Maria Okuniewski

Profile Information
Assoc. Prof. Maria Okuniewski
Purdue University
Associate Professor
Purdue University

Dr. Okuniewski is an associate professor in the School of Materials Engineering at Purdue University. She also holds a courtesy appointment with the School of Nuclear Engineering. Previously, she spent nearly eight years at Idaho National Laboratory as a research and development scientist and engineer. Dr. Okuniewski was also an adjunct faculty member at Idaho State University in the Department of Nuclear Engineering and Health Physics. She obtained her Ph.D. and M.S. degrees in nuclear engineering at the University of Illinois at Urbana-Champaign.


The over-arching goals of her research focus on the nexus between microstructure, fabrication, properties, and performance of nuclear fuels and materials. This research aims to connect phenomena that span multiple spatial (atomistic to mesoscale) and temporal (picoseconds to seconds) scales. Her research utilizes both experimental and modeling techniques in a complementary fashion. She has been instrumental in developing new techniques and expanding the capabilities of existing techniques to apply to nuclear fuels and materials such as synchrotron X-ray diffraction and tomography, positron annihilation spectroscopy, nanoindentation, neutron diffraction, and focused ion beam/scanning electron microscopy applications.


During her career she has authored or co-authored 73 peer-reviewed journal articles or proceedings, 63 invited talks, 26 technical reports, and more than 100 conference presentation abstracts. She has received the American Nuclear Society’s Mary Jane Oestmann Professional Women’s Award, the DOE Fuel Cycle Research and Development Excellence Award, and the Oak Ridge Associated Universities Ralph E. Powe Junior Faculty Enhancement Award.

