Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Modeling
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

Thursday 2:00 PM
March 2, 2017
Room: Point Loma
Location: Marriott Marquis Hotel

Session Chair: Shenyang Hu, Pacific Northwest National Laboratory; David Andersson, Los Alamos National Laboratory

2:00 PM  
Density Functional Theory Investigation of Defect and Fission Gas Diffusion in U3Si2: David Andersson1; 1Los Alamos National Laboratory
    One of the options being explored as an accident tolerant nuclear fuel is the U3Si2 compound, which benefits from high thermal conductivity (metallic) compared to the UO2 fuel currently used in reactors. The present study investigates point defect and fission gas diffusion in U3Si2 using density functional theory calculations. Point defect formation, Xe incorporation, migration of point defects and migration of Xe atoms were studied in a 2󫎾 supercell. The calculated defect formation and migration properties were used to estimate the Xe and U/Si self-diffusion coefficients. Our initial results show that Xe atoms diffuse much faster along the c axis than within the a-b plane of the U3Si2 structure. Diffusion along the c axis is faster than in UO2, while in-plane diffusion is somewhat slower. Similarly, point defect diffusion exhibits high diffusivities along the c axis and low diffusivities within the a-b plane of the U3Si2 structure.

2:20 PM  
A Grand-Potential Phase Field Model for Bubble Formation and Growth in U-Si Fuel: Karim Ahmed1; Larry Aagesen1; Daniel Schwen1; Yongfeng Zhang1; 1Idaho National Laboratory
    We present a novel grand-potential phase field model for investigating bubble formation, growth and coarsening in U-Si fuels. The grand-potential formulation has two major advantages. First, it decouples the interfacial and bulk properties, which allows us to use larger interface width for better computational efficiency. Second, since in this formulation the primary variable is the chemical potential and not concentration, it avoids the numerical issues that may arise from dealing with the extremely low vacancy and gas atoms concentrations. The model also considers the effects of bubble pressure and surface stress on the kinetics of growth. The numerical implementation of the model was carried out using the mesoscale simulator MARMOT developed at Idaho National Laboratory. The bubble coarsening rate is used to estimate the swelling rate in U-Si fuels and compared to experiments.

2:40 PM  
A Modified Embedded-Atom Method Interatomic Potential for U-Si: Benjamin Beeler1; Michael Baskes2; David Andersson3; Yongfeng Zhang1; 1Idaho National Laboratory; 2University of California, San Diego; 3Los Alamos National Laboratory
    Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). U-Si fuels benefit from higher thermal conductivity and higher fissile density compared to UO2. In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical Modified Embedded Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential accurately describes not only the primary phase of interest (U3Si2), but also a variety of U-Si phases across the composition spectrum. A ternary U-Si-Xe potential is also developed for the investigation of fission gas behavior in U-Si systems.

3:00 PM  
Cluster Dynamics Modeling of Cu Precipitation Hardening in Reactor Pressure Vessel Steels: Xian-Ming Bai1; Huibin Ke2; Pritam Chakraborty3; Yongfeng Zhang3; 1Virginia Tech; 2University of Wisconsin - Madison; 3Idaho National Laboratory
    Iron-based reactor pressure vessels (RPVs) in light water reactors are used as barrier structures for preventing the release of radiation to environment. The irradiation-induced supersaturated point defects accelerate the precipitation of low-solubility alloy elements or impurities such as copper in RPV steels. These precipitates are obstacles for the dislocation motion and cause hardening and embrittlement of RPV steels, which are the primary concerns for the consideration of the life extension of reactors. Here we use cluster dynamics simulations to model the radiation-induced precipitation of copper clusters in both high- and low- copper steels. The model can be applied to both steels under different irradiation conditions with only minor parameter adjustment. Finally the obtained number density and cluster size distribution of copper precipitates are used as inputs for Orowan’s looping model to predict the precipitation hardening at different neutron irradiation doses in the low-Cu RPV steel.

3:20 PM  
Monte Carlo Modeling of Recrystallization Processes in α-Uranium: Matthew Steiner1; Rod McCabe2; Elena Garlea3; Sean Agnew1; 1University of Virginia; 2Los Alamos National Laboratory; 3Y-12 National Security Complex
    Starting from electron backscattered diffraction (EBSD) data obtained from a warm clock-rolled α-uranium deformation microstructure, a Potts Monte Carlo model was used to simulate recrystallization while testing a number of different conditions for the assignment of recrystallized nuclei within the microstructure. Nucleation at non-twin high-angle grain boundary sites that exhibit a large degree of intra-granular misorientation produces the closest match to the experimental recrystallization texture. The ability of Monte Carlo modeling to simulate recrystallization processes in α-uranium promises wider applicability of the modeling technique to improve the understanding of recrystallization in other materials.

