Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface: Metallic Fuels
Sponsored by: TMS Structural Materials Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Nuclear Materials Committee
Program Organizers: Yi Xie, Purdue University; Miaomiao Jin, Pennsylvania State University; Jason Harp, Oak Ridge National Laboratory; Fabiola Cappia, Idaho National Laboratory; Jennifer Watkins, Idaho National Laboratory; Michael Tonks, University of Florida
Tuesday 2:30 PM
March 21, 2023
Room: 26B
Location: SDCC
Session Chair: Joshua White, Los Alamos National Laboratory
2:30 PM Introductory Comments
2:35 PM Invited
The Evolution of the Microstructure of Low-enriched Uranium Fuels During Irradiation in the Advanced Test Reactor: Dennis Keiser1; Brandon Miller1; Jan-Fong Jue1; Adam Robinson1; Charlyne Smith1; 1Idaho National Laboratory
The Material Management and Minimization Program is developing low enriched uranium fuels for application in research and test reactors. Different types of fuels are being developed, including a U-Mo monolithic fuel, a U-Mo dispersion fuel, and a uranium-silicide disperison fuel. To successfully qualify these fuel types, it is important to determine how the fuel microstructures evolve during irradiation. Microstructural characterization has been performed on samples from the different fuel types irradiated under different conditions in the Advanced Test Reactor. Techniques like scanning electron microscopy, transmission electron microscopy, atom probe tomography, electron energy loss spectroscopy, and/or electron backscattered electron diffraction have helped uncover the microstructure of the different fuel materials after irradiation. This presentation will discuss how the characterization results for the different fuels can be used to improve understanding of phenomena like recrystallization, grain growth, fission gas bubble growth, solid fission product phase development, radiation stability, amorphization, and swelling behavior.
3:00 PM Invited
Characterization of Crystal Structure Evolution in U-2wt.%Zr Using Neutron Diffraction with Particular Focus on the Beta-Uranium Phase: Sven Vogel1; Michael Benson2; Jason M. Harp3; Yi Xie4; 1Los Alamos National Laboratory; 2Idaho National Laboratory; 3Oak Ridge National Laboratory; 4Purdue University
Studies of the U-Zr phase diagram are motivated by alloys such as U-10wt.%Zr being leading fuel candidates for sodium-cooled fast reactors due to high fissile density, high thermal conductivity, ease of fabrication, and good compatibility with coolants. In the Zr-lean part of the phase diagram, the occurrence of beta-uranium, with its large tetragonal unit cell containing 30 atoms, is reported. However, experimental characterizations of such alloys with diffraction methods are very rare. Here, we present results on thermal cycling of U-2Zr characterized by time-of-flight neutron powder diffraction, during which the beta-uranium phase was observed. The data is compared with results from U-0.5Ti, an analog to pure uranium with the addition of a small amount of titanium atoms preventing the otherwise inevitable grain growth, inhibiting accurate characterization with diffraction methods.
3:25 PM Invited
Lower Length Scale Fuel Performance Modeling of U-Mo Fuel: Benjamin Beeler1; Bei Ye2; Zhi-Gang Mei2; Yongfeng Zhang3; Shenyang Hu4; Maria Okuniewski5; Sourabh Kadambi6; Linu Malakkal6; 1North Carolina State University; 2Argonne National Laboratory; 3University of Wisconsin-Madison; 4Pacific Northwest National Laboratory; 5Purdue University; 6Idaho National Laboratory
A monolithic fuel design with a U-Mo alloy has been selected as the fuel for conversion of the United States High-Performance Research Reactors (HPRRs). The accurate prediction of fuel evolution under irradiation requires the implementation of correct thermodynamic properties into mesoscale and continuum fuel performance modeling codes. As the physics that governs microstructural evolution span various time and spatial scales, understanding the fuel behavior inevitably involves atomic to mesoscale resolution that can be difficult to determine experimentally. Microstructural-level modeling and simulations can be used to develop physics-based materials models that can provide physical understanding to inform fabrication process control, as well as a valuable feedback mechanism between post-irradiation examination results and fabrication parameters. This presentation outlines recent lower-length scale modeling efforts that have been applied to address critical microstructural questions and provide mechanistic inputs for existing fuel performance codes to improve the descriptive and predictive capability at the engineering scale.
