Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface: Uranium Carbides, Nitrides and Silicides
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 8:00 AM
March 21, 2023
Room: 26B
Location: SDCC
Session Chair: Jennifer Watkins, Idaho National Laboratory
8:00 AM Introductory Comments
8:05 AM Invited
Accelerating the Qualification of Nuclear Fuels Through Advanced Characterization and Multiscale Modeling: Joshua White1; 1Los Alamos National Laboratory
Qualification of nuclear fuels for use in modern nuclear reactors currently requires lengthy, and costly, irradiation test campaigns to provide assurance of the reactor performance under a variety of scenarios. The Nuclear Regulatory Commission is interested in advanced modeling and simulation methods to expedite the qualification of fuels by utilizing improved mechanistic models combined with separate effects experiments to decrease the number of required data sets to commission fuels for use. The use of innovative machine learning (ML) tools coupled with state-of-the-art experimental measurements will be used to benchmark the models as well as reduce the uncertainty, providing confidence in fuel performance in-pile. This talk will focus on UN and UC, two potential high-impact fuels for microreactors and gas-cooled reactors. The developed ML models will be integrated with multiphysics engineering-scale neutronics and reactor simulations that ultimately define safety margins for novel nuclear fuels within specific reactor conditions.
8:30 AM Invited
Chemical Structures and Thermodynamics of Uranium Nitride and Uranium Carbide: Xiaofeng Guo1; Vitaliy Goncharov1; Juejing Liu1; Arjen van Veelen2; Joshua White2; Hongwu Xu2; 1Washington State University; 2Los Alamos National Laboratory
In the U.S. and many other countries, carbide and nitride matrices have received considerable attention as advanced nuclear fuel types. Compared to UO2, both of UC and UN have the advantages of high thermal conductivity, high melting point, and high fissile density. A fundamental understanding of the material chemistry and thermodynamic properties of UC and UN is critical for predicting their behavior under reactor or extreme conditions. In this work, we investigated (1) the local structures by X-ray diffraction and X-ray absorption fine structure, (2) bulk thermal oxidations by TGA – DSC, and thermochemical reactions, including the enthalpy of oxidation and standard enthalpy of formation of UN and UC by high temperature drop solution calorimetry. These updated understanding of UN and UC have two implications: (1) enable thermodynamic modeling and DFT computation for U-C, U-N, and U-C-N; and (2) a foundation for future studies on UC-, and UN-derived waste forms.
8:55 AM
Fabrication and Characterization of Uranium Carbide: Adrien Terricabras1; Arjen van Veelen1; Erofili Kardoulaki1; Scarlett Widgeon Paisner1; Timothy Coons1; Joshua White1; 1Los Alamos National Laboratory
Uranium carbide have been investigated as a potential fuel candidate for advanced reactors and nuclear thermal propulsion due to its attractive properties. Historical data showed large uncertainties due to impurities and sintering issues. Minute differences in feedstock preparation will result in hypo- or hyper-stoichiometric UC, greatly affecting its physical, thermal, and mechanical properties and overall composition. The lack of recent experimental data on well characterized feedstock and the discrepancies between existing data hinders the development of accurate models. Two different UC feedstocks were prepared using arc melting and carbothermic reduction. Chemical composition was determined using X-ray diffraction, and carbon/oxygen content was measured using combustion analysis. Thermo-physical properties were characterized using laser flash analysis, differential scanning calorimetry and dilatometry. Extended X-ray absorption fine-structure measurements were conducted to study the local structure pre and post sintering. Results from this study will be discussed and compared to historical work published on uranium carbide.
9:15 AM
Nuclear Fuels and Interfaces for Advanced Specialty Microreactors: Erofili Kardoulaki1; Najeb Abdul-Jabbar1; Josh White1; Scarlett Widgeon-Paisner1; Maria Kosmidou1; Mehadi Hassan1; Ken McClellan1; 1Los Alamos National Laboratory
Multiple agencies are investing heavily in specialty microreactor technologies to support terrestrial power applications where grid support is limited and extraterrestrial applications supporting manned missions for deep space exploration. Microreactors present unique design and materials challenges and must be commercially attractive in order to succeed. In all cases fuels that enable high thermal conductivity throughout the reactor lifetime and can withstand exposure to extreme environments (e.g. H2 and graphite exposure) are key. In many cases carburization is a key consideration, due to the graphite core that is used to provide increased moderation, and protective coatings have to be employed to ensure safe and robust operation. Fuel-cladding systems in contact with graphite, i.e. UN-Zr-C, will be discussed in the context of materials performance optimization to provide robust systems that will retain fission products during normal and off-normal operation, promote heat transfer, and ensure mechanical integrity.
