Gary Was

Profile Information
Professor Gary Was
University of Michigan
University of Michigan

Professor Was received his ScD from MIT in 1980.  He is the Walter J. Weber, Jr. Professor of Sustainable Energy, Environmental and Earth Systems Engineering, and holds appointments in Nuclear Engineering and Radiological Sciences, and Materials Science and Engineering at the University of Michigan.  He has held positions as Director of the Michigan Memorial Phoenix Energy Institute, Associate Dean of the College of Engineering and Chair of the Nuclear Engineering and Radiological Sciences Department.  Professor Was’ research is focused on materials for advanced nuclear energy systems and radiation materials science, including environmental effects on materials, radiation effects, ion beam surface modification of materials and nuclear fuels. Most recently his group has led the development of proton irradiation as a technique for emulating neutron irradiation effects in reactor structural materials and has conducted some of the first stress corrosion cracking experiments of austenitic and ferritic alloys in supercritical water. 

During his tenure at the University of Michigan, Professor Was has graduated 39 Ph.D. students, created several graduate level courses dealing primarily with irradiation effects on materials, ion beam modification of materials and nuclear fuels.  He served as chair of the Nuclear Engineering Department Heads Organization and co-authored the first ASEE report on Manpower in the Nuclear Industry.  He has helped to organize more than a dozen technical symposia and is a member of the American Society for Engineering Education, Materials Research Society, ASM International, The Minerals, Metals and Materials Society, the NACE International, Sigma Xi and Tau Beta Pi. He is director of three major laboratories at the University of Michigan: the Michigan Ion Beam Laboratory for Surface Modification and Analysis, the High Temperature Corrosion Laboratory, and the Irradiated Materials Testing Laboratory. Professor Was received the Presidential Young Investigator award from NSF in 1985 and in 1994 he received the Excellence in Research Award from the College of Engineering.  He was awarded the Champion H. Matthews Award from TMS, the Outstanding Achievement Award and Special Achievement Award by the Materials Science and Technology Division of the American Nuclear Society, the 2008 Henry Marion Howe Medal from ASM, the Lee Hsun Award from the Chinese Academy of Sciences, the Mishima Award from ANS and the Glenn Murphy Award from ASEE.  He is a Fellow of The Minerals, Materials and Metals Society (TMS), the Materials Research Society, ASM International, NACE International and the American Nuclear Society. Professor Was has published over 270 technical articles in referred, archival journals, presented over 400 conference papers, and delivered over 200 invited talks and seminars, published a graduate level textbook on Radiation Materials Science in 2007 and a second edition in 2016. He serves as Editor-in-Chief of the Journal of Nuclear Materials.


Alloys, Corrosion, DAMAGE, Ion-Irradiation, Irradiated Microstructure, Metallurgy, Stress Corrosion Cracking
"A methodology for customizing implantation profiles of light ions using a single thin foil energy degrader" Stephen Taller, Fabian Naab, Gary Was, Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms Vol. 478 2020 274-283 Link
A method was developed to quantify the spatial distribution and implantation depth of energy-degraded light ions with a thin foil rotating energy degrader for use during multiple ion beam irradiation. The methodology covers three physical phenomena: ions passing through the thin foil, ions travelling through the vacuum to the target, and ion implantation into the target, and accounts for the distribution of ions both in depth and in plane. The processes of energy straggling and scattering were calculated using SRIM. The effects of raster-scanning, and the geometry of the system were implemented in scripts handling the SRIM output files. Elastic backscattering (EBS) using 2.38 MeV H+ protons was used to measure the helium depth profiles after implantation with and without thin foil energy degradation. Defect analysis with transmission electron microscopy confirmed the implantation profiles measured with EBS and calculated with SRIM.
"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
"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.
"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.
"Evolution dependence of vanadium nitride nanoprecipitates on directionality of ion irradiation" Bong Goo Kim, Lizhen Tan, Gary Was, Journal of Nuclear Materials Vol. 495 2017 425-430 Link
The influence of the directionality of Fe2+ ion irradiation on the evolution of vanadium nitride platelet–shaped nanoprecipitates at 500 °C was investigated in a ferritic alloy using transmission electron microscopy. When the ion-irradiation direction was approximately aligned with the initial particle length, particles grew longer and sectioned into shorter lengths at higher doses, resulting in increased particle densities. As ion-irradiation direction deviated from particle-length direction, some particles sectioned lengthwise and then dissolved, resulting in decreased particle densities. Surviving particles were transformed into parallelograms with a different orientation relationship with the matrix. Nanoprecipitate evolution dependence on beam-nanoprecipitate orientation is a process that may be different from reactor irradiation.