3:50 PM Break
4:05 PM
Microstructure and Phase Evolutions of U-Zr System in Thermal Cycling Neutron Diffraction Experiments: Yi Xie1; Sven Vogel2; Michael Benson3; Jason Harp4; 1Purdue University; 2Los Alamos National Laboratory; 3Idaho National Laboratory; 4Oak Ridge National Laboratory
The U-10Zr (wt.%) fuel is the leading fuel candidate for sodium-cooled fast reactor due to high fissile density, high thermal conductivity, ease of fabrication, and good compatibility with coolants. The U-10Zr fuel is anticipated to be fabricated in rod geometry by extrusion technology. Understanding the microstructure and phase evolutions as a result of irradiation and thermal cycling at elevated temperatures is of paramount importance for the development and ultimately licensing of this fuel form. The compositions with higher Zr contents are also important as during reactor operation the Zr redistributes and results in a Zr-rich fuel center region as high as U-50Zr (wt.%). We carried out thermal cycling neutron diffraction experiments on the extruded U-10Zr, as-cast U-35Zr and U-50Zr, analyzed the evolutions of microstructure and phase from the time-of-flight neutron diffraction data, with a focus on lattice parameter, atom ordering, unit cell volume, texture, and weight fractions.
4:25 PM
The Fabrication, Advanced Characterization, Advanced Test Reactor Irradiation, Post Irradiation Examination, and Materials Informatics for Annular U-10Zr Metallic Fuel: Tiankai Yao1; Mukesh Bachhav1; Fei Xu1; Luca Capriotti1; Benson Michael1; Lingfeng He1; Jason Harp2; 1Idaho National Laboratory; 2Oak Ridge National Laboratory
To develop sodium cooled fast reactors metal fuel with capability of ultrahigh burnup of up to 30 – 40 at% (300 – 400 MWd/KgM), the Department of Energy Advanced Fuel Campaign has explored novel metallic fuel concepts. The fuel was fabricated, cladded with HT-9 at the Materials & Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory to a peak burnup of 3.3% FIMA. The irradiation lasted for ~120 days in the reactor. This study presents the advanced characterization of both archived fresh fuel and ATR irradiated fuel using state of the art materials characterization method. We also applied computer vision machine learning algorithms to add informatics to the irradiation effects for metallic fuels, including the pore statistics and fuel cladding chemical interaction thickness distribution. Our works showcased that INL is the ideal place to test and develop fuel concepts for advanced SFR reactors.
4:45 PM
Scaling laws in nanoindentation investigation of metallic uranium alloys: Tianyi Chen1; Gavin Vandenbroeder1; Tzu-Yi Chang1; David Frazer2; Yushu Dewen2; Stephanie Pitts2; Tiankai Yao2; 1Oregon State University; 2Idaho National Laboratory
Understanding the creep on nuclear fuel and structural materials is essential for safe reactor operation. The challenges with conventional mechanical testing are the long time-scales and high radiation levels from the large samples. Nanoindentation measurements hold promise in obtaining microstructure-specific material properties at significantly reduced times. However, work still needs to be done to allow information extraction from nanoindentation measurements for mechanistic model input. Combining modeling and nanoindentation measurements at different loading rates, the deformation and evolution between plastic and elastic deformation were examined to illustrate the governing spatial and temporal scaling laws under nanoindentation stress relaxation. This understanding gives insights into the deformation process and promotes the application of novel nanoindentation methodologies to accelerated materials testing.
