Kevin Field

Profile Information
Name
Dr. Kevin Field
Institution
University of Michigan
Position
Associate Professor
h-Index
19
ORCID
0000-0002-3105-076X
Expertise
Irradiated Concrete, Microstructural Analysis
Publications:
"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
"ACCELERATED IRRADIATION TESTING OF MINIATURE NUCLEAR FUEL AND CLADDING SPECIMENS" Christian Petrie, Takaaki Koyanagi, Richard Howard, Kevin Field, Joseph Burns, Kurt Terrani, OSTI.govI Vol. 2018 Link
"Application of NSUF Capabilities Towards Understanding the Emulation of High Dose Neutron Irradiations with Ion Beams" Kevin Field, Stephen Taller, Christopher Ulmer, Zhijie Jiao, Tarik Saleh, Arthur Motta, Gary Was, Transactions of the American Nuclear Society Vol. 116 2017 Link
"Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys" Chad Parish, Kevin Field, Alicia Certain, Janelle Wharry, Journal of Materials Research Vol. 30 2015 1275-1289 Link
This paper provides an overview of advanced scanning transmission electron microscopy (STEM) techniques used for characterization of irradiated BCC Fe-based alloys. Advanced STEM methods provide the high-resolution imaging and chemical analysis necessary to understand the irradiation response of BCC Fe-based alloys. The use of STEM with energy dispersive x-ray spectroscopy (EDX) for measurement of radiation-induced segregation (RIS) is described, with an illustrated example of RIS in proton- and self-ion irradiated T91. Aberration-corrected STEM-EDX for nanocluster/nanoparticle imaging and chemical analysis is also discussed, and examples are provided from ion-irradiated oxide dispersion strengthened (ODS) alloys. Finally, STEM techniques for void, cavity, and dislocation loop imaging are described, with examples from various BCC Fe-based alloys.
"Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature" Todd Allen, Zhijie Jiao, Djamel Kaoumi, Janelle Wharry, cem topbasi, Aaron Kohnert, Leland Barnard, Alicia Certain, Kevin Field, Gary Was, Dane Morgan, Arthur Motta, Brian Wirth, Yong Yang, Journal of Materials Research Vol. 30 2015 1246-1274 Link
Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. The studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.
"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
"Direct Experimental Evidence for Differing Reactivity Alterations of Minerals following Irradiation: The Case of Calcite and Quartz" Isabella Pignatelli, Yann Le Pape, Kevin Field, Scientific Reports Vol. 6 2016 20155
Concrete, used in the construction of nuclear power plants (NPPs), may be exposed to radiation emanating from the reactor core. Until recently, concrete has been assumed immune to radiation exposure. Direct evidence acquired on Ar+-ion irradiated calcite and quartz indicates, on the contrary, that, such minerals, which constitute aggregates in concrete, may be significantly altered by irradiation. More specifically, while quartz undergoes disordering of its atomic structure resulting in a near complete lack of periodicity, calcite only experiences random rotations, and distortions of its carbonate groups. As a result, irradiated quartz shows a reduction in density of around 15%, and an increase in chemical reactivity, described by its dissolution rate, similar to a glassy silica. Calcite however, shows little change in dissolution rate - although its density noted to reduce by ˜9%. These differences are correlated with the nature of bonds in these minerals, i.e., being dominantly ionic or covalent, and the rigidity of the mineral’s atomic network that is characterized by the number of topological constraints (nc) that are imposed on the atoms in the network. The outcomes have major implications on the durability of concrete structural elements formed with calcite or quartz bearing aggregates in nuclear power plants.
"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 exposure environment on surface decomposition of SiC–silver ion implantation diffusion couples" Todd Allen, Kevin Field, Tyler Gerczak, Guiqiu Zheng, Journal of Nuclear Materials Vol. 456 2015 281-286 Link
SiC is a promising material for nuclear applications and is a critical component in the construction of tristructural isotropic (TRISO) fuel. A primary issue with TRISO fuel operation is the observed release of 110mAg from intact fuel particles. The release of Ag has prompted research efforts to directly measure the transport mechanism of Ag in bulk SiC. Recent experimental efforts have focused primarily on Ag ion implantation designs. The effect of the thermal exposure system on the ion implantation surface has been investigated. Results indicate the utilization of a mated sample geometry and the establishment of a static thermal exposure environment is critical to maintaining an intact surface for diffusion analysis. The nature of the implantation surface and its potential role in Ag diffusion analysis are discussed.
"Effect of exposure environment on surface decomposition of SiC–silver ion implantation diffusion couples" Tyler Gerczak, Guiqiu Zheng, Kevin Field, Todd Allen, Journal of Nuclear Materials Vol. 456 2015 281-286
"Emulation of fast reactor irradiated T91 using dual ion beam irradiation" Stephen Taller, Zhijie Jiao, Kevin Field, Gary Was, Journal of Nuclear Materials Vol. 527 2019 Link
Dual ion irradiations using 5 MeV defocused Fe2+ ions and co-injected He2+ ions were conducted on a ferritic-martensitic steel alloy, T91, in the temperature range of 406 °C–570 °C over a damage range of 14.6–35 dpa followed by characterization of the microstructure using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). Dislocation loops were observed to increase in diameter and decrease in density with temperature until only network dislocations were observed at the highest temperatures of 520 °C and 570 °C. Swelling exhibited the expected bell-shaped trend with temperature following the number density of cavities, peaking at 460 °C and with a bimodal size distribution except at 520 °C and 570 °C. Nickel- and silicon-rich clusters formed under dual ion irradiations near the surface at all but the highest temperatures of 520 °C and 570 °C. Very little Cr and Si segregation was observed at lath boundaries while Ni enriched at all temperatures examined. Segregation of Cr and Ni appeared to saturate by 17 dpa, while Si enriched up to 35 dpa. The dislocation and cavity microstructures of dual ion irradiated T91 and T91 irradiated in the BOR-60 fast reactor matched extremely well using a temperature shift of +60–70 °C. However, segregation to grain boundaries and formation of nickel-silicon rich clusters were minimal in the dual ion irradiated T91 and less than that in T91 irradiated in the BOR-60 fast reactor.
