| Year | Reference | Link |
|---|---|---|
| FY2026 | Adkins, C., Goodson, M., Meyer, M., Wright, K., Hisle, E., Yao, T., Capriotti, L., Aitkaliyeva, A., and Pavlov, T. (2026). Modern measurement of local radial thermal conductivity of irradiated he-bonded U-10 wt. % Zr annular fuel: A cross-correlative study with electron probe microanalysis techniques. Journal of Nuclear Materials, 624, 156487. | Publication |
| FY2026 | Armstrong, R., Pacheco Duarte, J., Corradini, M., and Jensen, C. (2026). Review of ECCS acceptance criteria and experimental basis evolution toward fuel fragmentation, relocation, and dispersal studies. Nuclear Science and Engineering, 1 -18. | Publication |
| FY2026 | Cai, L., Xu, F., Xian, M., Tang, Y., Sun, S., and Stempien, J. D. (2026). RU-net for automatic characterization of TRISO fuel cross sections. Materials Characterization, 232, 116019. | Publication |
| FY2026 | Gorton, J. P., Parker, T. D., Mulligan, P. L., Petrie, C. M., McDuffee, J. L., and Howard, R. H. (2026). Leveraging the High Flux Isotope Reactor for nuclear fuel development: A review of experiments, facilities, and capabilities. Nuclear Engineering and Design, 454, 114915. | Publication |
| FY2026 | Hirschhorn, J., Jaradat, M., Sweet, R., Woolstenhulme, N., Demkowicz, P., Reger, D., Balestra, P., and Strydom, G. (2026). Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT. Nuclear Engineering and Design, 449. | Publication |
| FY2026 | Moussaoui, M. A., Anderson, K. S., Yoo, J., and Woolstenhulme, N. E. (2025). Device for steam cladding oxidation testing at TREAT. Nuclear Engineering and Design, 445, 114441. | Publication |
| FY2026 | Papesch, C., Wright, K., Squires, L., Trowbridge, T., Root, J., Cowan, B., Porter, D., Hartmann, T., and Hilton, B. (2025, October). Material characterization report of EBR II Mark IV fuel element (INL/RPT-25-88203, Rev. 1). Idaho National Laboratory. | Publication |
| FY2026 | Saleh, T.A., Eftink, B.P., Romero, T.J. et al. Tensile properties of the neutron irradiated HT-9 ACO-3 duct. Sci Rep 15, 43236 (2025). | Publication |
| FY2026 | Schulthess, J. L., Howard, C. B., Mauseth, T. J., and Bearcroft, C. (2026). Micro-tensile testing of neutron-irradiated Al/Zr and Zr/U-Mo diffusion bonds. Journal of Nuclear Materials, 619, 156296. | Publication |
| FY2026 | Schulthess, J. L., Williams, W. J., Jue, J.-F., Nielson, R. B., Merickel, J., and Fu, Y. (2025). A critical review of the history of fabricating monolithic U-Mo fuel plates. Journal of Nuclear Materials, 156220. | Publication |
| FY2026 | Schulthess, J., Pido, S., Noei, N. et al. Understanding Oxide -Metal Interactions During Hot Isostatic Pressing to Diffusion Bond Aluminum Alloy 6061 Plates. Metall Mater Trans A (2025). | Publication |
| FY2025 | Bawane, K.K., Yang, G., Yao, T., Xu, F., Xu, P., Gonderman, S. and J. Gazza (2025). Microstructure Analysis of Silicon Carbide Cladding Using 4D-STEM. Paper presented at M&M 2025. | |
| FY2025 | Beausoleil, G. L., Curnutt, B., Moorehead, M. and Bascom, A. (2025). Multi-principal element alloys for fast reactor cladding applications. Nuclear Engineering and Technology, 57(4), 103303. | Publication |
| FY2025 | Bermudez, S., Erdogan, F., Davis, V., Rojas, J.V. and R.V. Umretiya (2025). Effect of nickel on the FeCrAl alloy oxidation resistance in steam environment at high temperature (1000° C). Nuclear Materials and Energy : 101972. | Publication |
| FY2025 | Cakmak, E., Cinbiz, M. N., Arregui-Mena, J. D., Deck, C. and T. Koyanagi (2025). Damage Progression and Failure of SiC/SiC Composite Tubes under Hard-Contact Radial Expansion. Composites Part B: Engineering, 112869. | Publication |
| FY2025 | Cappia F., Colldeweih, A., Frazer, D., Hansen, R., Petersen, P., Stockwell, J., Anderson, S., Charbeneau, J., Kamerman, D. (2024). Effect of Metal Contaminants on Cr Coating Performance after Irradiation in the Advanced Test Reactor. TopFuel 2024 Conference Proceeding. Grenoble, France. | |
| FY2025 | Capps, N., Yan, Y., Harp, J., Ridley, M. and R. Salko Jr. (2024). Recent High Burnup LOCA Testing at Oak Ridge National Laboratory (ORNL/SPR-2024/3544). Oak Ridge National Laboratory, Oak Ridge, TN. | Publication |
| FY2025 | Capriotti, L., Di Lemma, F., Salvato, D., Xu, F., Tang, Y., Paaren, K.M., Swearingen, A.L., Jensen, C.B., Wang, Y. and D.L. Porter (2025). An Integrated Approach to Examining Fuel-Cladding Chemical Interaction in HT9/U-10Zr Metallic Fast Reactor Fuels: Coupling Machine Learning with Electron Microscopy and Local Mechanical Properties Analysis. Journal of Nuclear Materials, p.156092. | Publication |
| FY2025 | Carvajal, J. (2025). In-Rod Sensor System Irradiation Test Results with Segmented Fuel Assembly, 14th International Topical Meeting on Nuclear Plant Instrumentation, Control and Human-Machine Interface Technologies (NPICandHMIT 2025), Chicago. | |
| FY2025 | Cervenka, P., Seshadri A., Sevecek M., Cvrcek L. and K. Shirvan (2024). Development of PVD Cr-(Nb) coated fuel cladding with enhanced accident tolerance, Presented at the Nuclear Materials Conference. | |
| FY2025 | Chavez, R. (2025). Fluid Dynamics and Thermal Effects of Flow Over a Sphere at High Pressures and Graphitic Dust Behavior in Square Channels, PhD Dissertation, Texas A & M University. | |
| FY2025 | Chavez, R., Anand, N.K. and Hassan, Y. and S. Girimaji (2024) Flow Over a Sphere at Elevated Pressures: An Analysis of the Near-Wake Using Spectral Proper Orthogonal Decomposition Physics of Fluids, November 2024, Vol. 36, 115155 (1-17) Issue 11, selected as Editor âs Pick. | Publication |
| FY2025 | Chavez, R., Anand, N.K. and Y. Hassan (2025). High-Pressure Experimental Analysis of Thermal Effects on Near-Wake Turbulence and Energy Distribution of Flow over a Heated Sphere, Paper presented at the NURETH 21 Annual Meeting. | |
| FY2025 | Chen, D., Burns, J., Wright, K. E., Salvato, D., Yao, T. and L. Capriotti (2025). Transmission electron microscopy characterization of fuel cladding chemical interaction between minor actinides bearing U-Pu-Zr fuel and AIM1 cladding. Journal of Nuclear Materials, 607, 155667. | Publication |
| FY2025 | Chikhalikar, A.S., Qu, H., Abouelella, H., Nagothi, B., Rajendran, R., Roy, I., Umretiya, R., Hoffman, A. and R. Rebak, . Effect of Al content on steam oxidation behavior for ferritic Fe-21Cr-xAl alloys. Journal of Nuclear Materials 598 (2024): 155179. | Publication |
| FY2025 | Chuirazzi, W., Bush, J., Gross, B., Bryant, M., Clark, K., Cook, M., Burtenshaw, J., Price, J., Morankar, S., Blattner, M., Landon, R., Galloway, K., Stanger, J., Stamos, R., Duke, J., Watt, C. and J. Stempien (2025). Strategy to safely enable X-ray computed tomography examination of highly radioactive tristructural isotropic nuclear fuel. Nuclear Engineering and Technology, 57(10), 103726. | Publication |
| FY2025 | Colldeweih A., Kamerman, D., Matos, M., Bawane, K., J. Stockwell, J., A. Pradhan, A., Hansen, R., Cappia, F. and D. Lutz (2024) Corrosion of Neutron Irradiated FeCrAl in the ATR Water Loop. TopFuel 2024 Conference Proceeding. Grenoble, France. | |
| FY2025 | Colldeweih A., P. Petersen, M. Matos, J. Stockwell, R. Hansen, D. Kamerman, D. Lutz and F. Cappia (2025). Post irradiation examinations of FeCrAl cladding in PWR conditions. Journal of Nuclear Materials Vol. 603, 155402 | Publication |
| FY2025 | Dabney, T., Sasidhar, K.N., Willing, E., Eftink, B., Li, N., Maier, B., Walters, J. and K. Sridharan (2025). Performance of Cold Spray Cr Coatings on Zr-alloy Fuel Cladding , Symposium on Solid-state Processing and Manufacturing for Extreme Environment Applications: Integrating Insights and Innovations, The Metallurgical Society (TMS) Annual Conference, Las Vegas, NV, March 2025. | |
| FY2025 | Dabney, T., Sasidhar, K.N., Willing, E., Lukas, C., Quillin, K., Yeon, H. and K. Sridharan (2025). Microstructural Evolution in Ion Irradiated Cold Spray Cr Coated Zr-alloy , Journal of Nuclear Materials, vol. 606, 155652 | Publication |
| FY2025 | Dhulipala, S. L. N., Simon, P.-C. A., Demkowicz, P. A., Hirschhorn, J. A. and S. R. Novascone (2025). Unpacking model inadequacy: The quantification of silver release from TRISO fuel by considering empirical and mechanistic approaches. Journal of Nuclear Materials, 610, 155795. | Publication |
| FY2025 | Dolley, E. J., Zhang, W., Zorn, G., Sand, T. and R.B. Rebak (2024) Enhanced mechanical properties and wear resistance of FeCrAl alloys at~ 300 C and Higher temperatures. JOM 76, no. 8 (2024): 4123-4130. | Publication |
| FY2025 | Downey C.M., Oldham N., Fleming A., Chapman D., Mata Cruz A. and K. Ellis (2024). Design of a First-of-a-kind Instrumented Advanced Test Reactor Irradiation Capsule Experiment for in Situ Thermal Conductivity Measurements of Metallic Fuel. Prog Nucl Energy.175:105325. | Publication |
| FY2025 | Espersen, J. I., Garrison, B. E., Cervenka, P., Seshadri, A., Linton, K., Shirvan, K., Capps N.A and N.R. Brown (2025). The impact of chromium coatings on Zircaloy cladding deformation behavior under reactivity-initiated accident-like mechanical loading conditions. Journal of Nuclear Materials, 155910. | Publication |
| FY2025 | Hansen R., Colldeweih, A., Petersen, P., Stockwell, J., Charboneau, J., Albuquerque, L., Baird, K., Kamerman, D. and F. Cappia (2024) Examinations of Cr-Coated M5 Cladding Irradiated at the INL Advanced Test Reactor. TopFuel 2024 Conference Proceeding. Grenoble, France. | |
| FY2025 | Harp, J., Yan, Y., Morris, R., Baldwin, C., Jones, M. and N. Capps (2024). Development of Fission Gas Release Cabilities to Study High Burnup Commercial Fuel Performance under Loss of Coolant Accident Conditions. Proc. TopFuel 2024, Grenoble, France. | |
| FY2025 | Hawkes, G., Pham, B. and C. Otani (2024). Thermal Model of the AGR-5/6/7 Experiment with Offset Gas Gaps. Nuclear Science and Engineering, 1-26. | Publication |
| FY2025 | Hirschhorn J.A., Aagesen L.K., Jiang C. and G.L. Beausoleil (2025). Development and preliminary validation of a mechanistic multiscale model for fuel-cladding chemical interaction in metallic nuclear fuels. Nucl Eng Des ;432:113811. | Publication |
| FY2025 | Joyce, L., Umretiya, R.V., Qu, H., Shang, Z. and Y. Xie (2025). Oxidation behaviour of PM-C26M FeCrAl alloy in low-temperature steam 400-900 ° C. Nuclear Materials and Energy : 101953. | Publication |
| FY2025 | Joyce, L., Wang, P., Umretiya, R.V., Hoffman, A. and Y. Xie (2024). Oxide Layers in Ni-doped FeCrAl Alloy in 320 ° C Radioactive Hydrogenated Water. Journal of Nuclear Materials 593: 154987. | Publication |
| FY2025 | Jung, W., Dunbar, C., Jo, J.Y., Sridharan, K. and H. Yeom (2025). Thermal Response and Mechanical Integrity of High Temperature Cr-coated Zr cladding under Multiple Quench Tests, Symposium on Microstructural, Mechanical, and Chemical Behavior of Solid Nuclear Fuel and Fuel-Cladding Interface II, The Metallurgical Society (TMS) Annual Conference, Las Vegas, NV, March 2025. | |
| FY2025 | Kancharla R.R, Chuirazzi W.C, Kane J.J et al. (2025). X-ray computed tomography of deconsolidated TRISO particles from the AGR-5/6/7 irradiation experiment capsule 1 compact. J Nucl Mater. ; 607:155704. doi:10.1016/j.jnucmat.2025.155704. | Publication |
| FY2025 | Karlsson, T. Y. (2025). Fuel Qualification: Near-Term Activities and Needs for Molten Salt Fuels. Presented at the EPRI Advanced Reactor Workshop. | |
| FY2025 | Kilby S.M, Marshall M.A, Choe D.O. et al. (2024). Design of Mini-Plate-1 Irradiation Test for Qualification of High-Density, Low-Enriched U-10Mo Monolithic Fuel. JOM. | Publication |
| FY2025 | Kosmidou, M., Broussard, A., Lian, J. and E. Kardoulaki (2025). Filling of data gaps for the development of ceramic fuels, pp. 23.Materials in Nuclear Energy Systems (MiNES) 2025 Conference. | |
| FY2025 | Li, N., Xie, D., Kim, H., Dabney, T., Eftink, B., Sridharan, K., Graening, T., Nelson, A., Fensin, S.and S. Maloy (2025). In Situ Micro-Cantilever Beam Bending Tests to Assess the Adhesion Strength of Cr Coatings on Zry-4, Symposium on Mechanical Behavior Related to Interface Physics IV, The Metallurgical Society (TMS) Annual Conference, Las Vegas, NV, March 2025. | |
| FY2025 | Mauseth, T. J., Dunzik-Gougar, M. L., Teng, F., Shah, S., Bawane, K. K., Pradhan, A., Cai, L., Bachhav, M. and J.D. Stempien (2025). Correlative Atom Probe Tomography of the Buffer-IPyC Interlayer Region of TRISO-coated Particles. Presented at the 2025 Nuclear Science User Facilities (NSUF) Annual Program Review. | Publication |
| FY2025 | Mauseth, T. J., Dunzik-Gougar, M. L., Teng, F., Shah, S., Bawane, K. K., Pradhan, A., Cai, L., Bachhav, M. and J.D. Stempien (2025). Microstructural Characterization of AGR-2 TRISO Particle Buffer, IPyC, and Buffer-IPyC Interfaces. Presented at the 2025 Seventh International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-7). | |
| FY2025 | Mauseth, T. J., Teng, F., Cai, L. and J.D. Stempien (2024). Micro-Tensile Properties of Irradiated AGR-2 TRISO Fuel Pyrolytic Carbon (PyC) and Silicon Carbide (SiC) Coatings. Presented at the 2024 Workshop on Storage and Transportation of TRISO and Metal Spent Nuclear Fuels. | Publication |
| FY2025 | Mauseth, T. J., Teng, F., Cai, L., and J.D. Stempien (2024). Fracture Behavior Considerations for the TRISO Particle Matrix. Presented at the 2024 Workshop on Storage and Transportation of TRISO and Metal Spent Nuclear Fuels. | Publication |
| FY2025 | Mauseth, T. J., Teng, F., Cai, L., Laug, D.V. and J.D. Stempien (2024). Micro-tensile Properties of Fueled Irradiated AGR-2 TRISO-coated Particle Buffer, IPyC, and SiC Interlayer Regions. Presented at the 2024 Nuclear Materials (NuMat) Conference. | Publication |
| FY2025 | Mauseth, T., Dunzik-Gougar, M. L. and F. Teng (2025). Micro-tensile Characteristics of As-fabricated and Irradiated AGR-2 TRISO Fuel Particle Buffer, IPyC, and Buffer-IPyC Interlayer Regions. Journal of Nuclear Materials, 156086. | Publication |
| FY2025 | Meehan N.A., Gorton J.P., Capps N.A. and N.R. Brown (2025). Identifying high-impact and high-uncertainty parameters in MiniFuel model predictions. Journal of Nuclear Materials, 2025;609:155745. doi:10.1016/j.jnucmat.155745. | Publication |
| FY2025 | Middlemas, S., and C. Adkins (2025). A critical analysis of U-Pu-Zr phase transitions using calorimetric, microstructural, and phase equilibria data. Journal of Nuclear Materials, 612, 155778. | Publication |
| FY2025 | Mondal, S., Chatterjee, A., Roy, R., Muntaha, M.A., Wharry, J.P., Qu, H.J. and R. Umretiya. Synergistic Roles of Cr and Mo in Low Temperature Steam Oxidation of FeCrAl Alloys. Corrosion Science (2025): 113107. | Publication |
| FY2025 | Nagothi, B.S., Qu, H., Zhang, W., Umretiya, R.V., Dolley, E.and R.B. Rebak (2024). Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding. Materials 17, no. 7: 1633. | Publication |
| FY2025 | Nelson M., Samuha S., Kombaiah B., Kamerman D. and P. Hosemann (2024). Enhanced Stress Relaxation Behavior Via Basal dislocation activity in Zircaloy-4 cladding. Journal of Nuclear Materials; 601:155337. | Publication |
| FY2025 | Nicodemo G., Zullo G., Cappia F., Van Uffelen P., De Lara A., Luzzi L. and D. Pizzocri (2024). Chromia-doped UO2 fuel: An engineering model for chromium solubility and fission gas diffusivity. Journal of Nuclear Materials. 601:155301. | Publication |
| FY2025 | Petersen, P. G., Hansen, R. S., Cappia, F., Kamerman, D., Baird, K. and C. Christensen (2024). Design and Evaluation of a Ring Tension Test Grip for Remote Mechanical Testing of Irradiated Tubular Specimens. Journal of Testing and Evaluation, 52(6), 3326-3345. | Publication |
| FY2025 | Pham, B. T., Hawkes, G. L., Lybeck, N. J., Otani, C. and P.A. Demkowicz (2025). Uncertainty Quantification of Calculated Fuel Temperature for the AGR-5/6/7 Irradiation Experiment. Paper presented at the NURETH 21 Annual Meeting. | |
| FY2025 | Pradhan A, Xu F, Salvato D, et al. (2024). Characterization of Fuel Cladding Chemical Interaction on a High Burnup U-10Zr Metallic Fuel via Electron Energy Loss Spectroscopy Enhanced by Machine Learning. Mater Charact. 2024;218(1):114524. | Publication |
| FY2025 | Probert A., Swearingen A., Schulthess J., Capriotti L., Jensen C. and A. Aitkaliyeva (2025). Comparative Post-irradiation Examination of High Burnup U-19Pu-10Zr: Assessing Steady-state Irradiation Behavior Against Historical and Modeled Fuel Performance. Journal of Nuclear Materials.; 610:155782. | Publication |
| FY2025 | Qu, H., Yin, L., Larsen, M., and R.B. Rebak (2024). Distinctive oxide films develop on the surface of fecral as the environment changes for nuclear fuel cladding. Corrosion and Materials Degradation 5, no. 1: 109-123. | Publication |
| FY2025 | Qu, H.J., Chikhalikar, A.S., Abouelella, H., Roy, I., Rajendran, R., Nagothi, B.S., Umretiya, R., Hoffman, A.K. and R.B. Rebak (2024). Effect of molybdenum on the oxidation resistance of FeCrAl alloy in lower temperature (400° C) and higher temperature (1200° C) steam environments. Corrosion Science 229 (2024): 111870. | Publication |
| FY2025 | Rajendran, R., Chikhalikar, A.S., Roy, I., Abouelella, H., Qu, H.J., Umretiya, R.V., Hoffman, A.K., and R.B. Rebak (2024). Effect of aging and a segregation on oxidation and electrochemical behavior of FeCrAl alloys. Journal of Nuclear Materials 588: 154751. | Publication |
| FY2025 | Ravi, S.K., Comlek, Y., Pathak, A., Gupta, V., Umretiya, R., Hoffman, A., Pilania, G. et al. (2025) Interpretable multi-source data fusion through Latent Variable Gaussian Process. Engineering Applications of Artificial Intelligence 145: 110033. | Publication |
| FY2025 | Riet, A. A. and J.D. Stempien (2025). Use of Constrained Gamma Emission Computed Tomography to Evaluate Fission Product Distributions in High-Temperature Materials from a TRISO Fuel Irradiation. Nuclear Science and Engineering, 1-12. | Publication |
| FY2025 | Rittenhouse J., Pradhan A., Kamerman D.W, Burns J., Xu F., Wen H. and T. Yao (2025) Site-specific Nanoscale Characterization of Zirconium Hydrides in the Hydride Rim Structure of Hydrogen-charged Zircaloy-4 Cladding. Mater Charact ;224:115006. | Publication |
| FY2025 | Roy, R., Chatterjee, A., Mondal, S., Muntaha, M.A., Wharry, J.P., Qu, H.J. and R. Umretiya.(2025). Sequential oxidation and hydrothermal corrosion of FeCrAl alloys at BWR top-of-core conditions. Corrosion Science: 112965. | Publication |
| FY2025 | Salvato, D., Nguyen, B.-P., Wang, Y., Di Lemma, F. G., Capriotti, L., Aitkaliyeva, A. and T. Yao, (2025). TEM Characterization of Two Variants of Fuel Cladding Chemical Interaction in a HT-9 Clad U-10Zr Fuel. Variant 1: FCCI with a Zr Rind. Journal of Nuclear Materials, 614, 155855. | Publication |
| FY2025 | Seo S., Folsom C., Jensen C. et al. (2024). International Fuel Performance Study of Fresh Fuel Experiments for PCMI Effects During RIA Experiments. Nuclear Engineering and Design; 430:113673. | Publication |
| FY2025 | Seshadri A., Cervenka P., Fazi A., Sevecek M., Carpenter D., Cetiner N., Motta A., Ishak C., Fei Z., Raiman S., Xu P. and K. Shirvan. In-pile hydrothermal corrosion behavior of Zirconium Alloys with and without ATF Coatings, Presented at 21st ASTM International Symposium on Zirconium in the Nuclear Industry. | |
