Advanced Characterization and Modeling of Nuclear Fuels: Microstructure, Thermo-physical Properties: On-Demand Oral Presentations
Sponsored by: TMS Structural Materials Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Energy Committee, TMS: Nanomechanical Materials Behavior Committee, TMS: Nuclear Materials Committee
Program Organizers: David Frazer, Idaho National Laboratory; Fabiola Cappia, Idaho National Laboratory; Tsvetoslav Pavlov, Idaho National Laboratory; Peter Hosemann

Monday 8:00 AM
March 14, 2022
Room: Nuclear Materials
Location: On-Demand Room


High-resolution Thermal Conductivity and Thermal Boundary Resistance Mapping in TRISO: Dong Liu1; 1University of Bristol
    An accurate measurement of the local thermal conductivity (TC) and thermal boundary resistance (TBR) is critical for understanding the heat transfer in nuclear fuels. In particularly for TRISO (tristructural isotropic coated particles), due to its multi-layer coating system, these are very challenging to measure due to the large range of thermal conductivity (TC) values for SiC, PyC and buffer layer from a few W/mK to about 20 W/mK, and the sizable TBRs between each of the layers. In this work, a thermal reflectance-based method has been developed to measure/map the thermal conductivity of each individual layers in TRISO coatings on polished cross-sections as well as in un-polished whole particles. The interfaces between coatings, especially when voids/pores are present (e.g., between the IPyC and Buffer layer), the TBR is increased. A numerical model has been built to simulate the heat flow in TRISO based on the experimental measurements.

Structural Analysis of the IPyC/SiC Interface of AGR-2 Irradiated and Safety Tested TRISO Fuel: Rachel Seibert1; Tyler Gerczak1; Jesse Werden1; Darren Skitt1; John Hunn1; 1Oak Ridge National Laboratory
    Tristructural-isotropic (TRISO) coated particle fuel from the second Advanced Gas Reactor Fuel Qualification and Development Program irradiation experiment (AGR-2) was studied using focused-ion beam/scanning electron microscopy serial sectioning to analyze the inner pyrocarbon (IPyC)/silicon carbide (SiC) interface. Shared areas, interfaces, and pore/fission product shapes, sizes, and locations were quantitatively analyzed through both automated and manual batch processing of image stacks collected on 12 m x 12 m x 12 m regions of interest. This analysis focused on quantifying the nature of the IPyC/SiC interface to understand its role on structural changes and fission product interactions from the PyC and into the SiC layer under both irradiation and after upper margin safety testing (1800C, 300 h).

A Combined Molecular and Cluster Dynamics Approach to Determine Radiation Enhanced Diffusion in UMo Alloys: Benjamin Beeler1; Park Gyuchul2; Maria Okuniewski2; Shenyang Hu3; Zhi-Gang Mei4; 1North Carolina State University; 2Purdue University; 3Pacific Northwest National Laboratory; 4Argonne National Laboratory
    Under the United States High-Performance Research Reactor (HPRR) program, a number of research reactors are planned to undergo a conversion to U-Mo monolithic fuel. The accurate prediction of fuel evolution under irradiation requires implementation of correct thermodynamic and kinetic properties into fuel performance modeling. One such property where there exists incomplete data is the diffusion of relevant species under irradiation. Fuel performance swelling predictions rely on an accurate representation of diffusion in order to determine the rate of fission gas swelling and local microstructural evolution. In this work, molecular dynamics simulations are combined with cluster dynamics calculations to determine the radiation-enhanced diffusion of U, Mo, and Xe as a function of temperature and fission rate. In combination with previous studies on intrinsic diffusion and radiation-driven diffusion in U-Mo alloys, this study completes the multi-component diffusional picture for the U-Mo system.

Comparison of Observations from the Microstructure of Two High Burnup Fuel Samples Operated at Different Linear Heat Generation Rates: Jason Harp1; Tyler Gerczak1; 1Oak Ridge National Laboratory
    The formation of high burnup structure and high burnup fuel fragmentation (HBFF) behavior of light water reactor irradiated UO2 fuel is governed in part by the final linear heat generation rate of the fuel. While this behavior is empirically known, the phenomenological mechanism that drives this difference in behavior is still being actively researched. Current understanding of HBFF behavior during LOCA suggests that there are microstructural features that develop at different power conditions that lead to HBFF susceptibility. To better understand these features, microstructural characterization was performed on samples of light water reactor fuel irradiated at different linear heat generation rates above and below the HBFF pulverization threshold established in literature. The data collected during theses exams will be discussed and connected to the current understanding of HBFF available in literature.