3:40 PM Break

4:00 PM  
Continuum-level Modeling of Irradiation Damage Cascades with Explicit Microstructure Representation: Jesse Carter1; Jared Tannenbaum1; Richard Smith1; 1Bettis Laboratory, BMPC
    Irradiation damage modeling of nuclear reactor materials in-pile is typically done without spatial dependence, the so-called “mean-field” rate theory, or in recent studies, using a single spatial dimension. However, irradiation damage events and/or microstructural sinks are still often considered uniform in space and/or time. This work is aimed at incorporating explicit spatial and temporal dependence on both sources and sinks of irradiation-induced defects at conditions relevant to LWR's at length scales that microstructural features can be resolved. Modeling in 1D, 2D, and 3D is facilitated by the MOOSE finite-element multiphysics framework. Results show a departure in equilibrium defect values from that predicted by rate theory and the deviation tends to increase as more spatial dimensions are included. Additionally, the spatial domain can be informed by TEM examinations of irradiated material by including features such as network dislocations, dislocation loops, and grain boundaries. Preliminary results using these simulated microstructures are discussed.

4:20 PM  
Phase Field Modeling of PWR Cladding Corrosion with the HOGNOSE Code: Andrew Dykhuis1; Michael Short1; 1Massachusetts Institute of Technology
    Cladding corrosion remains an area of concern for fuel rods in pressurized water reactors (PWRs). Although cladding performance has improved in recent years with continued alloy development, mechanistic modeling is not common. To address this, the authors have developed a finite-element model that models the implications of charged particle transport, including assessing charge neutrality in the oxide. This model, HOGNOSE, accounts for variations in corrosion behavior by considering the effects of alloying elements on material properties related to electron, oxygen ion, and oxygen vacancy transport. Excessive oxygen transport shifts the electrostatic potential in the system so that oxygen transport is reduced and electron transport is favored. Radiation damage in intermetallics can lead to doping of the oxide and a change in charged species transport properties. HOGNOSE is therefore able to accurately capture the initial kinetics of Zircaloy-4 before the kinetic transition and address a key factor in irradiation-enhanced corrosion.

4:40 PM  
Thermodynamic Modeling and Continuum Scale Fuel Performance Simulations: Jacob McMurray1; Srdjan Simunovic1; Theodore Besmann2; Benjamin Gaston2; Markus Piro3; 1Oak Ridge National Laboratory; 2University of South Carolina; 3Canadian Nuclear Laboratories
     Continuum scale simulations of oxide based fuel are currently addressing the coupled heat and mass transport phenomenon. The approach uses thermodynamics for key inputs such as chemical potentials, in order to understand how oxygen redistributes under a temperature gradient during burn-up and how this affects both the properties of the fuel, in particular thermal expansion, and the temperature gradient itself since thermal conductivity of the dominant fluorite urania phase is influenced by defects and stoichiometry. Efforts to model and simulate behavior of fuel with burn-up using an expanding thermodynamic data set for oxide fuel with fission products developed and integrated with an equilibrium solver, THERMOCHIMICA, in the MOOSE-BISON fuel performance code will be presented. This research was sponsored by the U.S. Department of Energy through the Office of Nuclear Energy–Fuel Cycle R&D and Nuclear Energy Advanced Modeling and Simulation Programs.

5:00 PM  
Exposing the Mechanisms of Pellet-Cladding Interaction Using Atomistic Simulation: Adam Plowman1; C.T. Gillen1; Alistair Garner1; P. Wiringgalih1; Michael Preuss1; Philipp Frankel1; Christopher Race1; 1University of Manchester
    Pellet-cladding interaction (PCI) is a fuel rod failure mode in light water nuclear reactors, which is thought to be driven by stress corrosion cracking (SCC) induced by aggressive species, most likely iodine, released from the fuel pellets during fission. An incomplete understanding of the mechanisms associated with PCI has led to conservative restrictions on reactor operation. Uncovering the mechanisms of PCI would allow reactors to respond to energy demand changes more quickly, without compromising safety. In tandem with new experimental observations, we are using atomistic simulation to improve our mechanistic understanding of PCI. We present a systematic study of zirconium grain boundary properties, including cleavage energies, undertaken using density functional theory. We have further studied the thermodynamics of impurities (including iodine) in these boundaries and their effect on grain boundary cohesion. We compare our results with new experimental data on iodine-induced SC cracks in commercial Zr alloys.