"High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation" Gary Was,, Technical Report Vol. 2018 Link
"Insights into the stress corrosion cracking of solution annealed alloy 690 in simulated pressurized water reactor primary water under dynamic straining" wenjun kuang, Miao Song, Gary Was, Acta Materialia Vol. 151 2018 321-333
The intergranular attack near both stagnant and active stress corrosion crack (SCC) tips of solution annealed alloy 690 were characterized after constant extension rate tensile tests in simulated pressurized water reactor (PWR) primary water containing 18 cc H2/kg H2O. In both cracks, an intergranular oxide, composed of NiO and Cr2O3, was observed beyond the crack tip with an adjacent Cr depleted grain boundary migration zone. The stagnant crack has a compact Cr2O3 covering the crack tip and the adjacent grain boundary migration zone is deep and free of oxidation, while the active crack has porous penetrative oxide extending into the migration zone. The ability to form a compact Cr2O3 film at the crack tip is important to the IGSCC resistance and may be dictated by the efficiency of Cr diffusion up the migrated grain boundary. When a compact Cr2O3 film does not form at the crack tip, the Cr depleted grain boundary migration zone is subject to penetrative oxidation via the inward diffusion of oxygen, making the zone susceptible to crack propagation. Thus, even a high Cr content cannot guarantee the immunity of a Ni base alloy to SCC during dynamic straining.
"Irradiation assisted stress corrosion cracking of commercial and advanced alloys for light water reactor core internals" mi wang, Miao Song, Calvin Lear, Gary Was, Journal of Nuclear Materials Vol. 515 2019 52-70 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.
"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
"Multiple ion beam irradiation for the study of radiation damage in materials" Stephen Taller, David Woodley, Elizabeth Getto, Anthony Monterrosa, Zhijie Jiao, Ovidiu Toader, Fabian Naab, Thomas Kubley, Shyam Dwaraknath, Gary Was, Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms Vol. 412 2017 1-10 Link
The effects of transmutation produced helium and hydrogen must be included in ion irradiation experiments to emulate the microstructure of reactor irradiated materials. Descriptions of the criteria and systems necessary for multiple ion beam irradiation are presented and validated experimentally. A calculation methodology was developed to quantify the spatial distribution, implantation depth and amount of energy-degraded and implanted light ions when using a thin foil rotating energy degrader during multi-ion beam irradiation. A dual ion implantation using 1.34 MeV Fe+ ions and energy-degraded D+ ions was conducted on single crystal silicon to benchmark the dosimetry used for multi-ion beam irradiations. Secondary Ion Mass Spectroscopy (SIMS) analysis showed good agreement with calculations of the peak implantation depth and the total amount of iron and deuterium implanted. The results establish the capability to quantify the ion fluence from both heavy ion beams and energy-degraded light ion beams for the purpose of using multi-ion beam irradiations to emulate reactor irradiated microstructures.
"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
"Radiation tolerance of commercial and advanced alloys for core internals: a comprehensive microstructural characterization" Miao Song, Calvin Lear, Chad Parish, Mi Wang, Gary Was, Journal of Nuclear Materials Vol. 510 2018 396-413 Link
Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo-type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.
"Resolution of the carbon contamination problem in ion irradiation experiments" Stephen Taller, Gary Was, Zhijie Jiao, Anthony Monterrosa, David Woodley, Dylan Jennings, Thomas Kubley, Fabian Naab, Ovidiu Toader, Ethan Uberseder, Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms Vol. 412 2017 58-65 Link
The widely experienced problem of carbon uptake in samples during ion irradiation was systematically investigated to identify the source of carbon and to develop mitigation techniques. Possible sources of carbon included carbon ions or neutrals incorporated into the ion beam, hydrocarbons in the vacuum system, and carbon species on the sample and fixture surfaces. Secondary ion mass spectrometry, atom probe tomography, elastic backscattering spectrometry, and principally, nuclear reaction analysis, were used to profile carbon in a variety of substrates prior to and following irradiation with Fe2+ ions at high temperature. Ion irradiation of high purity Si and Ni, and also of alloy 800H coated with a thin film of alumina eliminated the ion beam as the source of carbon. Hydrocarbons in the vacuum and/or on the sample and fixtures was the source of the carbon that became incorporated into the samples during irradiation. Plasma cleaning of the sample and sample stage, and incorporation of a liquid nitrogen cold trap both individually and especially in combination, completely eliminated the uptake of carbon during heavy ion irradiation. While less convenient, coating the sample with a thin film of alumina was also effective in eliminating carbon incorporation.
