Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Fuels I
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Ramprashad Prabhakaran, Pacific Northwest National Laboratory; Dennis Keiser, Idaho National Laboratory; Raul Rebak, GE Global Research; Clarissa Yablinsky, Los Alamos National Laboratory
Monday 8:30 AM
February 27, 2017
Location: Marriott Marquis Hotel
Session Chair: Ramprashad Prabhakaran, Pacific Northwest National Laboratory; Dennis Keiser, Idaho National Laboratory
Results of Microstructural Characterization Focused on the U-10Mo/Zr Diffusion Barrier Interface in Irradiated Monolithic Fuel Plates: Dennis Keiser1; Jan-Fong Jue1; Brandon Miller1; Jian Gan1; Adam Robinson1; James Madden1; Assel Aitkaliyeva1; 1Idaho National Laboratory
The Materials Management and Minimization Reactor Conversion (MMMRC) Program is focused on developing low enriched uranium fuels for application in research and test reactors that currently employ high enriched uranium. One fuel type being developed is a monolithic fuel comprised of U-10Mo foil encased in Al-6061 cladding with a Zr diffusion barrier in between. To understand the irradiation performance of this fuel at the interface with the Zr diffusion barrier, samples from different irradiated fuel plates have been characterized using scanning electron microscopy (SEM) and transmission electron microscopy (TEM), and some of the characterization samples were produced using a focused ion beam (FIB). This presentation will discuss the microstructures observed for different irradiated fuel plate samples where focus will be given on the microstructures observed at different regions of the U-10Mo/Zr interface. Features like observed fission product phases, porosity, and grain structures will be discussed.
Nanoscale Structural and Compositional Analysis of U-10Mo Fuels: Arun Devaraj1; Vineet Joshi1; Libor Kovarik1; Saumyadeep Jana1; Bruce Arey1; Curt Lavender1; 1Pacific Northwest National Laboratory
Low enriched uranium 10-weight percent molybdenum alloy (LEU-Mo) in a thin sheet or foil form is considered as a potential replacement for the high enriched uranium fuel currently used in the fleet of United States High Performance Research Reactors. As-cast and homogenized U-10Mo alloy fuel exhibit metastable γ-UMo phase along with impurity phases like UC and MoSi2. Hence to understand the composition and structure of both the major phases and minor impurity phases in U10Mo alloys as a function of different processing steps, we conducted detailed transmission electron microscopy (TEM) and atom probe tomography (APT) studies. These detailed microstructural characterization efforts revealed both the structure and composition of all different phases at near atomic spatial resolution, providing a comprehensive understanding of the microstructure evolution of U-10Mo alloys. Also this work aided in understanding various phase transformation pathways that the alloy would take on various processing stages during fuel fabrication.
Recrystallization Texture in U10Mo Alloy: Karun Kalia1; David Field1; Vineet Joshi2; 1Washington State University; 2Pacific Northwest National Laboratory
Recrystallization texture of U10Mo alloy is studied. Uranium with 10% Molybdenum alloy is a nuclear fuel in which recrystallization texture plays a key role in its efficiency. Grain structure, grain size and texture effects the swelling kinetics of the grains in U10Mo alloy when consumed as a fuel. Hot rolled samples with 15% reduction per pass at 650⁰C with intermediate annealing at 700⁰C are prepared to analyze the texture and recrystallization behavior after deformation. Two samples hot rolled in four passes are prepared at different initial homogenization temperatures to check the change in resulting texture. Additional samples are prepared to study the recrystallization behavior by annealing the samples at 700⁰C in an inert atmosphere. EBSD (electron backscatter diffraction) images are collected and later analysis is done with help to quantify the microstructures.
Electron Backscatter Diffraction Analysis of Irradiated U-Mo Plate Fuel for the US High Performance Research Reactor Development Program: Bjorn Westman1; Brandon Miller2; Julie Tucker1; 1Oregon State University; 2Idaho National Laboratory
The US High Performance Research Reactor (HPRR) Fuel Development Program has been tasked with the development of new Low-Enriched Uranium (LEU) nuclear fuels to replace existing High-Enriched Uranium (HEU) fuels currently in use throughout the world. An important part of the program is an effort to understand the microstructure of irradiated fuel, including fuel grain size and orientation, phase distribution, and defect production such as fission products and fission gas pores. Variations in these parameters are believed to strongly affect the stability of the fuel. In support of this goal, Oregon State University (OSU) is helping increase Idaho National Laboratory’s (INL) Electron Backscatter Diffraction (EBSD) sample preparation and analysis techniques. This presentation will focus on preliminary results of EBSD analysis on previously irradiated fuel samples from the Reduced Enrichment for Research and Test Reactors (RERTR) program. Results will include patterns for both monolithic and dispersion U-Mo plate fuels and analysis.
