David Frazer

Profile Information
Name
Dr. David Frazer
Institution
Idaho National Laboratory
Position
Staff member
h-Index
ORCID
0000-0001-5139-858X
Expertise
Cladding, Mechanical Properties, Nanoindentation, Nuclear Fuel
Publications:
"High-Temperature Nanoindentation of SiC/SiC Composites" David Frazer, JOM Vol. 72 2020
The results of high-temperature nanoindentation testing on both a control and a neutron-irradiated silicon carbide matrix silicon carbide fiber composite sample are presented. The mechanical properties of the chemical vapor-infiltrated matrix were observed to have slightly increased in hardness and slightly decreased in elastic modulus after irradiation. Tyranno SA3 fiber behavior results are inconclusive, possibly because residual graphite in the fibers resulting from the manufacturing process produced a large scatter in the data. This work also demonstrates the capability to measure the individual components of fabricated composites at elevated temperature, which should provide inputs for modeling the macro-scale behavior of the composites.
"Mechanical characteristics of SiC coating layer in TRISO fuel particles" Thak Sang Byun, David Frazer, Peter Hosemann, John Hunn, Maria Okuniewski, Kurt Terrani, Gokul Vasudevamurthy, J. N. Matros, Brian Jolly, Journal of Nuclear Materials Vol. 442 2013 133-142 Link
Tristructural isotropic (TRISO) particles are considered as advanced fuel forms for a variety of fission platforms. While these fuel structures have been tested and deployed in reactors, the mechanical properties of these structures as a function of production parameters need to be investigated in order to ensure their reliability during service. Nanoindentation techniques, indentation crack testing, and half sphere crush testing were utilized in order to evaluate the integrity of the SiC coating layer that is meant to prevent fission product release in the coated particle fuel form. The results are complimented by scanning electron microscopy (SEM) of the grain structure that is subject to change as a function of processing parameters and can alter the mechanical properties such as hardness, elastic modulus, fracture toughness and fracture strength. Through utilization of these advanced techniques, subtle differences in mechanical properties that can be important for in-pile fuel performance can be distinguished and optimized in iteration with processing science of coated fuel particle production.
"Microstructural and nanomechanical characterization of in-situ He implanted and irradiated fcc materials" David Frazer, Peter Hosemann, Djamel Kaoumi, Ce Zheng, Microscopy & Microanalysis Vol. 23 (Suppl 1) 2017 756-757 Link
"Nano- and micro-indentation testing of sintered UO2 fuel pellets with controlled microstructure and stoichiometry" David Frazer, Journal of Nuclear Materials Vol. 516 2019 169-177 Link
Dense nanocrystalline and microcrystalline UO2 samples with controlled grain structure and stoichiometry were prepared by high energy ball milling and spark plasma sintering (SPS). Nano-indentation and micro-indentation testing were performed at different temperatures of 25 °C, 300 °C, and 600 °C in order to study the mechanical properties of the sintered fuels as functions of grain structure and temperature. Nanocrystalline UO2 display higher hardness than microcrystalline counterpart, consistent with the Hall-Petch strengthening mechanism. Greater Young's modulus and fracture toughness are also identified for the nanocrystalline UO2, and hardness and Young's modulus decrease with temperature, suggesting better ductility of oxide fuels at high temperature and small length scale. Hyper-stoichiometric UO2 specimen displays higher hardness and fracture toughness than stoichiometric UO2, due to the impediment of the crack propagation by the oxygen interstitial atoms. These results are useful in understanding the mechanical properties of the high burn-up structure (HBS) formed in nuclear fuels during reactor operation, and also provide critical experimental data as the input for the development and validation of the MARMOT fracture model of nuclear fuels.
"Twin boundary-accelerated ferritization of austenitic stainless steels in liquid lead–bismuth eutectic" David Frazer, Peter Hosemann, Konstantina Lambrinou, Erich Stergar, Scripta Materialia Vol. 118 2016 37-40 Link
Exposure of austenitic stainless steels to liquid lead-bismuth eutectic with low concentration of dissolved oxygen typically results in selective leaching of highly-soluble alloying elements and ferritization of the dissolution-affected zone. In this work, focused ion beam, transmission electron backscatter diffraction and scanning transmission electron microscopy were utilised to elucidate early-stage aspects of the dissolution corrosion process of cold-worked austenitic stainless steels exposed to static, oxygen-poor liquid lead-bismuth eutectic at 450C for 1000 hours. It was found that deformation-induced twin boundaries in the cold-worked steel bulk provide paths of accelerated penetration of the liquid metal into the steel bulk.
NSUF Articles:
DOE Awards 33 Rapid Turnaround Experiment Research Proposals - Projects total approximately $1.5 million These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE Office of Nuclear Energy. Monday, May 14, 2018 - Calls and Awards
RTE 1st Call Awards Announced - Projects total approximately $1.4 million These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE-NE. Friday, February 8, 2019 - Calls and Awards
DOE awards 39 RTE Projects - Projects total approximately $1.3 million Thursday, February 1, 2018 - Calls and Awards
RTE 2nd Call Awards Announced - Projects total approximately $1.6 million These project awards went to principal investigators from 26 U.S. universities, eight national laboratories, two British universities, and one Canadian laboratory. Tuesday, May 14, 2019 - Calls and Awards