Advanced Fuels, Diffraction, Diffraction Contrast, Electron Microscopy, Energy Dispersive X-Ray Spectroscopy (EDXS), Nuclear Fuel, Positron Annihilation Spectroscopy, Synchrotron, Synchrotron Scattering, Tomography, Transmission Electron Microscopy (TEM), X-Ray Computed Tomography
"3-D reconstruction and microstructural characterization of neutron-irradiated U-10Zr fuel using FIB-SEM serial sectioning" Nicole Rodriguez Perez, Jonova Thomas, Daniel Murray, Maria Okuniewski, MRS Advances Vol. 8 2023 14-20 Link
Focused ion beam-scanning electron microscopy serial sectioning was applied to characterize the three-dimensional (3-D) porosity and phase regions of a neutron-irradiated U-10 wt% Zr fuel. The specimen was removed from an intermediate radial region of a fuel pin irradiated to 5.7 at.% burn-up. Backscattered electron imaging and energy-dispersive spectroscopy were performed on each serial section, allowing for the characterization of microstructural morphology and composition. Porosity size followed a lognormal distribution, ranging from 1.46 × 10–4 to 25.58 μm3 with a total porosity volume fraction of 13.02%. Distinctive microstructural regions were identified by composition and porosity: (1) a Zr-rich region with an average composition of 28.4 wt% Zr and a local porosity fraction of 6.88%, and (2) a U-rich region with an average composition of 97.0 wt% U and a local porosity fraction of 14.11%; subdivided into U-rich—high porosity (16.68%) and U-rich—low porosity (8.04%) regions. The detailed 3-D compositional and porosity regions can improve nuclear fuel performance codes.
"An atomistic study of defect energetics and diffusion with respect to composition and temperature in U and U-Mo alloys" Gyuchul Park, Benjamin Beeler, Maria Okuniewski, Journal of Nuclear Materials Vol. 552 2021 Link
"An in-situ neutron diffraction study of crystallographic evolution and thermal expansion coefficients in U-22.5at.%Zr during annealing" Walter Williams, Maria Okuniewski, Sven Vogel, Jianzhong Zhang, JOM Vol. 72 2020 2042–2050 Link
A uranium-22.5at.% zirconium (U-10 wt.% Zr) ingot was manufactured by traditional arc-casting. The sample was characterized with the High-Pressure-Preferred Orientation (HIPPO) time-of-flight neutron diffractometer with in-situ heating reaching 900°C. The experiment investigated phase transitions, lattice parameter ratios, and linear coefficients of thermal expansion. Contradictory to commonly referenced phase diagrams, this work has shown that the β-U+γ’’-UZr dual phase region is not present in U-22.5at.% Zr. Rather, only two phase transformations were observed upon cooling. These transformations occurred at 655°C and 600°C and corresponded to γ-UZrα-U+γ’’-UZrα-U+δ-UZr2, respectively. The linear coefficients of thermal expansion were measured for α-U, δ-UZr2, and γ-UZr. The thermal expansion of α-U was shown to be anisotropic in nature with a significant lattice contraction in the 010b direction. To a lesser extent, a similar anisotropy was observed in δ-UZr2¬ with a contraction in the 001c direction.
"Atom probe study of irradiation-enhanced a' precipitation in neutron-irradiated Fe–Cr model alloys" Wei-Ying Chen, Jian Gan, Stuart Maloy, Kun Mo, Maria Okuniewski, James Stubbins, Yinbin Miao, Yaqiao Wu, Carolyn Tomchik, Journal of Nuclear Materials Vol. 462 2015 242-249 Link
Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe–Cr model alloys (10–16 at.%) irradiated at 300 and 450 °C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0–1.3 nm were observed. The precipitate density varied significantly from 1.1 × 1023 to 2.7 × 1024 1/m3, depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe–10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.
"Calculation of the displacement energy of alph and gamma uranium" Yongfeng Zhang, Maria Okuniewski, Chaitanya Deo, Journal of Nuclear Materials Vol. 508 2018 181--194 Link
"Impact of Fabrication Techniques on U-10wt%Mo Fuel Microstructure Irradiated to Low Burnup" Sukanya Majumder, Gyuchul Park, Daniel Murray, Benjamin Beeler, Maria Okuniewski, Gyuchul Park, ANS Annual Meeting Vol. 2023 383-385
"Kinetics of the Migration and Clustering of Extrinsic Gas in bcc Metals" Maria Okuniewski, James Stubbins, Chaitanya Deo, Srinivasan Srivilliputhur, Michael Baskes, Stuart Maloy, M. R. James, ASTM Special Technical Publication Vol. 1492 2008 177-189 Link
We study the mechanisms by which gas atoms such as helium and hydrogen diffuse and interact with other defects in bcc metals and investigate the effect of these mechanisms on the nucleation of embryonic gas bubbles. Large quantities of helium and hydrogen are produced due to spallation and transmutation in structural materials in fusion and accelerator-driven reactors. The long time evolution of the extrinsic gas atoms and their accumulation at vacancies is studied using a kinetic Monte Carlo algorithm that is parameterized by the migration energies of the point defect entities. First-order reaction kinetics are observed when gas clusters with vacancies. If gas-gas clustering is allowed, mixed-order diffusion limited kinetics are observed. When dissociation of gas from clusters is allowed, gas-vacancy clusters survive to steady state while gas-gas clusters dissolve. We obtain cluster size distributions and reaction rate constants that can be used to quantify microstructural evolution of the irradiated metal.
"Mechanical characteristics of SiC coating layer in TRISO fuel particles" Thak Sang Byun, David Frazer, Peter Hosemann, John Hunn, Maria Okuniewski, Kurt Terrani, Gokul Vasudevamurthy, J. N. Matros, Brian Jolly, Journal of Nuclear Materials Vol. 442 2013 133-142 Link
Tristructural isotropic (TRISO) particles are considered as advanced fuel forms for a variety of fission platforms. While these fuel structures have been tested and deployed in reactors, the mechanical properties of these structures as a function of production parameters need to be investigated in order to ensure their reliability during service. Nanoindentation techniques, indentation crack testing, and half sphere crush testing were utilized in order to evaluate the integrity of the SiC coating layer that is meant to prevent fission product release in the coated particle fuel form. The results are complimented by scanning electron microscopy (SEM) of the grain structure that is subject to change as a function of processing parameters and can alter the mechanical properties such as hardness, elastic modulus, fracture toughness and fracture strength. Through utilization of these advanced techniques, subtle differences in mechanical properties that can be important for in-pile fuel performance can be distinguished and optimized in iteration with processing science of coated fuel particle production.