Thermal Conductivity Measurement of Microstructures in Irradiated Nuclear Fuels: Yinbin Miao1; Lakshmi Amulya Nimmagadda2; Jingyi Shi3; Kun Mo1; Bei Ye1; Shipeng Shu1; Peter Mouche1; Winfried Petry3; Sanjiv Sinha2; Abdellatif Yacout1; 1Argonne National Laboratory; 2University of Illinois Urbana-Champaign; 3Technical University of Munich
    Thermal conductivity is an important thermophysical property of nuclear fuels. Advanced nuclear fuels intrinsically consist of lower-length-scale structures with enhanced properties. Microstructural modifications caused by the harsh environment in reactors (e.g., high-energy radiation and elevated temperatures) lead to changes in the thermophysical properties. Accurate measurement of thermal conductivity of nuclear fuels relevant to lower-length-scale features can be an element in improving our understanding of their in-pile behavior. The suspended bridge method, originally developed for nanomaterials, was adopted to measure the thermal conductivity of fuel samples of reduced size. Measurements were performed on prepared miniature specimens of common nuclear fuels materials. The results were compared against the literature values of bulk specimens to validate the method. In addition, a series of radiation-induced microstructures obtained from ion-irradiated nuclear fuel samples were also measured for thermal conductivity. The advantages of utilizing the suspended bridge method for nuclear fuel and materials research are discussed.

High-throughput Viscosity Measurements of Molten Salts for Molten Salt Reactors: Alexander Levy1; Haoxan Yan1; Federico Coppo1; Uday Pal1; Karl Ludwig1; Adam Powell2; 1Boston University; 2Worchester Polytechnic Institute
    There is considerable variability in the reported physical properties of molten salts used in nuclear reactors. These properties are important for the design of the reactors and in understanding the structure of the salts. Existing methods to measure these properties are expensive and time consuming. A high throughput method to measure one of these properties, viscosity, as a function of temperature will be presented. The method utilizes a glove box compatible high temperature furnace environment. The device is essentially a falling ball viscometer. The sensors are a pair of titanium electromagnetic coils acting as metal detectors to sense a metallic ball falling at terminal velocity through the molten salt. Viscosity is calculated with the measured terminal velocity using Stokes Law and accounting for wall effects. Two candidate salts, FLiNaK and NaF-ZrF4, were used for the viscosity measurements. The system will later be adapted for high-throughput measurement of other properties.

Mesoscale Model of Gas Bubble Evolution and Creep in Monolithic UMo Fuels: Shenyang Hu1; Benjamin Beeler2; 1Pacific Northwest National Laboratory; 2Idaho National Laboratory
    In monolithic UMo fuels large volumetric swelling combined with the cladding constraint generates a stress field. In turn, the stress affects the evolution of radiation defects and gas bubbles. In this work, we developed a mesoscale model of the dynamic interaction among the generation, diffusion and reactions of radiation defects, nucleation and growth of gas bubbles, and elastic-plastic deformation in polycrystalline UMo. A microstructure-dependent cluster dynamics model describing the concentration evolution of radiation defects; a phase-field model describing the nucleation and growth of nonequilibrium gas bubbles, and the crystal plasticity theory describing elastic-plastic deformation are linked by stress-enhanced diffusion of radiation defects and the absorption and/or emission of radiation defects at the interface of gas bubbles. The predicted effect of the cladding constraint (overall pressure) on the evolution of radiation defects and gas bubbles, total swelling, and creep will be reported.

Synergistic Electron/Thermal Microscope for High-throughput Screening of Emerging Nuclear Materials: Yuzhou Wang1; Cody Dennett1; Zilong Hua1; Robert Schley1; Daniel Murray1; Geoffrey Beausoleil II1; David Hurley1; 1Idaho National Laboratory
    The ever-increasing demands of next generation nuclear reactors requires the development of new materials with better performance. To expedite the discovery and qualification of emerging nuclear materials, advanced techniques that are able to characterize microstructure and thermophysical properties in high-throughput manner is needed. Under the Nuclear Materials Discovery and Qualification initiative (NMDQi), we construct a synergistic electron/thermal microscope that can characterize local microstructure, chemistry, and thermal properties correlatively. The thermal microscope is based on the square-pulse thermoreflectance technique, where the rising surface temperature after square-wave laser excitation is compared with a continuum heat transport model to extract interested thermophysical properties, including thermal diffusivity and thermal conductivity. The thermal microscope is fiberized and miniaturized to fit inside the vacuum chamber of a scanning electron microscope and the results of both techniques are linked through spatial correlation. We demonstrate such instrument on advanced fuels and alloys to show its great applicability.

Investigation of Damage Structure Evolution on Proton Irradiated Zr-alloys of Various Compositions Using Synchrotron X-ray Diffraction and TEM: mer Ko1; Tamas Ungr1; Rebecca Jones2; Hattie Xu1; Robert Harrison1; Michael Preuss1; Philipp Frankel1; 1The University of Manchester; 2Rolls-Royce
    Irradiation induces displacement of atoms leading to vacancies, interstitials and eventually to dislocation loops in core materials, which affect the mechanical performance of these materials. Proton irradiation is increasingly used as a surrogate to costly neutron irradiation for studying irradiation-induced damage in materials and its effect on microstructure. Several different proton irradiated Zr binary, commercial Zircaloy-2 and Low-Sn ZIRLO were analysed using Synchrotron XRD in combination with sophisticated line profile analysis and TEM imaging methods. Significantly higher dislocation densities were observed in ZrNb binary alloys compared to non-Nb containing alloys. Additionally, pure Zr or Zr only containing interstitial elements were found to have a high fraction of <a> type dislocation loops but overall lower dislocation density compared to the other alloys.