5:05 PM
Creep Testing of 70% Theoretical Density U10Zr: Jake Fay1; Fidelma Di Lemma2; Luca Capriotti3; Jie Lian1; 1Rensselaer Polytechnic Institute; 2Idaho National Laboratory ; 3Idaho National Laboratory
Metallic uranium alloys such as U10Zr are a promising nuclear fuel type in advanced reactor designs with advantages of improved thermal conductivity and being able to operate to high burnups (>10 at%). These fuels however exhibit very different swelling behaviors compared to ceramic fuels which produces mechanical interactions between the fuel and cladding that are not fully understood. This project aims to develop a better understanding of these mechanical interactions by measuring the creep behavior of porous U10Zr. U10Zr samples will be sintered to 70% theoretical density, characteristic of the density of U10Zr fuels after initial swelling. Compressive creep testing will then be performed with and without constraint to mimic the conditions experienced at the top of the fuel column. The resulting creep parameters determined from this study will be valuable for better understanding the mechanical interactions between metallic fuel and cladding and for improving modeling capabilities for metallic fuels.
5:25 PM
Characterization of U-10Mo Fuel Exposed to Intermediate Temperature Irradiation Conditions at the High Flux Isotope Reactor: Peter Doyle1; Jason Harp1; Dylan Richardson1; Tash Ulrich1; Ian Greenquist1; Andrew Nelson1; Rachel Seibert1; Grant Helmreich1; Randy Fielding2; Caleb Massey1; 1Oak Ridge National Laboratory; 2Idaho National Laboratory
Metallic uranium alloys have been explored since the beginning of nuclear power due to their high uranium density and thermal conductivity. U-10Mo has been of specific interest because with 10 weight % Mo, the γ-U crystal structure is indefinitely stabilized in reactor service. Historically, radiation data on U-10Mo fuels has been collected in fast reactor or lower temperature research reactor conditions. This work will use the MiniFuel device at the High Flux Isotope Reactor at Oak Ridge National Laboratory to explore radiation effects in U-10Mo fuel samples exposed at intermediate temperatures not previously investigated. Test conditions are targeted at four temperatures between 250 and 500 ℃ and 3 target burnup conditions, up to 3.5%FIMA. After the reactor experiment, samples will be examined for fission gas release, swelling, and microstructural changes. Comparison of this data to pre-irradiation examinations will provide insight into the suitability of U-10Mo fuels for broader applications.
5:45 PM
Magnetism and Finite Temperature Effects in δ-UZr2: A Density Functional Theory Analysis: Shehab Shousha1; Benjamin Beeler1; 1North Carolina State University
Uranium-Zirconium (U-Zr) fuels have been historically utilized in liquid metal fast breeder reactors and are candidates for microreactors and advanced reactor designs. However, the phase transformations of the U-Zr alloys, in particular, the phase stability and the chemical ordering of the δ-UZr2 phase, are not well understood. In this work, 0 K density functional theory (DFT) and ab-initio molecular dynamics (AIMD) were used to investigate the stoichiometric UZr2 phase, with and without electron localization corrections. DFT typically overestimates cell volume and predicts a ferromagnetic δ-UZr2 phase, contradicting experiments. The Hubbard U term enhances the ability to calculate formation enthalpies but gives unrealistic magnetic properties. It was found that through the inclusion of temperature via AIMD, both the magnetic state and the volume of UZr2 are more accurately predicted. Through the utilization of AIMD, thermodynamic quantities of the δ-UZr2 phase are predicted for the first time with an accurate magnetic state.
6:05 PM
Molecular Dynamics Based Microstructural Evaluation of the Surviving Defects in α-U Induced by a Single Displacement Cascade: Khadija Mahbuba1; Benjamin Beeler1; Andrea Jokisaari1; 1North Carolina State University
A single crystal of anisotropic α-U experiences expansion in the [010] direction and contraction in the [100] direction under irradiation, which is in contradiction with the directionally anisotropic thermal expansion. Though the temperature-dependent irradiation growth of α-U was first observed in the early sixties, the mechanism is not still well understood. It is believed that the increased temperature from the displacement cascade leads to anisotropic expansion, which subsequently results in preferential nucleation of point defects in an orientation which leads to the observed irradiation growth response. To validate this hypothesis, an atomistic study of displacement cascades of α-U is conducted. Primary knock-on atoms (PKAs) with both variable kinetic energy (varying from 1keV to 50keV) and incident direction are deployed to perform displacement cascade simulations. The population and structure of the surviving defects are evaluated, and the temperature dependence of the induced irradiation growth is analyzed.