"Solute segregation and precipitation across damage rates in dual-ion–irradiated T91 steel" Stephen Taller, Valentin Pauly, Zhijie Jiao, Rigel Hanbury, Gary Was, JNM Vol. 563 2022 Link
"Technical Aspects of Delivering Simultaneous Dual and Triple Ion Beams to a Target at the Michigan Ion Beam Laboratory " Ovidiu Toader, Gary Was, Fabian Naab, Ethan Uberseder, Thomas Kubley, Stephen Taller, Physics Procedia Vol. 90 2017 385-390 Link
The Michigan Ion Beam Laboratory (MIBL) at the University of Michigan in Ann Arbor, Michigan, USA, plays a significant role in supporting the mission of the U.S. DOE Office of Nuclear Energy. MIBL is a charter laboratory of the NSUF (National Scientific User Facility – US DoE) and hosts users worldwide. The laboratory has evolved from a single accelerator laboratory to a highly versatile facility with three accelerators (3 MV Tandem, a 400 kV Ion Implanter and a 1.7 MV Tandem), seven beam lines and five target chambers that together, provide unique capabilities to capture the extreme environment experienced by materials in reactor systems. This capability now includes simultaneous multiple (dual, triple) ion irradiations, an irradiation accelerated corrosion cell, and soon, in-situ dual beam irradiation in a transmission electron microscope (TEM) for the study of radiation damage coupled with injection of transmutation elements. The two beam lines that will connect to the 300 kV FEI Tecnai G2 F30 microscope are expected to be operational by the end of 2017. Multiple simultaneous ion beam experiments involving light and heavy ions are already in progress. This paper will outline the current equipment and will focus on the new capability of running dual and triple ion beam experiments.
"Understanding bubble and void nucleation in dual ion irradiated T91 steel using single parameter experiments" Stephen Taller, Gary Was, Acta Materialia Vol. 198 2020 47-60 Link
Ferritic-martensitic steels are attractive candidates for structural materials in next generation nuclear reactor systems due to their resistance to radiation induced swelling. Cavity and dislocation loop evolution was characterized in dual ion irradiated T91 steel in three separate irradiation campaigns examining single parameter dependencies of temperature, helium co-injection rate, and damage rate. Irradiations resulted in bimodal cavity size distributions across nearly all ranges of experimental parameters. It was determined that irradiation temperature and helium co-injection rate are stronger influences on bubble stability and the transition from bubbles to voids than is the irradiation damage rate. At low helium injection rates all helium is in vacancy clusters that evolve into bubbles or voids. At high helium injection rates, bubbles become saturated with helium resulting in accumulation of helium at other traps such as dislocation loops. At intermediate levels of He that should aid in the nucleation of bubbles and enhance swelling, the high density of sinks in the F-M microstructure suppresses bubble nucleation and therefore, the onset of swelling. At high enough temperatures, helium is only in bubbles as other strong helium traps, such as dislocation loops, did not form. The mechanism of bubble to void transition was found to shift from being driven by the accumulation of helium to the critical bubble at low damage rates to being driven by spontaneous formation by stochastic vacancy fluctuation at high damage rates.
"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]
"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)
"Technical Aspects of Delivering Simultaneous Dual and Triple Ion Beams to a Target at the Michigan Ion Beam Laboratory" Stephen Taller, Ovidiu Toader, Gary Was, Conference on the Application of Accelerators in Research and Industry, CAARI 2016 October 30-6, (2016)
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 Facility Highlight: University of Michigan - UM offers researchers access to its Ion Beam Lab, materials complex The University of Michigan's two NSUF partner facilities offers NSUF users access to extensive capabilities to study the effects of radiation as well as conduct high temperature mechanical property, corrosion, and stress corrosion cracking experiments on neutron irradiated materials in an aqueous environment to characterize the fracture surfaces after failure. Wednesday, October 9, 2019 - Facility Highlight, Newsletter
CINR Awards Announced - Eight projects were selected Projects will take advantage of NSUF capabilities to investigate important nuclear fuel and material applications. Thursday, June 27, 2019 - Calls and Awards
Michigan Center for Materials Characterization joins University of Michigan NSUF partner facilities - NSUF welcomes a new partner facility The Michigan Center for Materials Characterization, also known as (MC)2, has been approved to become a University of Michigan Nuclear Science User Facilities (NSUF) partner facility. Monday, August 2, 2021 - Newsletter