Ceramic Materials for Nuclear Energy Research and Applications: Alternate and Doped Fuels - Modeling and Experiments
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee, TMS: Energy Committee
Program Organizers: Walter Luscher, Pacific Northwest National Laboratory; Xian-Ming Bai, Virginia Polytechnic Institute and State University; Lingfeng He, North Carolina State University; Sudipta Biswas, Idaho National Laboratory; Simon Middleburgh, Bangor University

Thursday 2:00 PM
March 23, 2023
Room: 28B
Location: SDCC

Session Chair: David Bai, Virginia Tech; Miaomiao Jin, The Pennsylvania State University

2:00 PM  Invited
Atomistic Investigation of Radiation-induced Defects in ThO2: Miaomiao Jin¹; Hamdy Arkoub¹; Lingfeng He²; Chao Jiang²; Marat Khafizov³; David Hurley²; ¹Pennsylvania State University; ²Idaho National Laboratory; ³Ohio State University
    Radiation-induced defects and their evolution constitute the foundation of a high-fidelity description of radiation-assisted microstructure evolution. With molecular dynamics simulations of primary radiation damage, the production of defects is elaborated in ThO2. Notably, vacancy clusters approach being charge neutral, and interstitial clusters can embrace a high symmetry with a cuboctahedral (COT) structure. The relative stability of three different COT structures is compared based on molecular dynamics and density functional theory. To relate to experiments, the properties of experimentally observed dislocation loops (perfect and faulted) are inspected. Importantly, the atomistic details of the loop unfaulting process are captured in the fluorite oxide systems. These results facilitate a fundamental understanding of radiation-induced defects in fuel oxides and also a required input for mesoscale simulations of microstructure evolution.

2:30 PM  Invited
Hidden Defect Evolution Mechanism in ThO2 Revealed by Atomistic Modeling: Chao Jiang¹; Lingfeng He¹; Cody Dennett¹; Marat Khafizov²; James Mann³; David Hurley¹; ¹Idaho National Laboratory; ²The Ohio State University; ³United States Air Force
    Irradiation-generated point defect clusters can significantly impact the physical and mechanical properties of materials. However, direct experimental visualization of these small-scale defects via high-resolution scanning transmission electron microscopy remains a challenge. In this presentation, we employ thorium dioxide (ThO2) with the fluorite structure as a model system to demonstrate the synergetic usage of ab initio calculations and kinetic Monte Carlo simulations as a powerful tool for identifying small defect complexes in irradiated materials. In addition to providing quantitative insights into defect evolution in ThO2 under irradiation, this study reveals the unexpected role of bound anti-Schottky defect clusters in mediating defect transport. Remarkably, despite their short lifetime due to poor thermal stability against dissociation, the transient formation of bound anti-Schottky defects under irradiation and their subsequent migration represent the dominant mechanism behind the growth of large interstitial loops that have been experimentally observed in ThO2.

3:00 PM
Cluster Dynamics Modeling of Defects and Fission Gas in Gd Doped UO₂ under Irradiation: Vancho Kocevski¹; Michael Cooper¹; David Andersson¹; ¹Los Alamos National Laboratory
    Uranium dioxide (UO₂) is the primary nuclear fuel in LWRs, and its excess reactivity can be controlled by adding burnable absorbers, such as Gd₂O₃ forming UO₂/Gd₂O₃ system. These burnable absorbers have large neutron absorption cross-section lowering the high reactivity of reactors initial fuel load. However, there is limited understanding of how added Gd₂O₃ influences the properties of UO₂ during burnup. To understand the behavior of defects and fission gas in UO₂/Gd₂O₃ system during burnup, we use cluster dynamics modeling supported by density functional theory calculations. First, the formation energies of two Gd point defects and 13 bound defects were calculated. The temperature-dependent defect concentrations were evaluated using the defect formation energies and the entropies obtained using UO₂-Gd pair potential. Subsequently, the defect energies and entropies were introduced in the cluster dynamics code CENTIPEDE for modeling the influence of Gd on the defects and fission gas behavior in UO₂/Gd₂O₃ during burnup.