| FY2025 | Shirvan K., Cervenka P., Fazi A. and A. Seshadri (2025). Experimental Investigation of CrNb Coatings for PWRs and BWRs. Paper at the TopFuel 2025: Nuclear Reactor Fuel Performance Conference. | |
| FY2025 | Singh G., Yu J., Xu F., Yao T. and P. Xu (2024). Multiscale Modeling of Silicon Carbide Cladding for Nuclear Applications: Thermal Performance Modeling. Energies. 2024; 17(23):6124. | Publication |
| FY2025 | Skerjanc, W. F., Jiang, W., Demkowicz, P. A. and J.D. Stempien (2025). Evaluation of AGR-3/4 In-pile Silver Release Predictions Against Post-irradiation Examination measurements. Journal of Nuclear Materials, 615, 155942. | Publication |
| FY2025 | Sridharan, K. Maier, B., Dabney, T., Willing, E., Pocquette, N. Lukas, C., Anderson, N. and H. Yeom (2025). Cold Spray Materials Deposition Technology for Nuclear Energy Systems, Symposium on Advances in Materials Deposition by Cold Spray and Related Technologies, The Metallurgical Society (TMS) Annual Conference, Las Vegas, NV, March 2025. | |
| FY2025 | Umretiya, R.V, Qu, H., Yin, L., Jurewicz, T.B., Gupta, V.K., Drobnjak, M., Knussman, M. Hoffman, A.K. and R.B. Rebak (2024). Corrosion behavior of additively manufactured FeCrAl in out-of-pile light water reactor environments, npj Mater Degrad 8, 88. | Publication |
| FY2025 | Umretiya, R.V., Chikhalikar, A., Elward, B., Moreira, T.A., Anderson, M., Rebak, R.B. and J.V. Rojas (2024). The Effect of Ramp Heating on the Microstructure and Surface Chemistry of APMT FeCrAl Alloy. Nuclear Materials and Energy 38: 101567. | Publication |
| FY2025 | Walter, J., Roberts, E., Fredrick, K., Viands, D. and X. Huang (2025). The Effect of Chromium Coating Microstructure and Oxide Films on Hydrogen Uptake in Zirconium-alloy Nuclear Fuel Cladding, 21st International Symposium on Zirconium in the Nuclear Industry, Aix-en-Provence, France. | |
| FY2025 | Wang, Y., Burns, J., Yao, T. and L. Capriotti (2024). Transmission Electron Microscopy Characterization of Fuel Cladding Chemical Interaction (FCCI) in ATR-irradiated HT9 clad U-10M (10M = 5Mo-4.3Ti-0.7Zr wt%) metallic fuel, Journal of Nuclear Materials, Volume 599, 2024, 155209, ISSN 0022-3115. | Publication |
| FY2025 | Wang, Y., Howard, C., Xu, F., Salvato, D., Bawane, K., Murray, D., Frazer, D., Anderson, S., Yao, T., Yeo, S., Kim, J-H, Lee, B-O, Kim, J., Fielding, R. and L. Capriotti (2024). Microstructural and micromechanical characterization of Cr diffusion barrier in ATR irradiated U-10Zr metallic fuel, Journal of Nuclear Materials, Volume 599, 2024, 155231, ISSN 0022-3115. | Publication |
| FY2025 | Woolstenhulme, N. et al. (2025). SPARC - Plans for a New Critical Experiment Facility with a Horizontal Split Table (INL/RPT-25-84855). Idaho National Laboratory, Idaho Falls, ID. | Publication |
| FY2025 | Woolstenhulme, N., Martin, N., DeHart, M., Percher, C., Cutler, T., Wieselquist, W. (2025). SPARC, an Effort to Reestablish a Horizontal Split Table Critical Facility for HALEU Experiments and Beyond. Paper presented at the NCSD 2025 Annual Meeting. | |
| FY2025 | Worrall, M., Woolstenhulme, N., Downey, C., Jesse, C., Murdock, C. and M. Tippet (2024). Fast Neutron Irradiation Capability in Existing Thermal Test Reactors, Annals of Nuclear Energy, Volume 207, 110731, ISSN 0306-4549. | Publication |
| FY2025 | Yang, G., Nguyen, B.-P., Rittenhouse, J. E., Xu, F., Gonderman, S., Gazza, J., Xu, P. and T.Yao (2025). Investigating Grain Structure and Microcracking in SiCf-SiCm Composites Using 4D-STEM. Materials Characterization, 225, 115165. | Publication |
| FY2025 | Yang, Y., Weicheng Z. and C. Massey (2025). Computational Design of Improved Fast Reactor Cladding (ORNL/TM-2025/3953), Oak Ridge National Laboratory, Oak Ridge, TN. | Publication |
| FY2025 | Yuan, G., Cook, D.H., Barnard, H., Lahoda, E., Xu, P., Ritchie, R.O. and D. Liu (2025). Improved Damage Tolerance of SiC-Based Nuclear Fuel Cladding with Novel Multi-Layered SiC Coating Design at 1200 °C, Materials and Design, Volume 256, August 2025, 114260. | Publication |
| FY2025 | Zhang, J., Xu, P., Sevecek, M., Sim, K.S. and A. Khaperskaia (2025). Contribution of IAEA Coordinated Research Projects to Light Water Reactors Advanced Technology Fuel Testing and Simulation, Nuclear Engineering and Design 418, 112910. | Publication |
| FY2025 | Zhang, S., Ma, Z., Xu, P. (2024). Incorporating A Risk-Informed, Performance-Based Concept into Nuclear Fuel and Materials Development for Advanced Reactors, 2024 ANS Annual Meeting. | |
| FY2025 | Zhao, L., Wang, Y., and F. Xu (2025). Accurate Segmentation of Localized Fuel Cladding Chemical Interaction Layers in SEM Micrographs with Deep Learning Method. Scientific Reports, 15, 28878. | Publication |
| FY2025 | Zhao, L., Xu, F., Porter, D. L. and Y. Wang (2025). Quantification of line dislocations in FFTF irradiated HT9 cladding by deep learning method. Materials Characterization, 227, 115322. | Publication |
| FY2024 | Anderson KS, Hale DD, Schulthess JL, Arrowood MM. A standard capsule design for structural material testing in the Advanced Test Reactor. Nucl Eng Des. 2023;414:112630. | Publication |
| FY2024 | Beck PM, Hayne ML, Liu C, Valdez J, Nizolek T, Briggs SA, Maloy SA, Saleh TA, Eftink BP. Mandrel diameter effect on ring-pull testing of nuclear fuel cladding, J Nucl Mater. 2024;596:155087. | Publication |
| FY2024 | Gribok AV, Di Lemma FG, Fay J, Porter DL, Paaren KM, Capriotti L. Qualification and Quantification of Porosity at the Top of the Fuel Pins in Metallic Fuels Using Image Processing. Energies. 2024; 17(9):1990. | Publication |
| FY2024 | Hansen RS, Kamerman DW, Petersen PG, Cappia F. Evaluation of the ring tension test (RTT) for robust determination of material strengths. Int J Solids Struct. 2023;282:112471. | Publication |
| FY2024 | Hu C, Le J-L, Koyanagi T, Labuz JF. Experimental investigation of probabilistic failure of SiC/SiC composite tubes under multiaxial loading. Compos Struct. 2024;335:118002. | Publication |
| FY2024 | Kamerman D, Bachhav M, Yao T, Pu X, Burns J. Formation and characterization of hydride rim structures in Zircaloy-4 nuclear fuel cladding tubes. J Nucl Mater. 2023;586:154675. | Publication |
| FY2024 | Kamerman D. The deformation and burst behavior of Zircaloy-4 cladding tubes with hydride rim features subject to internal pressure loads. Eng Fail Anal. 2023;153:07547. | Publication |
| FY2024 | Koyanagi T, Hawkins C, Lamm B, Lara-Curzio E, Katoh Y, Deck C. Mechanical degradation of duplex SiC-fiber reinforced SiC matrix composite tubes under a controlled high-temperature steam environment. Ceram Int. 2024. | Publication |
| FY2024 | Koyanagi T, Hu X, Petrie CM, Singh G, Ang C, Deck CP, Kim W-J, Kim D, Sauder C, Braun J, Katoh Y. Hermeticity of SiC/SiC composite and monolithic SiC tubes irradiated under radial high-heat flux. J Nucl Mater. 2024;588:154784. | Publication |
| FY2024 | Lu C, Kardoulaki E, Stauff NE, Cuadra A. The Use of High-Density UN Fuel in Heat-Pipe Microreactors. Nucl Technol. 2024:1-18. | Publication |
| FY2024 | Martin N, Seo S, Prieto SB, Jesse C, Woolstenhulme N. Reactor physics characterization of triply periodic minimal surface-based nuclear fuel lattices. Prog Nucl Energy. 2023;165:104895. | Publication |
| FY2024 | Middlemas S, Janney DE, Adkins C, Bawane K. Determining the effects of U/Pu ratio on subsolidus phase transitions in U-Pu-Zr metallic fuel alloys. J Nucl Mater. 2024;591:154909. | Publication |
| FY2024 | Nelson M, Samuha S, Kamerman D, Hosemann P. Temperature-Dependent Mechanical Anisotropy in Textured Zircaloy Cladding. J Nucl Mater. | Publication |
| FY2024 | Paaren K, Gale M, Wootan D, Medvedev P, Porter D. Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data. Energies. 2023;16:7600. | Publication |
| FY2024 | Paaren KM, Christian S, Capriotti L, Aitkaliyeva A, Porter D. Comparison of Zirconium Redistribution in BISON EBR-II Models Using FIPD and IMIS Databases with Experimental Post Irradiation Examination. Energies. 2023;16(19):6817. | Publication |
| FY2024 | Patnaik S, Beausoleil II GL, Capriotti L. Fission accelerated steady-state post irradiation examinations Part II. Nucl Eng Technol. 2024. | Publication |
| FY2024 | Salvato D, Paaren KM, Hirschhorn JA, Aagesen LK, Xu F, Di Lemma FG, Capriotti L, Yao T. The effect of temperature and burnup on U-10Zr metallic fuel chemical interaction with HT-9: A SEM-EDS study. J Nucl Mater. 2024;591:154928. | Publication |
| FY2024 | Terricabras AJ, Drewry SM, Campbell K, et al. Performance and properties evolution of near-term accident tolerant fuel: Cr-doped UO2. J Nucl Mater. 2024;594:155022. | Publication |
| FY2024 | Williams WJ, Yao T, Pu X, Capriotti L. Characterization of micro-burnup treat irradiated U-22.5 at.% Zr and U-52.8 at.% Zr foils by transmission electron microscopy and X-ray diffraction. J Nucl Mater. 2023;585:154644. | Publication |
| FY2024 | Xu F, Yao T, Xu P, et al. Multi-Scale Characterization of Porosity and Cracks in Silicon Carbide Cladding after Transient Reactor Test Facility Irradiation. Energies. 2024;17(1):197. | Publication |
| FY2024 | Yan Y, Harp J, Le Coq A, Massey C, Linton K. High-temperature steam oxidation study of irradiated FeCrAl defueled specimens. Journal of Nuclear Materials. 2024 Mar 1;590:154868. | Publication |
| FY2023 | Beausoleil G, Capriotti L, Curnutt B, Fielding R, Hayes S, Wachs D. FAST irradiations and initial post irradiation examinations Part I. Nucl Eng Technol. 2022;54(11):4084-4094. ISSN 1738-5733 | Publication |
| FY2023 | Benson MT, Yao T, Zelina JN, Teng F, Murray D, Di Lemma F, Williams WJ, Zhang J, Zhuo W. The formation mechanism of the Zr rind in U-Zr fuels. J Nucl Mater. 2022;572:154057. ISSN 0022-3115. | Publication |
| FY2023 | Cappia F, Wright K, Frazer D, Bawane K, Kombaiah B, Williams W, Finkeldei S, Teng F, Giglio J, Cinbiz MN, Hilton B, Strumpell J, Daum R, Yueh K, Jensen C, Wachs D. Detailed characterization of a PWR fuel rod at high burnup in support of LOCA testing. J Nucl Mater. 2022;569:153881. ISSN 0022-3115. | Publication |
| FY2023 | Capriotti L, Di Lemma FG, Harp JM. Testing fast reactor fuels in a thermal reactor: Comparison of transmutation metallic fuel alloys behavior by scanning electron microscopy. J Nucl Mater. 2023;575:154221. ISSN 0022-3115. | Publication |
| FY2023 | Di Lemma FG, Yao T, Salvato D, Capriotti L, Teng F, Jokisaari AM, Beeler BW, Wang Y, Jensen CJ. Microstructural and phase changes in alpha uranium investigated via in-situ studies and molecular dynamics. J Nucl Mater. 2023;577:154341. ISSN 0022-3115. | Publication |
| FY2023 | Folsom CP, Armstrong RJ, Woolstenhulme NE, Fleming AD, Hill CM, Jensen CB, Wachs DM. Design of separate-effects In-Pile transient boiling experiments at the TREAT Facility. Nucl Eng Des. 2022;397:111919. ISSN 0029-5493. | Publication |
| FY2023 | Folsom CP, Schulthess JL, Kamerman DW, Hansen RS, Woolstenhulme NE, Jensen CB, Astle LA, Giraldo LO, Fleming A, Wachs DM. Resumption of water capsule reactivity-initiated accident testing at TREAT. Nucl Eng Des. 2023;413:112509. ISSN 0029-5493. | Publication |
| FY2023 | Hanson WA, Cappia F, White JT, McClellan KJ, Harp JM. Post-irradiation examination of low burnup U3Si5 and UN-U3Si5 composite fuels. J Nucl Mater. 2023;578:154346. ISSN 0022-3115. | Publication |
| FY2023 | Hu C, Labuz JF, Koyanagi T, Le J-L. Mechanistic Modeling of Lifetime Distribution of SiC/SiC Composite Claddings. J Am Ceram Soc. December 2022. | Publication |
| FY2023 | Kamerman D, Nelson M. Multiaxial Plastic Deformation of Zircaloy-4 Nuclear Fuel Cladding Tubes. Nucl Technol. February 2023. | Publication |
| FY2023 | Kane K, Bell S, Capps N, Garrison B, Shapovalov K, Jacobsen G, Deck C, Graening T, Koyanagi T, Massey C. The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts. J Nucl Mater. 2023;574:154152. ISSN 0022-3115. | Publication |
| FY2023 | Koyanagi T, Karakoc O, Hawkins C, Lara-Curzio E, Deck C, Katoh Y. Stress rupture of SiC/SiC composite tubes under high-temperature steam. Int J Appl Ceram Technol. 2023. ISSN 1546-542X. | Publication |
| FY2023 | Schulthess JL, Spencer BW, Petersen PG, Woolstenhulme NE, Ban D, Frazer D, Sudderth L, Hamilton S, Jewell JK, Mariani RD. Experimental results of conductive inserts to reduce nuclear fuel temperature during nuclear volumetric heating. J Nucl Mater. 2023;574:154176. ISSN 0022-3115. | Publication |
| FY2023 | U.S. Department of Energy. (2023). Alternate fuels: Thorium and Uranium-233. Thorium Energy Alliance. | Publication |
| FY2023 | Wang Y, Miller BD, Harp JM, Salvato D, Capriotti L, Yao T. Transmission electron microscopy characterization of the fuel-cladding chemical interactions in HT9 cladded U-10Zr fuel. J Nucl Mater. 2022;572:153990. ISSN 0022-3115. | Publication |
| FY2023 | Williams WJ, Vogel SC, Okuniewski MA. Phase transformations and thermal expansion coefficients of unirradiated U-X wt.% Zr (X = 6, 10, 20, 30) measured via neutron diffraction. J Nucl Mater. 2023;579:154380. ISSN 0022-3115. | Publication |
| FY2023 | Woolstenhulme N, Chapman D, Cordes N, Fleming A, Hill C, Jensen C, Schulthess J, Ramirez M, Linton K, Schappel D, Vasudevamurthy G. TREAT testing of additively manufactured SiC canisters loaded with high density TRISO fuel for the Transformational Challenge Reactor project. J Nucl Mater. 2023;575:154204. ISSN 0022-3115. | Publication |
| FY2023 | Xu F, Cai L, Salvato D, et al. Advanced characterization-informed machine learning framework and quantitative insight to irradiated annular U-10Zr metallic fuels. Sci Rep. 2023;13:10616. | Publication |
| FY2023 | Yan Y, Graening T, Nelson AT. Hydriding, Oxidation, and Ductility Evaluation of Cr-Coated Zircaloy-4 Tubing. Metals. 2022;12(12):1998. | Publication |
| FY2023 | Yuan G, Forna-Kreutzer JP, Xu P, Gonderman S, Deck C, Olson L, Lahoda E, Ritchie RO, Liu D. In situ high-temperature 3D imaging of the damage evolution in a SiC nuclear fuel cladding material. Mater Des. 2023;227:111784. ISSN 0264-1275. | Publication |
| FY2022 | Cocke, C.K., Rollett, A.D., Lebensohn, R.A. et al. The AFRL Additive Manufacturing Modeling Challenge: Predicting Micromechanical Fields in AM IN625 Using an FFT-Based Method with Direct Input from a 3D Microstructural Image, Integr Mater Manuf Innov Volume 10 (2021) 157 | Publication |
| FY2022 | Copeland-Johnson, T.M., Nyamekye, C.K.A., Ecker, L., Bowler, N., Smith, E.A., Rebak, R.B. and S. K. Gill. Analysis of Inconel 600 Oxidized under Loss-of-Coolant Accident Conditions: A Multi-modal Approach, Corrosion Science Volume 195 (2022) 109950, | Publication |
| FY2022 | Evans, K.J. and R. B. Rebak. Hydrogen Permeation in FeCrAl APMT Alloy for Accident Tolerant Fuel Cladding, Corrosion Journal, Volume 78 (May 2022) 449 | Publication |
| FY2022 | Garud, Y.S., Hoffman, A.K. and R. B. Rebak. Hydrogen Isotopes Permeation in Clean or Unoxidized FeCrAl Alloys: A Review, Metallurgical and Materials Transactions A, 53, 773-793 (2022). | Publication |
| FY2022 | Hoffman, A. K., Cappia, F., Burns, J., He, L., Umretiya, R., Gupta, V., Massey, C., Harp, J.and R. B. Rebak. FeCrAl Fuel Clad Chemical Interaction in Light Water Reactor Environment, in Transactions of the ANS Winter 2021 meeting, Washington DC, USA. December 2021 Volume 125 (2021) 515 | Publication |
| FY2022 | Huang, S., Dolley, E., An, K., Yu, D., Crawford, C., Othon, M.A., Spinelli, I., Knussman, M.P. and R. B. Rebak. Microstructure and Tensile Behavior of Powder Metallurgy FeCrAl Accident Tolerant Fuel Cladding, Journal of Nuclear Materials Volume 560 (2022) 153524 | Publication |
| FY2022 | Kamerman, D., Cappia, F., Wheeler, K., Petersen, P., Rosvall, E., Dabney, T., Yeom, H., Sridharan, K., Sevecek, M. and J. Schulthess. Development of Axial and Ring Hoop Tension Testing Methods for Nuclear Fuel Cladding Tubes, Nuclear Materials and Energy, Volume 31 (2022) | Publication |
| FY2022 | Kane K, Bell S, Garrison B, Ridley M, Gussev M, Linton K, Capps N. Quantifying deformation during Zry-4 burst testing: a comparison of BISON and a combined in-situ digital image correlation and infrared thermography method. J Nucl Mater. 2022;572:154063. | Publication |
| FY2022 | Kocevski, V., Cooper, M.W.D., Claisse, A.J., Andersson and D.A. Hide. Development and Application of a Uranium Mononitride (UN) Potential: Thermomechanical Properties and Xe Diffusion, Journal of Nuclear Materials, Volume 562 (April 2022) | Publication |
| FY2022 | Koyanagi, T. Wang, H., Arregui Mena, JD., Petrie, C.M., Deck, C.P., Kim, W-J., Kim, D., Sauder, D., Braun, J.and Y. Katoh. Thermal Diffusivity and Thermal Conductivity of SiC Composite Tubes: The Effects of Microstructure and Irradiation, Journal of Nuclear Materials, Volume 557 (December 2021) | Publication |
| FY2022 | Kumagai, T., Pachaury, Y., Maccione, R., Wharry, J.P and A. El-Azab. An Atomistic Investigation of Dislocation Velocity in Body-centered Cubic FeCrAl Alloys , Materialia Volume 18 (2021) 101165 | Publication |
| FY2022 | Liu, J. et al. Structural and Phase Evolution in U3Si2 During Steam Corrosion, Corrosion Science, Volume 204 (2022) 110373 | Publication |
| FY2022 | Macisaac, M. Bavdekar, S. Subhash, G. Nance, J. Sankar, B. V., Kim, N-H. and G. Subhash. A Novel Rotating Flexure-Test Technique for Brittle Materials with Circular Geometries, Experimental Techniques Volume 12 (2022) | Publication |
| FY2022 | Mirmohammad, H. and O. Kingstedt. Theoretical Considerations for Transitioning the Grid Method Technique to the Microscale, Exp Mech Volume 61 (2021) 753. | Publication |
| FY2022 | Mirmohammad, H., Gunn, T. and O.T. Kingstedt. In-Situ Full-Field Strain Measurement at the Sub-grain Scale Using the Scanning Electron Microscope Grid Method, Exp Tech Volume 45 (2021) 109. | Publication |
| FY2022 | Nagaraju, H. T., Subhash, G., Kim, N-H, Haftka, R.and B. Sankar. Effect of Curvature on Extensional Stiffness Matrix of 2-D Braided Composite Tubes, Composites Part A: Applied Science and Manufacturing Volume 147(2021) 106422 | Publication |
| FY2022 | Nance J.R., Subhash, G. Sankar, B., Haftka, R., Kim, N-H, Deck, C. and S. Oswal. Measurement of Residual Stress in Silicon Carbide Fibers of Tubular Composites Using Raman Spectroscopy, Acta Materialia Volume 217(2021) 117164 | Publication |
| FY2022 | Nance J.R., Subhash, G. Sankar, B., Kim, N-H, Deck C. and S. Oswald. Influence of Weave Architecture on Mechanical Response of SiCf-SiCm Tubular Composites, Materials Today Communications Volume 33(2022) 104206 | Publication |
| FY2022 | Pachaury, Y., Kumagai, T., Wharry, J.P. and A. El-Azab. A Data Science Approach for Analysis and Reconstruction of Spinodal-like Composition Fields in Irradiated FeCrAl Alloys, Acta Materialia Volume 234 (2022) 118019 | Publication |
| FY2022 | Quillin, K., Yeom, H., Dabney, T., McFarland, M. and K. Sridharan. Experimental Evaluation of Direct Current Magnetron Sputtered and High-power Impulse Magnetron Sputtered Cr Coatings on SiC for Lightwater Reactor Applications, Thin Solid Films Volume 716 (2020) 138431 | Publication |
| FY2022 | Quillin, K., Yeom, H., Dabney, T., Willing, E. and K. Sridharan. Microstructural and Nanomechanical Studies of PVD Cr coatings on SiC for LWR Fuel Cladding Applications, Surface and Coatings Technology Volume 441 (2022) 128577 | Publication |
