matrix in five Zircaloy materials, using atom probe microanalysis. In all five materials the matrix was highly depleted with respect to alloying elements. Only 100-200 wt.ppm of Fe, Cr and Ni remained in the matrix, which is approximately 10% of the bulk composition. The measured concentrations correspond well to recent diffusion and solubility data for Zircaloys.
In the second part of this study the oxide microstructure was related to the corrosion resistance and hydrogen uptake of zirconium alloys with different composition and heat treatments.
The initial oxidation in air or water of Zr-4 needles was studied with atom probe analysis and transmission electron microscopy (TEM). The grain size of the thin oxide layers was approximately 5 nm, and the oxide composition was ZrO.
Thin foil cross-sections were prepared of the metal-oxide interface and of the oxide layers grown in static autoclave at 400 °C steam on Zircaloy-4 and Zr-0.5Sn- 0.53Nb. Oxides in the thickness range 1-42 µm were examined with TEM. For comparison, one low-annealed Zr-4 material with high uniform corrosion rate was also investigated.
In the metal-oxide interface of all materials, small columnar ZrO2 grains (Ã? 15 x 100 nm) were observed growing in direct contact with the metal. No amorphous layer was found in the oxide-metal interface or between columnar grains. Thin oxide layers (pre-transition, Ã? 1 µm) were dense, while thick (post transition) oxide layers contained some cracks close to equiaxed grains.
On the low-annealed Zr-4, thick oxide layers contained a high number of equiaxed grains, and intergranular pores and cracks were frequent even close to the interface. The metal-oxide interface had a less ordered structure, with shorter columnar or equiaxed oxide grains.
In post-transition oxide layers of the Zr-Sn-Nb alloy (which had a lower relative hydrogen uptake) a crystalline 20-100 nm thick intermediate layer in the metal-oxide interface was found.