Kurt Terrani

Profile Information
Name
Kurt Terrani
Institution
Oak Ridge National Laboratory
h-Index
ORCID
0000-0001-5143-128X
Publications:
"A Challenge to Multivariate Statistical Analysis: Spent Nuclear Fuel" Philip Edmondson, Tyler Gerczak, Chad Parish, Kurt Terrani, Microscopy & Microanalysis Vol. 22 2016 Link
"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
"Accident Tolerant Fuel Cladding Tube Irradiations in the HFIR" Yutai Katoh, Christian Petrie, Kurt Terrani, Transactions of the American Nuclear Society Vol. 116 2017 Link
The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy (DOE) Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors (LWRs). Some of the leading candidates to replace traditional zirconium-based cladding are aluminaforming ferritic alloys (e.g., FeCrAl) and silicon carbide (SiC) composites. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative geometry in the High Flux Isotope Reactor (HFIR) under relevant LWR temperatures and in some cases under prototypic heat flux. These designs allow for post-irradiation examination (PIE) of cladding which closely resembles expected commercially viable geometries and microstructures. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost for moving advanced cladding materials closer to commercialization. This work describes the capsule designs that have been developed at ORNL, some initial results, and plans for future irradiations.
"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.
"Assembly of Rabbit Capsules for Irradiation of Pyrolytic Carbon / Silicon Carbide Diffusion Couples in the High Flux Isotope Reactor" Kory Linton, Tyler Gerczak, Kurt Terrani, Christian Petrie, OSTI.gov, Technical Report Vol. 2018 Link
Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept being considered for several advanced reactor applications and for accident-tolerant fuel for light water reactors. One of the aspects studied in the development of this advanced fuel concept is the release of specific fission products (Ag, Eu, and Sr). The silicon carbide (SiC) layer of TRISO fuel serves as the primary barrier to metallic fission products and actinides not retained in the fuel kernel. The goal of this project is to evaluate the effect of irradiation on the diffusion of these fission products in the SiC layer of the fuel. For this purpose, rabbit capsules containing small slab diffusion couple specimens have been assembled to be irradiated in the High Flux Isotope Reactor (HFIR). The diffusion couple specimens have been fabricated using similar processes and equipment as those used to make TRISO particles; the desired fission products have been implanted in the specimens using an ion accelerator. Moreover, the effect of temperature on the fission products diffusion will be studied separately by performing thermal experiments in the absence of irradiation. This report describes the irradiation experiment design concept, summarizes the irradiation test matrix, and reports on the successful assembly of two rabbit capsules that will be irradiated in the HFIR.
"Combining Transmission Kikuchi Diffraction and Scanning Transmission Electron Microscopy for Irradiated Materials Studies" Philip Edmondson, Chad Parish, Kurt Terrani, Kun Wang, Xunxiang Hu, Rachel Seibert, Yutai Katoh, Microscopy & Microanalysis Vol. 23 2017 2218-2219 Link
"Diffusion of fissile inventory in AGR-1 Triso fuel particles as a function of temperature & silver retention" Rachel Seibert, Jeff Terry, Tyler Gerczak, Kurt Terrani, John Hunn, Fred Montgomery, Charles Baldwin, Transactions of the American Nuclear Society Vol. 118 2018 1486-1487
"Evaluating the Irradiation Effects on the Elastic Properties of Miniature Monolithic SiC Tubular Specimens" Yutai Katoh, Christian Petrie, Kurt Terrani, Gyanender Singh, Takaaki Koyanagi, Journal of Nuclear Materials Vol. 499 2018 107-110 Link
The initial results of a post-irradiation examination study conducted on CVD SiC tubular specimens irradiated under a high radial heat flux are presented herein. The elastic moduli were found to decrease more than that estimated based on previous studies. The significant decreases in modulus are attributed to the cracks present in the specimens. The stresses in the specimens, calculated through finite element analyses, were found to be greater than the expected strength of irradiated specimens, indicating that the irradiation-induced stresses caused these cracks. The optical microscopy images and predicted stress distributions indicate that the cracks initiated at the inner surface and propagated outward.