| FY2022 | Rebak, R.B. Innovative Accident Tolerant Nuclear Fuel Materials Will Help Extending the Life of Light Water Reactors, KOM Corrosion and Material Protection Journal Volume 66 (2022) 36. | Publication |
| FY2022 | Rebak, R.B., Dolley, E.J., Zhang, W., Umretiya, R.V. and A. K. Hoffman. Enhanced Mechanical Properties of Iron-Chromium-Aluminum Cladding for Light Water Reactor Fuels, In Proceedings of ASME 2022 PVP Conference, Las Vegas, US. July 2022, | Publication |
| FY2022 | Rebak, R.B., Jurewicz, T.B., Hoffman, A.K., Yin, L., Amroussia, A., Umretiya, R.V. and R. M. Fawcett. Zinc Additions Reduces Dissolution Rate of FeCrAl Fuel Cladding, in Transactions of ANS Winter 2021 meeting, Washington DC, US. December 2021. Volume 125 (2021) 513. | Publication |
| FY2022 | Rebak, R.B., Jurewicz, T.B., Larsen, M. and L. Yi. Zinc water chemistry reduces dissolution of FeCrAl for nuclear fuel cladding, Corrosion Science 198 (2022) 110156. | Publication |
| FY2022 | Rebak, R.B., Umretiya, R.V., Hoffman, A.K., Yin, L., Amroussia, A. and D. R. Lutz. Reprocessing Capabilities of FeCrAl-Clad Used Fuel, in Transactions of the ANS Winter 2021 meeting, Washington DC, December 2021, Volume 125 (2021) 181. | Publication |
| FY2022 | Rebak, R.B., Yin, L., Jurewicz, T.B. and A. K. Hoffman. Acid Dissolution Behavior of Ferritic FeCrAl Tubes Candidates for Nuclear Fuel Cladding, Corrosion Journal, Volume 77 (2021) 1321. | Publication |
| FY2022 | Rebak, R.B., Yin, L., Larsen, M., Umretiya, R.V. and A. K. Hoffman. Mitigating LWR IronClad Fuel Cladding Dissolution Using Zinc Water Chemistry, Paper PVP2022-80559 in Proceedings of ASME 2022 PVP Conference, July 2022, Las Vegas | Publication |
| FY2022 | Sankar, B. V., Thandaga Nagaraju, H., Kim, N-H. and G. Subhash. An Extrapolation Method to Remove Spurious Stress Concentration in Pixel-based Meshes, Composite Structures Volume 290 (2022) 115522 | Publication |
| FY2022 | Schoell, R., Kabel, J., Lam, S., Sharma, A., Michler, J., Hosemann, P. and D. Kaoumi. Corrosion Behavior of a Series of Combinatorial Physical Vapor Deposition Coatings on SiC in a Simulated Boiling Water Reactor Environment, Journal of Nuclear Materials (2022) | Publication |
| FY2022 | Smith, A. J., Maxwell, H. L., Mirmohammad, H., Kingstedt, O. T. and R.B. Berke. A Novel Variable Extensometer Method for Measuring Ductility Scaling Parameters from Single Specimens. ASME. J. Appl. Mech, Volume 89 (2022) 031006 | Publication |
| FY2022 | Sun T, Cho J, Shang Z, Niu T, Ding J, Wang J, Wang H, Zhang X. Deformation mechanism in nanolaminate FeCrAl alloys by in situ micromechanical strain rate jump tests at elevated temperatures. Scripta Materialia. 2022;215:114698 | Publication |
| FY2022 | Sun T, Shang Z, Cho J, Ding J, Niu T, Zhang Y, Yang B, Xie D, Wang J, Wang H, Zhang X. Ultra-fine-grained and gradient FeCrAl alloys with outstanding work hardening capability. Acta Materialia. 2021;215:117049. | Publication |
| FY2022 | Warren, P., Warren, G., Wu, Y.Q., Burns, J., Dubey, M. and J.P. Wharry. Method for fabricating depth-specific TEM in situ tensile bars, JOM Volume 72 (2020) 2057 | Publication |
| FY2022 | Wei, B.Q., Xie, D.Y., Wu, W.Q. Shao, L and J Wang. Quantifying the Glide Resistance to Dislocations in Proton-Irradiated FeCrAl Alloy, JOM (2022) | Publication |
| FY2022 | Xi, J., Liu, C., Morgan, D. and I. Szlufarska, An unexpected role of H during SiC corrosion in water, Journal Phys. Chem. C, Volume 124 (2020) 9394 | Publication |
| FY2022 | Xi, J., Liu, C., Morgan, D. and I. Szlufarska, Deciphering water-solid reactions during hydrothermal corrosion of SiC, Acta Materialia Volume 209 (2021) 116803 | Publication |
| FY2022 | Xie, D.Y., Wei, B., Wu, W.Q. and J Wang. Crystallographic Orientation Dependence of Mechanical Responses of FeCrAl Micropillars, Crystals Volume 10 (2020) 943 | Publication |
| FY2022 | Xu, S., Xie, D., Liu, G., Ming, K. and J Wang. Quantifying the resistance to dislocation glide in single phase FeCrAl alloy, International Journal of Plasticity Volume 132 (2020) 102770 | Publication |
| FY2022 | Yang, K., Kardoulaki, E., Zhao, D., Gong, B., Broussard, A., Metzger, K., White, J.T., Sivack, M.R., McClellan, K.J., Lahoda, E.J. and J. Lian, Cr-incorporated uranium nitride composite fuels with enhanced mechanical performance and oxidation resistance, Journal of Nuclear Materials Volume 559 (2022) | Publication |
| FY2022 | Yang, K., Kardoulaki, E., Zhao, D., Gong, B., Broussard, A., Metzger, K., White, J.T., Sivack, M.R., McClellan, K.J., Lahoda, E.J. and J. Lian, UN and U3Si2 Composites Densified by Spark Plasma Sintering for Accident-Tolerant Fuels, Ceramics International (December 2021) | Publication |
| FY2022 | Yarrington JD, Schulthess JL, Parker SH, Argyle JM, Turner CG, Stanek JD, Christensen CL. Advanced autonomous welding for refabrication and follow-on testing of previously irradiated nuclear fuel. Nucl Technol. 2022;209(2):127-143 | Publication |
| FY2022 | Zhang, B., Study of Reference Burnup Steps Optimization in Fuel Segment Data File Generation for NEXUS/ANC9 Code System, in Proceedings of 2022 PHYSOR Conference, Pittsburgh, Pennsylvania, US. May 2022 | Publication |
| FY2021 | Balke T, Long AM, Vogel SC, Wohlberg B, Bouman CA. Hyperspectral neutron CT with material decomposition. 2021 IEEE International Conference on Image Processing (ICIP); 2021; Anchorage, AK, USA. pp. 3482-3486 | Publication |
| FY2021 | Beausoleil, G. L., Petrie, C., Williams, W., Jokisaari, A., Capriotti, L., Novascone, S., Kerr, M. (2021). Integrating Advanced Modeling and Accelerated Testing for a Modernized Fuel Qualification Paradigm. Nuclear Technology, 207(10), 1491 1510. | Publication |
| FY2021 | Bess, J.D., Pope, C.L., Chipman, A.S., and Jensen, C.B. (2021). Utility of EBR-II Benchmark Model to Enable MOX Fuel Pin Characterization. Transactions of the American Nuclear Society, 124(1), 238-241. | Publication |
| FY2021 | Capps, N., Jensen, C., Cappia, F., Harp, J., Terrani, K., Woolstenhulme, N., and Wachs, D. (2021). A Critical Review of High Burnup Fuel Fragmentation, Relocation, and Dispersal under Loss-Of-Coolant Accident Conditions. Journal of Nuclear Materials, 546, 152750. | Publication |
| FY2021 | Chaari, N., Bischoff, J., Buchanan, K., Delafoy, C., Barberis, P., Augereau, J., and Nimishakavi, K. (2021). The Behavior of Cr-Coated Zirconium Alloy Cladding Tubes at High Temperatures. ASTM Symposia, 189-210. | Publication |
| FY2021 | Curnutt, R., Woolstenhulme, N., Nielsen, J., Oldham, N., Weaver, K., Jensen, C., and Fradeneck, A. (2022). A neutronics investigation simulating fast reactor environments in the thermal-spectrum advanced test reactor. Nuclear Engineering and Design, 387, 111623. | Publication |
| FY2021 | Duenas, A., Wachs, D., Mignot, G., Reyes, J. N., Wu, Q., and Marcum, W. (2021). Dynamical System Scaling Application to Zircaloy Cladding Thermal Response During Reactivity-Initiated Accident Experiment. Nuclear Science and Engineering, 196(2), 193 208. | Publication |
| FY2021 | Gong, B., Cai, L., Lei, P., Metzger, K.E., Lahoda, E.J., Boylan, F.A., Yang, K., Fay, J., Harp, J., and Lian, J. (2020). Cr-doped U3Si2 composite fuels under steam corrosion. Corrosion Science, 177, 109001. | Publication |
| FY2021 | Gonzales, A., Watkins, J.K., Wagner, A.R., Jaques, B.J., and Sooby, E.S. (2021). Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium silicide. Journal of Nuclear Materials, 553, 153026. | Publication |
| FY2021 | Gouws, A., Hagen, D., Chen, A., Kardoulaki, E., Beaman, J.J., and Kovar, D. Onset of selective laser flash sintering of AlN. United States. | Publication |
| FY2021 | Harp, J.M., Morris, R.N., Petrie, C.M., Burns, J.R., and Terrani, K.A. (2021). Postirradiation examination from separate effects irradiation testing of uranium nitride kernels and coated particles. Journal of Nuclear Materials, 544, 152696. | Publication |
| FY2021 | Ingraci Neto, R.R., McClellan, K.J., Byler, D.D., and Kardoulaki, E. (2021). Controlled current-rate AC flash sintering of uranium dioxide. Journal of Nuclear Materials, 547, 152780. | Publication |
| FY2021 | Kardoulaki, E., Frazer, D.M., White, J.T., Carvajal, U., Nelson, A.T., Byler, D.D., Saleh, T.A., Gong, B., Yao, T., Lian, J., and McClellan, K.J. (2021). Fabrication and thermophysical properties of UO2-UB2 and UO2-UB4 composites sintered via spark plasma sintering. Journal of Nuclear Materials, 544, 152690. | Publication |
| FY2021 | Lee, D., Elward, B., Brooks, P., Umretiya, R., Rojas, J., Bucci, M., Rebak, R.B., and Anderson, M. (2021). Enhanced flow boiling heat transfer on chromium coated zircaloy-4 using cold spray technique for accident tolerant fuel (ATF) materials. Applied Thermal Engineering, 185, 116347. | Publication |
| FY2021 | Moorehead, M., Nelaturu, P., Elbakhshwan, M., Parkin, C., Zhang, C., Sridharan, K., Thoma, D.J., and Couet, A. (2021). High-throughput ion irradiation of additively manufactured compositionally complex alloys. Journal of Nuclear Materials, 547, 152782. | Publication |
| FY2021 | Mouche, P.A., Koyanagi, T., Patel, D., and Katoh, Y. (2021). Adhesion, structure, and mechanical properties of Cr HiPIMS and cathodic arc deposited coatings on SiC. Surface and Coatings Technology, 410, 126939. | Publication |
| FY2021 | Parkin, C., Moorehead, M., Elbakhshwan, M., Hu, J., Chen, W.-Y., Li, M., He, L., Sridharan, K., and Couet, A. (2020). In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation. Acta Materialia, 198, 85-99. | Publication |
| FY2021 | Petrie, C.M., Burns, J.R., Raftery, A.M., Nelson, A.T., and Terrani, K.A. (2019). Separate effects irradiation testing of miniature fuel specimens. Journal of Nuclear Materials, 526, 151783. | Publication |
| FY2021 | Radhakrishnan M, Kombaiah B, Bachhav MN, Nizolek TJ, Wang YQ, Knezevic M, Mara N, Anderoglu O. Layer dissolution in accumulative roll bonded bulk Zr/Nb multilayers under heavy-ion irradiation. J Nucl Mater. 2021;557:153315, | Publication |
| FY2021 | Rietema, C.J., Hassan, M.M., Anderoglu, O., Eftink, B.P., Saleh, T.A., Maloy, S.A., Clarke, A.J., and Clarke, K.D. (2021). Ultrafine intralath precipitation of V(C,N) in 12Cr-1MoWV (wt.%) ferritic/martensitic steel. Scripta Materialia, 197, 113787. | Publication |
| FY2021 | Rietema, C.J., Walker, M.A., Jacobs, T.R., Clarke, A.J., and Clarke, K.D. (2021). High-throughput nitride and interstitial nitrogen analysis in ferritic/martensitic steels via time-of-flight secondary ion mass spectrometry. Materials Characterization, 179, 111357. | Publication |
| FY2021 | Roache, D.C., Bumgardner, C.H., Harrell, T.M., Price, M.C., Jarama, A., Heim, F.M., Walters, J., Maier, B., and Li, X. (2022). Unveiling damage mechanisms of chromium-coated zirconium-based fuel claddings at LWR operating temperature by in-situ digital image correlation. Surface and Coatings Technology, 429, 127909. | Publication |
| FY2021 | Wang, H., Gould, B., Moorehead, M., Haddad, M., Couet, A., and Wolff, S.J. (2022). In situ X-ray and thermal imaging of refractory high entropy alloying during laser directed deposition. Journal of Materials Processing Technology, 299, 117363. | Publication |
| FY2021 | Williams, W.J., Okuniewski, M.A., and Vogel, S.C. et al. (2020). In Situ Neutron Diffraction Study of Crystallographic Evolution and Thermal Expansion Coefficients in U-22.5 at.%Zr During Annealing. JOM, 72, 2042 2050. | Publication |
| FY2021 | Xie, Y., Vogel, S.C., Harp, J.M., Benson, M.T., and Capriotti, L. (2021). Microstructure Evolution of U Zr System in A Thermal Cycling Neutron Diffraction Experiment: Extruded U 10Zr (wt. %). Journal of Nuclear Materials, 544, 152665. | Publication |
| FY2021 | Yang, J., Kardoulaki, E., Zhao, D., Broussard, A., Metzger, K., White, J.T., Sivack, M.R., McClellan, K.J., Lahoda, E.J., and Lian, J. (2021). Uranium nitride (UN) pellets with controllable microstructure and phase fabrication by spark plasma sintering and their thermal-mechanical and oxidation properties. Journal of Nuclear Materials, 557, 153272. | Publication |
| FY2021 | Yin, L., Jurewicz, T.B., Larsen, M., Drobnjak, M., Graff, C.C., Lutz, D.R., and Rebak, R.B. (2021). Uniform corrosion of FeCrAl cladding tubing for accident tolerant fuels in light water reactors. Journal of Nuclear Materials, 554, 153090. | Publication |
| FY2020 | Agarwal, S. et al. Revealing irradiation damage along with the entire damage range in ion-irradiated SiC/SiC composites using Raman spectroscopy. Journal of Nuclear Materials 526 (2019): 151778 | Publication |
| FY2020 | Ali, A., Kim, H.-G., Hattar, K., Briggs, S., Park, D. J., Park, J. H., and Lee, Y. Ion irradiation effects on Cr-coated zircaloy-4 surface wettability and pool boiling critical heat flux. Nucl. Eng. Des. 362 (2020): 110581 | Publication |
| FY2020 | Baker, J. L., Wang, G., Ulrich, T. L., White, J. T., Batista, E. R., Yang, P., Roback, R. C., Park, C., and Xu, H. High-Pressure Structural Behavior and Elastic Properties of U3Si5: A Combined Synchrotron XRD and DFT Study. Journal of Nuclear Materials (2020) | Publication |
| FY2020 | Beausoleil GL, Petrie C, Williams W, Jokisaari A, Capriotti L, Novascone S, Kerr M. Integrating advanced modeling and accelerated testing for a modernized fuel qualification paradigm. Nucl Technol. 2021;207(10):1491-1510 | Publication |
| FY2020 | Brown, N. R., Garrison, B. E., Lowden, R. R., Cinbiz, M. N., and Linton, K. D. Mechanical failure of fresh nuclear grade iron chromium aluminum (FeCrAl) cladding under simulated hot zero power reactivity-initiated accident conditions. Journal of Nuclear Materials (2020):152352 | Publication |
| FY2020 | Bumgardner, C. H., Heim, F. M., Roache, D. C., Jarama, A., Xu, P., Lu, R., Lahoda, E. J., Croom, B. P., Deck, C. P., and Li, X. Unveiling hermetic failure of ceramic tubes by digital image correlation and acoustic emission. Journal of the American Ceramic Society (2019) | Publication |
| FY2020 | Burns, J. R., Hernandez, R., Terrani, K. A., Nelson, A. T., and Brown, N. R. Reactor and fuel cycle performance of light water reactor fuel with 235U enrichments above 5%. Annals of Nuclear Energy, 142 (2020): 107423 | Publication |
| FY2020 | Capps, N., Sweet, R., Wirth, B. D., Nelson, A., Terrani, K. A. Development and demonstration of a methodology to evaluate high burnup fuel susceptibility to pulverization under a loss of coolant transient. Nuclear Engineering and Design 366 (2020): 110744, ISSN 0029-5493 | Publication |
| FY2020 | Capps, N., Yan, Y., Raftery, A., Burns, Z., Smith, T., Terrani, K. A., Yueh, K., Bales, M., and Linton, K. D. Integral LOCA fragmentation test on high-burnup fuel. Nuclear Eng. And Design 367 (2020): 110811 | Publication |
| FY2020 | Capriotti, L., and Harp, J. M. Characterization of a minor actinides bearing metallic fuel pin irradiated in EBR-II. Journal of Nuclear Materials 539 (2020): 152279 | Publication |
| FY2020 | Chichester, H. J. M., Hilton, B. A., Hayes, S. L., Capriotti, L., Medvedev, P. G., and Porter, D. L. (2020). Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels. Journal of Nuclear Materials, 542, 152480. | Publication |
| FY2020 | Cui, Y., Aydogan, E., Gigax, J. G., Wang, Y., Misra, A., Maloy, S. A., Li, N. (2021). In situ micro-pillar compression to examine radiation-induced hardening mechanisms of FeCrAl alloys. Acta Materialia, 202, 255-265. | Publication |
| FY2020 | Dabney, T., Johnson, G., Yeom, H., Maier, B., Walters, J., and Sridharan, K. Experimental Evaluation of Cold Spray FeCrAl Alloys Coated Zirconium-alloy for Potential Accident Tolerant Fuel Cladding. Nuclear Materials and Energy 21 (2019): 100715 | Publication |
| FY2020 | Deng, P., Karadge, M., Rebak, R. B., Gupta, V. K., Prorok, B. C., and Lou, X. The origin and formation of oxygen inclusions in austenitic stainless steels manufactured by laser powder fusion. Additive Manufacturing 35 (2020):101334 | Publication |
| FY2020 | Doyle, P. J. et al. Evaluation of the effects of neutron irradiation on first-generation corrosion mitigation coatings on SiC for accident-tolerant fuel cladding. Journal of Nuclear Materials (2020): 152203 | Publication |
| FY2020 | Doyle, P. J. et al. The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC. Journal of Nuclear Materials (2020):152190 | Publication |
| FY2020 | Doyle, P. J., Zinkle, S., and Raiman, S. S. Hydrothermal corrosion behavior of CVD SiC in high temperature water. Journal of Nuclear Materials (2020):152241 | Publication |
| FY2026 | Khan, S. A., Schulthess, J. L., Burns, J., Pu, X., Woolstenhulme, N., Charit, I., and O'Brien, R. (2026). Microscopic analysis of irradiated UN -Mo -W for space nuclear propulsion. Nuclear Engineering and Design, 446(Part A), 114546. | Publication |
| FY2020 | Eftink, B. P., Quintana, M. E., Romero, T. J., Xu, C., Hoelzer, D. T., Saleh, T. A., and Maloy, S. A. Shear Punch Testing of Neutron-Irradiated HT-9 and 14YWT. JOM 72 (2020) | Publication |
| FY2020 | Evitts, L. J., Middleburgh, S. C., Kardoulaki, E., Ipatova, I., Rushton, M. J. D., and Lee, W. E. Influence of boron isotope ratio on the thermal conductivity of uranium diboride (UB2) and zirconium diboride (ZrB2). Journal of Nuclear Materials (2020):1 7. | Publication |
| FY2020 | Gigax, J., Torrez, A., McCulloch, Q., Kim, H., Li, N., and Maloy, S. Sizing up mechanical testing: Comparison of microscale and mesoscale mechanical testing techniques on a FeCrAl welded tube. J. Mater. Res. (2020) | Publication |
| FY2026 | Petrie, C. M., Doyle, P., Chandler, D., Russell, N. G., Le Coq, A., Geringer, J. W., Deck, C. P., Koyanagi, T., and Katoh, Y. (2026). Radiation-induced bowing of SiC/SiC composites under neutron flux gradients - Integral experimental data for model validation. Journal of Nuclear Materials, 619, 156290. | Publication |
| FY2020 | Gong, B., Yao, T., Lei, P., Lu, C., Metzger, K. E., Lahoda, E. J., Boylan, F. A., Mohamad, A., Harp, J., Nelson, A. T., and Lian, J. U3Si2 and UO2 composites densified by spark plasma sintering for accident tolerant fuels. Journal of Nuclear Materials 534 (2020): 152147 | Publication |
| FY2020 | Gorton, J. P., Lee, S. K., Lee, Y., and Brown, N. R. Comparison of experimental and simulated critical heat flux tests with various cladding alloys: Sensitivity of iron-chromium-aluminum (FeCrAl) to heat transfer coefficients and material properties. Nucl. Eng. Des. 353 (2019): 110295 | Publication |
| FY2020 | Harp, J. M., Capriotti, L., Porter, D. L., and Cole, J. I. U-10Zr and U-5Fs: Fuel/cladding chemical interaction behavior differences. Journal of Nuclear Materials 528 (2020): 151840 | Publication |
| FY2020 | He, M., and Lee, Y. Application of Deep Belief Network for Critical Heat Flux Prediction on Microstructure Surfaces. Nuclear Technology 206 (2020):358 374 | Publication |
| FY2020 | He, M., and Lee, Y. Application of machine learning for prediction of critical heat flux: He, M., and Lee, Y. Revisiting heater size sensitive pool boiling critical heat flux using neural network modeling: Heater length of the half of the Rayleigh-Taylor Instability Wavelength maximizes CHF. Therm. Sci. Eng. Prog. 14 (2019): 100421 | Publication |
