Joshua Taylor White

Continuing to refine our knowledge of the evolving mechanical properties of nuclear fuel over the entire fuel service cycle is necessary to understand the pellet-clad mechanical interaction that occurs in the fuel rods during the operation. A challenge with measuring the mechanical properties of irradiated fuels is their high levels of radioactivity that usually require the use of hot cells making testing time consuming and expensive. Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young ' s modulus, hardness, and creep stress exponents. Increasing the mixed oxide fuels mechanical properties database through a variety of testing techniques should enhance modelers ' abilities to predict failure mechanisms in the fuel/clad interface. A current challenge to testing mixed oxide fuels is the plutonium component in the fuel. Mixed fluorite type oxides with ceria (CeO 2 ) can be used as a surrogate for mixed oxide fuels. In this study, (U,Ce)O 2 solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (U x-1 ,Ce x )O 2 compounds ranging from x equals 0.1 to 0.3 in equal steps at temperatures up to 800 degrees C: their Young ' s modulus, hardness, and creep stress exponents were evaluated. The Young ' s modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800 degrees C give stress exponent values, n = 4.7 - 6.9, that suggests dislocation motion as the deformation mechanism.

•UO2 grain size was enlarged using Cr2O3 dopant up to 7800 ppm (pre-sintering).•∼48 % reduction in Cr content measured post-sintering.•Cr incorporation in the UO2 lattice confirmed via XRD.•Thermal conductivity decreases at/above 2500 ppm in contrast to undoped UO2. Chromium-doped UO2 fuel has received significant interest due to the ability for chromium to produce pellets with large average grain size (>30 μm), which has shown to increase fission gas retention during operation. Sintering of chromium-doped UO2 pellets was pursued with oxygen potential and sintering atmosphere controlled to tailor the final microstructure of the material. Chromium additions in this study ranged from 750 to 7800 ppm. Cr concentrations were studied pre and post sintering using Inductively Coupled Plasma Optical Emission Spectroscopy (ICP-OES). Effects of chromium content on lattice parameter and microstructure were examined with X-ray diffraction (XRD) and scanning electron microscopy (SEM). Contraction of the UO2 lattice parameter was observed, as well as enlargement of grain size with increasing chromium content up to 4900 ppm Cr2O3. SEM indicated Cr incorporation within the matrix and the formation of chromium oxide precipitates throughout the microstructure at high Cr concentrations. Evaluation of thermophysical properties of Cr-doped UO2 pellets were conducted up to 1200 °C to illustrate their evolution with increased dopant concentration and microstructural changes. The results show that grain size is maximized at 52 μm with Cr2O3 concentration equal to 4900 ppm; however, grain size decreases at higher Cr2O3 concentrations. No significant changes were observed in specific heat capacity, linear thermal expansion, and coefficient of thermal expansion compared to undoped UO2. The thermal conductivity also decreased through the incorporation of Cr2O3 dopants above 750 ppm and is shown to be ∼15 % lower than reported UO2 values.

The next generation of nuclear materials must withstand severe operating conditions including high temperatures and irradiation exposure. Oxide dispersion strengthened steels, especially 14YWT, have shown promise as a durable material under these conditions. Understanding the irradiation-enhanced creep of structural components is fundamental to evaluating their suitability for applications in reactor environments. Ion irradiations can be used to expedite irradiation testing, but they have a restricted depth of penetration, limiting the characterization of changes to the material’s properties. Small-scale mechanical testing combined with ion beam irradiations has the potential to evaluate the irradiation-enhanced creep of materials. In this study, in-situ transmission electron microscopy nanopillar creep studies on 14YWT were performed with simultaneous ion beam irradiation. The irradiation increases the measured strain rate by an order of magnitude. Variable temperature ex-situ nanoindentation creep studies were performed between room temperature and 1073 K on control samples of 14YWT; observations indicated that there was a change in the deformation mechanism between 873 K and 1073 K, which agrees well with macro-scale mechanical testing. These results validate continued research into applying these meso-scale testing techniques to nuclear materials in the future.