NSUF - Nuclear Science User Facilities

Lance Snead

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.
"Advanced synchrotron characterization techniques for fusion materials science" David Sprouster, J Trelewicz, Lance Snead, Daniel Morrall, Takaaki Koyanagi, X Hu, Chad Parish, Lizhen Tan, Yutai Katoh, Brian Wirth, Journal of Nuclear Materials Vol. 543 2020 152574 Link
"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
"Dimensional stability and anisotropy of SiC and SiC-based composites in transition swelling regime" Yutai Katoh, Takaaki Koyanagi, Lance Snead, Joel McDuffee, Ken Yueh, Journal of Nuclear Materials Vol. 499 2017 471-479 Link
Swelling, or volumetric expansion, is an inevitable consequence of the atomic displacement damage in crystalline silicon carbide (SiC) caused by energetic neutron irradiation. Because of its steep temperature and dose dependence, understanding swelling is essential for designing SiC-based components for nuclear applications. In this study, swelling behaviors of monolithic CVD SiC and nuclear grade SiC fiber – SiC matrix (SiC/SiC) composites were accurately determined, supported by the irradiation temperature determination for individual samples, following neutron irradiation within the lower transition swelling temperature regime. Slightly anisotropic swelling behaviors were found for the SiC/SiC samples and attributed primarily to the combined effects of the pre-existing microcracking, fiber architecture, and specimen dimension. A semi-empirical model of SiC swelling was calibrated and presented. Finally, implications of the refined model to selected swelling-related issues for SiC-based nuclar reactor components are discussed.
"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.
"Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa" Caen Ang, Yutai Katoh, Lance Snead, Kurt Terrani, Journal of Nuclear Materials Vol. 499 2018 242-247 Link
Five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700 °C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Thermal resistivity increase as a result of irradiation was ~20% higher when compared to CVD-SiC.
"Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum" Lauren Garrison, Yutai Katoh, Takaaki Koyanagi, Lance Snead, Kiran Kumar, Taehyun Hwang, Xunxiang Hu, Journal of Nuclear Materials Vol. 490 2017 66-74 Link
Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.
"Stability of MX-type strengthening nanoprecipitates in ferritic steels under thermal aging, stress and ion irradiation" Yutai Katoh, Lizhen Tan, Lance Snead, Thak Sang Byun, Acta Materialia Vol. 71 2014 11–19 Link
The stability of MX-type precipitates is critical to retain mechanical properties of both reduced activation ferritic–martensitic (RAFM) and conventional FM steels at elevated temperatures above 500 C. The stability of TaC, TaN and VN nanoprecipitates under thermal aging (600 and 700 C), creep (600 C) and ion irradiation (Fe ion, 500 C) conditions was systematically studied in this work. The statistical particle evolution in density and size was characterized using transmission electron microscopy. Nanoprecipitate stability under the studied conditions manifested differently through either dissolution, reprecipitation, growth or fragmentation, with TaC exhibiting the greatest stability followed by VN and TaN in sequence. Nanoprecipitate evolution phenomena and mechanisms and the apparent disagreement of this interpretation with published literature on the subject are discussed. These findings not only help understanding the degradation mechanisms of RAFM and conventional FM steels at elevated temperatures and under stress and irradiation, but should also prove beneficial to the development of advanced RAFM steels.
"Stability of the Strengthening Nanoprecipitates in Reduced Activation Ferritic Steels Under Fe2+ Ion Irradiation" Yutai Katoh, Lance Snead, Lizhen Tan, Journal of Nuclear Materials Vol. 445 2014 104-110 Link
The stability of MX-type precipitates is critical to retain mechanical properties of both reduced activation ferritic–martensitic (RAFM) and conventional FM steels at elevated temperatures. Radiation resistance of TaC, TaN, and VN nanoprecipitates irradiated up to ~49 dpa at 500 °C using Fe2+ is investigated in this work. Transmission electron microscopy (TEM) utilized in standard and scanning mode (STEM) reveals the non-stoichiometric nature of the nanoprecipitates. Irradiation did not alter their crystalline nature. The radiation resistance of these precipitates, in an order of reduced resistance, is TaC, VN, and TaN. Particle dissolution, growth, and reprecipitation were the modes of irradiation-induced instability. Irradiation also facilitated formation of Fe2W type Laves phase limited to the VN and TaN bearing alloys. This result suggests that nitrogen level should be controlled to a minimal level in alloys to gain greater radiation resistance of the MX-type precipitates at similar temperatures as well as postpone the formation and subsequent coarsening of Laves phase.
"Stored Energy Release in Neutron Irradiated Silicon Carbide" Lance Snead, Journal of Nuclear Materials Vol. 514 2019 181-188 Link
"X-ray characterization of anisotropic defect formation in SiC under irradiation with applied stress" David Sprouster, Takaaki Koyanagi, Lance Snead, Yutai Katoh, Scripta Materialia Vol. 197 2021 113785 Link
"X-ray characterization of anisotropic defect formation in SiC under irradiation with applied stress" David Sprouster, Takaaki Koyanagi, Lance Snead, Yutai Katoh, Scripta Materialia Vol. 197 2021 113785 Link
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
"Stability of MX Nanoprecipitates in Ferritic Steels Under Thermal, Stress, and Ion Irradiation" Yutai Katoh, Lance Snead, Lizhen Tan, Gary Was, 16th International Conference on Fusion Reactor Materials (ICFRM-16) October 20-26, (2013)
"Topological and atomic investigation of nuclear graphite using multi-scale x-ray scattering" David Sprouster, Lance Snead, Boris Khaykovich, Yutai Katoh, Anne Campbell, 45th International Conference and Expo on Advanced Ceramics and Composites (ICACC2021) February 8-11, (2021) Link
NSUF Articles:
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
NSUF awards 28 Rapid Turnaround Experiment proposals - Approximately $1.74M has been awarded. The new call closes June 28. Thursday, June 1, 2023 - Calls and Awards

About Us

The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear energy user facility. Through peer-reviewed proposal processes, the NSUF provides researchers access to neutron, ion, and gamma irradiations, post-irradiation examination and beamline capabilities at Idaho National Laboratory and a diverse mix of university, national laboratory and industry partner institutions.

Privacy and Accessibility · Vulnerability Disclosure Program