Advanced Characterization and Modeling of Nuclear Fuels: Microstructure, Thermo-physical Properties: Nuclear Fuels Microstructure-Modeling
Sponsored by: TMS Structural Materials Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Energy Committee, TMS: Nanomechanical Materials Behavior Committee, TMS: Nuclear Materials Committee
Program Organizers: David Frazer, Idaho National Laboratory; Fabiola Cappia, Idaho National Laboratory; Tsvetoslav Pavlov, Idaho National Laboratory; Peter Hosemann
Tuesday 2:30 PM
March 1, 2022
Room: 202B
Location: Anaheim Convention Center
Session Chair: David Frazer, Idaho National Laboratory
2:30 PM Invited
Mesoscale Hybrid Model for Fission Gas Behavior in UO2: Coupling the Phase Field Method to Spatially Resolved Cluster Dynamics: Sophie Blondel1; David Andersson2; David Bernholdt3; Dong-Uk Kim4; Fande Kong5; Md Ali Muntaha4; Philip Roth3; Michael Tonks4; Brian Wirth1; 1University of Tennessee; 2Los Alamos National Laboratory; 3Oak Ridge National Laboratory; 4University of Florida; 5Idaho National Laboratory
Fission gas release within UO2 occurs as gas atoms diffuse through grains and arrive at grain boundary (GB) bubbles, GB bubbles grow and interconnect with grain edge bubbles, and finally grain edge tunnels grow and connect to free surfaces. We present a multi-physics simulation approach to investigate these mechanisms at the mesoscale. The fission gas production and intragranuar physics are included using spatially resolved cluster dynamics in the Xolotl code. Intergranular gas bubble growth and interconnection are included using the phase field method in the MARMOT fuel performance code. The two codes are loosely coupled through the MOOSE MultiApp system: MARMOT passes the GBs locations to Xolotl, and Xolotl passes fission gas rate at GBs surfaces to MARMOT. We have demonstrated this hybrid model behaves correctly using 2D simulations, and a model for fission gas outgazing is being added. Initial 3D results will also be presented.
3:00 PM
Adding Irradiation-assisted Grain Growth to the MARMOT Tool for UO2 Nuclear Fuel: Md Ali Muntaha1; Larry Aagesen2; Michael Tonks1; Zefeng Yu3; Arthur Motta3; 1University of Floirda; 2Idaho National Laboratory; 3Penn State University
MARMOT simulations have been extensively used to investigate grain-boundary migration and grain growth in UO2. While MARMOT was created to model irradiated fuel and cladding at the mesoscale, the current UO2 grain growth model in MARMOT only considers grain growth kinetics of fresh and unirradiated fuel. Thus, the model is missing a critical mechanism that could significantly reduce the accuracy of its predictions. In this study, we are expanding the capabilities of the MARMOT simulation tool to include the impact of irradiation on UO2 grain growth. We have implemented the irradiation effects into MARMOT considering thermal spike by coupling to a heat conduction simulation with a random heat source. We will validate the modeling predictions against in-situ-TEM experimental results for ion-irradiation conditions with nanocrystalline UO2 thin films. Once validated, we can employ the improved model to assess whether this missing mechanism is crucial for grain growth in LWR fuel pellets.
3:20 PM
Multiphysics Modeling of Fracture in Sintered Uranium Dioxide Pellets: Levi McClenny1; Mohammad Abdoelatef1; Moiz Butt1; Hari Krishnan1; Michal Pate1; Kay Yee1; Wen Jiang2; Karim Ahmed1; Sean McDeavitt1; 1Texas A&M University; 2Idaho National Laboratory
Commercial nuclear power plants extensively rely on fission energy from uranium dioxide fuel pellets to provide thermal energy to the coolant in current generation reactors. UO2 fuel incurs damage and fractures during operation due to large thermal gradients between the fuel and the coolant in the reactor core. In this work we describe an experimental study performed to understand the fuel fracturing behavior of sintered powder UO2 pellets when exposed to thermal shock conditions, as well as a computational fracture model which accurately predicts the experimental results. Experimental data was collected from multiple experiments by exposing UO2 pellets to a cartescal bath (600-900 C) which are subsequently quenched in sub-zero water. We exhibit that the fracture results gathered in the experimental setting are consistently recreated in a computational setting, demonstrating a reliable ability to computationally simulate thermal shock gradients and subsequent fracture mechanics in the primary fuel source of LWRs.
3:40 PM Invited
Modeling Irradiation-enhanced Diffusion in Advanced Ceramic Nuclear Fuels: Michael Cooper1; Christopher Matthews1; Vancho Kocevski1; Christopher Stanek1; David Andersson1; 1Los Alamos National Laboratory
Nuclear fuel performance and the degradation of fuel properties are governed, in many respects, by the formation and diffusion of point defects and clusters in the lattice. For example, the diffusion of fission gas and vacancies through the lattice controls fission gas swelling and release, which are key performance metrics that also impact thermal transport. During reactor operation, the concentration and diffusion of defects can be enhanced through irradiation processes. In this work, we represent atomic scale calculations of the diffusion mechanisms of host and impurity (Xe) defects in ceramic nuclear fuels, such as doped UO2, UN, and UC. The atomic scale predictions of the stability and diffusivity of point defect and clusters in these systems have then been implemented in cluster dynamics simulations to predict irradiation-enhanced defect concentration and diffusivities. The importance of the in-reactor conditions and of including various defects will be discussed.
4:10 PM Break
4:30 PM
Centipede: A New Tool for Calculating Irradiation Enhanced Transport of Defects in Nuclear Fuel: Christopher Matthews1; Michael Cooper; Romain Perriot1; Xiang-Yang Liu1; Chris Stanek1; David Andersson1; 1Los Alamos National Laboratory
The supersaturation of defects in irradiated nuclear fuel complicates the transport of defects. Coupled with the impacts of temperature and chemistry, the behavior of the defects drives much of the microstructural evolution and subsequent macroscopic performance of fuels. Calculations of defects have traditionally relied on rate theory to estimate defect concentrations to provide meaningful information to fuel performance simulations, albeit with limited scope, and subsequently, poor comparison to experimental measurements. The cluster dynamics code Centipede has been developed to incorporate the hundreds of defects that must be tracked in order to account for the impact of large defect clusters. Through the utilization of an extensive database of atomistic calculations, the mechanistically calculated irradiation enhancement of fission gas in UO2 has been shown to compare favorably with experimental measurements. Centipede has been further refined using machine learning and has been applied to advanced fuels for which data is absent.