| FY2020 | He, M., and Lee, Y. Application of machine learning for prediction of critical heat flux: Support vector machine for data-driven CHF look-up table construction based on sparingly distributed training data points. Nucl. Eng. Des. 338 (2018):189 198 | Publication |
| FY2020 | Heim FM, Daspit JT, Li X. Quantifying the effect of tow architecture variability on the performance of biaxially braided composite tubes. Compos Part B Eng. 2020;201:108383 | Publication |
| FY2020 | Heim, F. M., Daspit, J. T., Holzmond, O. B., Croom, B. P., and Li, X. Analysis of tow architecture variability in biaxially braided composite tubes. Composites Part B: Engineering 190 (2020): 107938 | Publication |
| FY2020 | Jo, H., Yeom, H., Gutierrez, E., Sridharan, K., and Corradini, M. Evaluation of Critical Heat Flux of ATF Candidate Coating Materials in Pool Boiling. Nuclear Engineering and Design 354 (2019): 110166 | Publication |
| FY2020 | Johnson, K. E., Adorno, D. L., Kocevski, V., Ulrich, T. L., White, J. T., Claisse, A., McMurrary, J. W., and Besmann, T. M. Impact of Fission Product Inclusion on Phase Development in U3Si2 Fuel. Journal of Nuclear Materials 537 (2020): 152235 | Publication |
| FY2020 | Kane, K. A., Lee, S. K., Bell, S. B., Brown, N. R., and Pint, B. A. Burst behavior of nuclear grade FeCrAl and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions. Journal of Nuclear Materials 539 (2020): 152256 | Publication |
| FY2020 | Kardoulaki, E., White, J. T., Byler, D. D., Frazer, D. M., Shivprasad, A. P., Saleh, T. A., Gong, B., Yao, T., Lian, J., and McClellan, K. J. Thermophysical and mechanical property assessment of UB2 and UB4 sintered via spark plasma sintering. J. Alloys Compd. 818 (2020): 1 14. | Publication |
| FY2020 | Kocevski, V., Lopes, D. A., Claisse, A. J., and Besmann, T. M. Understanding the interface interaction between U3Si2 fuel and SiC cladding. Nature Communications 11 (1) (2020): 1-8 | Publication |
| FY2020 | Koyanagi, T., Katoh, Y., and Nozawa, T. Design and strategy for next-generation silicon carbide composites for nuclear energy. Journal of Nuclear Materials (2020):152375 | Publication |
| FY2020 | Le Coq, A. G., Morris, R. N., Petrie, C. M., and Burns, J. R. Post-Irradiation Examination Results of Miniature Fuel Specimens Irradiated in the High Flux Isotope Reactor. Transactions of the American Nuclear Society 121 (2019):615-618 | Publication |
| FY2020 | Lee, S. K., Lee, Y., Brown, N. R., and Terrani, K. A. Elucidating the Impact of Flow on Material-Sensitive Critical Heat Flux and Boiling Heat Transfer Coefficients: An Experimental Study with Various Materials. International J. Heat Mass Transf. 158 (2020): 119970 | Publication |
| FY2020 | Lee, S. K., Liu, M., Brown, N. R., Terrani, K. A., and Lee, Y. Effect of Heater Material and Thickness on the Steady-State Flow Boiling Critical Heat Flux. Nuclear Technology 206 (2020): 339 346 | Publication |
| FY2020 | Lee, S. K., Liu, M., Brown, N. R., Terrani, K. A., Blandford, E. D., Ban, H., Jensen, C. B., and Lee, Y. Comparison of steady and transient flow boiling critical heat flux for FeCrAl accident tolerant fuel cladding alloy, Zircaloy, and Inconel. Int. J. Heat Mass Transf. 132 (2019): 643 654 | Publication |
| FY2020 | Losko, A. S., Daemen, L., Hosemann, P., Nakotte, H., Tremsin, A., Vogel, S. C., Wang, P., and Wittman, F. H. Separation of Uptake of Water and Ions in Porous Materials Using Energy Resolved Neutron Imaging. JOM (2020): 1-8 | Publication |
| FY2020 | McCulloch, Q., Gigax, J., and Hosemann, P. Femtosecond laser ablation for mesoscale specimen evaluation. JOM 72(4) (2020): 1694 | Publication |
| FY2020 | McKinney, C., Gerczak, T. J., and Harp, J. Sample Preparation for 3D Characterization of Irradiated Fuel. United States: N. p., 2020. Web. | Publication |
| FY2020 | Mouche, P. A. et al. Characterization of PVD Cr, CrN, and TiN coatings on SiC. Journal of Nuclear Materials 527 (2019): 151781 | Publication |
| FY2020 | Mouche, P. A., and Terrani, K. A. Steam pressure and velocity effects on high temperature silicon carbide oxidation. Journal of the American Ceramic Society 103.3 (2020): 2062-2075 | Publication |
| FY2020 | Peterson, N. E., Malta, D., Vogel, S. C., Clausen, B., Jana, S., Joshi, V. V., and Agnew, S. R. The role of ternary alloying elements in eutectoid transformation of U 10Mo alloy part II. In and ex-situ neutron diffraction-based assessment of eutectoid phase transformation kinetics in U-9.8 Mo-0.2 X alloy (X= Cr, Ni or Co). Journal of Nuclear Materials 540 (2020):152383 | Publication |
| FY2020 | Petrie, C. M., Le Coq, A., Richardson, D., Hobbs, C., Helmreich, G., Burns, J., and Harp, J. Monolithic ATF MiniFuel Sample Capsules Ready for HFIR Insertion. United States: N. p., 2020. Web. | Publication |
| FY2020 | Raiman, S. S., Field, K. G., Rebak, R. B., Yamamoto, Y., and Terrani, K. A. Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding. Journal of Nuclear Materials 536. | Publication |
| FY2020 | Rebak, R. B., Yin, L., and Andresen, P. L. Resistance of ferritic FeCrAl alloys to stress corrosion cracking for light water reactor fuel cladding applications. Corrosion Journal, NACE International | Publication |
| FY2020 | Reed, B., Wang, R., Lu, R. Y., and Qu, J. (2021). Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel. Wear, 466-467, 203578 | Publication |
| FY2020 | Schulthess, J., Woolstenhulme, N., Craft, A., Kane, J., Boulton, N., Chuirazzi, W., Winston, A., Smolinski, A., Jensen, C., Kamerman, D., and Wachs, D. Non-Destructive Post-irradiation Examination Results of the First Modern Fueled Experiments in TREAT. Journal of Nuclear Materials 541 (2020): 152442 | Publication |
| FY2020 | Sooby Wood, E., Moczygemba, C., Robles, G., Acosta, Z., Brigham, B. A., Grote, C. J., Metzger, K. E., and Cai, L. High temperature steam oxidation dynamics of U3Si2 with alloying additions: Al, Cr, and Y. Journal of Nuclear Materials 533 (2020) | Publication |
| FY2020 | Su, G. Y., Wang, C., Zhang, L., Seong, J. H., Phillips, B., Kommayosula, R., and Bucci, M. Investigation of flow boiling heat transfer and boiling crisis on a rough surface using infrared thermometry. International Journal of Heat and Mass Transfer 160 (2020): 120134 | Publication |
| FY2020 | Terrani, K. A., Jolly, B. C., and Harp, J. M. Uranium nitride tristructural-isotropic fuel particle. Journal of Nuclear Materials 531 (2020): 152034 | Publication |
| FY2020 | Ulrich, T. L., Vogel, S. C., Lopes, D. A., Kocevski, V., White, J. T., Sooby, E. S., and Besmann, T. M. Phase stability of U5Si4, Usi, and U2Si3 in the uranium silicon system. Journal of Nuclear Materials 540 (2020): 152353 | Publication |
| FY2020 | Ulrich, T. L., Vogel, S. C., White, J. T., Andersson, D. A., Wood, E. S., and Besmann, T. M. High temperature neutron diffraction investigation of U3Si2. Materialia 9 (2020):100580 | Publication |
| FY2020 | Umretiya, R. V., Elward, B., Lee, D., Anderson, M., Rebak, R. B., and Rojas, J. V. Mechanical and chemical properties of PVD and cold spray Cr-coatings on Zircaloy-4. Journal of Nuclear Materials 541 (2020): 152420 | Publication |
| FY2020 | Umretiya, R. V., Vargas, S., Galeano, D., Mohammadi, R., Castano, C. E., and Rojas, J. V. Effect of surface characteristics and environmental aging on wetting of Cr-coated Zircaloy-4 accident tolerant fuel cladding material. Journal of Nuclear Materials (2020): 152163 | Publication |
| FY2020 | Vogel, S. C., Bourke, M. A., Craft, A. E., Harp, J. M., Kelsey, C. T., Lin, J., Long, A. M., Losko, A. S., Hosemann, P., McClellan, K. J., and Roth, M. Advanced Postirradiation Characterization of Nuclear Fuels Using Pulsed Neutrons. JOM 72(1) (2020): 187-196 | Publication |
| FY2020 | Vogel, S. C., Fernandez, J. C., Gautier, D. C., Mitura, N., Roth, M., and Schoenberg, K. F. Short-Pulse Laser-Driven Moderated Neutron Source. EPJ Web of Conferences 231 (2020): 01008). EDP Sciences | Publication |
| FY2020 | Woolstenhulme, N., Jensen, C., Folsom, C., Armstrong, R., Yoo, J., and Wachs, D. (2020). Thermal-hydraulic and engineering evaluations of new LOCA testing methods in TREAT. Nuclear Technology, 207(5), 637-652 | Publication |
| FY2020 | Yao, T., Gong, B., Lei, P., Lu, C., Xu, P., Lahoda, E., and Lian, J. (2020). UO2 + 5 vol% ZrB2 nano composite nuclear fuels with full boron retention and enhanced oxidation resistance. Ceramics International, 46(17), 26486-26491 | Publication |
| FY2020 | Yeom H, Gutierrez E, Jo H, Zhou Y, Mondry K, Sridharan K, Corradini M. Pool boiling critical heat flux studies of accident tolerant fuel cladding materials. Nucl Eng Des. 2020;370:110919 | Publication |
| FY2019 | Abdul-Jabbar, N. M., and White, J. T. (2019). Processing and characterization of U3Si2 at the MiniFuel scale (Report No. LA-UR-19-26376). Los Alamos National Laboratory. | Publication |
| FY2019 | Abdul-Jabbar, N. M., Grote, C. J., and White, J. T. (2019). Assessment of thermophysical property characterization of MiniFuel scale geometries (Report No. LA-UR-19-24287). Los Alamos National Laboratory. | Publication |
| FY2019 | Ang, C., Carpenter, D., Terrani, K., and Katoh, Y. (2018, November). Preliminary characterization and projections of PVD coatings on SiC cladding for light water reactors. In Proceedings of the 42nd International Conference on Advanced Ceramics and Composites, Ceramic Engineering and Science Proceedings 3, 119. John Wiley and Sons. | Publication |
| FY2019 | Ang, C., Kemery, C., and Katoh, Y. (2018). Electroplating chromium on CVD SiC and SiCf-SiC advanced cladding via PyC compatibility coating. Journal of Nuclear Materials, 503, 245-249. | Publication |
| FY2019 | Aydogan, E., Rietema, C. J., Carvajal-Nunez, U., Vogel, S. C., Li, M., and Maloy, S. A. (2019). Effect of high-density nanoparticles on recrystallization and texture evolution in ferritic alloys. Crystals, 9(3), 172. | Publication |
| FY2019 | Aydogan, E., Weaver, J. S., Carvajal-Nunez, U., Schneider, M. M., Gigax, J. G., Krumwiede, D. L., Hosemann, P., Saleh, T. A., Mara, N. A., Hoelzer, D. T., Hilton, B., and Maloy, S. A. (2019). Response of 14YWT alloys under neutron irradiation: A complementary study on microstructure and mechanical properties. Acta Materialia, 167, 181-196. | Publication |
| FY2019 | Beausoleil, G. L., Povirk, G. L., and Curnutt, B. J. (2020). A revised capsule design for the accelerated testing of advanced reactor fuels. Nuclear Technology, 206(3), 444-457. | Publication |
| FY2019 | Beausoleil, G., Cappia, F., Emerson, L. A., Murray, D., Sell, D., Christensen, C., Winston, A., Wahlquist, D., Bachhav, M., and Miller, B. (2019). Separate effects testing in TREAT for ATF fuels. Portions of this work were presented in a poster session at The Mineral, Metals, and Materials Society (TMS) 2019 annual meeting in San Antonio, TX. | |
| FY2019 | Benson, M. T., Xie, Y., He, L., Tolman, K. R., King, J. A., Harp, J. M., Mariani, R. D., Hernandez, B. J., Murray, D. J., and Miller, B. D. (2019). Microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln. Journal of Nuclear Materials, 518, 287-297. | Publication |
| FY2019 | Bess, J. D., Woolstenhulme, N. E., Jensen, C. B., Parry, J. R., and Hill, C. M. (2019). Nuclear characterization of a general-purpose instrumentation and materials testing location in TREAT. Annals of Nuclear Energy, 124, 270-294. | Publication |
| FY2019 | Braase, L., and Richardson, K. (2019). GAIN-EPRI-NEI Advanced Fuels Workshop Summary Report (INL/EXT-19-55476). Idaho National Laboratory. | Publication |
| FY2019 | Burns, J. R., Petrie, C. M., and Chandler, D. (2019). Burnup calculation methodology for a small-scale fuel irradiation experiment in the High Flux Isotope Reactor (HFIR). Transactions of the American Nuclear Society, 120, 841-844. | Publication |
| FY2019 | Curnutt, B. J., and Beausoleil, G. L. (2019, June). The Fission Accelerated Steady State Test (FAST) - A revised capsule design for the accelerated testing of advanced reactor fuels. In Proceedings of the American Nuclear Society (ANS) Annual Meeting, Minneapolis, MN. | Publication |
| FY2019 | Czerniak, L., Lahoda, E., Sivack, M., Lyons, J., Byers, W., Wang, G., Oelrich, R., Xu, P., and Lu, R. (2019, September). Development of silicon carbide as a nuclear fuel cladding. Submitted to TopFuel 2019, Seattle, WA. | |
| FY2019 | Dabney, T., Johnson, G., Maier, B., Yeom, H., and Sridharan, K. (2019, June). Development of cold spray FeCrAl coatings for accident tolerant fuel. In Proceedings of the 2019 American Nuclear Society Annual Conference, 120(1), 387-390, Minneapolis, MN. | Publication |
| FY2019 | Di Lemma, F., et al. (2019, November 17-21). Metallic fast reactor separate effect studies for fuel safety. Paper presented at the American Nuclear Society Winter Meeting, Washington, D.C. | |
| FY2019 | Di Lemma, F., et al. (2019, October 6-10). Metallic fast reactor separate effect studies for fuel safety. Paper presented at MiNES (Materials in Nuclear Energy Systems), Baltimore, MD. | |
| FY2019 | Franceschini, F., Kucukboyaci, V., Stucker, D., Lahoda, E. J., and Karouta, Z. (2019, September 22-27). Modeling of Westinghouse advanced fuels EnCore and ADOPT with the CASL tools. In Proceedings of TopFuel 2019, Seattle, WA, 153-160. | Publication |
| FY2019 | Frazer, D., White, J. T., and Saleh, T. A. (2019). Nanomechanical properties of high uranium density fuels (Report No. NTRD-M3FT-19LA020201026, LA-UR-19-27315). Los Alamos National Laboratory. | |
| FY2019 | Garrison, B., Howell, M., Cinbiz, M. N., Gussev, M., and Linton, K. (2019, September 22-27). Length-dependence of severe accident test station integral testing. Results from this work presented presented at the 2019 ANS Top Fuel Conference, Seattle WA. | Publication |
| FY2019 | Gigax, J. G., Kim, H., Aydogan, E., Price, L. M., Wang, X., Maloy, S. A., Garner, F. A., and Shao, L. (2019). Impact of composition modification induced by ion beam Coulomb-drag effects on the nanoindentation hardness of HT9. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 444, 68-73. | Publication |
| FY2019 | Grote, C. (2019, July 17). Assessment of viability of scaled annular pellet fabrication technologies (Report No. LA-UR-19-27197). Los Alamos National Laboratory. | Publication |
| FY2019 | Heim, F. M., Croom, B. P., Bumgardner, C., and Li, X. (2018). Scalable measurements of tow architecture variability in braided ceramic composite tubes. Journal of the American Ceramic Society, 101(9), 4297-4307. | Publication |
| FY2019 | Jiao, Z., Taller, S., Field, K., Yeli, G., Moody, M. P., and Was, G. S. (2018). Microstructure evolution of T91 irradiated in the BOR60 fast reactor. Journal of Nuclear Materials, 504, 122-134. | Publication |
| FY2019 | Jolly, B., Kato, Y., Lowden, R., Nelson, A., Schumacher, A., and Cooley, K. (2019). Development of vapor processing capability for advanced SiC/SiC composites (Milestone Report No. ORNL/SPR-2019/1125). Oak Ridge National Laboratory. | |
| FY2019 | Lin, Y. P., Fawcett, R. M., DeSilva, S. S., Lutz, D. R., Yilmaz, M. O., Davis, P., Rand, R. A., Cantonwine, P. E., Rebak, R. B., Dunavant, R., and Satterlee, N. (2018, September 30-October 4). Path towards industrialization of enhanced accident tolerant fuel. Paper A0141 presented at TopFuel 2018, Prague, European Nuclear Society. | Publication |
| FY2019 | Lu, R. Y., Walters, J. L., and Qu, J. (2019, September). Assessment of wear coefficients of accident tolerance fuel claddings with coated materials. TopFuel 2019, Seattle, WA. | |
| FY2019 | Lyons, J. L., Partezana, J., Byers, W. A., Wang, G., Parsi, A., Walters, J., Romero, J., Mueller, A. J., Shah, H., and Oelrich, R. Jr. (2019, September 22-27). Westinghouse chromium-coated zirconium alloy cladding development and testing. In Proceedings of Top Fuel 2019 (pp. 8-14), Seattle, WA. | Publication |
| FY2019 | Maier, B. R., Yeom, H., Johnson, G., Dabney, T., Hu, J., Baldo, P., Li, M., and Sridharan, K. (2018). In situ TEM investigation of irradiation-induced defect formation in cold spray Cr coatings for accident tolerant fuel applications. Journal of Nuclear Materials, 512, 320-323. | Publication |
| FY2019 | Maier, B., Yeom, H., Johnson, G., Dabney, T., Walters, J., Xu, P., Romero, J., Shah, H., and Sridharan, K. (2019). Development of cold spray chromium coatings for improved accident tolerant zirconium-alloy cladding. Journal of Nuclear Materials, 519, 247-254. | Publication |
| FY2019 | Massey, C. P., Dryepondt, S. N., Edmondson, P. D., Frith, M. G., Littrell, K. C., Kini, A., Gault, B., Terrani, K. A., and Zinkle, S. J. (2019). Multiscale investigations of nanoprecipitate nucleation, growth, and coarsening in annealed low-Cr oxide dispersion strengthened FeCrAl powder. Acta Materialia, 166, 1-17. | Publication |
| FY2019 | Massey, C. P., Hoelzer, D. T., Seibert, R. L., Edmondson, P. D., Kini, A., Gault, B., Terrani, K. A., and Zinkle, S. J. (2019). Microstructural evaluation of a Fe-12Cr nanostructured ferritic alloy designed for impurity sequestration. Journal of Nuclear Materials, 522, 111-122. | Publication |
| FY2019 | Matthews, C., Bieberdorf, N., Capolungo, L., and Andersson, D. (2019). Combined visco-plasticity and swelling in metallic nuclear fuel (Report No. LA-UR-19-25483). Los Alamos National Laboratory. | |
| FY2019 | Oelrich, R., Karoutas, Z., Xu, P., Romero, J., Shah, H., Walters, J., Lahoda, E., Sivack, M., Lyons, J., Czerniak, L., Boylan, F., ?vali, R., Bowman, A., Limb ¤ck, M., Claisse, A., and Wright, J. (2019, September 22-27). Overview of Westinghouse lead EnCore accident tolerant fuel program. In Proceedings of Top Fuel 2019 (pp. 192-196), Seattle, WA. | Publication |
| FY2019 | Petrie, C. M., Burns, J. R., Raftery, A. M., Nelson, A. T., and Terrani, K. A. (2019). Separate effects irradiation testing of miniature fuel specimens. Journal of Nuclear Materials, 526, 151783. | Publication |
| FY2019 | Petrie, C. M., Burns, J., Morris, R., and Terrani, K. A. (2017). Miniature fuel irradiations in the High Flux Isotope Reactor. In Proceedings of the 40th Enlarged Halden Programme Group Meeting, Lillehammer, Norway. | Publication |
| FY2019 | Prakash, N., Matthews, C., Versino, D., and Unal, C. (2019). A general constitutive framework for the combined creep, plasticity, and swelling behavior of nuclear fuels in an implicit hypoelastic formulation (Report No. LA-UR-20166). Los Alamos National Laboratory. | Publication |
| FY2019 | Rebak, R. B., Blair, R. J., and Gupta, V. K. (2019). Corrosion evaluation of iron-chromium-aluminum alloys in used fuel cooling pools. Paper No. C2019-12944, 1-14. NACE International. Nashville, TN. | Publication |
| FY2019 | Rebak, R. B., Gupta, V. K., Drobnjak, M., Keck, D. J., and Dolley, E. J. (2018, September 30-October 4). Overcoming sensitization in welds using FeCrAl alloys. Paper A0052 presented at TopFuel 2018, Prague, European Nuclear Society. | Publication |
| FY2019 | Rebak, R. B., Huang, S., Schuster, M., Buresh, S. J., and Dolley, E. J. (2019, July). Fabrication and mechanical aspects of using FeCrAl for light water reactor fuel cladding. Paper PVP2019-93128 presented at the PVP ASME Conference, San Antonio, TX. | Publication |