"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.
"Microscopy of Plasma-Materials Interactions in Tungsten for Fusion Power" Kevin Field, Yutai Katoh, Chad Parish, Microscopy & Microanalysis Vol. 22 2016 1462-1463 Link
"Microstructural evolution of neutron-irradiated T91 and NF616 to ~4.3 dpa at 469 °C" Kevin Field, Bong Goo Kim, Lizhen Tan, Yong Yang, Sean Gray, Meimei Li, Journal of Nuclear Materials Vol. 493 2017 12-20 Link
Ferritic-martensitic steels such as T91 and NF616 are candidate materials for several nuclear applications. This study evaluates radiation resistance of T91 and NF616 by examining their microstructural evolutions and hardening after the samples were irradiated in the Advanced Test Reactor to ∼4.3 displacements per atom (dpa) at an as-run temperature of 469 °C. In general, this irradiation did not result in significant difference in the radiation-induced microstructures between the two steels. Compared to NF616, T91 had a higher number density of dislocation loops and a lower level of radiation-induced segregation, together with a slightly higher radiation-hardening. Unlike dislocation loops developed in both steels, radiation-induced cavities were only observed in T91 but remained small with sub-10 nm sizes. Other than the relatively stable M23C6, a new phase (likely Sigma phase) was observed in T91 and radiation-enhanced MX → Z phase transformation was identified in NF616. Laves phase was not observed in the samples.
"Microstructure evolution of T91 irradiated in the BOR60 fast reactor" Zhijie Jiao, Stephen Taller, Kevin Field, G. Yeli, M.P. Moody, Gary Was, Journal of Nuclear Materials Vol. 504 2018 122-134 Link
"Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: Effect of temperature" Ce Zheng, Elaina Reese, Kevin Field, Tian Liu, Emmanuelle Marquis, Stuart Maloy, Djamel Kaoumi, Journal of Nuclear Materials Vol. 528 2019 Link
The ferritic/martensitic steel HT9 was irradiated in the BOR-60 reactor at 650, 690 and 730 K (377, 417 and 457 °C) to doses between ∼14.6–18.6 displacements per atom (dpa). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the influence of irradiation temperature on microstructure evolution. Mn/Ni/Si-rich (G-phase) and Cr-rich (αʹ) precipitates were observed within martensitic laths and at various defect sinks at 650 and 690 K (377 and 417 °C). For both G-phase and αʹ precipitates, the number density decreased while the size increased with increasing temperature. At 730 K (457 °C), within martensitic laths, a very low density of large G-phase precipitates nucleating presumably on dislocation lines was observed. No αʹ precipitates were observed at this temperature. Both a <100> and a/2 <111> type dislocation loops were observed, with the a <100> type being the predominant type at 650 and 690 K (377 and 417 °C). On the contrary, very few dislocation loops were observed at 730 K (457 °C), and the microstructure was dominated by a/2 <111> type dislocation lines (i.e., dislocation network) at this temperature. Small cavities (diameter < 2 nm) were observed at all three temperatures, whereas large cavities (diameter > 2 nm) were observed only at 690 K (417 °C), resulting in a bimodal cavity size distribution at 690 K (417 °C) and a unimodal size distribution at 650 and 730 K (377 and 457 °C). The highest swelling (%) was observed at 690 K (417 °C), indicating that the peak of swelling happens between 650 and 730 K (377 and 457 °C).
"Post irradiation examination of nanoprecipitate stability and α′ precipitation in an oxide dispersion strengthened Fe-12Cr-5Al alloy" Caleb Massey, Philip Edmondson, Kevin Field, David Hoelzer, Kurt Terrani, Steven Zinkle, Scripta Materialia Vol. 162 2018 94-98 Link
"Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9wt.% Cr model ferritic/martensitic steel" Todd Allen, Heather Chichester, Kevin Field, Brandon Miller, Kumar Sridharan, Journal of Nuclear Materials Vol. 445 2013 143-148 Link
Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.
Presentations:
"Advancements in FeCrAl Alloys for Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors" Kevin Field, Maxim Gussev, Lance Snead, Kurt Terrani, 2016 ANS Annual Meeting June 12-16, (2016) Link
"Application of NSUF Capabilities Towards Understanding the Emulation of High Dose Neutron Irradiations with Ion Beams" Kevin Field, Zhijie Jiao, Tarik Saleh, Stephen Taller, Gary Was, 2017 ANS Annual Meeting [unknown]
"Complementary techniques for quantification of a' phase precipitation in neutron-irradiated Fe-Cr-Al model alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, Microscopy & Microanalysis 2016 July 24-28, (2016) Link
"Dependencies of a' Embrittlement in Neutron-Irradiated Model Fe-Cr-Al Alloys" Philip Edmondson, Kevin Field, Kumar Sridharan, 2016 ANS Annual Meeting June 12-16, (2016) Link
"Irradiation and PIE of ATF cladding materials in HFIR" Kevin Field, Yutai Katoh, Takaaki Koyanagi, Christian Petrie, Advanced Fuels Campaign Integration Meeting (2017) March 1-2, (2017)
"Nanoscale analysis of neutron irradiated ODS 14YWT ferritic alloy" Maria A Auger, David Hoelzer, Kevin Field, European MRS 2019 May 27-31, (2019)