3:20 PM Break

3:40 PM  Invited
Susceptibility of Nuclear Fuel Ceramics to Oxidation and Hydridization during Off Normal Events: Elizabeth Sooby¹; Adrian Gonzales¹; Geronimo Robles¹; Joshua White²; ¹University of Texas at San Antonio; ²Los Alamos National Laboratory
    Ceramic and functional ceramic nuclear fuel compounds, namely uranium mononitride (UN), uranium silicide (U₃Si₂), uranium monocarbide, and uranium diboride (UB₂), have been recently assessed as candidate fuels for several reactor designs. For water-cooled reactors, susceptibility to corrosion during a cladding breach is of peak concern, and to that effect, U₃Si₂ has shown rapid pulverization in both flowing steam and pressurized water at temperatures as low as 350°C. However, UN displays a more graduate oxidation kinetic with onsets of reaction exceeding 600°C for high-purity, high-density samples exposed to steam. Furthermore, hydrogen exposure and potential hydrogen absorption can facilitate fuel degradation. Presented will be the latest experimental work assessing the thermochemical behavior and observed mechanisms for ceramic fuel failure of high uranium density compounds exposed to both oxidizing and reducing atmospheres.

4:10 PM
Development of UC/UO₂ Composite Fuels for Light Water Reactors: Scarlett Widgeon Paisner¹; Joshua White¹; Ian Porter²; Russell Fawcett²; ¹Los Alamos National Laboratory; ²Global Nuclear Fuels
    Development of accident tolerant fuels are important to improve the safety and economics of light water reactors (LWRs). Currently, UO₂ is used in commercial reactors; however, the low thermal conductivity causes high fuel centerline temperatures under normal operation. Increasing the thermal conductivity can lead to improvements in steady-state and transient performance while reducing the stored energy of the reactor core, thus improving safety under operation. UO₂ composite fuels can also increase the overall uranium density, therefore increasing the ²³⁵U loadings in a bundle to provide additional energy output. The work presented here will detail the development of UC/UO₂ composite samples and includes investigation of the thermophysical properties, chemical stability at elevated temperature, oxidation performance, and microstructure of sintered pellets. The data shows a 25% improvement of the thermal conductivity with UC additions as little as 10 wt%, which can enable an easier licensing path relative to alternative new fuel concepts.

4:30 PM
Atomistic and Mesoscale Modeling of Fission Gas and Fission Products Diffusivity in TRISO Fuel Kernels: Xiang-Yang Liu¹; Christopher Matthews¹; Wen Jiang²; Michael Cooper¹; Jason Hales²; David Andersson¹; ¹Los Alamos National Laboratory; ²Idaho National Laboratory
    TRISO fuel particles are candidates for use in next generation reactors including gas reactors, fluoride salt- cooled high temperature reactors, and micro-reactors. A fundamental understanding of the fission gas and fission products diffusivity in the UCO fuel kernels in irradiation conditions is important for accurate fuel performance prediction of the TRISO fuels. Typically, UCO fuel kernels used in TRISO fuels are a mixture of uranium carbides and uranium dioxides. In this study, based on the atomistic modeling of the fuel kernel properties, cluster dynamics simulations have been carried out to predict the diffusivity of fission gas xenon and fission products such as Ag in UCO fuel kernels. And the application of the diffusivity model in fuel performance simulations is also demonstrated. Finally, the complexity in the modeling of TRISO fuel kernels is further discussed, including fuel chemistry, different forms of uranium carbides, and higher burn-up issues.

4:50 PM
Modelling the Melting Temperature of CrUO₄ to Assess its Behaviour during the Sintering of Cr-doped UO₂: Sarah Vallely¹; Conor Galvin²; Michael Cooper²; Simon Middleburgh¹; ¹Bangor University; ²Los Alamos National Laboratory
    Cr₂O₃ is added to UO₂ during fabrication to enhance grain growth. The large grains are anticipated to improve the mechanical properties of the fuel and increase fission gas retention. Previous atomic scale simulations and experimental work predict that when Cr₂O₃ is added to hyper-stoichiometric UO₂, the excess lattice oxygen atoms are transferred to form a secondary CrUO₄ phase. Limited data published on CrUO₄ means that its behaviour during the sintering and irradiation of Cr-doped UO₂ is largely unknown. Despite literature reporting the decomposition temperature of CrUO₄, the solidus and liquidus of the compound is yet to be investigated. This work uses classical molecular dynamics to predict the melting temperature of CrUO₄ to assess the possibility of such a phase contributing to liquid phase sintering in Cr-doped UO₂, promoting grain growth. The formation energy of CrUO₄ is also assessed using DFT to predict the UO_2+x non-stoichiometry at which it forms.