9:35 AM Break
9:50 AM
Chemical Interaction and Compatibility of Uranium Nitride and Alumina Forming Austenitic Alloys: Andre Broussard1; Dong Zhao1; Jie Lian1; Bruce Pint2; Jiheon Jun2; Jason Harp2; Erofili Kardoulaki3; 1Rensselaer Polytechnic Institute; 2Oak Ridge National Laboratory; 3Los Alamos National Laboratory
Uranium Nitride (UN) and Alumina Forming Austenitic Alloys (AFAs) have been identified as a potential fuel-cladding combination for Westinghouse’s lead cooled fast reactor (LFR). The chemical interaction between UN (monolithic and with expected secondary phases), preoxidized (alumina scaling on the surface), and as-cast (no alumina scaling) AFA, was investigated by utilizing a diffusion couple setup consisting of spark plasma sintered UN pellets and AFA coupons prepared by ORNL. The diffusion couples were isothermally annealed at both 550℃ and 750℃ for 500 and 1000 hours under inert conditions. Preoxidized AFA exhibits a significant reduction of interaction with UN compared to as-cast AFA due to the protective alumina scaling. Interaction between as-cast AFA and UN exhibits the formation of aluminum nitride and a suspected uranium-iron ternary phase.
10:10 AM
Modeling of Fission Gas Behavior in Uranium Nitride Fuel: Jason Rizk1; Christopher Matthews1; Michael Cooper1; Anders Andersson1; 1Los Alamos National Laboratory
Uranium nitride fuel is a candidate for accident tolerant fuel designs and microreactors due to its high uranium density and high thermal conductivity. Its fission gas and swelling behaviors need to be better understood before its widespread use. Uranium nitride, like uranium carbide, exhibits “breakaway swelling,” which is an accelerated swelling rate once conditions exceed temperature and burnup thresholds. Free energy cluster dynamics simulations are used to calculate irradiation-enhanced Xe diffusion and track irradiation damage. The diffusion coefficients are utilized in fuel performance simulations to calculate fission gas release and swelling. The fuel performance simulations are compared to experimental results, and a sensitivity analysis and uncertainty quantification are performed. This work advances the fuel performance modeling capabilities of uranium nitride fuel, and provides insight into the “breakaway swelling” phenomenon.
10:30 AM
High Temperature Steam Oxidation Performance of Alloyed, High Density Fuel Composite: U3Si2 + 50wt% UB2: Geronimo Robles1; Joshua White2; Scarlett Widgeon Paisner2; Elizabeth Sooby1; 1University of Texas at San Antonio; 2Los Alamos National Laboratory
Recently, advanced nuclear reactor fuel development has investigated high density fuel (HDF) compounds and composites like UN/U3Si2 and UB2/U3Si2 as drop-in replacements of traditional UO2 for their superior uranium density and thermal conductivity. Despite improved fuel structural stability, economy, and safety of these HDF’s, mitigating the energetic pulverization of U3Si2 when exposed to high temperatures and oxidants possible during off-normal conditions remains paramount. Techniques like compositing U3Si2 with UB2 at concentrations <10wt% has achieved a delay in oxidation by >100°C and alloying and oxide dispersion strengthening (ODS) U3Si2 have also shown delays. Reported here, U3Si2 + 50wt% UB2 HDF composites are fabricated with Al and Al2O3 additives. As fabricated and heat-treated samples are oxidized and characterized to examine the impact of complex microstructure on performance in high temperature oxidizing atmospheres. Fabricability challenges for each composition are reported. SEM/EDS and p-XRD analysis and thermogravimetric data related changes to performance.
10:50 AM
Assessment of High-density Fuels During Hydrogen Interaction: Adrian Gonzales1; Elizabeth Sooby1; Joshua White2; 1The University of Texas at San Antonio; 2Los Alamos National Laboratory
High-density fuels (HDF) are nuclear fuels that contain a higher uranium density than traditional uranium dioxide. Research of HDF focusing on the oxidation of U3Si2, UN, UC, and UB2 in air, water, and steam environments concluded uranium forms UO2 and U3Si2 forms a hydride. However, minimal research has been done on the interaction between HDF and hydrogen. Typical oxidation characterization involves using XRD and SEM images which can be difficult in detecting hydrogen absorption. The Sieverts technique is implemented in the present study to detect hydrogen absorption by monitoring pressure drops during the hydrogen interaction with HDF. The results presented cover a temperature range from 100 to 500 °C in up to atmospheric pressures for 24-96 hour dwells. Additionally, a LECO instrument is used for elemental analysis of the hydrogen content after testing. The presented data will be discussed in relation to LWR fuel performance interactions with hydrogen containing atmospheres.
11:10 AM
Assessing the Influence of Microstructure on Uranium Hydride Size Distributions via Small Angle Neutron Scattering: Zachary Harris1; Elena Garlea2; Tasha Boyd2; Lisa DeBeer-Schmitt3; Kenneth Littrell3; Sean Agnew4; 1University of Pittsburgh; 2Y-12; 3Oak Ridge National Laboratory; 4University of Virginia
The effect of microstructure on the internal hydriding behavior of both cast (1 mm grain size) and rolled (25 m grain size) uranium containing hydrogen concentrations between 0 and 1.8 wppm were evaluated via small angle neutron scattering (SANS). Increasing hydrogen content up to 1.8 wppm in the cast uranium only weakly affected the average uranium hydride (UH3) precipitate size, calculated from the SANS data. Conversely, the UH3 phase fraction was found to strongly depend on the hydrogen content in the same as-cast samples. A substantially reduced UH3 particle size distribution was observed in the rolled uranium relative to cast uranium containing the same nominal hydrogen content. It is hypothesized that the suppression of UH3 formation in the rolled uranium is driven by increased hydrogen trapping at grain boundaries, and theoretical calculations that account for trap density, potency, and hydrogen diffusion kinetics support this hypothesis.