Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface: Uranium Dioxide Fuels II
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
Monday 2:00 PM
March 20, 2023
Room: 26B
Location: SDCC
Session Chair: Mia Jin, Pennsylvania State University
2:00 PM Introductory Comments
2:05 PM Invited
Multiphysics Modeling of Nuclear Fuels at the Mesoscale: Karim Ahmed1; 1Texas A&M University
The drastic change in the microstructure of nuclear fuels was linked to several degradation mechanisms affecting their performance and integrity during normal operation, transients, and accidents. It is therefore crucial to understand the irradiation-induced co-evolution of microstructure and thermo-mechanical properties of nuclear fuels. In this talk, I will discuss how we can employ a multi-physics modeling approach at the mesoscale to achieve that goal. Particularly, I will elaborate on the best practices of utilizing spatially resolved rate-theory, phase-field, and finite-element modeling methods to simulate radiation damage and effects in selected nuclear fuels. Specific examples will include modeling the mechanical properties and fracture of UO2 and TRISO particles and thermal properties of U-Zr, UO2-BeO, and UO2-Mo fuels. Moreover, the utilization of Machine Learning techniques to derive reduced order models will also be presented in this talk.
2:30 PM
Comparing the Impact of Thermal Stresses and Bubble Pressure on Intergranular Fracture in UO2 Using 2D Phase Field Fracture Simulations: Shuaifang Zhang1; Wen Jiang2; Kyle Gamble2; Michael Tonks3; 1Oak Ridge National Laboratory; 2Idaho National Laboratory; 3University of Florida
UO2 fuel fragmentation and pulverization during loss-of-coolant accidents (LOCAs) is an ongoing safety concern that has gained importance due to recent interest in increasing burnup limits for light water reactor fuel. In this work, we investigate the importance of bubble pressure on fragmentation using 2D phase field fracture simulations of UO2 polycrystals. A stress is applied that is representative of the internal stresses observed in macroscale BISON simulations of LOCA tests. Cracks do not form in UO2 polycrystals with the applied stress and no porosity. Crack nucleation and propagation do occur throughout stressed polycrystals with unpressurized intergranular voids. They also occur in polycrystals with and without applied stress that have pressurized bubbles. Our results indicate that bubble pressure may not be necessary to initiate fragmentation in polycrystals with intergranular porosity under LOCA conditions.
2:50 PM
Extended Defect Coalescence in Kr Irradiated UO2 During High Temperature Annealing: Joshua Ferrigno1; Chang-Yu Hung2; Lingfeng He3; Marat Khafizov1; 1The Ohio State University; 2Johns Hopkins University; 3Idaho National Laboratory
Transmission electron microscopy (TEM) taken of 0.7 MeV-Kr irradiated UO2, after isochronal annealing (1000°C-1600°C, 1 hour) showed reduction in density and increase in size of interstitial dislocation loops and fission gas bubbles at increasing temperatures. A Rate theory (RT) model accounting for point defect recombination and absorption in loops and bubbles, and extended defect merging was employed to explore the significance of extended defect merging in UO2. Two primary mechanisms were considered for extended defect growth, the first looks at the interactions of the point defect populations with extended defects while the second examines solely the interaction of extended defects merging with another. Comparison of the models yielded the activation energy and confirmed that little interaction takes place with the point defect population during annealing due to the predomination of defect merging. This suggests that traditional Rate Theory models require consideration of separate defect mechanisms during annealing, not considered during irradiation.
3:10 PM
Modeling Stoichiometry Controlled Defect Dependent Densification in UO2±X: Brandon Battas1; Michael Cooper2; Michael Tonks1; 1University of Florida; 2Los Alamos National Laboratory
Many common nuclear fuels are processed from powders, and as such retain significant porosity. Therefore, it is important to understand how the fuel will sinter in a reactor setting, especially as that happens concurrent to other in reactor microstructural evolution. One aspect that affects the densification of sintered ceramic fuels, such as UO2±X, is the defect concentration. The defect concentration in the material helps determine the diffusivity, and the densification is largely diffusion controlled. This study focuses on mesoscale modeling of the effect of varying defect concentration in UO2±X, which is largely controlled by factors such as the stoichiometry of the fuel or the addition of irradiation to a system, on densification. We used atomistic simulations to quantify the impact of stoichiometry on defect concentrations, then used that data to model the densification in mesoscale using MARMOT.
3:30 PM Break
3:45 PM
Modeling Fission Gas Release Behavior from Microcracking and Thermal Diffusion at High Burnup in UO2 Fuel in BISON: Oliver Baldwin1; Walter Brinkley1; Jonathan Norman1; Nathan Capps2; Brian Wirth1; 1University of Tennessee; 2Oak Ridge National Laboratory
Fission gas release (FGR) is a behavior of interest for fuel performance modeling, particularly at high burnup. High burnup fuels exposed to a temperature transient exhibit a burst release mechanism of this accumulated fission gas. Burst release of fission gas is attributed to micro-cracking and interconnecting between fission gas bubbles and ultimately extensive fuel fragmentation. The finite element fuel performance code, BISON calculates fission gas release through a module referred to as Sifgrs (Simple Integration Fission Gas Release and Swelling). This module contains a combination of mechanistic and empirical models describing the fission gas behavior in UO2 fuel. Additional high burnup transient fission gas release data has been acquired and was used to improve the micro-cracking correlation. This model will be implemented as a new empirical model in BISON to account for high burnup fission gas release mechanisms.
