Steven Zinkle

Profile Information

Name: Dr. Steven Zinkle
Institution: University of Tennessee-Knoxville
Position: Professor
Affiliation: University of Tennessee, Knoxville
h-Index
ORCID: 0000-0003-2890-6915
Expertise: Irradiation-Induced Degradation, Structural Damage

Publications:

		"Accident tolerant fuels for LWRs: A perspective" Lance Snead, Kurt Terrani, Steven Zinkle, Jess Gehin, Larry Ott, Journal of Nuclear Materials Vol. 448 2014 374–379 Link The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.
		"Advanced oxidation-resistant iron-based alloys for LWR fuel cladding" Lance Snead, Kurt Terrani, Steven Zinkle, Journal of Nuclear Materials Vol. 448 2014 420–435 Link Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (~0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.
		"Designing Radiation Resistance in Materials for Fusion Energy" Lance Snead, Steven Zinkle, Annual Review of Materials Research Vol. 44 2014 241-267 Link Proposed fusion and advanced (Generation IV) fission energy systems require high-performance materials capable of satisfactory operation up to neutron damage levels approaching 200 atomic displacements per atom with large amounts of transmutant hydrogen and helium isotopes. After a brief overview of fusion reactor concepts and radiation effects phenomena in structural and functional (nonstructural) materials, three fundamental options for designing radiation resistance are outlined: Utilize matrix phases with inherent radiation tolerance, select materials in which vacancies are immobile at the design operating temperatures, or engineer materials with high sink densities for point defect recombination. Environmental and safety considerations impose several additional restrictions on potential materials systems, but reduced-activation ferritic/martensitic steels (including thermomechanically treated and oxide dispersion–strengthened options) and silicon carbide ceramic composites emerge as robust structural materials options. Materials modeling (including computational thermodynamics) and advanced manufacturing methods are poised to exert a major impact in the next ten years.
		"Development of novel Cu-Cr-Nb-Zr alloys with the aid of computational thermodynamics" Ying Yang, Ling Wang, Lance Snead, Steven Zinkle, Materials & Design Vol. 156 2018 370-380 Link
		"Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys" Yutai Katoh, Chad Parish, Lizhen Tan, Steven Zinkle, Kinga Unocic, Sosuke Kondo, Lance Snead, David Hoelzer, Journal of Nuclear Materials Vol. 483 2017 21-34 Link We irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ~50 dpa, ~15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ~8 nm, ~1021 m-3 (CNA), and of ~3 nm, 1023 m-3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces in all alloys survived ~50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.
		"High pressure synthesis of a hexagonal close-packed phase of the high-entropy alloy CrMnFeCoNi" Cameron Tracy, Sulgiye Park, Dylan Rittman, Steven Zinkle, Hongbin Bei, Maik Lang, Rodney Ewing, Wendy Mao, Nature Communications Vol. 8 2017 Link High-entropy alloys, near-equiatomic solid solutions of five or more elements, represent a new strategy for the design of materials with properties superior to those of conventional alloys. However, their phase space remains constrained, with transition metal high-entropy alloys exhibiting only face- or body-centered cubic structures. Here, we report the high-pressure synthesis of a hexagonal close-packed phase of the prototypical high-entropy alloy CrMnFeCoNi. This martensitic transformation begins at 14?GPa and is attributed to suppression of the local magnetic moments, destabilizing the initial fcc structure. Similar to fcc-to-hcp transformations in Al and the noble gases, the transformation is sluggish, occurring over a range of >40?GPa. However, the behaviour of CrMnFeCoNi is unique in that the hcp phase is retained following decompression to ambient pressure, yielding metastable fcc-hcp mixtures. This demonstrates a means of tuning the structures and properties of high-entropy alloys in a manner not achievable by conventional processing techniques.