Mesoscale Modeling of Effective Thermal Conductivity in U-Zr Fuels with Heterogeneous Phases: Weiming Chen1; Xian-Ming Bai1; 1Virginia Polytechnic Institute and State University
    Uranium-zirconium (U-Zr) alloys are promising candidate fuels for next-generation fast reactors. Depending on the temperature, U-Zr fuels can contain multiple phases, which affect many fuel properties including thermal conductivity. Many existing U-Zr thermal conductivity models are empirical and do not contain the microstructure- or phase-dependent information. In this work, we used mesoscale modeling in the MOOSE framework to model the effective thermal conductivities of U-Zr alloys containing alpha-U + delta-UZr2 lamellar structures. We found that the interface thermal resistance (Kapitza resistance) should be properly taken into account in order to predict the effective thermal conductivity of such a heterogeneous microstructure accurately. Based on the mesoscale modeling results, we developed a temperature- and geometry-dependent Kapitza resistance model for the lamella structure. By including this model in the mesoscale modeling, the effective thermal conductivities of a number of lamellar structures can be reasonably predicted at a wide range of temperatures.

Correlative APT-TEM Investigation of Defects’ Influence on Thermal Diffusivity in ThO2 Nuclear Fuel: Amrita Sen1; Mukesh Bachhav2; Cody Dennett2; James Mann3; Janelle Wharry1; 1Purdue University; 2Idaho National Laboratory; 3Air Force Research Laboratory
    The objective of this study is to investigate the influence of irradiation and chemical doping on the thermal properties of model ceramic oxide nuclear fuel ThO2. Phonon-mediated thermal transport is key to the performance, efficiency, and safety of nuclear fuels. But the relationship between irradiation, microstructure, and thermal transport, in oxide fuels is not yet well understood. In this study, we investigate these mechanisms through correlative atom probe tomography (APT) and transmission electron microscopy (TEM). Preliminary APT studies of proton irradiated ThO2 show that small-scale off-stoichiometry, doping, and extended irradiation-induced defects alter the thermal tails of APT spectra, from which the influence on thermal diffusivity can be ascertained. Coupling APT with correlative TEM will verify the influence of specific microstructural features. Complementary Raman spectroscopy and X-ray diffraction (XRD) results will also be presented. Implications of findings on ceramic nuclear fuel efficiency and performance will be discussed.

Evolution of the Internal Layer Structure in Irradiated TRISO Fuel: Tyler Gerczak1; John Hunn1; Grant Helmreich1; Rachel Seibert1; John Stempien2; Darren Skitt1; Brian Eckhart1; Andrew Kercher1; 1Oak Ridge National Laboratory; 2Idaho National Laboratory
    Tristructural-isotropic (TRISO) coated particle fuel is a robust fuel form originally developed for high temperature gas-cooled reactors (HTGRs). Each layer in the TRISO particle architecture is designed to provide specific functionality and irradiation response. The evolution of the TRISO particle layers after irradiation were observed through multiscale post-irradiation examination (PIE) (e.g., coupled microcomputed X-ray tomography with electron microscopy). Radiation-induced changes in TRISO fuel from the second Advanced Gas reactor Fuel Qualification and Development program irradiation experiment (AGR-2) will be presented. The presentation will focus on the impact of fast fluence, irradiation temperature, and post-irradiation high-temperature exposures on the change in buffer structure (e.g., densification and fracture) and its subsequent influence on fission product transport and local SiC layer corrosion.

Experimentally Validated Model for Investigating High-burnup Structure Formation in U-Mo Fuels: Sudipta Biswas1; Charlyne Smith2; Brandon Miller1; Dennis Keiser1; Assel Aitkaliyeva2; 1Idaho National Laboratory; 2University of Florida
    High-burnup regions in irradiated fuels exhibit a fine-grained microstructure with large pores, known as a “high-burnup structure.” In this study, we utilize a grand-potential-based phase-field modeling approach to investigate the mechanisms behind such structure formation in U-Mo fuels. It is hypothesized that irradiation-induced damage of existing grains leads to the reorganization of dislocations into low-angle grain boundaries (LAGBs), creating subgrains within the existing damaged grains. Over time, these subgrains lead to the formation of new grains with high-angle grain boundaries (HAGBs). The model employs a fission-density-based discrete nucleation algorithm to simulate the recrystallization process. It is observed that the first LAGB-to-HAGB transition occurs at a fission density of 2.31021 fissions/cm3. With increasing fission density, the fraction of damaged residual grains with LAGB subgrains decreases. Additionally, the presence of fission gas pores accelerates the HAGB grain nucleation. The model prediction is validated against the experimental characterization, using electron microscopy techniques.