Ceramic Materials for Nuclear Energy Research and Applications: Non-oxide Ceramics for Nuclear Applications II
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
Wednesday 2:00 PM
March 1, 2017
Location: Marriott Marquis Hotel
Session Chair: Andrew Nelson, Los Alamos National Laboratory
2:00 PM Invited
Multi-scale Coupled Radiation Damage and Heat Transport Modeling for Dispersed Nuclear Fuels: Daniel Schwen1; Sebastian Schunert1; 1Idaho National Laboratory
Modeling nuclear fuel in operation is a challenging multi-physics and multi-scale problem. The accumulated radiation damage, thermal and mechanical loading, and chemical changes experienced during the operational lifetime of the fuel have a strong influence on its material properties both directly and indirectly by impacting the microstructural evolution. This causes a feedback through the modified the neutron transport characteristics of the material. A system of particular interest is the dispersed fuel in INL’s transient test reactor (TREAT). Based on INL’s Multiphysics Object-Oriented Simulation Environment (MOOSE), a damage production model has been developed that can use neutron reaction data obtained through a concurrent neutron transport simulation to obtain spatially resolved recoil energy and mass distributions to initiate cascade simulations using a binary collision Monte Carlo code. The discrete cascade data is used as a source term in a phase field microstructure simulation component and to model thermal conductivity changes.
2:30 PM Invited
Neutron Irradiated SiC Advanced Analysis to Understand Fission Product Transport: Safety Tested TRISO Coated Particles: Isabella van Rooyen1; Tom Lillo1; Karen Wright1; Jeffery Aguiar1; Terry Holesinger1; 1Idaho National Laboratory
Transport of metallic fission products through intact SiC layers of TRISO coated particles is studied to answer the poorly understood transport phenomena which has not been directly correlated with simulated out-of-reactor research experiments. Recent nano-scaled studies on neutron irradiated TRISO coated particles from the AGR-1 experiment provided new insights to these phenomena. Specifically in this study safety tested Variant 3 type UCO fueled TRISO coated particles were examined using Scanning Electron Microscopy, Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy, High Resolution Electron Microscopy and Electron Probe Micro-Analyzer to provide information on fission product distribution and structure. Compact 4-3-3 achieved 18.63% FIMA (average volume average temperature of 1094°C) and was then exposed subsequently to a safety test at 1600°C. The two particles selected for this comparative analysis are particle AGR1-433-003 with a measured-to-calculated retained 110mAg of <22% and particle AGR1-433-007, which has a measured-to-calculated retained 110mAg of 100%.
Processing Routes for Improving Purity and Theoretical Density of UN Microspheres: Jacob McMurray1; Terry Lindemer1; Rodney Hunt1; Jack Collins1; Chinthaka Silva1; Jim Kiggans1; Kurt Terrani1; 1Oak Ridge National Laboratory
Uranium nitride (UN) offers higher heavy metal density when compared to urania that could prove attaractive in ceratin reactor platforms with high moderator to fuel ratio. An effort is ongoing to produce UN microshperes that may be used as fuel kernels in TRISO particles or feedstock for solid pellet production. The developed process to date yields nearly 90% theoretical density (TD) UCyN1-y (solid solution of UN and UC) kernels starting from internal gelation feedstock spheres composed of hydrated UO3 and C. With y values varying between 0.2 and 0.4, this work reports on efforts to increase the N content by implementation of an HCN isothermal stripping step. Alternatively, the feedstock C content can be adjusted to achieve similar N enrichments. The effects of Hot Isostatic Pressing along with feedstock C content and air dired gel microstructure on the final density, microstructure, and phase will also be discussed. Research supported by the US Department of Energy, Office of Nuclear Energy, Advanced Fuels Campaign.
Evolution of Irradiation Defects in Ti2AlC Ceramics During Heavy Ion Irradiation: Bai Cui1; Fei Wang1; Qing Su1; Michael Nastasi1; 1University of Nebraska–Lincoln
Ti2AlC is a promising material for nuclear fuel cladding applications due to its irradiation tolerance and oxidation resistance in high temperature environments. However, the fundamental process of the defect generation, clustering, and annihilation in Ti2AlC in the irradiation environment is not well understood. This study examines the evolution of irradiation defects under Kr ion irradiation at room and elevated temperatures by using the IVEM-Tandem Facility. The dependence of defect size and density on the irradiation dose and temperature was characterized by in-situ transmission electron microscopy (TEM) experiments. The localized disorder of atomic positions in the irradiated Ti2AlC was investigated in high-resolution TEM studies. These results suggest that although this material shows exceptional resistance to amorphization over a high level of irradiation dose, the localized nanoscale defect was accumulated in the microstructures which may cluster to larger defects, or annihilate at elevated temperatures.