| FY2019 | Rebak, R. B., Jurewicz, T. B., and Dolley, E. J. (2018, September 30-October 4). Assessing the electrochemical behavior of ferritic FeCrAl in high temperature water. Paper A0053 presented at TopFuel 2018, Prague, European Nuclear Society. | Publication |
| FY2019 | Rebak, R. B., Jurewicz, T. B., and Kim, Y.-J. (2019). Electrochemical behavior of accident tolerant fuel cladding materials under simulated light water reactor conditions. In ASTM STP 1609: Advances in electrochemical techniques for corrosion monitoring (pp. 231-243). | Publication |
| FY2019 | Richardson, M. D., Helmreich, G. W., Raftery, A. M., and Nelson, A. T. (2019). Resolution capabilities for measurement of fuel swelling using tomography (Report No. ORNL/SPR-2019/1071). Oak Ridge National Laboratory. | Publication |
| FY2019 | Schley, R. S., Hurley, D. H., Hua, Z., and Reese, S. J. (2019, February 9-14). In-pile instrument to measure changes in grain microstructure. In Proceedings of Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies (NPICandHMIT 2019) (pp. 1135-1142), Orlando, FL. | Publication |
| FY2019 | Schuster, M., Dolley, E. J., Jurewicz, T. B., and Rebak, R. B. (2019, August 18-22). Environmental degradation resistance of ATF FeCrAl cladding tube specimens during the fuel cycle. In Proceedings of the 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (pp. 331-338), Boston, MA. | Publication |
| FY2019 | Seibert, R. L., Burns, J. R., Kiggans, J. O., and Terrani, K. A. (2019). Fabrication of fully ceramic microencapsulated compacts for miniature fuel specimen irradiation. Transactions of the American Nuclear Society, 121(1), 741-743. | Publication |
| FY2019 | Seibert, R. L., Kiggans, J. O., and Terrani, K. A. (2019, April). Fabrication of fully ceramic microencapsulated fuel pellets for HFIR irradiation (Report No. ORNL/SPR-2019/1133). Oak Ridge National Laboratory. | |
| FY2019 | Seibert, R. L., Terrani, K. A., Kiggans, J. O., McMurray, J. W., Jolly, B. C., Petrie, C. M., and Nelson, A. T. (2019, January). Fabrication and irradiation test plan for fully ceramic microencapsulated fuels (Report No. ORNL/TM-2019/1088). Oak Ridge National Laboratory. | Publication |
| FY2019 | Taller, S., Jiao, Z., Field, K., and Was, G. S. (2019). Emulation of fast reactor irradiated T91 using dual ion beam irradiation. Journal of Nuclear Materials, 527, 151831. | Publication |
| FY2019 | Ulrich, T. L., Vogel, S. C., White, J. T., Andersson, D. A., Wood, E. S., and Besmann, T. M. Temperature-dependent crystal structure of U3Si2 by high temperature neutron diffraction. Acta Materialia. | |
| FY2019 | Vogel, S. C., Wilson, T. L., and White, J. T. (2018, August 17). Crystal structure evolution of U-Si nuclear fuel phases as a function of temperature (Report No. LA-UR-18-28584). Los Alamos National Laboratory. | Publication |
| FY2019 | Vogel, S. C., Wilson, T. L., Wood, E. S., White, J. T., and Besmann, T. M. (2019, September 22-27). Temperature-dependent crystal structure of U3Si2 by high-temperature neutron diffraction. In Global 2019 Proceedings (pp. 1062-1069), Seattle, WA. | Publication |
| FY2019 | Williams, W. J., Hale, C., Sikik, E., Sprenger, M., Borghmans, G., Wachs, D. M., Van den Berghe, S., Okuniewski, M. A., Maddock, T., and Boer, B. (2019). Thermal-hydraulics and neutronics overview of the DISECT experiment. Transactions of the American Nuclear Society, 120(1), 348-351. | Publication |
| FY2019 | Williams, W. J., Wachs, D. M., Okuniewski, M. A., and van den Berghe, S. (2020). Assessment of swelling and constituent redistribution in uranium-zirconium fuel using phenomena identification and ranking tables (PIRT). Annals of Nuclear Energy, 136, 107016. | Publication |
| FY2019 | Wilson, T. L., Besmann, T. M., Vogel, S. C., and White, J. T. (2019). Crystal structure characterization of uranium-silicides accident tolerant fuel by high temperature neutron diffraction. In Advances in X-ray Analysis (Vol. 63). Proceedings of the 68th Denver X-ray Conference, Volume 63, Lombard, Illinois, U.S.A., August 5-9, 2019. | Publication |
| FY2019 | Wood, E. S., Moczygemba, C., Robles, G., Nesloney, S., Grote, C., Cai, L., Xu, P., and Lahoda, E. (2019, September). Fabrication and steam oxidation testing of alloyed uranium silicide fuels. Submitted to TopFuel 2019, Seattle, WA. | |
| FY2019 | Woolstenhulme, N., Baker, C., Bess, J., Chapman, D., Dempsey, D., Hill, C., Jensen, C., and Snow, S. (2018). New capabilities for in-pile separate effects tests in TREAT. In Transactions of the American Nuclear Society Summer Meeting, Philadelphia, PA. | |
| FY2019 | Woolstenhulme, N., Baker, C., Jensen, C., Chapman, D., Imholte, D., Oldham, N., Hill, C., and Snow, S. (2019). Development of irradiation test devices for transient testing. Nuclear Technology, 205(10), [Special issue on restarting transient reactor test facility]. | Publication |
| FY2019 | Woolstenhulme, N., Bess, J., Calderoni, P., Heidrich, B., Hurley, D., Jensen, C., Schley, R., and Tsai, K. (2019, June 9-13). Overview of I2 irradiation deployment activities in TREAT. In Proceedings of the American Nuclear Society Annual Meeting, 120(1), 280-282. | Publication |
| FY2019 | Woolstenhulme, N., Fleming, A., Holschuh, T., Jensen, C., Kamerman, D., and Wachs, D. (2020). Core-to-specimen energy coupling results of the first modern fueled experiments in TREAT. Annals of Nuclear Energy, 140, 107117. | Publication |
| FY2019 | Wozniak, N. R., White, J. T., Nolen, B. P., and Wermer, J. R. (2019, February 22). Assessment of feedstock synthesis routes for high density fuels (Report No. FT-19LA02020102). | |
| FY2019 | Xie, Y., Benson, M. T., He, L., King, J. A., Mariani, R. D., and Murray, D. J. (2019). Diffusion behaviors between metallic fuel alloys with Pd addition and Fe. Journal of Nuclear Materials, 525, 111-124. | Publication |
| FY2019 | Yeom, H., Dabney, T., Johnson, G., Maier, B., and Sridharan, K. (2019). Oxidation of cold spray Cr coatings in high temperature steam environments. In Proceedings of the 2019 American Nuclear Society Annual Conference, 120(1), 383-386. | Publication |
| FY2019 | Zheng, C., Ke, J.-H., Maloy, S. A., and Kaoumi, D. (2019). Correlation of in-situ transmission electron microscopy and microchemistry analysis of radiation-induced precipitation and segregation in ion irradiated advanced ferritic/martensitic steels. Scripta Materialia, 162, 460-464. | Publication |
| FY2018 | Ševeček, M., Gurgen, A., Seshadri, A., Che, Y., Wagih, M., Phillips, B., Champagne, V., and Shirvan, K. (2018). Development of Cr cold spray-coated fuel cladding with enhanced accident tolerance. Nuclear Engineering and Technology, 50(2), 229-236. | Publication |
| FY2018 | Arndt, J. L., Lahoda, E. J., and Oelrich, R. L. (2018, June 3-6). Westinghouse accident tolerant fuel and its benefits for CANDU reactors. In Proceedings of the 38th Annual Conference of the Canadian Nuclear Society, Saskatoon, SK, Canada. | Publication |
| FY2018 | Aydogan, E., El-Atwani, O., Takajo, S., Vogel, S. C., and Maloy, S. A. (2018). High temperature microstructural stability and recrystallization mechanisms in 14YWT alloys. Acta Materialia, 148, 467-481. | Publication |
| FY2018 | Benson, M. T., He, L., King, J. A., and Mariani, R. D. (2018). Microstructural characterization of annealed U-12Zr-4Pd and U-12Zr-4Pd-5Ln: Investigating Pd as a metallic fuel additive. Journal of Nuclear Materials, 502, 106-112. | Publication |
| FY2018 | Benson, M. T., He, L., King, J. A., Mariani, R. D., Winston, A. J., and Madden, J. W. (2018). Microstructural characterization of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln. Journal of Nuclear Materials, 508, 310-318. | Publication |
| FY2018 | Benson, M. T., King, J. A., and Mariani, R. D. (2018). Investigation of tin as a fuel additive to control FCCI. In TMS 2018 147th Annual Meeting and Exhibition Supplemental Proceedings (pp. 695-702). The Minerals, Metals and Materials Series. Springer, Cham. | Publication |
| FY2018 | Benson, M. T., Xie, Y., King, J. A., Tolman, K. R., Mariani, R. D., Charit, I., Zhang, J., Short, M. P., Choudhury, S., Khanal, R., and Jerred, N. (2018). Characterization of U-10Zr-2Sn-2Sb and U-10Zr-2Sn-2Sb-4Ln to assess Sn+Sb as a mixed additive system to bind lanthanides. Journal of Nuclear Materials, 510, 210-218. | Publication |
| FY2018 | Bischoff, J., Delafoy, C., Vauglin, C., Barberis, P., Roubeyrie, C., Perche, D., Duthoo, D., Schuster, F., Brachet, J.-C., Schweitzer, E. W., and Nimishakavi, K. (2018). AREVA NP's enhanced accident-tolerant fuel developments: Focus on Cr-coated M5 cladding. Nuclear Engineering and Technology, 50(2), 223-228. | Publication |
| FY2018 | Capps, N., Mai, A., Kennard, M., and Liu, W. (2018). PCI analysis of Zircaloy coated clad under LWR steady state and reactor startup operations using BISON fuel performance code. Nuclear Engineering and Design, 332, 383-391. | Publication |
| FY2018 | Carvajal-Nunez, U., et al. (2018). Mechanical properties of uranium silicides by nanoindentation and finite elements modeling. The Journal of The Minerals, Metals, and Materials Society, 70, 203-208. | Publication |
| FY2018 | Che, Y., Pastore, G., Hales, J., and Shirvan, K. (2018). Modeling of Cr2O3-doped UO2 as a near-term accident tolerant fuel for LWRs using the BISON code. Nuclear Engineering and Design, 337, 271-278. | Publication |
| FY2018 | Chipaux, R., Cecilia, G., Beauvy, M., and Troc, R. (1986). Capacite thermique a haute temperature de UBe13, ThBe13 et UB4. Journal of Less-Common Metals, 121, 347-351. | |
| FY2018 | Cinbiz, M. N., Brown, N. R., Lowden, R. R., Gussev, M., Linton, K., and Terrani, K. A. (2018). Report on design and failure limits of SiC/SiC and FeCrAl ATF cladding concepts under RIA (ORNL/LTR-2018/521 NT-M3FT-2018-204032). Oak Ridge National Laboratory. | Publication |
| FY2018 | Csontos, A., Whitt, J., Carmack, J., Gavrilas, M., Williams, J., and Lahoda, E. (2018, June 18-21). Panel discussion on ATF implementation in the nuclear industry. In Proceedings of the American Nuclear Society (ANS) Meeting, Philadelphia, PA. | |
| FY2018 | Dahl, P., Kaus, I., Zhao, Z., Johnsson, M., Nygren, M., Wiik, K., Grande, T., and Einarsrud, M.-A. (2007). Densification and properties of zirconia prepared by three different sintering techniques. Ceramics International, 33(8), 1603-1610. | Publication |
| FY2018 | Dancausse, J.-P., Gering, E., Heathman, S., Benedict, U., Gerward, L., Olsen, S. S., and Hulliger, F. (1992). Compression study of uranium borides UB2, UB4 and UB12 by synchrotron X-ray diffraction. Journal of Alloys and Compounds, 189(2), 205-208. | Publication |
| FY2018 | Deck, C. P., Khalifa, H. E., Jacobsen, G. M., Shedder, J., Zhang, J., Bacalski, C., Stone, J. G., Shih, C., Vasudevamurthy, G., Lahoda, E., and Back, C. A. (2018, January 23). Development of engineered SiC-SiC accident tolerant fuel cladding. In Proceedings of the 42nd International Conference and Expo on Advanced Ceramics and Composites, Daytona Beach, Florida. | Publication |
| FY2018 | Deck, C. P., Khalifa, H. E., Jacobsen, G., Sheeder, J., Shapovalov, K., Gonderman, S., Song, E., Gazza, J., Back, C. A., Xu, P., Boylan, F., and Jacko, R. (2018, September 30 - October 4). Demonstration of engineered multi-layered SiC-SiC cladding with enhanced accident tolerance. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic. | Publication |
| FY2018 | Demuynck, M., Erauw, J.P., Van der Biest, O., Delannay, F., and Cambier, F. (2012). Densification of alumina by SPS and HP: A comparative study. Journal of the European Ceramic Society, 32(9), 1957-1964. | Publication |
| FY2018 | Deng, Y., Shirvan, K., Wu, Y., and Su, G. (2018). Probabilistic view of SiC/SiC composite cladding failure based on full core thermo-mechanical response. Journal of Nuclear Materials, 507, 24-37. | Publication |
| FY2018 | Dryepondt, S., Massey, C. P., Gussev, M. N., Linton, K. D., and Terrani, K. A. (2018). Tensile strength and steam oxidation resistance of ODS FeCrAl sheet and tubes (ORNL Report TM-2018/870). Oak Ridge National Laboratory. | Publication |
| FY2018 | Dryepondt, S., Unocic, K. A., Hoelzer, D. T., Massey, C. P., and Pint, B. A. (2018). Development of low-Cr ODS FeCrAl alloys for accident-tolerant fuel cladding. Journal of Nuclear Materials, 501, 59-71. | Publication |
| FY2018 | Fink, J. K. (2000). Thermophysical properties of uranium dioxide. Journal of Nuclear Materials, 279(1), 1-18. | Publication |
| FY2018 | Franceschini, F., King, J., Lahoda, E., Oelrich, B., and Ray, S. (2018, April 22-28). Implementation of Westinghouse ATF to extend cycle length of current PWR to 24-month cycles. In Reactor physics paving the way towards more efficient systems (PHYSOR 2018). Cancun, Q. R. (Mexico): Sociedad Nuclear Mexicana. | Publication |
| FY2018 | Gofryk, K. (2018). Effect of off-stoichiometry and grain size on thermal transport in uranium dioxide. Talk given at the Nuclear Materials Conference (NuMat 2018), Seattle, WA. | |
| FY2018 | Gurgen, A., and Shirvan, K. (2018). Estimation of coping time in pressurized water reactors for near term accident tolerant fuel claddings. Nuclear Engineering and Design, 337, 38-50. | Publication |
| FY2018 | Jossou, E., Malakkal, L., Szpunar, B., Oladimeji, D., and Szpunar, J. A. (2017). A first principles study of the electronic structure, elastic and thermal properties of UB2. Journal of Nuclear Materials, 490, 41-48. | Publication |
| FY2018 | Karoutas, Z., Luangdilok, W., Shockling, M., Schneider, R., Lahoda, E., and Xu, P. (2018). Update on Westinghouse benefits of ENCORE ® fuel. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic, Sep 30 - Oct 4, 2018. | Publication |
| FY2018 | Koyanagi, T., Katoh, Y., Singh, G., Petrie, C., Deck, C., and Terrani, K. (2018, January 23). Post-irradiation examination of SiC tubes neutron irradiated under a radial high heat flux. In Proceedings of the 42nd International Conference and Expo on Advanced Ceramics and Composites, Daytona Beach, Florida. | Publication |
| FY2018 | Lahoda, E. (2017, November 1). Approaches for accelerating licensing of ATF products. Presentation at the American Nuclear Society, Washington, D.C. | |
| FY2018 | Lahoda, E. (2017, October 10). Westinghouse accident tolerant fuel materials. Presentation at the Materials Science and Technology Meeting, Pittsburgh, PA. | |
| FY2018 | Lin, Y.-P., Fawcett, R. M., Desilva, S., Luz, D. R., Yilmaz, M. O., Davis, P., Rand, R., Cantonwine, P. E., Rebak, R. B., Dunavant, R., and Satterlee, N. (2018, September 30-October 4). Path towards industrialization of enhanced accident tolerant fuel. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic. | Publication |
| FY2018 | Long, Y., Kersting, P. J., Linsuain, O., Crede, T. M., and Oelrich, R. L. (2018, September 30-October 4). Fuel performance analysis of EnCore ® fuel. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic. | Publication |
| FY2018 | Maier, B. R., Yeom, H., Johnson, G. O., Dabney, T., Walters, J., Romero, J., Shah, H., Xu, P., and Sridharan, K. (2018). Development of cold spray coatings for accident-tolerant fuel cladding in light water reactors. Journal of Minerals, Metals, and Materials Society (JOM), 70(2), 198-202. | Publication |
| FY2018 | Massey, C. P., Dryepondt, S. N., Edmondson, P. D., Terrani, K. A., and Zinkle, S. J. (2018). Influence of mechanical alloying and extrusion conditions on the microstructure and tensile properties of low-Cr ODS FeCrAl alloys. Journal of Nuclear Materials, 512, 227-238. | Publication |
| FY2018 | Matthews, C., Stevens, G., and Unal, C. (2018, June 17-21). Calibration of Zr redistribution models for metallic fuel in BISON. In Transactions of the American Nuclear Society Annual Meeting, Philadelphia, PA. | Publication |
| FY2018 | McMurray, J. W., and Besmann, T. M. (2018). Thermodynamic modeling of nuclear fuel materials. In W. Andreoni and S. Yip (Eds.), Handbook of materials modeling. Springer | Publication |
| FY2018 | McMurray, J. W., Kiggans, J. O., Helmreich, G. W., and Terrani, K. A. (2018). Production of near-full density uranium nitride microspheres with a hot isostatic press. Journal of the American Ceramic Society, 101(10), 4492-4497. | Publication |
| FY2018 | Muta, H., Kurosaki, K., Uno, M., and Yamanaka, S. (2008). Thermal and mechanical properties of uranium nitride prepared by SPS technique. Journal of Materials Science, 43, 6429-6434. | Publication |
| FY2018 | Oelrich, R., Ray, S., Karoutas, Z., Xu, P., Romero, J., Shah, H., Lahoda, E., and Boylan, F. (2018, September 30-October 4). Overview of Westinghouse lead accident tolerant fuel program. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic. | Publication |
| FY2018 | Oelrich, R., Xu, P., Lahoda, E., and Deck, C. (2018, June 18-21). Update on Westinghouse EnCore ® accident tolerant fuel program. In Proceedings of the American Nuclear Society (ANS) Meeting, 118(1), 1311-1313, Philadelphia, PA. | Publication |
| FY2018 | Pal, S., Alam, M. E., Maloy, S. A., Hoelzer, D. T., and Odette, G. R. (2018). Texture evolution and microcracking mechanisms in as-extruded and cross-rolled conditions of a 14YWT nanostructured ferritic alloy. Acta Materialia, 152, 338-357. | Publication |
| FY2018 | Petrie, C. M., Burns, J. R., Morris, R. N., and Terrani, K. A. (2018). Accelerated irradiation testing of miniature fuel specimens. Transactions of the American Nuclear Society, 118, 1476-1479. | Publication |
| FY2018 | Petrie, C. M., Burns, J. R., Morris, R. N., Smith, K. R., Le Coq, A. G., and Terrani, K. A. (2018). Irradiation of miniature fuel specimens in the High Flux Isotope Reactor (Report No. ORNL/SR-2018/844). Oak Ridge National Laboratory. | |
| FY2018 | Petrie, C. M., Koyanagi, T., Howard, R. H., Field, K. G., Burns, J. R., and Terrani, K. A. (2018, September 30-October 4). Accelerated irradiation testing of miniature nuclear fuel and cladding specimens. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic. | Publication |
| FY2018 | Raftery, A. M., Morris, R. N., Smith, K. R., Helmreich, G. W., Petrie, C. M., Terrani, K. A., and Nelson, A. T. (2018). Development of a characterization methodology for post-irradiation examination of miniature fuel specimens (Report No. ORNL/SPR-2018/918). Oak Ridge National Laboratory. | Publication |
| FY2018 | Ray, S. (2017, October 31). The need for hot cells for nuclear RandD - The role of hot cells in new fuel development. Presentation at the American Nuclear Society, Washington, D.C. | |
| FY2018 | Rebak, R. B. (2018). Versatile oxide films protect FeCrAl alloys under normal operation and accident conditions in light water power reactors. JOM, 70, 176-185. | Publication |
| FY2018 | Rebak, R. B., Gupta, V. K., and Larsen, M. (2018). Oxidation characteristics of two FeCrAl alloys in air and steam from 800 °C to 1300 °C. JOM, 70, 1484-1492. | Publication |
| FY2018 | Scott, S. M., Yao, T., Lu, F., Xin, G., Zhu, W., and Lian, J. (2017). Fabrication of lanthanum-doped thorium dioxide by high-energy ball milling and spark plasma sintering. Journal of Nuclear Materials, 485, 207-215. | Publication |
