Exploring a Surrogate of Pellet-Cladding Interaction—Characterization and Oxidation Behavior

The thermal gradient at which the nuclear fuel is subjected during operation is one of the reasons of the deformation caused in the pellet, which could affect to its interaction with the cladding. Thus, a potential pellet-cladding contact could take place because of a decrease in the cladding diameter (creep-down due to pressure from the coolant) and an increase in the pellet diameter (thermal expansion, swelling due to the inclusion of solid fission products in the matrix and inter- and intra-granular accumulation of fission gases in pores). Contact between the pellet and the cladding, through a bonding layer called Fuel Cladding Chemical Interaction (FCCI) layer, first occurs at the inter-pellet spaces while a continuing pellet fragmentation (radial and axial cracks) takes place simultaneously. Previous studies conducted on fuel with burnups higher than 55 MWdxkgU^-1 showed the formation of a restructured region, named as “rim” structure or High Burn-up Structure (HBS) at the periphery of UO₂ pellets [1]. The risk of enhanced Pellet-Cladding Interaction (PCI) grows with burnup. As the burnup increases, some effects turn out to appear in this new region, such as smaller grains (0.1 μm vs initial grains of ~10 μm diameter), lattice contraction controlled by recrystallization, higher porosity, hardness decrease and the closure of pellet-cladding gap in fuel rods [2]. The FCCI layer consists in principle of two regions, one closer to the cladding (cubic polycrystalline ZrO₂) and a second nearer to the fuel pellet, characterized by an interphase formed by cubic solid solutions of (U,Zr)O₂ and an amorphous phase, with variable relative concentrations of U and Zr.<br/>PCI is considered as a mid-priority process in the investigation of the safety approach for LWR reactor fuel due to the possibility of cladding failure during a power transient. In addition, it could be one of the causes identified as leading to potential fuel failure. Understanding the potential chemical oxidation resistance of UO₂ (matrix fuel) to U₃O₈ as a consequence of ZrO₂/Zr system in case of undetected damaged cladding (zirconium alloy), a potential air intrusion is relevant in terms of assessing fuel cladding integrity.<br/>To gain insight into the PCI response to oxidation, in the present work, a collection of Zr-doped UO₂ (0, 20, 40, 80 and 100 %) pellets were prepared via solid-state synthesis by mimicking the chemical bonding between ZrO₂ and UO₂. After sintering, the Zr-doped UO₂ monoliths were characterized by (i) surface morphology and average grain size calculation by SEM; (ii) BET Specific Surface Area (SSA) with N₂; (iii) evaluation of the purity and the crystalline structure by both XRD and Rietveld Quantitative Phase Analysis (RQPA); (iv) Raman spectroscopy; and (v) geometrical and experimental density obtained by Archimedean immersion.<br/>Theinterpretation of these characterizations shows non-uniform distribution of Zr in the UO₂ matrix and ZrO₂ segregation in grain boundaries, presumably because the solubility limit has been reached in the fabrication procedure. Controlled air oxidation of Zr-doped UO₂ samples has been monitored by thermogravimetric analysis (TGA). Based on those results, we observe a profound effect of delayed oxidation with the addition of Zr to UO₂, and then, an increased resistance of UO₂ to oxidation to U₃O₈ when compared to pure UO₂ pellet_.<br/>[1] K. Lassmann, C. T. Walker, J. van de Laar, and F. Lindström, "Modelling the high burnup UO2 structure in LWR fuel," <i>Journal of Nuclear Materials, </i>vol. 226, pp. 1-8, 1995/10/01/ 1995.<br/>[2] U. S. NRC, "Standard Review Plan for Dry Cask Storage Systems Final Report," Washington2010.