"Evaluation of Irradiation-Induced Strain in SiC Tubes by a Combination of Experiment and Simulation" Takaaki Koyanagi, Yutai Katoh, Christian Petrie, Kurt Terrani, Transactions of the American Nuclear Society Vol. 118 2018 Link
"Evaluation of microstructure stability at the interfaces of Al-6061 welds fabricated using ultrasonic additive manufacturing" Niyanth Sridharan, Maxim Gussev, Chad Parish, Dieter Isheim, Davud Seidman, Kurt Terrani, Sudarsanam Babu, Materials Characterization Vol. 139 2018 249-258 Link
Ultrasonic additive manufacturing (UAM) is a solid-state additive manufacturing process that uses fundamental principles of ultrasonic welding and sequential layering of tapes to fabricate complex three-dimensional (3-D) components. One of the factors limiting the use of this technology is the poor tensile strength along the z-axis. Recent work has demonstrated the improvement of the z-axis properties after post-processing treatments. The abnormally high stability of the grains at the interface during post-weld heat treatments is, however, not yet well understood. In this work we use multiscale characterization to understand the stability of the grains during post-weld heat treatments. Aluminum alloy (6061) builds, fabricated using ultrasonic additive manufacturing, were post-weld heat treated at 180, 330 and 580 °C. The grains close to the tape interfaces are stable during post-weld heat treatments at high temperatures (i.e., 580 °C). This is in contrast to rapid grain growth that takes place in the bulk. Transmission electron microscopy and atom-probe tomography display a significant enrichment of oxygen and magnesium near the stable interfaces. Based on the detailed characterization, two mechanisms are proposed and evaluated: nonequilibrium nano-dispersed oxides impeding the grain growth due to grain boundary pinning, or grain boundary segregation of magnesium and oxygen reducing the grain boundary energy.
"Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux" Christian Deck, Yutai Katoh, Takaaki Koyanagi, Christian Petrie, Joel McDuffee, Kurt Terrani, Journal of Nuclear Materials Vol. 491 2017 94-104 Link
The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300–350 °C under a representative high heat flux (~0.66 MW/m2) during one cycle of irradiation in an un-instrumented “rabbit” capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb the expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. The success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.
"Fully Ceramic Microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Sensitivity of reactor behavior during design basis accidents to fuel properties and the potential impact of the SiC defect annealing process" Takaaki Koyanagi, Yutai Katoh, Kurt Terrani, Nicholas Brown, Nuclear Engineering and Design Vol. 345 2019 125-147 Link
"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
"Irradiation effects on thermal properties of LWR hydride fuel" Mehdi Balooch, Donald Olander, Kurt Terrani, David Carpenter, Gordon Kohse, Dennis Keiser, Mitch Meyer, Journal of Nuclear Materials Vol. 486 2017 381-390 Link
Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.
"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.
"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.
"Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up" John Hunn, Rachel Seibert, Kurt Terrani, Jeff Terry, Daniel Velazquez, Charles Baldwin, Fred Montgomery, Journal of Nuclear Materials Vol. 500 2017 316-326 Link
The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2?=?x?=?3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700?°C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.
"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.
"Micro-mechanical evaluation of SiC-SiC composite interphase properties and debond mechanisms" Mehdi Balooch, Peter Hosemann, Cameron Howard, Yutai Katoh, Takaaki Koyanagi, Yong Yang, Joey Kabel, Kurt Terrani, Composites Part B: Engineering Vol. 131 2017 173-183 Link
SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. These results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. These results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.
"Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic" Mehdi Balooch, Edgar Buck, Andy Casella, Peter Hosemann, Donald Olander, Dave Senor, Kurt Terrani, Journal of Nuclear Materials Vol. 486 2017 391-401 Link
A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.
