Ceramic Materials for Nuclear Energy Research and Applications: Fuel Performance Modeling and Fundamental Defect Science in Ceramics
Sponsored by: TMS Structural Materials Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Energy Committee, TMS: Nuclear Materials Committee
Program Organizers: Xian-Ming Bai, Virginia Tech; Yongfeng Zhang, Idaho National Laboratory; Maria Okuniewski, Purdue University; Donna Guillen, Idaho National Laboratory; Marat Khafizov, Ohio State University; Thierry Wiss, European Commission- JRC -Institute of Transuranium Elements – Germany

Monday 2:00 PM
February 27, 2017
Room: Palomar
Location: Marriott Marquis Hotel

Session Chair: Michael Tonks, Penn State University; Chris Stanek, Los Alamos National Laboratory


2:00 PM  Invited
Highlights of Ceramic Nuclear Fuel Research within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program: Chris Stanek1; 1Los Alamos National Laboratory
    The US Department of Energy – Office of Nuclear Energy program Nuclear Energy Advanced Modeling and Simulation (NEAMS) is developing a mechanistic computational toolset for nuclear fuel design and/or analysis. Multiscale materials modeling of ceramic nuclear fuel is an important element of this approach in order to allow the transition from empirical to more mechanistic models. By design, atomic and mesoscale models are necessarily connected to the development of an advanced fuel performance code. In this talk, several highlights of ceramic nuclear fuel research within this approach will be provided. Specifically, recent developments pertaining to thermal conductivity and fission product behavior in uranium dioxide will be discussed. Also, extension of this work to uranium silicide-based fuels will be presented.

2:30 PM  
Modeling the Effect of Percolation on Fission Gas Release in UO2 Nuclear Fuels: Larry Aagesen1; Daniel Schwen1; 1Idaho National Laboratory
    The release of fission gases from UO2 fuel pellets has an important impact on the performance of the fuel element. The process of accumulation and release of fission gases occurs in three stages: 1) the diffusion of fission gas atoms to grain boundaries and accumulation there, 2) nucleation and growth of fission gas bubbles at grain boundaries, and 3) percolation of grain boundary bubbles, leading to connection to a free surface and release of the fission gases to the plenum. A phase-field model of fission gas bubbles has been developed, and is used to investigate the percolation process at the microstructural level. Additionally, the physics of percolation have been incorporated to an engineering-scale model in the BISON code. Results from both length scales and a strategy for linking the two will be discussed.

2:50 PM  
Irradiation-induced Recrystallization in UO2: A Phase Field Study: Karim Ahmed1; Xianming Bai1; Yongfeng Zhang1; Daniel Schwen1; Cody Permann1; Bulent Biner1; 1Idaho National Laboratory
    We present a quantitative phase field model for investigating the irradiation-induced recrystallization. The model takes into consideration the evolution of dislocation density and the migration of grain boundaries. As such, the model captures both the nucleation of subgrains and their subsequent growth into stable grains. The nucleation of new subgrains is attributed to the high deformation energy of the existing grains due to the presence of high dislocation density. The numerical implementation of the model was carried out using MARMOT, the mesoscale simulator developed at Idaho National Laboratory. The model predicts a threshold dislocation density for nucleation of the subgrains that depends on the surface energy and elastic modulus of the material and irradiation temperature. The model has been applied to investigate the recrystallization in UO2. The generalization of the model to capture all the features of the high burn-up structure formation in nuclear fuel is discussed.

3:10 PM  
Sensitivity Analysis and Uncertainty Quantification of the MARMOT Mesoscale Fuel Performance Code: Marina Sessim1; Michael Tonks1; Jie Lian2; 1Pennsylvania State University; 2Rensselaer Polytechnic Institute
    MARMOT is a mesoscale fuel performance code under development by the Nuclear Energy Advanced Modeling and Simulation Program. It predicts the coevolution of microstructure and properties in reactor fuel and cladding materials. While MARMOT is a powerful tool already in use to inform the development of mechanistic materials models of fuel and cladding behavior, it is critical to understand the propagation of uncertainty in the model predictions. In this study we use the DAKOTA software to evaluate the sensitivity of effective thermal conductivity and fracture behavior calculations of UO2 using MARMOT to a wide range of parameters. Furthermore, we investigate the impact of parameter uncertainty, including distributions, means and standard deviations, on the predicted responses. Understanding the model uncertainty is a critical first step in validating the thermal conductivity and fracture models in MARMOT, such that a statistical comparison can be made between the simulation results and experimental data.

3:30 PM Break