Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials: Fuels & Claddings I
Sponsored by: TMS Structural Materials Division, TMS: Mechanical Behavior of Materials Committee, TMS: Nuclear Materials Committee
Program Organizers: Dong Liu, University of Oxford; Peng Xu, Idaho National Laboratory; Simon Middleburgh, Bangor University; Christian Deck, General Atomics; Erofili Kardoulaki, Los Alamos National Laboratory; Robert Ritchie, University of California, Berkeley
Monday 8:30 AM
February 28, 2022
Room: 204A
Location: Anaheim Convention Center
Session Chair: Peng Xu, Idaho National Laboratory; Erofili Kardoulaki, Los Alamos National Laboratory
8:30 AM Invited
Development of Advanced Nuclear Fuels for Current and Next Generation Reactors: Joshua White1; Tarik Saleh1; Kenneth McClellan1; 1Los Alamos National Laboratory
The Advanced Fuels Campaign has focused on the development of next generation nuclear fuels using a pragmatic approach to improve the safety and economics of nuclear reactors. Fuel development has utilized a multifaceted approach to fabricate and characterize fuel options for current and next generation reactors. This includes efforts to assess the thermophysical properties for accurate fuel performance code modeling, fabrication of test articles for small scale accelerated irradiation studies, evaluation of industrial scale up, and degradation mechanism under accident scenarios. The focus of this talk will detail the motivation as well as research and development into fuels of interest within the AFC including doped oxide fuels and high uranium density fuels as well as prospects of increasing the burn up limits for current generation light water reactors.
9:00 AM
Fabrication and Thermophysical Properties of (U,Zr)C; A High Uranium Density Fuel Candidate for Nuclear Thermal Propulsion Reactors: Erofili Kardoulaki1; Brian Taylor2; Jhonathan Rosales2; Tim Coons1; Darrin Byler1; Ken McClellan1; 1Los Alamos National Laboratory; 2NASA
Advanced fuels with high melting points (>3000 K), high thermal conductivity, and ability to sustain exposure to corrosive environments are of interest for advanced nuclear reactors for terrestrial and space applications. These include advanced microreactor concepts for powering remote communities as well as space nuclear propulsion reactors to support manned Mars missions. One of the primary fuel candidates for these reactors is UN, however, adverse reaction of UN with hot hydrogen can prove detrimental. Another advanced fuel of interest is a mixed carbide such as (U,Zr)C, which can provide the same benefits as UN but has an even higher melting point and could resist hot hydrogen corrosion. In this work we have synthesized high purity (U,Zr)C via a carbothermic reduction route and have fabricated high density pellets via spark plasma sintering. Characterization of the pellets and measured thermophysical properties are presented here, confirming significant benefits of this fuel.
9:20 AM
High Temperature Mechanical Testing of Uranium Fuel Pellets: Tarik Saleh1; James Valdez1; Michael Torrez1; Scarlett Widgeon Paisner1; Kathryn Metzger2; Joshua White1; 1Los Alamos National Laboratory; 2Westinghouse Electric Corporation
In order to understand high temperature mechanical and creep behavior of fuel pellets at realistic reactor temperatures, 5x5 mm and larger fuel pellets were fabricated in the Fuels Research Laboratory at Los Alamos National Laboratory and tested in compression at temperatures of 1500-2000 °C. Data from these experiments will inform mechanical behavior models to elucidate the thermal portion of the creep behavior of fuel pellets in reactors during operation. This talk will discuss processing and fabrication of high density UN fuel pellets, along with traditional UO2 pellets. Once pellets were made, they were tested in compression in a specially modified load frame and furnace in high purity environments at elevated temperatures. Discussion of testing techniques along with results from these compression tests on UN and UO2 pellets will be presented.