"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
"Restructuring in high burnup UO2 studied using modern electron microscopy" Tyler Gerczak, Chad Parish, Philip Edmondson, Kurt Terrani, Journal of Nuclear Materials Vol. 509 2018 245-259 Link
Modern electron microscopy techniques were used to conduct a thorough study of an irradiated urania fuel pellet microstructure to attempt at an understanding of high burnup structure formation in this material. The fuel was irradiated at low power to high burnups in a light water reactor, proving ideal for this purpose. Examination of grain size and orientation with strict spatial selectivity across the fuel pellet radius allowed for capturing the progression of the restructuring process, from its onset to full completion. Based on this information, the polygonization mechanism was shown to be responsible for restructuring, involving formation of low-angle grain boundaries with their initiation occurring at the original high-angle grain boundaries of the as-fabricated pellet and at the gas bubble-matrix interfaces. The low-angle character of boundaries between the subdivided grains disappeared in the fully developed high burnup structure, likely due to creep deformation in the pellet.
"Structural Characterization of Fission Products in Irradiated TRISO Fuels using Transmission Kikuchi Diffraction, Transmission Electron Microscopy, and Synchrotron X-ray Absorption Spectroscopy" John Hunn, Chad Parish, Jeff Terry, Rachel Seibert, Charles Baldwin, Kurt Terrani, Microscopy & Microanalysis Vol. 23 2017 1118-1119 Link
Presentations:
" Hydride Microstructure at the Metal-Oxide Interface of a Zircaloy-4 Fuel Clad from the H. B. Robinson Nuclear Reactor" Mahmut Cinbiz, Kurt Terrani, 2017 ANS Annual Meeting [unknown]
"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
"Diffusion of fissile inventory in AGR-1 Triso fuel particles as a function of temperature & silver retention" Rachel Seibert, Jeff Terry, Tyler Gerczak, Kurt Terrani, John Hunn, Fred Montgomery, Charles Baldwin, The American Nuclear Society Annual Meeting June 17-21, (2018)
"Evaluation of Irradiation-Induced Strain in SiC Tubes by a Combination of Experiment and Simulation" Takaaki Koyanagi, Yutai Katoh, Gyanender Singh, Xunxiang Hu, Christian Petrie, Kurt Terrani, 2018 ANS Annual Meeting NFSM Poster Session June 17-21, (2018) Link
"Hydride Microstructure at the Metal-Oxide Interface of a Zircaloy-4 Fuel Clad from the H. B. Robinson Nuclear Reactor" Mahmut Cinbiz, Philip Edmondson, Kurt Terrani, American Nuclear Society June 11-15, (2017) Link
"Hydride Microstructure at the Metal-Oxide Interface of a Zircaloy-4 Fuel Clad from the H.B. Robinson Nuclear Reactor" Mahmut Cinbiz, Philip Edmondson, Kurt Terrani, 2017 ANS Annual Meeting June 11-15, (2017)
"Microstructural investigation of hydride reorientation in zirconium based spent nuclear fuel cladding" Tyler Smith, Steven Zinkle, Kurt Terrani, NUMAT 2018 October 15-18, (2018)
"Study of gamma irradiation effect on the corrosion of zirconium alloy with scanning precession electron diffraction" Li He, Adrien Couet, Samuel Armson, Michael Preuss, Kurt Terrani, Midwest Microscopy and Microanalysis Society Meeting 2019 May 22-22, (2019)
"The Effect of Photon Irradiation on the Corrosion of Zirconium Alloys" Adrien Couet, Yalong He, Kurt Terrani, Samuel Armson, Michael Preuss, Taeho Kim, Mohamed Elbakhshwan, Li He, The 19th International Symposium on Zirconium in the Nuclear Industry May 20-23, (2019)
NSUF Articles:
U.S. Department of Energy Announces FY17 CINR FOA Awards - DOE selected 14 NSUF projects DOE selected five university, four national laboratory, and five industry-led projects that will take advantage of NSUF capabilities to investigate important nuclear fuel and material applications. Wednesday, September 20, 2017 - Calls and Awards
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