Preparation and characterization of UO2-ZrO2 model materials as analogues of spent nuclear fuel produced by co-precipitation. (IONMAT PROJECT. SUBTASK 1.3)

Abstract

UO2-ZrO2 mixed oxide materials are representative analogues of the interaction products formed at the fuel-cladding interface of spent nuclear fuel during reactor operation. The characterization of these phases is of considerable interest for understanding fuel performance and long-term behaviour during storage and disposal. In this work, a series of UO2-ZrO2 model materials covering the full compositional range from pure UO2 to pure ZrO2 (0, 20, 40, 60 and 100 wt.% ZrO2) were synthesized by an aqueous co-precipitation route within the framework of the IONMAT project. Following precipitation, the powders were subjected to calcination-reduction treatments under H2/Ar atmosphere, pressed into pellets and sintered at 1675 °C under reducing conditions. The obtained materials were characterized by XRD, Raman spectroscopy, helium pycnometry, BET surface area measurements and SEM-EDS. XRD analysis revealed a progressive structural evolution from cubic fluorite UO2 to monoclinic ZrO2 with increasing zirconia content. Intermediate compositions exhibited mixtures of cubic and tetragonal (U,Zr)O2 solid-solution phases, demonstrating partial mutual incorporation of uranium and zirconium within the oxide lattice. Raman spectroscopy confirmed this phase evolution and provided evidence of increasing local structural disorder and zirconia-rich phase formation as the zirconium content increased. Density measurements showed improved sintering behaviour for UO2-rich compositions, with pycnometric density increasing from approximately 5.5 g·cm-3 for pure ZrO2 to 9.7 g·cm-3 for pure UO2. Correspondingly, relative density increased and total porosity decreased significantly with increasing uranium content. BET measurements revealed low specific surface areas after sintering (0.19-0.36 m2·g-1), consistent with extensive densification. SEM-EDS observations confirmed the formation of consolidated polycrystalline microstructures, homogeneous uranium and zirconium distributions, and the progressive development of zirconia-rich regions in compositions with higher zirconium contents. Overall, the results demonstrate that aqueous co-precipitation is an effective route for producing homogeneous and well-characterized UO2-ZrO2 model materials. Compared with previously investigated dry powder processing, the wet route provides improved homogeneity and densification, making these materials suitable surrogates for the study of fuel-cladding interaction products and their long-term behaviour under nuclear waste management conditions.