Tyler Gerczak

ORCID: 0000-0001-9967-3579
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About
Contact & Profiles
Research Areas
  • Nuclear Materials and Properties
  • Nuclear reactor physics and engineering
  • Graphite, nuclear technology, radiation studies
  • Nuclear and radioactivity studies
  • Fusion materials and technologies
  • Advanced ceramic materials synthesis
  • Nuclear Physics and Applications
  • Radioactive element chemistry and processing
  • Nuclear materials and radiation effects
  • Silicon Carbide Semiconductor Technologies
  • Ion-surface interactions and analysis
  • Aluminum Alloys Composites Properties
  • Semiconductor materials and devices
  • Nuclear Engineering Thermal-Hydraulics
  • Silicon and Solar Cell Technologies
  • Advanced materials and composites
  • Radiation Effects in Electronics
  • Risk and Safety Analysis
  • Rocket and propulsion systems research
  • Material Properties and Applications
  • Granular flow and fluidized beds
  • High Temperature Alloys and Creep
  • Engineering Applied Research
  • Catalytic Processes in Materials Science
  • Mineral Processing and Grinding

Oak Ridge National Laboratory
2016-2025

Office of Scientific and Technical Information
2021-2023

Government of the United States of America
2023

National Technical Information Service
2021

University of Wisconsin–Madison
2008-2015

10.1016/j.jnucmat.2015.03.027 article EN publisher-specific-oa Journal of Nuclear Materials 2015-03-23

The nuclear security community has long been interested in the identification and quantification of material signatures to understand a material’s provenance, use, ultimate application. New forensics methods intended for non-traditional or advanced fuel applications require irradiation experiments demonstrate viability validity. Integral irradiations have historically required significant costs timelines design, irradiate, characterize. This paper describes how recently developed testbed can...

10.3389/fnuen.2023.1123134 article EN cc-by Frontiers in Nuclear Engineering 2023-02-03

10.1016/j.jnucmat.2020.152736 article EN publisher-specific-oa Journal of Nuclear Materials 2020-12-16

Abstract Surrogate tristructural‐isotropic (TRISO)‐coated fuel particles were oxidized in 0.2 kPa O 2 at 1200–1600°C to examine the behavior of SiC layer and understand mechanisms. The thickness microstructure resultant SiO layers analyzed using scanning electron microscopy, focused ion beam, transmission microscopy. majority surface comprised smooth, amorphous with a constant indicative passive oxidation. apparent activation energy for oxide growth was 188 ± 8 kJ/mol consistent across all...

10.1111/jace.19032 article EN publisher-specific-oa Journal of the American Ceramic Society 2023-02-03

10.1016/j.jnucmat.2020.152185 article EN publisher-specific-oa Journal of Nuclear Materials 2020-04-30
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