"Microstructural analysis of a Zr rich specimen from the central radial region of a neutron irradiated U-10Zr fuel" Nicole Rodriguez Perez, Jonova Thomas, Maria Okuniewski, Transactions of the American Nuclear Society Vol. 128 2023 376-379 Link
"Nano-mechanical Property Assessment of a Neutron Irradiated HT-9 Steel Cladding and a Fuel-cladding Chemical Interaction Region of a Uranium-10wt.% Zirconium Nuclear Fuel" Maria Okuniewski, MRS Advances Vol. 6 2021 1037-1042 Link
"Neutron irradiation effects in Fe and Fe-Cr at 300°C" Wei-Ying Chen, Yinbin Miao, Jian Gan, Maria Okuniewski, Stuart Maloy, James Stubbins, Acta Materialia Vol. 111 2016 407-416 Link
Fe and Fe-Cr (Cr = 10-16 at.%) specimens were neutron-irradiated at 300 °C to 0.01, 0.1 and 1 dpa. The TEM observations indicated that the Cr significantly reduced the mobility of dislocation loops and suppressed vacancy clustering, leading to distinct damage microstructures between Fe and Fe-Cr. Irradiation-induced dislocation loops in Fe were heterogeneously observed in the vicinity of grown-in dislocations, whereas the loop distribution observed in Fe-Cr is much more uniform. Voids were observed in the irradiated Fe samples, but not in irradiated Fe-Cr samples. Increasing Cr content in Fe-Cr results in a higher density, and a smaller size of irradiation-induced dislocation loops. Orowan mechanism was used to correlate the observed microstructure and hardening, which showed that the hardening in Fe-Cr can be attributed to the formation of dislocation loops and α′ precipitates.
"Thermotransport in γ(bcc) U-Zr Alloys: A Phase-Field Model Study" Rashmi Mohanty, J. Bush, Maria Okuniewski, Yongho Sohn, Journal of Nuclear Materials Vol. 414 2011 211-216 Link
Atomic transport in the presence of a temperature gradient, commonly known as thermotransport or the thermomigration phenomenon, was simulated for U–Zr alloys using a phase-field model derived from irreversible thermodynamics. The free energy of the U–Zr system, a necessary ingredient for the phase-field-model, was directly incorporated from the available thermodynamic database. Kinetic parameters such as atomic mobility and heat of transport terms were obtained from experimental values reported in the literature. The model was applied to a single-phase (bcc-γ phase) alloy and to a diffusion couple consisting of two single-phase (bcc-γ phase) alloys of different compositions, both subjected to a constant temperature gradient. Constituent redistribution in the absence and presence of a compositional gradient was examined. An enrichment of Zr with a corresponding depletion of U was observed at the hot end of the initially homogeneous single-phase alloy. A similar atomic transport behavior was observed in the diffusion couple, where the magnitude and direction of the final composition gradient was dictated by the combined influence of atomic mobility and heat of transport terms.
"Assessment of irradiation damage and chemical interactions in neutron irradiated U-10Zr fuel and HT9 cladding with high-energy X-rays" Jonova Thomas, Sri Tapaswi Nori, Alejandro Figueroa, Ran Ren, Peter Kenesei, Jon D. Almer, Jason Harp, Maria Okuniewski, MRS Fall Meeting November 26-1, (2017)
"Assessment of Radiation Damage and Microstructural Changes in Neutron Irradiated U-10Zr Fuels with High Energy X-Rays" Jonova Thomas, Sri Tapaswi Nori, Alejandro Figueroa, Maria Okuniewski, Peter Kenesei, Jun-Sang Park, Jon Almer, Jason Harp, ANS Conference on Embedded Topical Nuclear Fuels & Structural Materials for Next Generation Nuclear Reactors June 17-21, (2018)
"Characterization of microstructural evolution in metallic fuels occurring over multiple length scales" Maria Okuniewski, Jonova Thomas, Sri Tapaswi Nori, Gyuchul Park, Walter Williams, NUMAT 2018 October 15-18, (2018)
"Comparison of SXRD and Microstructure of Electron-beam Welded RPV Steels" Jasmyne Emerson, Grayson Nemets, Elliot Marrero, MEHMET TOPSAKAL, Simerjeet Gill, Janelle Wharry, Maria Okuniewski, NSLS-II, CFN, & LBMS Users’ Meeting April 24-27, (2023)
"High-Resolution TEM Characterization of Neutron-Irradiated U-10Mo Fuel in the Low Temperature and Low Burnup Regime" Sukanya Majumder, Gyuchul Park, Tiankai Yao, Kaustubh Bawane, Cameron Howard, Kourtney Wright, Laura Hawkins, Brandon Miller, Jonova Thomas, Benjamin Beeler, Maria Okuniewski, Materials in Nuclear Energy Systems (MiNES) December 11-14, (2023)
"Impact of Fabrication Techniques on U-10wt%Mo Fuel Microstructure Irradiated to Low Burnup" Sukanya Majumder, Gyuchul Park, Daniel Murray, Benjamin Beeler, Maria Okuniewski, ANS Annual Meeting June 11-14, (2023)
"Localized Composition of Constituent Redistribution Regions in a Neutron Irradiated U-10wt.%Zr Fuel" Nicole Rodriguez Perez, Jonova Thomas, Maria Okuniewski, Materials in Nuclear Energy Systems (MiNES 2023) December 10-14, (2023)
"Microstructural analysis of a Zr rich specimen from the central radial region of a neutron irradiated U-10Zr fuel" Nicole Rodriguez Perez, Jonova Thomas, Maria Okuniewski, 2023 American Nuclear Society annual meeting (ANS 2023) June 11-14, (2023)
"Microstructural characterization of radiation damage in neutron irradiated U-10wt%Zr fuels" Jonova Thomas, Sri Tapaswi Nori, Alejandro Figueroa, Maria Okuniewski, Peter Kenesei, Jun Sang Park, Jason Harp, NuMat: The Nuclear materials Conference October 14-18, (2018)
"Synchrotron X-ray Diffraction Characterization of the Phase Transformation Behaviors Induced by Electron Beam Welding in SA508 Reactor Pressure Vessel Steels" Jasmyne Emerson, Grayson Nemets, Elliot Marrero, MEHMET TOPSAKAL, Simerjeet Gill, Janelle Wharry, Maria Okuniewski, Materials in Nuclear Energy Systems (MiNES 2023) December 11-14, (2023)