Chinthaka M. Silva

Synthesis of phase-pure U2N3 microspheres and its decomposition into UN

Uranium mononitride (UN) is important as a nuclear fuel. Fabrication of UN in its microspherical form also has its own merits since the advent of the concept of accident-tolerant fuel, where UN is being considered as a potential fuel in the form of TRISO particles. However, not many processes have been well established to synthesize kernels of UN. Therefore, a process for synthesis of microspherical UN with a minimum amount of carbon is discussed herein. First, a series of single-phased microspheres of uranium sesquinitride (U2N3) were synthesized by nitridation of UO2+C microspheres at a few different temperatures. Resulting microspheres were of low-density U2N3 and decomposed into low-density UN. The variation of density of the synthesized sesquinitrides as a function of its chemical composition indicated the presence of extra (interstitial) nitrogen atoms corresponding to its hyperstoichiometry, which is normally indicated as α-U2N3. Average grain sizes of both U2N3 and UN varied in a range of 1-2.5 μm. These also had a considerably large amount of pore spacing, indicating the potential sinterability of UN toward its use as a nuclear fuel.

Application of X-ray microcomputed tomography in the characterization of irradiated nuclear fuel and material specimens

X‐ray microcomputed tomography (μCT) was applied in characterizing the internal structures of a number of irradiated materials, including carbon‐carbon fibre composites, nuclear‐grade graphite and tristructural isotropic‐coated fuel particles. Local cracks in carbon‐carbon fibre composites associated with their synthesis process were observed with μCT without any destructive sample preparation. Pore analysis of graphite samples was performed quantitatively, and qualitative analysis of pore distribution was accomplished. It was also shown that high‐resolution μCT can be used to probe internal layer defects of tristructural isotropic‐coated fuel particles to elucidate the resulting high release of radioisotopes. Layer defects of sizes ranging from 1 to 5 μm and up could be isolated by tomography. As an added advantage, μCT could also be used to identify regions with high densities of radioisotopes to determine the proper plane and orientation of particle mounting for further analytical characterization, such as materialographic sectioning followed by optical and electron microscopy. In fully ceramic matrix fuel forms, despite the highly absorbing matrix, characterization of tristructural isotropic‐coated particles embedded in a silicon carbide matrix was accomplished using μCT and related advanced image analysis techniques.

Carbothermic Synthesis of ~820- m UN Kernels. Investigation of Process Variables

This report details the continued investigation of process variables involved in converting sol-gel-derived, urainia-carbon microspheres to ~820-μm-dia. UN fuel kernels in flow-through, vertical refractory-metal crucibles at temperatures up to 2123 K. Experiments included calcining of air-dried UO3-H2O-C microspheres in Ar and H2-containing gases, conversion of the resulting UO2-C kernels to dense UO2:2UC in the same gases and vacuum, and its conversion in N2 to in UC1-xNx. The thermodynamics of the relevant reactions were applied extensively to interpret and control the process variables. Producing the precursor UO2:2UC kernel of ~96% theoretical density was required, but its subsequent conversion to UC1-xNx at 2123 K was not accompanied by sintering and resulted in ~83-86% of theoretical density. Decreasing the UC1-xNx kernel carbide component via HCN evolution was shown to be quantitatively consistent with present and past experiments and the only useful application of H2 in the entire process.

Radiation-Induced Changes in Quartz, A Mineral Analog of Nuclear Power Plant Concrete Aggregates

Quartz single-crystal samples consisting of α-quartz crystal structure were neutron irradiated to fluences of 5 × 1018, 4 × 1019, and 2 × 1020 n/cm2 (E > 0.1 MeV) at two temperatures (52 and 95 °C). The changes in the α-quartz phase as a function of these two conditions (temperature and fluence) were studied using X-ray powder diffraction (XRD), Raman spectroscopy, and transmission electron microscopy (TEM), and the results acquired using these complementary techniques are presented in a single place for the first time. XRD studies showed that the lattice parameters of α-quartz increased with increasing neutron flux. The lattice growth was larger for the samples that were neutron irradiated at 52 °C than at 95 °C. Moreover, an amorphous content was determined in the quartz samples neutron irradiated at 4 × 1019 n/cm2, with the greater amount being in the 52 °C irradiated sample. Complete amorphization of quartz was observed at a fluence of 2 × 1020 n/cm2 (E > 0.1 MeV) using XRD and confirmed by TEM characterization and Raman spectroscopic studies. The cause for α-quartz lattice expansion and sample amorphization was also explored using XRD and Raman spectroscopic studies.

M3FT-15OR0202212: SUBMIT SUMMARY REPORT ON THERMODYNAMIC EXPERIMENT AND MODELING

Modeling the behavior of nuclear fuel with a physics-based approach uses thermodynamics for key inputs such as chemical potentials and thermal properties for phase transformation, microstructure evolution, and continuum transport simulations. Many of the lanthanide (Ln) elements and Y are high-yield fission products. The U-Y-O and U-Ln-O ternaries are therefore key subsystems of multi-component high-burnup fuel. These elements dissolve in the dominant urania fluorite phase affecting many of its properties. This work reports on an effort to assess the thermodynamics of the U-Pr-O and U-Y-O systems using the CALPHAD (CALculation of PHase Diagrams) method. The models developed within this framework are capable of being combined and extended to include additional actinides and fission products allowing calculation of the phase equilibria, thermochemical and material properties of multicomponent fuel with burnup.

Multi-tier Analysis of SiC Breaches in Safety-Tested AGR-1 TRISO Fuel Particles

Tristructural isotropic (TRISO) coated particle fuel development is being supported by the US Department of Energy, Office of Nuclear Energy. The development plan includes a series of irradiations to qualify TRISO fuel. The first irradiation, AGR-1, included fuel fabricated at Oak Ridge National Laboratory (ORNL) and irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL). The TRISO fuel design consisted of a uranium carbide/uranium oxide kernel surrounded by concentric coating layers of carbonaceous buffer, inner pyrolitic carbon (IPyC), silicon carbide (SiC), and outer pyrolitic carbon (OPyC). Particles were then over-coated with carbonaceous matrix material and pressed into a cylindrical compact, with each compact containing greater than 4100 TRISO particles. A total of 72 compacts were included in AGR-1 and the irradiation was completed in November 2009 after ~620 effective full power days [1].