| FY2018 | Seshadri, A., and Shirvan, K. (2018). Quenching heat transfer analysis of accident tolerant coated fuel cladding. Nuclear Engineering and Design, 338, 5-15. | Publication |
| FY2018 | Seshadri, A., Phillips, B., and Shirvan, K. (2018). Towards understanding the effects of irradiation on quenching heat transfer. International Journal of Heat and Mass Transfer, 127(Part B), 1087-1095. | Publication |
| FY2018 | Sheeder, J., Gonderman, S., Jacobsen, G., Khalifa, H. E., Shih, C., Song, E., Shapovalov, K., and Deck, C. P. (2018). Non-destructive evaluation of sealed SiC-SiC composite cladding structures using X-ray computed tomography, pycnometry, and helium leak testing. In Proceedings of the 42nd International Conference and Expo on Advanced Ceramics and Composites, Daytona Beach, FL, Jan 21-26, 2018. | Publication |
| FY2018 | Shrestha, K., Yao, T., Lian, J., Antonio, D., Sessim, M., Tonks, M. R., and Gofryk, K. (2019). The grain-size effect on thermal conductivity of uranium dioxide. Journal of Applied Physics, 126(12), 125116. | Publication |
| FY2018 | Squires, L. N., King, J. A., Fielding, R. S., and Lessing, P. (2018). Isolation of high purity americium metal via distillation. Journal of Nuclear Materials, 500, 26-32. | Publication |
| FY2018 | Unal, C., Stevens, G. N., and Matthews, C. (2018, September 30-October 4). Progressive Bayesian calibration of the BISON fuel performance capability. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic. | Publication |
| FY2018 | Wagih, M., Spencer, B., Hales, J., and Shirvan, K. (2018). Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings. Annals of Nuclear Energy, 120, 304-318. | Publication |
| FY2018 | Wang, J., Jo, H. J., and Corradini, M. L. (2018). Potential recovery actions from a severe accident in a PWR: MELCOR analysis of a station blackout scenario. Nuclear Technology, 204(1), 1-14. | Publication |
| FY2018 | Xu, P., Lahoda, E. J., Lyons, J., Deck, C. P., and Kohse, G. E. (2018, September 30-October 4). Status update on Westinghouse SiC composite cladding fuel development. In Proceedings of the TopFuel 2018 Conference, Prague, Czech Republic. | Publication |
| FY2018 | Xu, P., Lahoda, E., Boylan, F., and Oelrich, R. L. (2018, January 21-26). Status update on Westinghouse EnCore™ SiC/SiC composite cladding development. In Proceedings of the 42nd International Conference and Expo on Advanced Ceramics and Composites, Daytona Beach, FL. | Publication |
| FY2018 | Yao, T., Scott, S. M., Xin, G., and Lian, J. (2016). TiO2 doped UO2 fuels sintered by spark plasma sintering. Journal of Nuclear Materials, 469, 251-261. | Publication |
| FY2018 | Yeo, S., McKenna, E., Baney, R., Subhash, G., and Tulenko, J. (2013). Enhanced thermal conductivity of uranium dioxide-silicon carbide composite fuel pellets prepared by spark plasma sintering (SPS). Journal of Nuclear Materials, 433(1-3), 66-73. | Publication |
| FY2018 | Yeom, H., Dabney, T., Johnson, G., and others. (2019). Improving deposition efficiency in cold spraying chromium coatings by powder annealing. International Journal of Advanced Manufacturing Technology, 100, 1373-1382. | Publication |
| FY2018 | Yeom, H., Maier, B., Johnson, G., Dabney, T., Walters, J., and Sridharan, K. (2018). Development of cold spray process for oxidation-resistant FeCrAl and Mo diffusion barrier coatings on optimized ZIRLO™. Journal of Nuclear Materials, 507, 306-315. | Publication |
| FY2018 | Zalkin, A., and Templeton, D. H. (1953). The crystal structures of CeB4, ThB4, and UB4. Acta Crystallographica, 6(3), 269-272. | Publication |
| FY2017 | Alam, M. E., Pal, S., Fields, K., Maloy, S. A., Hoelzer, D. T., and Odette, G. R. (2016). Tensile deformation and fracture properties of a 14YWT nanostructured ferritic alloy. Materials Science and Engineering: A, 675, 437-448. | Publication |
| FY2017 | Alam, M. E., Pal, S., Maloy, S. A., and Odette, G. R. (2017). On delamination toughening of a 14YWT nanostructured ferritic alloy. Acta Materialia, 136, 61-73. | Publication |
| FY2017 | Aliberity, G., Kim, T. K., and Stauff, N. (2017). Assessment of AmBB performance and tradeoff of advanced fuel concepts (FY 2017, NTRD-FUEL-2017-00012). September 30, 2017. | |
| FY2017 | Ang, C. K., Kiggans, J., Kemery, C., O'Dell, S., Burns, J., Terrani, K. A., and Katoh, Y. (2016). Mechanical properties and characterization of coated SiC fuel cladding. Transactions of American Nuclear Society, 115(1), 433-435. | Publication |
| FY2017 | Ang, C., Katoh, Y., Kemery, C., Kiggans, J., and Terrani, K. (2016). Chromium-based mitigation coatings on SiC materials for fuel cladding. Transactions of American Nuclear Society, 114(1), 1095-1097. | Publication |
| FY2017 | Ang, C., Raiman, S., Burns, J., Hu, X., and Katoh, Y. (2017). Evaluation of the first generation dual-purpose coatings for SiC cladding (ORNL/SR-2017/318). Oak Ridge National Laboratory, Oak Ridge, TN. | Publication |
| FY2017 | Ang, C., Terrani, K., Burns, J., and Katoh, Y. (2016). Examination of hybrid metal coatings for mitigation of fission product release and corrosion protection of LWR SiC/SiC (ORNL/TM-2016/332). Oak Ridge National Laboratory, Oak Ridge, TN. | Publication |
| FY2017 | Antonio, D. J., Shrestha, K., Harp, J. M., Adkins, C. A., Zhang, Y., Carmack, J., and Gofryk, K. (2018). Thermal and transport properties of U3Si2. Journal of Nuclear Materials, 508, 154-158. | Publication |
| FY2017 | Aydogan, E., Almirall, N., Odette, G. R., Maloy, S. A., Anderoglu, O., Shao, L., Gigax, J. G., Price, L., Chen, D., Chen, T., Garner, F. A., Wu, Y., Wells, P., Lewandowski, J. J., and Hoelzer, D. T. (2017). Stability of nanosized oxides in ferrite under extremely high dose self ion irradiations. Journal of Nuclear Materials, 486, 86-95. | Publication |
| FY2017 | Aydogan, E., Maloy, S. A., Anderoglu, O., Sun, C., Gigax, J. G., Shao, L., Garner, F. A., Anderson, I. E., and Lewandowski, J. J. (2017). Effect of tube processing methods on microstructure, mechanical properties and irradiation response of 14YWT nanostructured ferritic alloys. Acta Materialia, 134, 116-127. | Publication |
| FY2017 | Benson, M. T., King, J. A., Mariani, R. D., and Marshall, M. C. (2017). SEM characterization of two advanced fuel alloys: U-10Zr-4.3Sn and U-10Zr-4.3Sn-4.7Ln. Journal of Nuclear Materials, 494, 334-341. | Publication |
| FY2017 | Besmann, T. M., Noorhoek, M. J., Wilson, T., Nelson, A. T., Wood, E. S., McMurray, J. W., Shin, D., Lahoda, E. J., and Middleburgh, S. C. (2017). Uranium silicide-nitride fuels: Thermochemical behavior and compatibility. In Proceedings of the International Conference and Exposition on Advanced Ceramics and Composites (ICACC), Daytona Beach, FL, January, 2017. | Publication |
| FY2017 | Bess, J. D., Hill, C. M., Woolstenhulme, N. E., and Jensen, C. B. (2017). Analyses supporting design review of TREAT multi-SERTTA experiment test vehicle. In Proceedings of the International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (MandC 2017), Jeju, Korea, Republic of, April 16-20, 2017. | Publication |
| FY2017 | Burr, P. A., Horlait, D., and Lee, W. E. (2016). Experimental and DFT investigation of (Cr,Ti)3AlC2 MAX phases stability. Materials Research Letters, 5(3), 144-157. | Publication |
| FY2017 | Cai, L., Xu, P., Atwood, A., Boylan, F., and Lahoda, E. J. (2017). Thermal analysis of ATF fuel materials at Westinghouse. In Proceedings of the International Conference and Exposition on Advanced Ceramics and Composites (ICACC), Daytona Beach, FL, January, 2017. | Publication |
| FY2017 | Carvajal-Nunez, U., Saleh, T. A., White, J. T., Maiorov, B., and Nelson, A. T. (2018). Determination of elastic properties of polycrystalline U3Si2 using resonant ultrasound spectroscopy. Journal of Nuclear Materials, 498, 438-444. | Publication |
| FY2017 | Carvajal-Nunez, U., White, J. T., Wood, E. S., Mara, N. A., and Nelson, A. T. (2016). Nanoscale mechanical behavior of uranium silicide compounds. In Top Fuel 2016: LWR Fuels with Enhanced Safety and Performance (pp. 1435-1441). | |
| FY2017 | Domitr, P., Cheng, L.-Y., Kohut, P., and Ramsey, J. (2017). Development of TRACE model based on TRAC-M input for PWR LOCA analysis. In Proceedings of the 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), Xi'an, China, September 3-8, 2017. | Publication |
| FY2017 | Doyle, P., Raiman, S., Rebak, R., and Terrani, K. (2017). Characterization of the hydrothermal corrosion behavior of ceramics for accident tolerant fuel cladding. In Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems â Water Reactors. | Publication |
| FY2017 | Dryepondt, S., Massey, C., and Gussev, M. N. (2017). Production and characterization of large batch FeCrAl ODS alloy (ORNL/TM-2017/445). Oak Ridge National Laboratory. | |
| FY2017 | Fernandez, J. C., Gautier, D. C., Huang, C., Palaniyappan, S., Albright, B. J., Bang, W., Dyer, G., Favalli, A., Hunter, J. F., Mendez, J., Roth, M., Swinhoe, M., Bradley, P. A., Deppert, O., Espy, M., Falk, K., Guler, N., Hamilton, C., Hegelich, B. M., ... Yin, L. (2017). Laser-plasmas in the relativistic transparency regime: Science and applications. Physics of Plasmas, 24(5), 056. | Publication |
| FY2017 | Field, K., Snead, M., Yamamoto, Y., and Terrani, K. (2017). Handbook of the materials properties of FeCrAl alloys for nuclear power production applications (ORNL/TM-2017/186, Rev. 1). Oak Ridge National Laboratory. | Publication |
| FY2017 | Gofryk, K. (2017, March). Unusual thermal behavior of uranium dioxide fuel. Invited talk at the 47èmes Journées des Actinides, Karpacz, Poland. | |
| FY2017 | Golovchenko, Y. M. (2011). Some results of developments and investigations of fuel pins with metal fuel for heterogeneous core of fast reactors of the BN-type. Energy Procedia, 7, 205-212. | Publication |
| FY2017 | Harp, J. M., Hayes, S. L., Medvedev, P. G., Porter, D. L., and Capriotti, L. (2017). Testing fast reactor fuels in a thermal reactor: A comparison report (INL/EXT-17-41677 [NTRDFUEL-2017-000148], Rev. 0). Idaho National Laboratory. | Publication |
| FY2017 | Harp, J. M., Porter, D. L., Miller, B. D., Trowbridge, T. L., and Carmack, W. J. (2017). Scanning electron microscopy examination of a Fast Flux Test Facility irradiated U-10Zr fuel cross section clad with HT-9. Journal of Nuclear Materials, 494, 227-239. | Publication |
| FY2017 | He, L., Harp, J. M., Hoggan, R. E., and Wagner, A. R. (2017). Microstructure studies of interdiffusion behavior of U3Si2/Zircaloy-4 at 800 and 1000 °C. Journal of Nuclear Materials, 486, 274-282. | Publication |
| FY2017 | Hill, C. M., Bess, J. D., and Woolstenhulme, N. E. (2017). Neutronics analysis of TREAT Multi-SERTTA calibration test vehicle (Multi-SERTTA CAL). Transactions of the American Nuclear Society, 116(1), 1073-1076. San Francisco, CA. | Publication |
| FY2017 | Hoggan, R., Harp, J., and He, L. (2017). Interdiffusion behavior of U3Si2 and FeCrAl via diffusion couple studies. Transactions of the American Nuclear Society, 116(1), 403-406. | Publication |
| FY2017 | Isler, J., Zhang, J., Mariani, R., and Unal, C. (2017). Experimental solubility measurements of lanthanides in liquid alkalis. Journal of Nuclear Materials, 495, 438-441. | Publication |
| FY2017 | J.Bischoff, C.Vauglin, P. Barberis, C., Roubeyrie, D. Perche, D. Duthoo,, F.Schuster, J-C. Brachet, K. Nimishakavi, AREVA's Enhanced Accident Tolerant, Fuel Developments: Focus on Cr-Coated, M5TM Cladding, 2017 Water Reactor Fuel, Performance Meeting, Korea, September 2017 | |
| FY2017 | Janney, D. E., and Sencer, B. H. (2017). Microstructure changes caused by annealing of U-Pu-Zr alloys. Journal of Nuclear Materials, 486, 66-69. | Publication |
| FY2017 | Jensen, C. B., Woolstenhulme, N. E., and Wachs, D. M. (2017, September 10-14). Fuel-coolant interaction results for high energy in-pile LWR fuels experiments. In Proceedings of Water Reactor Fuel Performance Meeting 2017, Jeju Island, South Korea. | Publication |
| FY2017 | Karoutas, Z. E., Schneider, R., LaBarge, N. R., Romero, J., and Xu, P. (2017, September 10-14). Westinghouse accident tolerant fuel plant benefits. Paper presented at the 2017 Water Reactor Fuel Performance Meeting, Jeju Island, South Korea. | |
| FY2017 | Khalifa, H., Deck, C., Jacobsen, G., Sheeder, J., Zhang, J., Bacalski, C., Stone, J., Shih, C., Vasudevamurthy, G., Shatoff, H., Huang, X., Bao, J., Jacko, R., Lahoda, E., and Back, C. (2017, January). Development of SiC-SiC composite accident tolerant fuel cladding. Paper presented at the ICACC, Daytona Beach, FL. | |
| FY2017 | Koyanagi, T., Katoh, Y., Singh, G., and Snead, M. (2017). SiC/SiC cladding materials properties handbook (ORNL/SPR-2017/385). Oak Ridge National Laboratory. | Publication |
| FY2017 | Li, X., Samin, A., Zhang, J., Unal, C., and Mariani, R. D. (2017). Ab-initio molecular dynamics study of lanthanides in liquid sodium. Journal of Nuclear Materials, 484, 98-102. | Publication |
| FY2017 | Matthews, C., Galloway, J., and Unal, C. (2017, June 11-15). Advanced simulation aided metallic fuel design. Paper presented at the ANS 2017 Summer Meeting, San Francisco. (LA-UR-17-2044). | |
| FY2017 | Matthews, C., Galloway, J., Unal, C., Novascone, S., and Williamson, R. (2017, June 26-29). BISON for metallic fuels modeling. Paper presented at the International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17), Yekaterinburg, Russian Federation. (IAEA-CN-245-366). | Publication |
| FY2017 | Matthews, C., Unal, C., Galloway, J., Keiser, D. D., and Hayes, S. L. (2017). Fuel-cladding chemical interaction in U-Pu-Zr metallic fuels: A critical review. Nuclear Technology, 198(3), 231-259. | Publication |
| FY2017 | Medvedev, P., Hayes, S., Bays, S., Novascone, S., and Capriotti, L. (2018). Testing fast reactor fuels in a thermal reactor. Nuclear Engineering and Design, 328, 154-160. | Publication |
| FY2017 | Miao, Y., Harp, J., Mo, K., Bhattacharya, S., Baldo, P., and Yacout, A. M. (2017). Short communication on In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures. Journal of Nuclear Materials, 484, 168-173. | Publication |
| FY2017 | Miao, Y., Harp, J., Mo, K., Zhu, S., Yao, T., Lian, J., and Yacout, A. M. (2017). Bubble morphology in U3Si2 implanted by high-energy Xe ions at 300 °C. Journal of Nuclear Materials, 495, 146-153. | Publication |
| FY2017 | Middleburgh, S., Lahoda, E., Luszck, K., Grimes, R., Andersson, D., Stanek, C., and Besmann, T. (2017, January). Ongoing work on modelling of UN-U3Si2 fuel. Paper presented at the ICACC, Daytona Beach, FL. | |
| FY2017 | Oelrich, R., Ray, S., Karoutas, Z., Lahoda, E., Boylan, F., Xu, P., Romero, J., and Shah, H. (2017, September 10-14). Overview of Westinghouse Lead Accident Tolerant Fuel Program. Paper presented at the 2017 Water Reactor Fuel Performance Meeting, Jeju Island, South Korea. | |
| FY2017 | Raiman, S., Doyle, P., Ang, C., and Terrani, K. (2017). Hydrothermal corrosion of SiC materials for accident tolerant fuel cladding with and without mitigation coatings. In Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (pp. 1475-1483). | Publication |
| FY2017 | Rebak, R. B., Gassmann, W. P., and Terrani, K. A. (2017, February 12-16). Managing nuclear power plant safety with FeCrAl alloy fuel cladding. Paper A0042 presented at IAEA Top Safe 2017, Vienna, Austria. | Publication |
| FY2017 | Rebak, R. B., Larsen, M., and Kim, Y.-J. (2017). Characterization of oxides formed on iron-chromium-aluminum alloy in simulated light water reactor environments. Corrosion Reviews, 35(3), 177-188. | Publication |
| FY2017 | Rebak, R. B., Terrani, K. A., Gassmann, W. P., and others. (2017). Improving nuclear power plant safety with FeCrAl alloy fuel cladding. MRS Advances, 2, 1217-1224. | Publication |
| FY2017 | Romero, J., Byers, W. A., Wang, G., Mueller, A., and Karoutas, Z. (2017, September 10-14). Simulated severe accident testing for evaluation of accident tolerant fuel. Paper presented at the 2017 Water Reactor Fuel Performance Meeting, Jeju Island, South Korea. | |
| FY2017 | Roth, M., Vogel, S. C., Bourke, M. A. M., Fernandez, J. C., Mocko, M. J., Glenzer, S., Leemans, W., Siders, C., and Haefner, C. (2017, April 19). Assessment of laser-driven pulsed neutron sources for poolside neutron-based advanced NDE-pathway to LANSCE-like characterization at INL (LA-UR-17-23190). | Publication |
| FY2017 | Saleh, T. A., Romero, T. J., Quintana, M. E., and Field, K. J. (2017). Mechanical properties of HFIR irradiated FeCrAl alloys. NTRandD milestone report NTRDFUEL-2017-000006, LA-UR-17-28992. | |
| FY2017 | Schneider, R., LaBarge, N. R., Van De Berg, H., Van Haltern, M., Lahoda, E., and Karoutas, Z. (2017, September 24-28). Estimating the benefits of accident tolerant fuel (ATF). Paper presented at PSA 2017, Pittsburgh, PA. | |
| FY2017 | Schuster, M., Crawford, C. J., and Rebak, R. B. (2017, March 26-30). Thermal shock resistance of FeCrAl alloys for accident tolerant fuel cladding application. In Proceedings of the CORROSION 2017. NACE-2017-8900 (pp. 1-15). AMPP. New Orleans, Louisiana, USA. | Publication |
| FY2017 | Shah, H., Romero, J., Xu, P., Maier, B., Johnson, G., Walters, J., Dabney, T., Yeom, H., and Sridharan, K. (2017, September 10-14). Development of surface coatings for enhanced accident tolerant fuel. Paper presented at the 2017 Water Reactor Fuel Performance Meeting, Jeju Island, South Korea. | Publication |
| FY2017 | Singh, G., Gonczy, S., Lara-Curzio, E., and Katoh, Y. (2017). Interlaboratory round robin axial tensile testing of tubular SiC/SiC specimens (ORNL/SR-2017/397). Oak Ridge National Laboratory. | Publication |
| FY2017 | Sooby Wood, E., White, J. T., and Nelson, A. T. (2017). Oxidation behavior of U-Si compounds in air from 25 to 1000 °C. Journal of Nuclear Materials, 484, 245-257. | Publication |
| FY2017 | Sooby Wood, E., White, J. T., and Nelson, A. T. (2017). The effect of aluminum additions on the oxidation resistance of U3Si2. Journal of Nuclear Materials, 489, 84-90. | Publication |
| FY2017 | Stauff, N., Kim, T. K., and Hayes, S. (2017, June). Tradeoff study of advanced transmutation fuels in sodium-cooled fast reactors. Paper presented at the International Conference on Fast Reactors and Related Fuel Cycles: FR-17, Yekaterinburg, Russian Federation. (CN245-152 PI-81 poster). | Publication |
| FY2017 | Stevens, G. N., Unal, C., Galloway, J., and Matthews, C. (2017, May 3-5). Progressively informed calibration of BISON nuclear fuel models. Paper presented at the 2017 ASME VandV Workshop, Las Vegas, NV. (LA-UR-17-23571). | Publication |
| FY2017 | Sun, Z., and Yamamoto, Y. (2017). Processability evaluation of a Mo-containing FeCrAl alloy for seamless thin-wall tube fabrication. Materials Science and Engineering: A, 700, 554-561. | Publication |
| FY2017 | Sun, Z., Bei, H., and Yamamoto, Y. (2017). Microstructural control of FeCrAl alloys using Mo and Nb additions. Materials Characterization, 132, 126-131. | Publication |
| FY2017 | Sun, Z., Chen, X., and Yamamoto, Y. (2017). Examination of powder metallurgy vs. induction melting for FeCrAl alloy production (ORNL/TM-2017/381). Oak Ridge National Laboratory. | |
