Materials in Nuclear Energy Systems (MiNES) 2021: Fuels and Actinide Materials- Oxide Fuels III
Program Organizers: Todd Allen, University of Michigan; Clarissa Yablinsky, Los Alamos National Laboratory; Anne Campbell, Oak Ridge National Laboratory
Thursday 3:30 PM
November 11, 2021
Room: Monongahela
Location: Omni William Penn Hotel
Session Chair: Michael Cooper, Los Alamos National Laboratory
3:30 PM Invited
Calculation of Irradiation Enhanced Diffusivities Using Centipede: Christopher Matthews1; Michael Cooper1; Romain Perriot1; Christopher Stanek1; David Andersson1; 1Los Alamos National Laboratory
The transport of defects in the supersaturated defect environment in irradiated nuclear fuel drives much of the important performance aspects of nuclear fuel. 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. Led by early indications that larger clusters could enable enhanced fission gas diffusion during irradiation, 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.
4:10 PM
Defect Clustering in UO2 Doped Systems Studied Using XAS and Neutron Scattering: Arjen van Veelen1; Joshua White1; Tashiema Ulrich1; Scarlett Widgeon Paisner1; Tarik Saleh1; 1Los Alamos National Laboratory
Uranium Dioxide (UO2) is the dominant fuel that powers nuclear reactors. UO2 fuel microstructure, in particular grain size, is known to affect both creep and fission gas release behavior which improves fuel lifetime and stability. Several additives have been recognized to enhance grain growth, of which Cr has been put forth by industry. The addition of Cr increases the grainsizes from 10 (± 3 µm) to 35 (± 5 µm). The dissolution of Cr into the UO2 lattice is strongly dependent on temperature and oxygen potential during sintering. Cr doping results in lattice contraction due to the introduction of structural defects. Additionally, Cr can potentially be accommodated on the uranium site at low temperatures and low concentrations. We used state-of-the-art X-ray Absorption Spectroscopy combined with X-ray and Neutron Diffraction to study the defect structures of Cr-doped and undoped UO2. We will discuss the Cr-doping effects on the UO2 defect chemistry.
4:30 PM
Dislocation Loop Evolution in Fluorite Oxides: Marat Khafizov1; Saqeeb Adnan1; Joshua Ferrigno1; Kaustubh Bawane2; Tiankai Yao2; Miaomiao Jin3; Chao Jiang2; Lingfeng He2; David Hurley2; 1Ohio State University; 2Idaho National Laboratory; 3Pennsylvania State University
We report on a combined experimental and modeling study focusing on dislocation loop evolution under irradiation in fluorite oxides. UO2, ThO2 and CeO2 were irradiated using a few MeV protons. Dislocation loops were characterized using transmission electron microscopy (TEM). A rate theory (RT) model is implemented to describe evolution of stoichiometric loops. Analysis of TEM results using RT model suggests that loop growth over 400-800 oC temperature range is governed by mobility of cation interstitials, whereas their nucleation is impacted by mobility of cation interstitials and anion defects. It was found that migration barrier for cation interstitials is proportional to the melting temperature of this oxides and migration barriers for other defects are consistent with atomic level simulations of defect energetics. This analysis provides a method to predict point defect concentrations, which impact the physical properties of these compounds, in particular thermal conductivity.
4:50 PM Cancelled
Grain Growth Kinetic Models for Accident Tolerant Oxide Fuel: Tashiema Ulrich1; Joshua White1; David Frazer2; 1Los Alamos National Laboratory; 2Idaho National Laboratory
Increasing demand for clean energy requires continuous improvement of the operating nuclear reactor fleet. For the last six decades, it has been demonstrated that increasing the grain size of UO2 has improved fission gas release behavior and is expected to increase the viscoplastic behavior of the fuel at operating temperatures. However, UO2 grain size increased by the addition of dopants complicates decoupling of dopant and large grain size effects. This work is aimed at developing a method that enhances the grain size of undoped UO2, which could be used as a reference for fission gas behavior in enhanced UO2 accident tolerant fuel. By changing the sintering dwell time and atmosphere, a 3 fold increase in grain size was achieved compared to standard UO2 manufacturing methods. These results will be discussed relative to the respective commercial dopants.This work was funded by the Department of Energy’s Accident Tolerant Fuel Campaign.