9:40 AM
NOW ON-DEMAND ONLY - Material Degradation Analysis through Machine Learning-based Information Extraction from Legacy SFR Metallic Fuel Performance Data: Zhi-Gang Mei1; Aaron Oaks1; Kun Mo1; Yinbin Miao1; Logan Ward1; Abdellatif Yacout1; 1Argonne National Laboratory
The FIPD database, the sodium-cooled fast reactors (SFR) metallic fuels irradiation & physics database, is established at Argonne National Laboratory and contains massive amounts of pin-by-pin post-irradiation examination (PIE) data with corresponding EBR-II operating condition parameters. Although the collected irradiation data have been widely accepted to qualify metallic fuels up to 10 at.% burnup, there are less extensive data at higher burnups. Machine learning (ML) approaches are capable of extracting complex relationships from the existing FIPD data to increase the confidence on the evaluation of fuel performance at high-burnups. As a case study for cladding degradation analysis, random-forest and deep neural network-based ML models were developed to predict the cladding strain with respect to a variety of irradiation parameters. Both models are found to provide more accurate predictions than empirical models. Finally, the potential application of ML to accelerate metallic fuel qualification using these legacy data will be discussed.
10:00 AM Break
10:20 AM
Investigating the Thermophysical Properties and Key Contributions to the Thermal Conductivity of Different Nitride Systems: Conor Galvin1; Nicholas Barron2; Navaratnarajah Kuganathan1; Michael Cooper3; Robin Grimes1; 1Imperial College London; 2National Nuclear Laboratory; 3Los Alamos National Laboratory
Nitrides are an attractive alternative fuel form to oxides, particularly for fast reactors. New and existing empirical potential models were employed to predict temperature dependent properties of UN, PuN and (U,Pu)N. The thermophysical properties were investigated and compared to available experimental values. If we are to understand how fuel performs under reactor operating conditions, it is not only crucial to understand the thermophysical properties but also what effect defects have on these properties – especially as defects accumulate as fuel burn-up proceeds. Therefore, this work was extended to examine hyper-stoichiometric UN by adding U vacancies and N interstitials. Aside from these temperature dependent properties; thermal conductivity is one of the main properties that governs nuclear fuel performance. Therefore, we also predict the thermal conductivity of nitrides. The thermal conductivity is broken down into its constituent parts and their influence compared to the thermal conductivity of oxides.
10:50 AM Invited
Small Scale Mechanical Testing of Irradiated Cladding and Fuel for Nuclear Reactors: David Frazer1; Fabiola Cappia1; Daniel Murray1; Cameron Howard1; Yachun Wang1; Fei Teng1; Jatuporn Burns1; 1Idaho National Laboratory
Evaluating the mechanical properties of nuclear fuels and cladding is important for multiple reasons. Firstly, it enables understanding the pellet-cladding mechanical interactions during operation. Secondly, a better understanding of the fuel’s properties would assist in evaluating fine fuel fragmentation for fuel burn up extensions. A main challenge with measuring these properties is the high levels of radioactivity of the samples, making them difficult to handle. Small-scale mechanical testing can be used to minimize the volume of material need to measure their mechanical properties allowing them to be performed cheaply and rapidly outside of hot cells. The small volume probed enables the ability to measure the heterogeneity in the microstructure of irradiated materials not captured with macro-scale techniques. In this work, a variety of small-scale mechanical testing techniques were applied on control and irradiated fuel and cladding to measure the mechanical properties such as Young’s modulusand hardness over temperature.
11:20 AM
Phase Field Fracture Study of the Effect of Gas Bubble on Fracture at U-Mo/Zr Interface: Aashique Rezwan1; Sean Masengale1; Benjamin Beeler2; Yongfeng Zhang1; 1University of Wisconsin Madison; 2Idaho National Laboratory
A zirconium diffusion barrier is commonly used between the U-Mo fuel plate and the aluminum cladding in U-Mo based research reactor fuels. This Zr layers incorporation leads to forming an interaction zone with multiple sub-layers, including a U-enriched one. During operation, a high density of fission gas bubbles generates in the U-enriched sub-layer, making it susceptible to fracture. Fracture initiation and propagation have been experimentally observed in the interaction zone paralleling the Zr/U-Mo interface. This study investigates the role of gas bubbles on fracture behavior using the phase-field fracture model. By varying gas bubble density, size, alignment, and gas pressure, a series of simulations are carried out to establish a correlation between the critical fracture stress and the varying microstructure. Such a correlation will help assess the mechanical integrity of U-Mo fuel at different burn-up levels.