| FY2017 | Unal, C., Matthews, C., Xiang, L., Isler, J., Zhang, J., and Galloway, J. (2017, June 11-15). A potential mechanism for lanthanide transport in metallic fuels. Transactions of the American Nuclear Society, 116, 501-503. San, Francisco, (LA-UR-17-20083). | Publication |
| FY2017 | Unal, C., Xiang, L., Isler, J., Matthews, C., Abid, S., Zhang, J., Galloway, J., and Mariani, R. (2017, June 26-29). Modeling of lanthanide transport in metallic fuels: Recent progresses. Paper presented at the International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17), Yekaterinburg, Russian Federation. (IAEA-CN-245-350, LA-UR-17-20106). | Publication |
| FY2017 | Wang, J., Mccabe, M., Wu, L., Dong, X., Wang, X., Haskin, T. C., and Corradini, M. L. (2017). Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY short term station black out. Nuclear Engineering and Design, 313, 458-469. | Publication |
| FY2017 | Wang, J., Toloczko, M. B., Bailey, N., Garner, F. A., Gigax, J., and Shao, L. (2016). Modification of SRIM-calculated dose and injected ion profiles due to sputtering, injected ion buildup and void swelling. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 387, 20-28. | Publication |
| FY2017 | Wang, J., Toloczko, M. B., Kruska, K., and others. (2017). Carbon contamination during ion irradiation - Accurate detection and characterization of its effect on microstructure of ferritic/martensitic steels. Scientific Reports, 7, 15813. | Publication |
| FY2017 | Wang, Y., Hurley, D. H., Luther, E. P., Beaux, M. F., Vodnik, D. R., Peterson, R. J., Bennett, B. L., Usov, I. O., Yuan, P., Wang, X., and Khafizov, M. (2018). Characterization of ultralow thermal conductivity in anisotropic pyrolytic carbon coating for thermal management applications. Carbon, 129, 476-485. | Publication |
| FY2017 | Xu, P., Lahoda, E., and Long, Y. (2017, January). Westinghouse accident tolerant fuel program update on SiC composite cladding development. Paper presented at ICACC, Daytona Beach, FL. | Publication |
| FY2017 | Xu, P., Lahoda, E., Jacko, R., Boylan, F., and Oelrich, R. (2017, September 10-14). Status of Westinghouse SiC composite cladding fuel development. Paper A0184 presented at the 2017 LWR Fuel Performance Meeting, Jeju Island, South Korea. | |
| FY2017 | Yamamoto, Y., and Sun, Z. (2017). Quality optimization of commercial FeCrAl tube production (ORNL/TM-2017/338). Oak Ridge National Laboratory. | Publication |
| FY2017 | Zapata-Solvas, E., Christopoulos, S.-R. G., Ni, N., Parfitt, D. C., Horlait, D., Fitzpatrick, M. E., Chroneos, A., and Lee, W. E. (2017). Experimental synthesis and density functional theory investigation of radiation tolerance of Zr3(Al1-xSix)C2 MAX phases. Journal of the American Ceramic Society, 100, 1377-1387. | Publication |
| FY2017 | Zapata-Solvas, E., Hadi, M. A., Horlait, D., Parfitt, D. C., Thibaud, A., Chroneos, A., and Lee, W. E. (2017). Synthesis and physical properties of (Zr1-x,Tix)3AlC2 MAX phases. Journal of the American Ceramic Society, 100, 3393-3401. | Publication |
| FY2016 | Alat, E., Motta, A. T., Comstock, R. J., Partezana, J. M., and Wolfe, D. E. (2015). Ceramic coating for corrosion (C3) resistance of nuclear fuel cladding. Surface and Coatings Technology, 281, 133-143. | Publication |
| FY2016 | Alva, L., Shapovalov, K., Jacobsen, G. M., Back, C. A., and Huang, X. (2015). Experimental study of thermo-mechanical behavior of SiC composite tubing under high temperature gradient using solid surrogate. Journal of Nuclear Materials, 466, 698-711. | Publication |
| FY2016 | Anderoglu, O. (2016). In situ synchrotron characterization of the field assisted sintering of UO2. In ANS 2016 - Poster presentation and extended abstract. | |
| FY2016 | Anderoglu, O., and Saleh, T. A. (2016). Microstructural investigation of ATR irradiated (290°C to 6 dpa) MA956. Submitted for milestone Microstructural, Analysis of ATR Irradiated FeCrAl ODS alloys, M3FT-16LA020202082. | |
| FY2016 | Aydogan, E., Pal, S., Anderoglu, O., Maloy, S. A., Vogel, S. C., Odette, G. R., Lewandowski, J. J., Hoelzer, D. T., Anderson, I. E., and Rieken, J. R. (2016). Effect of tube processing methods on the texture and grain boundary characteristics of 14YWT nanostructured ferritic alloys. Materials Science and Engineering: A, 661, 222-232. | Publication |
| FY2016 | Bacalski, C. F., Jacobsen, G. M., and Deck, C. P. (Sept. 12-14, 2016). Characterization of SiC-SiC accident tolerant fuel cladding after stress application. In American Nuclear Society TOP FUEL 2016 Proceedings (pp. 843-848). Boise, Idaho. | Publication |
| FY2016 | Baker, K. E., Ellis, K., and Glass, C. (April 2016). BISON fuel performance code examination of coating/clad interfaces for accident tolerant fuels irradiation testing. In TMS 2016 Meeting Conference Proceedings. | |
| FY2016 | Barrett, K. E., Durtschi, B. P., Daw, J., Unruh, T., Smith, J., Woolum, C. T., Davis, K. L., and Chichester, H. J. M. (2016). Sensor qualification tests in preparation for accident tolerant fuels water loop irradiation testing in the Advanced Test Reactor at the INL. In Enhanced Halden Reactor Project Meeting, Oslo, Norway, May 9-12, 2016. | Publication |
| FY2016 | Bentzel, G. W., Naguib, M., Lane, N. J., Vogel, S. C., Presser, V., Dubois, S., Lu, J., Hultman, L., Barsoum, M. W., and Caspi, E. N. (2016). High-temperature neutron diffraction, Raman spectroscopy, and first-principles calculations of Ti3SnC2 and Ti2SnC. Journal of the American Ceramic Society, 99(7), 2233-2242. | Publication |
| FY2016 | Besmann, T. (2014). Fiscal year 2014 summary report on thermodynamic assessment of advanced accident tolerant fuel compositions (Milestone No. M3FT-14OR02021810). | |
| FY2016 | Bess, J. D., Woolstenhulme, N. E., Hill, C. M., Jensen, C. B., and Snow, S. D. (2016). TREAT neutronics analysis and design support, part I: Multi-SERTTA. In Proceedings of the Top Fuel 2016 Conference, Boise, Idaho, USA, Sep 11-16, 2016. | Publication |
| FY2016 | Bess, J. D., Woolstenhulme, N. E., Hill, C. M., O'Brien, R. C., and Bays, S. E. (2016). TREAT Multi-SERTTA neutronics design and analysis support. In Proceedings of the PHYSOR-2016 Conference, Sun Valley, Idaho, USA, May 1-5, 2016. | Publication |
| FY2016 | Bess, J. D., Woolstenhulme, N. E., Hill, C. M., Snow, S. D., and Jensen, C. B. (2016). TREAT neutronics analysis and design support, part II: Multi-SERTTA-CAL. In Proceedings of the Top Fuel 2016 Conference, Boise, Idaho, USA, Sep 11-16, 2016. | Publication |
| FY2016 | Betzler, B. R., and Powers, J. J. (2016). A fully ceramic microencapsulated fuel for space reactor applications. In Proceedings of PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, Sun Valley, ID, USA, May 1-5, 2016. | Publication |
| FY2016 | Bischoff, J., Vauglin, C., Delafoy, C., Barberis, P., Perche, D., Guerin, B., Vassault, J. P., and Brachet, J. C. (2016). Development of Cr-coated zirconium alloy cladding for enhanced accident tolerance. In ANS Top Fuel 2016 Meeting Proceedings (pp. 1165-1171), Boise, ID, September 11-15, 2016. | Publication |
| FY2016 | Bragg-Sitton, S. M., and Carmack, W. J. (2014). Advanced Fuels Campaign: Light Water Reactor Accident Tolerant Fuel Performance Metrics (INL/EXT-13-29957, FCRD-FUEL-2013-000264). February 2014. | Publication |
| FY2016 | Bragg-Sitton, S. M., Todosow, M., Montgomery, R., Stanek, C. R., and Carmack, W. J. (2016). Metrics for the technical performance evaluation of light water reactor accident-tolerant fuel. Nuclear Technology, 195(2), 111-123. | Publication |
| FY2016 | Brown, N. R., Wysocki, A. J., and Terrani, K. A. (2016). Reactivity initiated accident simulation to inform transient testing of candidate advanced cladding. In Top Fuel 2016: LWR Fuels with Enhanced Safety and Performance (pp. 271-285). American Nuclear Society. Boise, ID, September 11-16, 2016. | Publication |
| FY2016 | Brown, N. R., Wysocki, A. J., Terrani, K. A., Ali, A., Liu, M., and Blandford, E. (June 2016). Survey of thermal-fluids evaluation and confirmatory experimental validation requirements of accident tolerant cladding concepts with focus on boiling heat transfer characteristics (FCRD Milestone M3FT-16OR020204032, ORNL/TM-2016-252). | Publication |
| FY2016 | Brown, N. R., Wysocki, A. J., Terrani, K. A., Xu, K. G., and Wachs, D. M. (2017). The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors. Annals of Nuclear Energy, 99, 353-365. | Publication |
| FY2016 | Byler, D., and Valdez, J. (2014). Report on synthesis of high-density ceramic composite materials with microstructural and chemical characterization (LA-UR-14-24678). | |
| FY2016 | Carmack, J. (2015). Enhanced accident tolerant fuels for LWRs technical review committee charter (FCRD-FUEL-2015-000021). March 2015. | |
| FY2016 | Cheng, B., Chou, P., Topbasi, C., Kim, Y., Armijo, S., Do, C., and Ring, P. (2016). Fabrication of rodlets with coated and lined Mo-alloy cladding for testing and irradiation. In ANS Top Fuel 2016 Meeting Proceedings (pp. 207-216), Boise, ID, September 11-15, 2016. | |
| FY2016 | Chichester, H. J. M., Core, G. M., Barrett, K. E., and Wachs, D. M. (2016). Irradiation testing strategy for U.S. accident tolerant fuels. In Enlarged Halden Programme Group Meeting 2016 Proceedings, May 2016. | |
| FY2016 | Cinbiz, M. N., Brown, N. R., Lowden, R. R., Erdman, D., and Terrani, K. A. (2016). Issue report documenting simulated PCI testing (FCRD Milestone M3FT-16OR020204031). September 9, 2016. | |
| FY2016 | Cinbiz, N., Brown, N., Terrani, K. A., Lowden, R. R., and Erdman, D. (2017). The mechanical response evaluation of advanced claddings during proposed reactivity initiated accident conditions. In X. Liu et al. (Eds.), Energy Materials 2017 (pp. 355-365). Springer, Cham. | Publication |
| FY2016 | Cologna, M., Rashkova, B., and Raj, R. (2010). Flash sintering of nanograin zirconia in andlt;5 s at 850°C. Journal of the American Ceramic Society, 93(11), 3556-3559. | Publication |
| FY2016 | Davis, C. B., and Woolstenhulme, N. E. (2016). Validation of a RELAP5-3D point kinetics model of TREAT. In Proceedings of the PHYSOR-2016 Conference, Sun Valley, Idaho, USA, May 1-5, 2016. | |
| FY2016 | Daw, J. E., Unruh, T. C., Durtschi, B. P., Barrett, K. E., Woolum, C. T., Smith, J. A., and Chichester, H. J. M. (2016). Instrumentation for accident tolerant fuel loop testing at ATR. In Enlarged Halden Programme Group Meeting 2016 Proceedings, May 2016. | |
| FY2016 | Dryepondt, S., Massey, C., and Edmonson, P. (2016). Oak Ridge National Laboratory report on 2nd generation ODS FeCrAl alloy development for accident-tolerant fuel cladding, ORNL-TM-2016/456, Oak Ridge National Laboratory, August 26, 2016. | Publication |
| FY2016 | Edmondson, P. D., Briggs, S. A., Yamamoto, Y., Howard, R. H., Sridharan, K., Terrani, K. A., and Field, K. G. (2016). Irradiation-enhanced precipitation in model FeCrAl alloys. Scripta Materialia, 116, 112-116. | Publication |
| FY2016 | Elbakhshwan, M. S., Gill, S. K., Motta, A. T., Weidner, R., Anderson, T., and Ecker, L. E. (2016). Sample environment for in situ synchrotron corrosion studies of materials in extreme environments. Review of Scientific Instruments, 87(10), 105122. | Publication |
| FY2016 | Elbakhshwan, M. S., Gill, S. K., Rumaiz, A. K., Bai, J., Ghose, S., Rebak, R. B., and Ecker, L. E. (2017). High-temperature oxidation of advanced FeCrNi alloy in steam environments. Applied Surface Science, 426, 562-571. | Publication |
| FY2016 | Field, K. G., and Howard, R. H. (2016). Status of FeCrAl ODS irradiations in the High Flux Isotope Reactor (FY-16 DOE-NE FCRD Report: ORNL/TM-2016/394). Oak Ridge National Laboratory. | Publication |
| FY2016 | Field, K. G., and Howard, R. H. (2016). Status report on irradiation capsules, designed to evaluate FeCrAl-UO2 interactions (FY-16 DOE-NE FCRD Report: ORNL/TM-2016/267). Oak Ridge National Laboratory. | Publication |
| FY2016 | Field, K. G., Barrett, K., Sun, Z., and Yamamoto, Y. (2016). Submission of FeCrAl feedstock for support of AFC ATF-2 irradiations (FY-16 DOE-NE FCRD Report: ORNL/TM-2016/394). September 2016. Oak Ridge National Laboratory. | Publication |
| FY2016 | Field, K. G., Briggs, S. A., Edmondson, P. D., Haley, J. C., Howard, R. H., Hu, X., Littrell, K. C., Parish, C. M., and Yamamoto, Y. (2016). Database on performance of neutron irradiated FeCrAl alloys (FY-16 DOE-NE FCRD Report: ORNL/TM-2016/335). Oak Ridge National Laboratory. | Publication |
| FY2016 | Folsom, C. P., Jensen, C. B., Williamson, R. L., Woolstenhulme, N. E., Ban, H., and Wachs, D. M. (2016). BISON modeling of reactivity-initiated accident experiments in a static environment. In Top Fuel 2016: LWR Fuels with Enhanced Safety and Performance (pp. 239-247). Boise, Idaho, USA, Sept. 11-16, 2016. | Publication |
| FY2016 | Ge, L., Subhash, G., Baney, R. H., Tulenko, J. S., and McKenna, E. (2013). Densification of uranium dioxide fuel pellets prepared by spark plasma sintering (SPS). Journal of Nuclear Materials, 435(1-3), 1-9. | Publication |
| FY2016 | George, N. M., Sweet, R. T., Maldonado, G. I., Wirth, B. D., Powers, J. J., and Worrall, A. (2015). Fuel performance calculations for FeCrAl cladding in BWRs. Transactions of the American Nuclear Society, 113, 553-556. | Publication |
| FY2016 | Hill, C. M., Woolstenhulme, N. E., Bess, J. D., Parry, J. R., and Housley, G. K. (2016, May 1-5). MCNP water physics scoping study to support accident tolerant fuel testing in TREAT. In Proceedings of PHYSOR 2016. American Nuclear Society, Sun Valley, Idaho. May 1-5, 2016 | Publication |
| FY2016 | Hill, C. M., Woolstenhulme, N. E., Parry, J. R., Bess, J. D., and Housley, G. K. (2016, September 11-16). TREAT neutronics analysis of water-loop concept accommodating LWR 9-rod bundle. In Proceedings of the Top Fuel 2016 Conference. Boise, Idaho, USA. Sep, 11-16, 2016 | Publication |
| FY2016 | Hu, X., Ang, C. K., Singh, G., and Katoh, Y. (2016). Technique development for modulus, microcracking, hermeticity, and coating evaluation capability characterization of SiC/SiC tubes ORNL/TM-2016/372, Oak Ridge National Laboratory. | Publication |
| FY2016 | Hull, L. C., Barrett, K. E., Lybeck, N., Johnson, N., Galbraith, S. G., Core, G. M., and Chichester, J. M. (2016, September). NDMAS implementation and data qualification for accident tolerant fuel experiments. Paper presented at the ANS Top Fuel 2016 Meeting, Boise, ID. Paper number 16-50375. | Publication |
| FY2016 | Janney, D. E., and Papesch, C. A. (2016). An introduction to the FCRD transmutation fuels handbook 2015. Transactions of the American Nuclear Society, 114(1), 1077-1080. Presented at the 2016 Transactions of the American Nuclear Society Annual Meeting (ANS 2016), New Orleans, United States, June 12-16, 2016. | Publication |
| FY2016 | Jensen, C. B., Davis, C. B., Woolstenhulme, N. E., O'Brien, R. C., and Wachs, D. M. (2016). Thermal-hydraulic response of the TREAT multi-vessel static environment test vehicle. In Proceedings of the PHYSOR-2016 Conference, Sun Valley, Idaho, USA, May 1-5, 2016. | Publication |
| FY2016 | Jensen, C. B., Folsom, C. P., Davis, C. B., Woolstenhulme, N. E., Bess, J. D., O'Brien, R. C., Ban, H., and Wachs, D. M. (2016). Thermal-hydraulic performance of the TREAT Multi-SERTTA for reactivity-initiated accident experiments. In Proceedings of ANS Top Fuel 2016 Conference, Boise, ID. Sep, 11-16, 2016. | Publication |
| FY2016 | Katoh, Y., Terrani, K. A., Koyanagi, T., Petrie, C. M., Singh, G., Snead, L. L., and Deck, C. (2016). Irradiation high heat flux synergism in silicon carbide-based fuel claddings for light water reactors. In Proceedings of Top Fuel 2016 (pp. 823-831). | Publication |
| FY2016 | Khafizov, M., Pakarinen, J., He, L., Henderson, H. B., Manuel, M. V., Nelson, A. T., Jaques, B. J., Butt, D. P., and Hurley, D. H. (2016). Subsurface imaging of grain microstructure using picosecond ultrasonics. Acta Materialia, 112, 209-215. | Publication |
| FY2016 | Khalifa, H. E., Sheeder, J. D., Jacobsen, G. M., and Deck, C. P. (2016). Radiation tolerant joining for silicon carbide-based accident tolerant fuel cladding. In Light Water Reactor (LWR) Fuel Performance Meeting (TopFuel), 2016, Boise, Idaho, September 12-14, 2016, paper number 17517. | Publication |
| FY2016 | Kiran Kumar, N. A. P., Stevens, J. N., Savela, M., Mays, B., and Strumpell, J. (2016). AREVA enhanced accident tolerant fuel program - current results and future plans. In ANS Top Fuel 2016 Meeting Proceedings, Boise, ID, September 11-15, (pp. 169-178). | Publication |
| FY2016 | Koyanagi, T., Lance, M. J., and Katoh, Y. (2016). Quantification of irradiation defects in beta-silicon carbide using Raman spectroscopy. Scripta Materialia, 125, 58-62. | Publication |
| FY2016 | Kristiansen, P. (2016, August). Preliminary neutronics calculations for the proposed accident tolerant fuel (ATF) test for DOE. Institutt for energiteknikk OECD, Halden Reactor Project, CP-NOTE, 16-22. | |
| FY2016 | Law, M., Carr, D. G., and Vogel, S. C. (2015). Materials for the nuclear energy sector. In Neutron applications in materials for energy. Springer International Publishing. | Publication |
| FY2016 | Liu, M., Ryals, M., Ali, A., Blandford, E. D., Jensen, C., Condie, K., Svoboda, J., and O'Brien, R. (2016). Development of electrical capacitance sensors for accident tolerant fuel (ATF) testing at the Transient Reactor Test (TREAT) Facility. In Proceedings of Test, Research and Training Reactors (TRTR) 2016 Conference, Albuquerque, NM. | Publication |
| FY2016 | Liu, Y., Bhamji, I., Withers, P. J., Wolfe, D. E., Motta, A. T., and Preuss, M. (2015). Evaluation of the interfacial shear strength and residual stress of TiAlN coating on ZIRLO™ fuel cladding using a modified shear-lag model approach. Journal of Nuclear Materials, 466, 718-727. | Publication |
| FY2016 | Losko, A. S., Vogel, S. C., Bourke, M. A., et al. (2016). Energy-resolved neutron imaging for interrogation of nuclear materials. In Proceedings of the Advances in Nuclear Nonproliferation Technology and Policy Conference (ANTPC), Santa Fe, NM, September 25-30, 2016. | |
| FY2016 | Losko, A. S., Vogel, S. C., Bourke, M. A., et al. (2016). Neutron characterization of UN/U-Si accident tolerant fuel prior to irradiation. In Proceedings of Top Fuel 2016, Boise, ID, 11-14 September 2016. | |
| FY2016 | Losko, A. S., Vogel, S. C., Bourke, M. A., Voit, S. L., McClellan, K. J., Mocko, M., Byler, D. D., Tremsin, A. S., and Hosemann, P. (2016). Characterization of fresh nuclear fuel using time-of-flight neutrons. Transactions of the American Nuclear Society, 114(1), 1083-1086. New Orleans, LA. June 12-16, 2016. | Publication |
| FY2016 | Maier, B. R., Garcia-Diaz, B. L., Hauch, B., Olson, L. C., Sindelar, R. L., and Sridharan, K. (2015). Cold spray deposition of Ti2AlC coatings for improved nuclear fuel cladding. Journal of Nuclear Materials, 466, 712-717. | Publication |
| FY2016 | Massey, C. P., Terrani, K. A., Dryepondt, S. N., and Pint, B. A. (2016). Cladding burst behavior of Fe-based alloys under LOCA. Journal of Nuclear Materials, 470, 128-138. | Publication |
| FY2016 | Nuclear Energy Agency. (2014). Uranium 2014: Resources, production and demand. OECD Publishing. 488 | Publication |