Eutectoid Transformation Kinetics of As-Cast U - 8 wt% Mo Established by In Situ Neutron Diffraction: Matthew Steiner1; Christopher Calhoun1; Robert Klein1; Ke An2; Elena Garlea3; Sean Agnew1; 1University of Virginia; 2Oak Ridge National Lab; 3Y12 National Security Complex
The α-phase transformation kinetics of as-cast U - 8 wt% Mo below the eutectoid temperature have been established by in situ neutron diffraction. α-phase weight fraction data acquired utilizing Rietveld refinement at five different isothermal hold temperatures can be modeled accurately utilizing a simple Johnson-Mehl-Avrami-Kolmogorov impingement-based theory, and the results are validated by a corresponding evolution in the γ-phase lattice parameter during transformation that follows Vegard’s law. Neutron diffraction data is used to produce a detailed Time-Temperature-Transformation diagram that improves upon inconsistencies in the current literature, exhibiting a minimum transformation start time of 40 minutes at temperatures between 500°C and 510°C. The transformation kinetics of U - 8 wt% Mo can vary significantly from as-cast conditions after extensive heat treatments, due to homogenization of the typical dendritic microstructure which possesses non-negligible solute segregation.
10:10 AM Break
10:30 AM Cancelled
Assessment of the Suppression Methods for Porosity Growth in U-Mo/Al Dispersion Fuel: Yeon Soo Kim1; Gwan Yoon Jeong2; Dong-Seong Sohn2; 1Argonne National Laboratory; 2UNIST
U-Mo/Al dispersion fuel has shown excessive fuel meat swelling or breakaway swelling at higher fission rate and burnup conditions. This breakaway swelling is believed to be a result from the combination of accelerated fuel swelling and interconnection of large pores formed outside of the U-Mo particles. Kim et al  developed a model for this pore growth using a mechanical equilibrium between gas pressure in the pore, interfacial energies, and the external stress at the pore surface. Performance improvement methods in view of the large pore growth were parametrically studied using the pore growth model by Kim . The methods studied include coating on the U-Mo particles to reduce interaction layer growth of U-Mo and Al, U-Mo particle size, and fission gas bubble growth (recrystallization) in the U-Mo particles.  Yeon Soo Kim et al., J. Nucl. Mater. 478 (2016) 275.
Microstructural Development of UMo-Al Dispersion Fuels after Thermal Annealing: Laura Jamison1; Bei Ye1; Sumit Bhattacharya2; Abdellatif Yacout1; 1Argonne National Laboratory; 2Argonne National Laboratory and Northwestern University
As part of the ongoing effort to convert high power research and test reactors from high-enriched to low-enriched fuels, uranium-molybdenum alloy fuel dispersed in aluminum matrix (UMo-Al dispersion) has become the leading candidate fuel. One of the primary barriers to qualification of the fuel is the formation of an interaction layer (IL) between UMo and Al, which can lead to breakaway swelling due to fission gas accumulation in the IL. Additionally, in order for swelling behavior to remain homogenous and predictable, it is important that the UMo be in the gamma-phase. In order to separate the effects of temperature and irradiation on the IL formation and UMo phase composition, an annealing study was conducted. UMo dispersion fuel samples, with and without coating, were annealed for 90 hours at 350°C. The microstructure of these samples will be compared to those previously irradiated with heavy ions with a similar thermal history.
Effect of Grain Morphology on Gas Bubble Swelling in UMo Fuels – A 3D Microstructure Dependent Booth Model: Shenyang Hu1; Curt Lavender1; Vineet Joshi1; 1Pacific Northwest National Laboratory
Gas bubble swelling is one of important design parameters of metallic nuclear fuels. This talk will present a three dimensional microstructure dependent swelling model. The model is extended from the Booth model in order to investigate the effect of heterogeneous microstructures and thermodynamic and kinetics properties on gas bubble swelling kinetics. The recrystallization, which is often observed in irradiated fuels, is described with a concept of “phase transition” and integrated into the swelling model. As an application of the model, the effect of grain morphology, fission gas diffusivity, spatial dependent fission rate, and recrystallization on swelling kinetics are simulated in UMo fuels. The potential application of the developed model in investigating the effect of other heterogeneities such as second phases and spatial dependent thermodynamic properties including diffusivity of fission gas, sink and source strength of defects on swelling kinetics will be discussed.
An Integrated Simulation for Deformation and Irradiation-Induced Grain Growth in, U-10 wt%Mo: William Frazier1; Vineet Joshi1; Shenyang Hu1; 1Pacific Northwest National Laboratory
U-Mo alloys are attractive low-enriched uranium (LEU) alternatives to current high-enriched uranium (HEU) fissile materials because of their high density, excellent irradiation performance, and good thermal conductivity, specifically when in the γ-U phase. However, grain growth, recrystallization, and phase transformation kinetics in the system are not yet entirely understood. Processing variations often result in different grain morphologies, α-U precipitation at the grain boundaries from Mo solute depletion, and abnormal grain growth (AGG). In order to insure consistent microstructure and performance of the alloy, and to identify processing pathways that avoid the detrimental effects of these phenomena, deformation-induced recrystallization was simulated using the Potts Model of grain growth. To this end, our simulation work uses inputs of both deformation simulations and experimental observations of strain accumulation in U-Mo samples to predict grain growth behavior of the alloy during stages of hot rolling, and then to predict resultant swelling response during irradiation.