Engineering topics

Finkeldei, Sarah C.

Publications and source records attributed to Finkeldei, Sarah C..

3 records

DOE OSTI2024

Using in situ UO 2 bicrystal sintering to understand grain boundary dislocation nucleation kinetics and creep

Capillary evolution at bicrystal UO 2 grain boundaries is characterized using in situ transmission electron microscopy. The discontinuous nature of the densification process, both particle rotation and axial strain, along with the large activation stress for densification support a hypothesis that grain boundary strain in UO 2 follows nucleation rate limited kinetics at low to intermediate stresses, that is, less than ≈ 10 8 PA. Further, the temperature dependence of the average activation stress for sintering agrees well with analysis of bulk sintering data and creep data reported within the literature when analyzed in the context of a grain boundary dislocation nucleation rate limited kinetic model.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Investigations towards incorporation of Eu 3+ and Cm 3+ during ZrO 2 crystallization in aqueous solution

Nuclear energy provides a widely applied carbon-reduced energy source. Following operation, the spent nuclear fuel (SNF), containing a mixture of radiotoxic elements such as transuranics, needs to be safely disposed of. Safe storage of SNF in a deep geological repository (DGR) relies on multiple engineered and natural retention barriers to prevent environmental contamination. In this context, zirconia (ZrO 2 ) formed on the SNF rod cladding, could be employed as an engineered barrier for immobilization of radionuclides via structural incorporation. This study investigates the incorporation of Eu 3+ and Cm 3+ , representatives for trivalent transuranics, into zirconia by co-precipitation and crystallization in aqueous solution at 80 °C. Complementary structural and microstructural characterization has been carried out by powder X-ray diffraction (PXRD), spectrum imaging analysis based on energy-dispersive X-ray spectroscopy in scanning transmission electron microscopy mode (STEM-EDXS), and luminescence spectroscopy. The results reveal the association of the dopants with the zirconia particles and elucidate the presence of distinct bulk and superficially incorporated species. Hydrothermal aging for up to 460 days in alkaline media points to great stability of these incorporated species after initial crystallization, with no indication of phase segregation or release of Eu 3+ and Cm 3+ over time. These results suggest that zirconia would be a suitable technical retention barrier for mobilized trivalent actinides in a DGR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Detailed characterization of a PWR fuel rod at high burnup in support of LOCA testing

Experimental investigations of the fuel microstructure and volatile fission products along the pellet of a high burnup specimen (local burnup 76 GWd/tHM) have been conducted to support future transient testing. Detailed microscopy examinations have been carried out at different length scales. Transmission electron microscopy has highlighted a significant amount of damage across the entire radius with the formation of networks of dislocations. The optical and scanning electron microscopy determined the formation of three zones in the pellet with different characteristics. An intermediate region with no grain subdivision, lower porosity than the central zone porosity and high retained fission gas in nanometric bubbles and in the matrix was present between r/r0 ˜ 0.55 and 0.8. The high retention of gas in this region might suggest that the region will be prone to fine fragmentation, in addition to the HBS. An abrupt transition in the structure was observed at mid radius, with a third region developing from the mid radius to the pellet center. In this part of the pellet, metallic and grey phases with size between hundreds of nanometers and a few micrometers have formed at grain boundaries. The majority of fission gas has been released from the grain matrix and the original grains have polygonised, forming sub-grain domains separated by low-angle grain boundaries. No final explanation can be given for the polygonization occurring in the center, but on the basis of the irradiation history and the analysis of all the post irradiation examination (PIE) data, it is postulated that the polygonization within the original grains is an effect of dynamic recovery occurring at high temperature in the fuel center.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