4:05 PM
Uncertainty Quantification of Thermal Performance of UO2 Fuel Pellets: Robert Annewandter1; 1Nuclear Futures Institute
Predicting thermal performance of a UO2-Zr system at high discharge burn up poses a challenge for nuclear fuel performance codes if they rely on empirical models. For a well established nuclear fuel performance code (TRANSURANUS) a statistical approach is adopted to quantify uncertainties in thermal transfer from pellet to cladding arising at high discharge burn-up. The adopted uncertainty quantification is supported by MD simulations to understand the thermal performance of High Burnup Structures in the rim region, and to quantify the effect of stoichiometry on thermal conductivity. Uncertainty modelling will be used to target the most rapid reduction of uncertainty of fuel behaviour possible.
4:25 PM
Atomistically-informed Cluster Dynamics Modelling of Defect Evolution in Irradiated ThO2: Sanjoy Mazumder1; Maniesha Singh1; Tomohisa Kumagai1; Anter El-Azab1; 1Purdue University
A cluster dynamics (CD) model has been developed to investigate the nucleation and growth of point defect clusters, i.e., interstitial prismatic loops and nanoscale and sub-nanoscale voids, in ThO2 during irradiation by energetic particles. The model considers cluster off-stoichiometry due to the asymmetry of point defect generation on the O and Th sublattices under irradiation, as well as the point defect diffusivities and the defect binding energies to clusters. The latter energies were established using detailed molecular dynamics simulations considering the statistical variability of cluster configuration. The predicted loop density and their average size were found to be in good agreement with reported experimental observations for proton irradiated ThO2 at 600oC. Further annealing of irradiated ThO2 at 1100oC enhance the kinetics of immobile vacancies resulting in void growth. The CD prediction closely corresponds to the TEM observations, additionally validating the model fidelity.
4:45 PM
Modeling Low-temperature Hydrided Zircaloy Cladding Failure Under a Reactivity-initiated Accident: Katheren Nantes1; Miaomiao Jin1; Arthur Motta1; 1Pennsylvania State University
A reactivity-initiated accident (RIA) caused by the loss of a control rod results in rapid fuel thermal expansion and pellet cladding mechanical interaction which may deform the Zircaloy cladding at relatively low temperatures, when hydrides are still present. The hydrides are known to embrittle Zircaloy cladding, and may cause it to fail earlier than predicted. To investigate the likelihood of low-temperature failures during an RIA, we utilize the fuel performance code BISON to simulate cladding deformation as a function of pulse/reactivity characteristics (energy deposition, pulse width, energy distribution of deposited energy) within the pellet at different burnups. The goal of the project is to develop a physically-based failure criterion that considers not only hydrogen concentration but also hydride distribution and the state of stress in the cladding imposed on the cladding.
5:05 PM
Diffusion Coefficients of Zr- and Cr-based Binary Systems for Simulation of Cr-coated Zircaloy Nuclear Fuel Cladding: Ella Pek1; Wei Zhong1; Ji-Cheng Zhao1; 1University of Maryland
Chromium-coated Zircaloy is one of the accident-tolerant fuel (ATF) claddings. A thin Mo, Nb, or Ta interlayer may be placed between Cr and Zircaloy to serve as a diffusion barrier to reduce reaction between Cr and Zircaloy under severe accident conditions. Diffusion coefficients of Zr-(Cr, Mo, Nb, Ta), Zircaloy-Cr, and Cr-(Mo, Nb, Ta) binary systems are essential for accurate prediction of the ATF lifetime. Diffusion coefficients of these binary systems at 6 temperatures, including both hcp and bcc phases, were measured using high-throughput diffusion multiples. These diffusion coefficients will help establish a reliable and comprehensive diffusion coefficient (atomic mobility) database for the 8 binary systems to fill a knowledge gap in diffusion data for the Cr-coated Zircaloy ATF.
5:25 PM
Molecular Dynamics Study of the Anisotropic Elastic Response of Defects in Alpha-Uranium: Yuhao Wang1; Benjamin Beeler2; Andrea Jokisaari3; 1University of Michigan; 2North Carolina State University; 3Idaho National Laboratory
Alpha-uranium is a highly anisotropic material that demonstrates a complicated mechanical behavior under irradiation. Understanding the elastic response of defects in alpha-uranium and its relationship with temperature will provide critical insight into the structural evolution of metallic fuels under irradiation, such as the anisotropic growth. In this work, molecular dynamics simulations are performed to calculate the elastic constant of alpha-uranium and the volumetric strain induced by single defects and small defect clusters. The homogeneous strain of different defects is calculated in both NVT and NPT ensembles, and the results of different ensembles are compared to validate the reliability of the method. Moreover, the elastic dipole tensor and the lambda-tensor are determined for each type of defect considered, and the anisotropy of these elastic response properties is analyzed. These results will contribute to mesoscale phase-field modeling with important atomistic inputs.