		"High temperature ion irradiation effects in MAX phase ceramics" Daniel Clark, Chad Parish, Maulik Patel, Steven Zinkle, Acta Materialia Vol. 105 2016 130–146 Link The family of layered carbides and nitrides known as MAX phase ceramics combine many attractive properties of both ceramics and metals due to their nanolaminate crystal structure and are promising potential candidates for application in future nuclear reactors. This investigation examines the effects of energetic heavy ion (5.8 MeV Ni) irradiations on polycrystalline samples of Ti3SiC2, Ti3AlC2, and Ti2AlC. The irradiation conditions consisted of midrange ion doses between 10 and 30 displacements per atom at temperatures of 400 and 700 °C, conditions relevant to application in future nuclear reactors and a relatively un-explored regime for this new class of materials. Following irradiation, a comprehensive analysis of radiation response properties was compiled using grazing incidence X-ray diffraction (XRD), nanoindentation, scanning electron microcopy (SEM), and transmission electron microscopy (TEM). In all cases, XRD and TEM analyses confirm the materials remain fully crystalline although the intense atomic collisions induce significant damage and disorder into the layered crystalline lattice. X-ray diffraction and nanoindentation show this damage is manifest in anisotropic swelling and hardening at all conditions and in all materials, with the aluminum based MAX phase exhibiting significantly more damage than their silicon counterpart. In all three materials there is little damage dependence on dose, suggesting saturation of radiation damage at levels below 10 displacements per atom, and significantly less retained damage at higher temperatures, suggesting radiation defect annealing. SEM surface analysis showed significant grain boundary cracking and loss of damage tolerance properties in the aluminum-based MAX phase irradiated at 400 °C, but not in the silicon counterpart. TEM analysis of select samples suggest that interstitials are highly mobile while vacancies are immobile and that all three materials are in the so-called point defect swelling regime between 400 and 700 °C. All results are consistent with previous work involving traditional and MAX phase ceramics. Results show the aluminum MAX phases are not fit for application near 400 °C and that the silicon MAX phase is more damage tolerant at 400–700 °C.
		"Influence of mechanical alloying and extrusion conditions on the microstructure and tensile properties of Low-Cr ODS FeCrAl alloys" Caleb Massey, Sebastien Dryepondt, Philip Edmondson, Kurt Terrani, Steven Zinkle, Journal of Nuclear Materials Vol. 512 2018 227-238 Link
		"Materials challenges in nuclear energy" Gary Was, Steven Zinkle, Acta Materialia Vol. 61 2013 735–758 Link Nuclear power currently provides about 13% of electrical power worldwide, and has emerged as a reliable baseload source of electricity. A number of materials challenges must be successfully resolved for nuclear energy to continue to make further improvements in reliability, safety and economics. The operating environment for materials in current and proposed future nuclear energy systems is summarized, along with a description of materials used for the main operating components. Materials challenges associated with power uprates and extensions of the operating lifetimes of reactors are described. The three major materials challenges for the current and next generation of water-cooled fission reactors are centered on two structural materials aging degradation issues (corrosion and stress corrosion cracking of structural materials and neutron-induced embrittlement of reactor pressure vessels), along with improved fuel system reliability and accident tolerance issues. The major corrosion and stress corrosion cracking degradation mechanisms for light-water reactors are reviewed. The materials degradation issues for the Zr alloy-clad UO2 fuel system currently utilized in the majority of commercial nuclear power plants are discussed for normal and off-normal operating conditions. Looking to proposed future (Generation IV) fission and fusion energy systems, there are five key bulk radiation degradation effects (low temperature radiation hardening and embrittlement; radiation-induced and -modified solute segregation and phase stability; irradiation creep; void swelling; and high-temperature helium embrittlement) and a multitude of corrosion and stress corrosion cracking effects (including irradiation-assisted phenomena) that can have a major impact on the performance of structural materials.
		"Phase instabilities in austenitic steels during particle bombardment at high and low dose rates" Samara Levine, C. Pareige, Philip Edmondson, Gary Was, Steven Zinkle, Arunodaya Bhattacharya, Materials & Design Vol. 217 2022 Link
		"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
		"Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments" Cheng Sun, Lin Shao, Steven Zinkle, Todd Allen, Haiyan Wang, Xinghang Zhang, Scientific Reports Vol. 5 2015 Link

Presentations:

	"Fundamental Aspects of Radiation Effects in Materials" Steven Zinkle, ANS Annual Meeting 2018 June 18-22, (2018)
	"Microstructural investigation of hydride reorientation in zirconium based spent nuclear fuel cladding" Tyler Smith, Steven Zinkle, Kurt Terrani, NUMAT 2018 October 15-18, (2018)

NSUF Articles:

U.S. DOE Nuclear Science User Facilities Awards 35 Rapid Turnaround Experiment Research Proposals - Awards total approximately $1.3 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, September 20, 2017 - Calls and Awards
DOE Awards 33 Rapid Turnaround Experiment Research Proposals - Projects total 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. Monday, June 18, 2018 - Calls and Awards
DOE awards 39 RTE Projects - Projects total approximately $1.3 million Thursday, February 1, 2018 - Calls and Awards

Accomplishments

NSUF Profile

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Authored an NSUF-supported publication

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Presented an NSUF-supported publication

NSUF Submitter

Submitted an RTE Proposal to NSUF

Accomplished Researcher

Awarded 3+ RTE Proposals

Top 5% Collaborator

Top 5% of all RTE Proposal collaborations

NSUF Reviewer

Reviewed an RTE Proposal

NSUF Supported Research

Investigation of neutron irradiation effects on Zr based bulk metallic glass (BMG) via advanced in situ mechanical testing and microstructural analysis. - FY 2018 RTE 3rd Call, #18-1469

Irradiation-induced precipitation/segregation in dual-phase Al0.3CoCrFeNi alloy - FY 2018 RTE 1st Call, #18-1207

Irradiation responses of ultrastrong nano precipitation martensite steel - FY 2017 RTE 3rd Call, #17-1029

NSUF Research Collaborations

Mechanical Properties Characterization and Atom Probe Tomography Preparation of BOR60 Neutron Irradiated T91 Steel Samples for Surrogate Ion-Irradiation Comparison - FY 2025 RTE 1st Call, #25-5225

The Role of Dislocation Cell Walls on Cavity Nucleation in Additively Manufactured 316H Steel - FY 2024 RTE 1st Call, #24-4838

Stability of VN, TaN, and TaC MX-type Precipitates in Ferritic Steels under Neutron Radiation - FY 2023 RTE 2nd Call, #23-4636

Electron tomography study of dislocation loops and precipitates in ion irradiated Fe-Cr alloys - FY 2023 RTE 1st Call, #23-4597

Examining microstructures and mechanical properties of neutron and ion irradiated T91, HT9 and 800H alloys - FY 2022 RTE 1st Call, #22-4456

Imaging of Irradiation Effects in Tantalum Alloys for Fast-Spectrum Self-Powered Neutron Detectors - FY 2021 RTE 1st Call, #21-4331

Phase stability of a’ precipitates in pre-aged Fe-25Cr model binary alloys after ion irradiation - FY 2020 RTE 2nd Call, #20-4192

Sink strength dependent coherency loss of precipitates during in-situ ion irradiation of fcc-structured model binary alloys - FY 2019 RTE 3rd Call, #19-2900

The influence of second phase precipitates on hydride reorientation in spent nuclear fuel cladding - FY 2019 RTE 3rd Call, #19-2845

Nano-precipitate Stability and a'-Precipitation in ODS and Wrought FeCrAl Alloys - FY 2019 RTE 2nd Call, #19-1747

Bubble formation of in-situ He-implanted 14YWT and CNA advanced nanostructured ferritic alloys - FY 2019 RTE 1st Call, #19-1634

Ion Irradiation for High Fidelity Simulation of High Dose Neutron Irradiation - FY 2019 RTE 1st Call, #19-1541

Microstructural investigation of hydride reorientation in zirconium based spent nuclear fuel cladding - FY 2018 RTE 3rd Call, #18-1542

Nano-precipitate Response to Neutron Irradiation in Model ODS FeCrAl Alloy 125YF - FY 2017 RTE 2nd Call, #17-961

Improving understanding of defect evolution in neutron-irradiated MAX phases - FY 2016 RTE 3rd Call, #16-713