Abstract
Characterisation of materials utilised for fuel cladding in nuclear reactors prior to service is integral in order to understand corrosion mechanisms which would take place in reactor. Zircaloy-4 is one such material of choice for nuclear fuel containment in Pressurised Water Reactors (PWRs). In particular, the metal-oxide interface has been a predominant focus of previous research, however, due to the complex oxidation process of zirconium cladding, there is still no clear understanding of what is present at the interface. Using Scanning Transmission Electron Microscopy (STEM) and Dual Electron Energy Loss Spectroscopy (DualEELS), we have studied the corrosion of this material under conditions similar to those that could be encountered in service. It is shown that under all conditions, whether during faster oxidation in the early stages, slow growth just prior to the transition to a new growth regime, or in the faster growth that happens after this transition, the surface of the metal below the scale is loaded with oxygen up to around 33 at%. Approaching transition, in conditions of slow growth and slow oxygen supply, an additional metastable suboxide is apparent with a thickness of tens of nm. By studying changes in both chemical composition and dielectric function of the material at the oxide scale – metal interface with nanometre resolution, quantitative mapping could be achieved, clearly showing that this is a suboxide composition of ZrO and a Zr oxidation state close to +2.
Highlights
Zirconium alloys are used for fuel cladding elements and other structural components within several commercial designs of nuclear reactor due to their low thermal neutron cross section [1], high corrosion resistance [2], good mechanical properties, and favourable chemical stability in highly aggressive environments
Water corrosion of the fuel containments has become a key factor in the limitation of the lifetime of fuel rods within nuclear reactors, and maintaining containment integrity under corrosion is critical to ensuring safe operation and preventing accidental release of radionuclides into the cooling
By placing a zirconium alloy in contact with hot water it reacts with the water resulting in oxidation of the surface and the liberation of hydrogen
Summary
Zirconium alloys are used for fuel cladding elements and other structural components within several commercial designs of nuclear reactor due to their low thermal neutron cross section [1], high corrosion resistance [2], good mechanical properties, and favourable chemical stability in highly aggressive environments. The association between the transition and lateral cracking in the oxide layer depicts some interaction between the mechanical behaviour of the system, and its corrosion kinetics, but does not provide a clear understanding of the morphology of the metal:oxide interface during the corrosion process, at the nanometre level Understanding why this transition behaviour happens is critical when modelling the rate of growth of oxide, and to the lifetime prediction of Zr clads, and to the safety of nuclear power reactors. For Zircaloy in particular, it was found that it follows a similar oxide sequence as that seen in crystal bar Zr, but with thicker intermediate layers of ZrO and Zr(O)sat They observed the same effect as Ni et al [15] whereby at low corrosion rates, the width of oxygen saturated zirconium was larger. The current paper applies DualEELS on a modern aberration-corrected STEM to the study of the evolution of the suboxide phase through the corrosion process, as well as revealing further details of its chemistry through the analysis of the details of the EEL spectra
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