Samuel A. Briggs

Profile Information
Dr. Samuel A. Briggs
Oregon State University
Assistant Professor

Assistant Professor in the School of Nuclear Science & Engineering at Oregon State University. Research interests include enabling next-generation nuclear reactor technologies through materials design and development. Expertise in microstructural characterization and microscopy of radiation damage in materials, with a focus on advanced steels and structural metals.

Graduated from the University of Wisconsin-Madison with Ph.D. in Nuclear Engineering & Engineering Physics while studying under Dr. Todd Allen/Kumar Sridharan. Dissertation research emphasizes characterization of precipitates in FeCrAl alloys using correlative microscopy techniques, including atom probe tomography, small-angle neutron scattering, and analytical transmission electron microscopy techniques. Previously worked at the In-situ Ion Irradiation Transmission Electron Microscopy (I3TEM) facility at Sandia National Laboratories studying microstructural evolution of materials exposed to extreme environments.

Graduated from Oregon State University with a B.S. in Nuclear Engineering and minors in Mathematics and Chemistry in 2011.

Atom Probe Tomography, chromium precipitates, Ferritic Martensitic Steels
"A combined APT and SANS investigation of a' phase precipitation in neutron-irradiated model FeCrAl alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, Kurt Terrani, Samuel A. Briggs, Kenneth Littrell, Yukinori Yamamoto, Richard Howard, Charles Daily, Acta Materialia Vol. 129 2017 217-228 Link
"Complementary Techniques for Quantification of a' Phase Precipitation in Neutron-Irradiated Fe-Cr-Al Model Alloys" Samuel A. Briggs, Philip Edmondson, Kevin Field, Kumar Sridharan, Yukinori Yamamoto, Kenneth Littrell, Charles Daily, Microscopy & Microanalysis Vol. 22 2016 1470-1471 Link
"Correlative Microscopy of Neutron-Irradiated Materials" Samuel A. Briggs, Kevin Field, Kumar Sridharan, Advanced Materials & Processes Vol. 174 2016 16-21 Link
Development of new, radiation-tolerant materials that maintain the structural integrity and safety margins over the course of a nuclear power reactor’s service life requires the ability to predict degradation phenomena.
"Dependencies of a' embrittlement in neutron-irradiated model Fe-Cr-Al alloys" Samuel A. Briggs, Philip Edmondson, Kevin Field, Kumar Sridharan, ANS Transactions Vol. 114 2016 1046-1047 Link
"Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys" Philip Edmondson, Kevin Field, Jack Haley, Steve Roberts, Kumar Sridharan, Samuel A. Briggs, Sergio Lozano-Perez, Acta Materialia Vol. 136 2017 390-401 Link
Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions in-situ with transmission electron microscopy to a dose of 2.5 displacements per atom (dpa) at 320 °C. In all cases, the microstructural damage consisted of dislocation loops with ½<111> and <100> Burgers vectors. The proportion of ½<111> dislocation loops varied from ~50% in the Fe-10%Cr-5%Al model alloy and the Fe-18Cr%-3%Al model alloy to a peak of ~80% in the model Fe-15%Cr-4.5%Al alloy. The dislocation loop volume density increased with dose for all alloys and showed signs of approaching an upper limit. The total loop populations at 2.5 dpa had a slight (and possibly insignificant) decline as the chromium content was increased from 10 to 15 wt %, but the Fe-18%Cr-3%Al alloy had a dislocation loop population ~50% smaller than the other model alloys. The largest dislocation loops in each alloy had image sizes of close to 20 nm in the micrographs, and the median diameters for all alloys ranged from 6 to 8 nm. Nature analysis by the inside-outside method indicated most dislocation loops were interstitial type.
"Effect of friction stir welding and self-ion irradiation on dispersoid evolution in oxide dispersion strengthened steel MA956 up to 25 dpa" Elizabeth Getto, Brad Baker, B. Tobie, Samuel A. Briggs, Khalid Hattar, K. Knipling, Journal of Nuclear Materials Vol. 515 2018 407-419 Link
"Irradiation-enhanced a' precipitation in model FeCrAl alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, Samuel A. Briggs, Yukinori Yamamoto, Richard Howard, Kurt Terrani, Scripta Materialia Vol. 116 2016 112-116 Link
Model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) have been neutron irradiated at ~ 320 to damage levels of ~ 7 displacements per atom (dpa) to investigate the compositional influence on the formation of irradiation-induced Cr-rich a' precipitates using atom probe tomography. In all alloys, significant number densities of these precipitates were observed. Cluster compositions were investigated and it was found that the average cluster Cr content ranged between 51.1 and 62.5 at.% dependent on initial compositions. This is significantly lower than the Cr-content of a' in binary FeCr alloys. Significant partitioning of the Al from the a' precipitates was also observed.
"Observations of defect structure evolution in proton and Ni ion irradiated Ni-Cr binary alloys" Samuel A. Briggs, Khalid Hattar, Janne Pakarinen, Kumar Sridharan, Mitra Taheri, Christopher Barr, Mahmood Mamivand, Dane Morgan, Journal of Nuclear Materials Vol. Volume 479 2016 48-58 Link
Two binary Ni-Cr model alloys with 5 wt% Cr and 18 wt% Cr were irradiated using 2 MeV protons at 400 and 500 °C and 20 MeV Ni4+ ions at 500 °C to investigate microstructural evolution as a function of composition, irradiation temperature, and irradiating ion species. Transmission electron microscopy (TEM) was applied to study irradiation-induced void and faulted Frank loops microstructures. Irradiations at 500 °C were shown to generate decreased densities of larger defects, likely due to increased barriers to defect nucleation as compared to 400 °C irradiations. Heavy ion irradiation resulted in a larger density of smaller voids when compared to proton irradiations, indicating in-cascade clustering of point defects. Cluster dynamics simulations were in good agreement with the experimental findings, suggesting that increases in Cr content lead to an increase in interstitial binding energy, leading to higher densities of smaller dislocation loops in the Ni-18Cr alloy as compared to the Ni-5Cr alloy.
"Effect of Friction Stir Welding on Microstructure Evolution on in situ and ex situ Self-Ion Irradiated MA956" Elizabeth Getto, Samuel A. Briggs, Khalid Hattar, Brad Baker, TMS 2018 March 11-15, (2018)
NSUF Articles:
U.S. DOE Nuclear Science User Facilities Awards 30 Rapid Turnaround Experiment Research Proposals - Awards total nearly $1.2 million The U.S. Department of Energy (DOE) Nuclear Science User Facilities (NSUF) has selected 30 new Rapid Turnaround Experiment (RTE) projects, totaling up to approximately $1.2 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. Wednesday, April 26, 2017 - Calls and Awards
DOE Awards 31 RTE Proposals, Opens FY-20 1st Call - Projects total $1.1 million; Next proposals due 10/31 Awards will go to 22 principal investigators from universities, six from national laboratories, and three from foreign universities. Tuesday, September 17, 2019 - Calls and Awards, Announcement