Materials in Nuclear Energy Systems (MiNES) 2021: Fuels and Actinide Materials- Metallic Fuels III
Program Organizers: Todd Allen, University of Michigan; Clarissa Yablinsky, Los Alamos National Laboratory; Anne Campbell, Oak Ridge National Laboratory

Wednesday 1:30 PM
November 10, 2021
Room: Grand Ballroom
Location: Omni William Penn Hotel

Session Chair: Assel Aitkaliyeva, University of Florida


1:30 PM  Invited
Constructing Multi-component Diffusion under Irradiation in U-Mo Alloys: Benjamin Beeler1; Gyuchul Park2; Maria Okuniewski2; Zhi-Gang Mei3; Shenyang Hu4; 1North Carolina State University; 2Purdue University; 3Argonne National Laboratory; 4Pacific Northwest 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 rate-theory 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.

2:10 PM  
Three-dimensional Characterization of Pore Evolution in High-burnup U-Mo: Alejandro Figueroa1; Daniel Murray2; Peter Kenesei3; Maria Okuniewski1; 1Purdue University; 2Idaho National Laboratory; 3Argonne National Laboratory
    Low-enrichment monolithic uranium-molybdenum-based fuels are of interest due to their potential to replace high-performance research and test reactors in conjunction with global nonproliferation efforts. Swelling in uranium-molybdenum in monolithic and dispersion fuel applications undergo a two-part swelling behavior, with an initial linear swelling rate followed by a subsequent increase in swelling due to irradiation-induced grain refinement. Understanding the porosity development as a function of fission density in this increased swelling rate regime is critical to understand the swelling behavior properly. This work investigates four separate locations on an edge-on miniplates that experienced a burnup between 5-10E21 fissions/cc. Analyzing the porosity in three dimensions utilizing micro-computed tomography creates a better understanding of porosity development as a fission density and rate function. This understanding is crucial for mechanistic modeling of fuel swelling in this increased swelling regime.

2:30 PM  
An Investigation of FCCI Using Diffusion Couple Test between UMTZ Alloys and Cladding: Weiqian Zhuo1; Huali Wu1; Michael Benson2; Jinsuo Zhang1; 1Virginia Tech; 2Idaho National Laboratory
    Two U-rich alloys containing additives Mo, Ti, Zr are proposed as fuel candidates, the alloys are U-1.5Mo-1.5Ti-7Zr (UMT7Z), and U-2.5Mo-2.5Ti-5Zr (UMT5Z) in wt.%. To investigate their fuel-cladding chemical interactions (FCCIs), the diffusion couple tests were performed between the fuel alloys and the cladding at 600℃. The as-cast fuel alloys and the pre-annealed alloys (annealed at 600C for 168h) were used for comparison because their microstructures and phases were different according to the scanning electron microscope (SEM) characterizations. The cladding materials were Fe and HT9. In those diffusion couples using as-cast alloys, Zr-rind was found at the fuel interfaces. The reaction product UFe2 was found at the HT9 side but not at the Fe side. The interface interaction of those diffusion couples using pre-annealed alloys is expected to be different from that of the as-cast alloys.

2:50 PM  
Transmission Electron Microscopy of the Uranium-22.5 Atom% Zirconium System Following Casting, Cold-working, and Annealing: Walter Williams1; Maria Okuniewski2; Fidelma Di Lemma3; Tiankai Yao3; 1Idaho National Laboratory/Purdue University; 2Purdue University; 3Idaho National Laboratory
    Transmission electron microscopy was performed on uranium-22.5atom% zirconium samples in as-cast, cold-worked, and post-annealed conditions. Samples were characterized through bright field, dark field, energy-dispersive X-ray spectroscopy (EDS), and selected-area electron diffraction (SAED). Phase morphology and bulk microstructure were measured throughout the material processing steps to quantify the evolution of the α-U and δ-UZr2¬ lamellar structure. The solubility of Zr in the α-U phase was measured, by semi-quantitative EDS, to be <15atom%. Additionally, EDS identified that the δ-UZr2 phase consists of ~66atom% Zr. Zirconium rich (U-80atom% Zr) inclusions were also identified through imaging and EDS. The crystal structure of each phase, including Zr inclusions, was identified with SAED. The SAED of the bulk microstructure indicates deformation-induced grain subdivision following cold-working, as well as recrystallization and growth during annealing. Zirconium inclusions were also observed to undergo a stress-induced phase transformation from a hexagonal to face-centered cubic crystal structure.

3:10 PM  
First-principles Study of the Interfaces between Gamma-U and Uranium Carbide: Zhi-Gang Mei1; Bei Yei1; Abdellatif Yacout1; Benjamin Beeler2; 1Argonne National Laboratory; 2North Carolina State University
    To understand the effect of uranium carbide formation on the mechanical properties of UMo alloy fuel, we investigated the interfaces formed between bcc gamma-U metal and uranium carbide using first-principles density-functional theory calculations. Two representative interfacial plane orientations, i.e., U(110)/UC(100) and U(100)/UC(100), were investigated for different potential terminations of UC. Calculations show that both lattice mismatch and interfacial bonding play crucial roles in determining the interfacial stability and adhesion strength of the gamma-U and UC interfaces. The effect of Mo alloy in gamma-U metal and non-stoichiometric of UC on the adhesion strength of gamma-U/UC interfaces were investigated. We studied the effect of the defect concentration and their locations with respect to the interface on the interfacial stability. Finally, the predicted interface models of gamma-U/UC interface were used to simulate the fracture toughness of gamma-UMo/UC interfaces.

3:30 PM Break