| FY2016 | O'Brien, R. C., Woolstenhulme, N. E., Folsom, C. P., Jensen, C., Wachs, D. M., and Beasley, A. A. (June 22-24). Resumption of transient testing at the Idaho National Laboratory TREAT reactor: Development of experimental and analytical capabilities in support of the Accident Tolerant Fuels campaign. Proceedings of OECD/NEA Workshop on Pellet Cladding Interaction (PCI) in Water Cooled Reactors, Lucca, Italy. | |
| FY2016 | Park, D., Mouche, P. A., Zhong, W., Han, X., Heuser, B. J., Mandapaka, K. K., and Was, G. S. (2016). TEM study of Zircaloy 2 with FeCrAl layer under simulated BWR environment. In Transactions of the American Nuclear Society, 114(1), 1059-1060. Poster presented at the 2016 ANS Annual Meeting, New Orleans, LA. | Publication |
| FY2016 | Pereira da Silva, J. G., Al-Qureshi, H. A., Keil, F., and Janssen, R. (2016). A dynamic bifurcation criterion for thermal runaway during the flash sintering of ceramics. Journal of the European Ceramic Society, 36(5), 1261-1267. | Publication |
| FY2016 | Petrie, C. M., and Terrani, K. A. (2016). Thermal analysis of a flexible rabbit design for irradiating PWR cladding. FY-16 DOE-NE FCRD Report: ORNL/TM-2016/197. Oak Ridge National Laboratory. | Publication |
| FY2016 | Petrie, C. M., Koyanagi, T., McDuffee, J. L., Deck, C. P., Katoh, Y., and Terrani, K. A. (2017). Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux. Journal of Nuclear Materials, 491, 94-104. | Publication |
| FY2016 | Powers, J. J. (2016, April). Preliminary neutronics assessment of fully ceramic microencapsulated fuel in high-temperature gas-cooled reactors. In 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016), San Francisco, California, April 17-20, 2016. | Publication |
| FY2016 | Powers, J. J., Worrall, A., Robb, K. R., George, N. M., and Maldonado, G. I. (2016). ORNL analysis of operational and safety performance for candidate accident tolerant fuel and cladding concepts. In Proceedings of IAEA Technical Meeting on Accident Tolerant Fuel Concepts for Light Water Reactors, IAEA-TECDOC-1797. International Atomic Energy Agency. | Publication |
| FY2016 | Rebak, R. B. (2015). Alloy selection for accident tolerant fuel cladding in commercial light water reactors. Metallurgical and Materials Transactions E, 2(4), 197-207. | Publication |
| FY2016 | Rebak, R. B., and Ellis, D. D. (2016). Passivation characteristics of ferritic stainless materials in simulated reactor environments. Paper 7452, Corrosion 2016. NACE International, Houston, TX. | Publication |
| FY2016 | Rebak, R. B., Kim, Y.-J., Gynnerstedt, J., Terrani, K. A., and Stachowski, R. E. (2016, September). Fabrication of FeCrAl cladding for accident tolerant fuel. Paper presented at Top Fuel 2016, Boise, Idaho. | Publication |
| FY2016 | Rebak, R. B., Terrani, K. A., and Fawcett, R. M. (2016). FeCrAl alloys for accident tolerant fuel cladding in light water reactors. In Proceedings of the ASME 2016 Pressure Vessels and Piping Conference, Volume 6B: Materials and Fabrication, Vancouver, British Columbia, Canada, July 17-21, 2016 (Paper No. PVP2016-63162, V06BT06A009). ASME. | Publication |
| FY2016 | Rebak, R. B., Terrani, K. A., Gassmann, W., Williams, J., Fawcett, R. M., and Stachowski, R. E. (2016). Minimizing risk in nuclear power plant operation by using accident tolerant FeCrAl cladding. Paper RISK16-8330, NACE International Corrosion Risk Management Conference, Houston, TX, May 23-25, 2016. | Publication |
| FY2016 | Reiche, H. M., and Vogel, S. C. (2016). In situ synthesis and characterization of uranium carbide using high temperature neutron diffraction. In Proceedings of Top Fuel 2016, Boise, ID, September 11-14, 2016. | Publication |
| FY2016 | Reiche, H. M., Vogel, S. C., and Tang, M. (2016). In situ synthesis and characterization of uranium carbide using high temperature neutron diffraction. Journal of Nuclear Materials, 471, 308-316. | Publication |
| FY2016 | Robb, K. R. (2015). FeCrAl accident tolerant fuel response during BWR severe accidents. In Proceedings of the 21st International Quench Workshop (QUENCH) (ISBN 978-3-923704-90-3), Karlsruhe, Germany, October 27-29, 2015. | |
| FY2016 | Robb, K. R., McMurray, J. W., and Terrani, K. A. (2016). M2FT-16OR020205042: Severe accident analysis of BWR core fueled with UO2/FeCrAl with updated materials and melt properties from experiments. ORNL/TM-2016/237. Oak Ridge National Laboratory, June 2016. | Publication |
| FY2016 | Saleh, T. A., Quintana, M. E., and Romero, T. J. (2016). Tensile tests from the StipV irradiation. Submitted for milestone: Complete and report on tensile testing of STIP V FeCrAl specimens (M3FT-16LA020202085). LA-UR-16-22503. March 30, 2016. | |
| FY2016 | Schappel, D., Terrani, K., Powers, J., Snead, L. L., and Wirth, B. D. (2016). Thermo mechanical analysis of fully ceramic microencapsulated fuel during in-pile operation. In Transactions of the 2016 LWR Fuel Performance Meeting (Top Fuel, 2016), Boise, ID, USA. | Publication |
| FY2016 | Shamma, M., Caspi, E. N., Anasori, B., Clausen, B., Brown, D. W., Vogel, S. C., Presser, V., Amini, S., Yeheskel, O., and Barsoum, M. W. (2015). In situ neutron diffraction evidence for fully reversible dislocation motion in highly textured polycrystalline Ti2AlC samples. Acta Materialia, 98, 51-63. | Publication |
| FY2016 | Singh, G., Sweet, R., Wirth, B. D., Terrani, K. A., and Katoh, Y. (2016). Bison modeling of SiC/SiC cladding including fuel-pellet interaction. ORNL/TM-216/449. Oak Ridge National Laboratory | |
| FY2016 | Squires, L. N., and Lessing, P. (2016). Direct chemical reduction of neptunium oxide to neptunium metal using calcium and calcium chloride. Journal of Nuclear Materials, 471, 65-68. | Publication |
| FY2016 | Stachowski, R. E., Rebak, R. B., Gassmann, W. P., and Williams, J. (2016). Progress of GE development of accident tolerant fuel FeCrAl cladding. In Top Fuel 2016, Boise, Idaho, September 2016. | Publication |
| FY2016 | Stauff, N. E., Fei, T., and Kim, T. K. (2016). Assessment of AmBB performance and tradeoff of advanced fuel concepts (FCRD-FUEL-2016-000223). September 30, 2016. | |
| FY2016 | Stauff, N. E., Fei, T., Kim, T. K., and Hayes, S. L. (2016). Am-bearing blanket transmutation strategies in sodium-cooled fast reactors. In Actinide and Fission Product Partitioning and Transmutation 14th Information Exchange Meeting (14IEMPT), San Diego, October 17-20, 2016. | |
| FY2016 | Stone, J. G., Schleicher, R., Deck, C. P., Jacobsen, G. M., Khalifa, H. E., and Back, C. A. (2015). Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding. Journal of Nuclear Materials, 466, 682-697. | Publication |
| FY2016 | Sweet, R. T., George, N. M., Terrani, K. A., and Wirth, B. D. (2016). Fuel performance analysis of FeCrAl cladding during LWR operation. In Top Fuel 2016 transactions, Boise, ID, 1485-1492. | |
| FY2016 | Terrani, K. A., et al. (2016). Characterization report on FeCrAl cladding for Halden irradiation, ORNL/TM2016/343, Oak Ridge National Laboratory, July 2016. | |
| FY2016 | Terrani, K. A., Pint, B. A., Kim, Y.-J., Unocic, K. A., Yang, Y., Silva, C. M., Meyer, H. M., and Rebak, R. B. (2016). Uniform corrosion of FeCrAl alloys in LWR coolant environments. Journal of Nuclear Materials, 479, 36-47. | Publication |
| FY2016 | Vogel, S. C., Bourke, M. A., Stanek, C. R., et al. (2016). Summary report of joint FCRD/NEAMS technical experts working meeting on neutron-based NDE. Report for FCRD program, June 3, 2016. | |
| FY2016 | Vogel, S. C., Losko, A. S., Bourke, M. A., McClellan, K. J., et al. (2016). Nondestructive examination of UN/U-Si fuel pellets using neutrons (preliminary assessment). Report for FCRD program, March 20, 2016 (LA-UR-16-22179). | Publication |
| FY2016 | Vogel, S. C., Losko, A. S., Bourke, M. A., McClellan, K. J., et al. (2016). Non-destructive pre-irradiation assessment of UN/U-Si LANL1 ATF formulation. Report for FCRD program (LA-UR-16-27110) September 15, 2016. | Publication |
| FY2016 | Woolstenhulme, N. E., Baker, C. C., Bess, J. D., Davis, C. B., Hill, C. M., Housley, G. K., Jensen, C. B., Jerred, N. D., O'Brien, R. C., Snow, S. D., and Wachs, D. M. (2016). Capabilities development for transient testing of advanced nuclear fuels at TREAT. In Proceedings of Top Fuel 2016 Conference, American Nuclear Society - ANS, Boise, ID (pp. 67-76). | Publication |
| FY2016 | Woolstenhulme, N. E., Bess, J. D., Davis, C. B., Housley, G. K., Jensen, C. B., O'Brien, R. C., and Wachs, D. M. (2016, May 15). TREAT irradiation vehicle designs, capabilities, and future plans. In Proceedings of the PHYSOR-2016 Conference, Sun Valley, Idaho, May 1-5, 2016. | |
| FY2016 | Woolum, C., Archibald, K., Moore, G., and Galbraith, S. (2016). Fabrication and qualification of small scale irradiation experiments in support of the Accident Tolerant Fuels Program. In TMS 2016: 145th Annual Meeting and Exhibition: Supplemental Proceedings. TMS (Ed.). | Publication |
| FY2016 | Wysocki, A., Brown, N. R., Terrani, K. A., and Wachs, D. M. (2016). Potential impact of cladding wettability on LWR transient progression. Transactions of the American Nuclear Society, 115, 473-477. Paper presented at the 2016 Transactions of the American Nuclear Society, ANS 2016, Las Vegas, United States, November 6-10, 2016. | Publication |
| FY2016 | Yamamoto, Y., Pint, B. A., Terrani, K. A., Field, K. G., Yang, Y., and Snead, L. L. (2015). Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors. Journal of Nuclear Materials, 467(Part 2), 703-716. | Publication |
| FY2016 | Yang, X.-d., Gao, J.-c., Wang, Y., and Chang, X. (2008). Low-temperature sintering process for UO2 pellets in partially-oxidative atmosphere. Transactions of Nonferrous Metals Society of China, 18(1), 171-177. | Publication |
| FY2016 | Yeom, H., Hauch, B., Cao, G., Garcia-Diaz, B., Martinez-Rodriguez, M., Colon-Mercado, H., Olson, L., and Sridharan, K. (2016). Laser surface annealing and characterization of Ti2AlC plasma vapor deposition coating on zirconium-alloy substrate. Thin Solid Films, 615, 202-209. | Publication |
| FY2016 | Zhong, W., Mouche, P. A., Han, X., Heuser, B. J., Mandapaka, K. K., and Was, G. S. (2016). Performance of iron-chromium-aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions. Journal of Nuclear Materials, 470, 327-338. | Publication |
| FY2015 | Angle, J. P., Nelson, A. T., Men, D., and Mecartney, M. L. (2014). Thermal measurements and computational simulations of three-phase (CeO2-MgAl2O4-CeMgAl11O19) and four-phase (3Y-TZP-Al2O3-MgAl2O4-LaPO4) composites as surrogate inert matrix nuclear fuel. Journal of Nuclear Materials, 454(1-3), 69-76. | Publication |
| FY2015 | Bailey, N. A., Stergar, E., Toloczko, M., and Hosemann, P. (2015). Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures. Journal of Nuclear Materials, 459, 225-234. | Publication |
| FY2015 | Baker, C., Housley, G. K., Imholte, D. D., and Woolstenhulme, N. E. (2015). HFEF support equipment determination report for the TREAT water loop (INL/LTD-15-36111). Idaho National Laboratory. | |
| FY2015 | Barabash, R. I., Voit, S. L., Aidhy, D. S., Lee, S. M., Knight, T. W., Sprouster, D. J., and Ecker, L. E. (2015). Cation and vacancy disorder in U1-yNdyO2.00-x alloys. Journal of Materials Research, 30(11), 3026-3040. | Publication |
| FY2015 | Barrett, K. E., Ellis, K. D., Glass, C. R., Roth, G. A., Teague, M. P., and Johns, J. (2015). Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests. Nuclear Engineering and Design, 294, 38-51. | Publication |
| FY2015 | Beasley, A., Hill, C., Housley, G., Jensen, C., O'Brien, R., and Woolstenhulme, N. (2015). TREAT water loop status report (INL/LTD-15-36768). Idaho National Laboratory. | |
| FY2015 | Brese, R. G., McMurray, J. W., Shin, D., and Besmann, T. M. (2015). Thermodynamic assessment of the U-Y-O system. Journal of Nuclear Materials, 460, 5-12. | Publication |
| FY2015 | Brown, N. R., Cheng, L.-Y., and Todosow, M. (2014). Uranium nitride composite fuels in a light water reactor: Advanced cladding, nodal core calculations, and transient analysis. Transactions of the American Nuclear Society, 111(1), 1367-1370. | Publication |
| FY2015 | Brown, N. R., Todosow, M., and Cuadra, A. (2015). Screening of advanced cladding materials and UN-U3Si5 fuel. Journal of Nuclear Materials, 462, 26-42. | Publication |
| FY2015 | Brown, N. R., Todosow, M., Cheng, L.-Y., and Cuadra, A. (2015). Screening of reactor performance and safety of fuel and cladding candidates with enhanced accident tolerance. In Proceedings of Top Fuel 2015 (pp. 10-20). Zurich, Switzerland, September 13-17, 2015. | Publication |
| FY2015 | Craft, A. E., Chichester, D. L., Papaioannou, G. C., and Williams, W. J. (2015). Qualification of a neutron computed radiography system. FY-15 status report (INL/LTD-15-36644). Idaho National Laboratory. | |
| FY2015 | Craft, A. E., Wachs, D. M., Okuniewski, M. A., Chichester, D. L., Williams, W. J., Papaioannou, G. C., and Smolinski, A. T. (2015). Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory. Physics Procedia, 69, 483-490. | Publication |
| FY2015 | Davis, C. B., and Woolstenhulme, N. E. (2015, August 10-12). Validation of a RELAP5-3D point kinetics model of TREAT. Paper presented at the 2015 International RELAP Users Group Meeting, Idaho Falls, ID. | Publication |
| FY2015 | Field, K. G., Hu, X., Littrell, K. C., Yamamoto, Y., and Snead, L. L. (2015). Radiation tolerance of neutron-irradiated model FeCrAl alloys. Journal of Nuclear Materials, 465, 746-755. | Publication |
| FY2015 | Galloway, J., and Unal, C. (2016). Accident-tolerant-fuel performance analysis of APMT steel clad/UO2 fuel and APMT steel clad/UN-U3Si5 fuel concepts. Nuclear Science and Engineering, 182(4), 523-537. | Publication |
| FY2015 | Galloway, J., Unal, C., Carlson, N., Porter, D., and Hayes, S. (2015). Modeling constituent redistribution in U-Pu-Zr metallic fuel using the advanced fuel performance code BISON. Nuclear Engineering and Design, 286, 1-17. | Publication |
| FY2015 | George, N. M., Powers, J. J., Maldonado, G. I., Worrall, A., and Terrani, K. A. (2015). Demonstration of a full-core reactivity equivalence for FeCrAl enhanced accident tolerant fuel in BWRs. In Proceedings of Advances in Nuclear Fuel Management V (ANFM V) (p. 31). Hilton Head Island, South Carolina, USA, March 29 - April 1, 2015. | Publication |
| FY2015 | George, N. M., Terrani, K., Powers, J., Worrall, A., and Maldonado, I. (2015). Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors. Annals of Nuclear Energy, 75, 703-712. | Publication |
| FY2015 | Gussev, M. N., Byun, T. S., Yamamoto, Y., Maloy, S. A., and Terrani, K. A. (2015). In-situ tube burst testing and high-temperature deformation behavior of candidate materials for accident tolerant fuel cladding. Journal of Nuclear Materials, 466, 417-425. | Publication |
| FY2015 | Harp, J. M., Lessing, P. A., and Hoggan, R. E. (2015). Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation. Journal of Nuclear Materials, 466, 728-738. | Publication |
| FY2015 | Hoelzer, D. T., Unocic, K. A., Sokolov, M. A., and Byun, T. S. (2016). Influence of processing on the microstructure and mechanical properties of 14YWT. Journal of Nuclear Materials, 471, 251-265. | Publication |
| FY2015 | Hu, X., Terrani, K. A., Wirth, B. D., and Snead, L. L. (2015). Hydrogen permeation in FeCrAl alloys for LWR cladding application. Journal of Nuclear Materials, 461, 282-291. | Publication |
| FY2015 | Jensen, C., Davis, C., and Woolstenhulme, N. (2015, June 7-11). Thermal analysis of TREAT experimental devices. Transactions of the American Nuclear Society, 112(1), 372. San Antonio TX. | Publication |
| FY2015 | Koyanagi, T., Kiggans, J., Shih, C., and Katoh, Y. (2015). Processing and characterization of diffusion-bonded silicon carbide joints using molybdenum and titanium interlayers. Ceramic Engineering and Science Proceedings, 35(7), 151-160. | Publication |
| FY2015 | Lim, H. C., K. Rudman, K. Krishnan, R. McDonald, P. Peralta, P. Dickerson, D. Byler, C. Stanek, K. J. McClellan. Microstructurally Explicit Study of Transport Phenomena In Uranium Oxide. In TMS 2014: 143rd Annual Meeting and Exhibition, Annual Meeting Supplemental Proceedings (pp. 1041-1047). Springer, Cham. | Publication |
| FY2015 | Lim, H. C., Rudman, K., Krishnan, K., McDonald, R., Peralta, P., Dickerson, P., Byler, D., Stanek, C., and McClellan, K. J. (2013). Microstructural effects on thermal conductivity of uranium oxide: A 3D multi-physics simulation. In Proceedings of the ASME 2013 International Mechanical Engineering Congress and Exposition, Volume 6B: Energy (Paper No. V06BT07A056). San Diego, California, USA, November 15-21, 2013. ASME. | Publication |
| FY2015 | Maloy, S. A., Saleh, T. A., Anderoglu, O., Romero, T. J., Odette, G. R., Yamamoto, T., Li, S., Cole, J. I., and Fielding, R. (2016). Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ~295°C to ~6.5 dpa. Journal of Nuclear Materials, 468, 232-239. | Publication |
| FY2015 | McMurray, J. W., Shin, D., and Besmann, T. M. (2015). Thermodynamic assessment of the U-La-O system. Journal of Nuclear Materials, 456, 142-150. | Publication |
| FY2015 | Nelson, A. T., White, J. T., Byler, D. D., Dunwoody, J. T., Valdez, J. A., and McClellan, K. J. (2014). Overview of properties and performance of uranium-silicide compounds for light water reactor applications. Transactions of the American Nuclear Society, 110(1), 987-989. | Publication |
| FY2015 | Parish, C. M., Field, K. G., Certain, A. G., and Wharry, J. P. (2015). Application of STEM characterization for investigating radiation effects in BCC Fe-based alloys. Journal of Materials Research, 30(9), 1275-1289. | Publication |
| FY2015 | Pint, B. A., Terrani, K. A., Yamamoto, Y., and Snead, L. L. (2015). Material selection for accident tolerant fuel cladding. Metallurgical and Materials Transactions E, 2, 190-196. | Publication |
| FY2015 | Pint, B. A., Unocic, K. A., and Terrani, K. A. (2015). Effect of steam on high temperature oxidation behaviour of alumina-forming alloys. Materials at High Temperatures, 32(1-2), 28-35. | Publication |
| FY2015 | Porter, D. L., Chichester, H. J. M., Medvedev, P. G., Hayes, S. L., and Teague, M. C. (2015). Performance of low smeared density sodium-cooled fast reactor metal fuel. Journal of Nuclear Materials, 465, 464-470. | Publication |
| FY2015 | Powers, J. J., Worrall, A., Robb, K. R., George, N. M., and Maldonado, G. I. ORNL analysis of operational and safety performance for candidate accident tolerant fuel and cladding concepts. In Accident tolerant fuel concepts for light water reactors: Proceedings of a technical meeting (pp. 253-273). IAEA-TECDOC-1797. International Atomic Energy Agency October 13-17, 2014 | Publication |
| FY2015 | Robb, K. R. (2015). Analysis of the FeCrAl accident tolerant fuel concept benefits during BWR station blackout accidents. In Proceedings of NURETH-16. Chicago, IL, USA, August 30-September 4, 2015. | Publication |
| FY2015 | Robb, K. R., and Powers, J. J. (2014, October 27-30). Predicted system response to station blackout severe accident in a boiling water reactor employing FeCrAl cladding [Poster presentation]. NuMat 14: The Nuclear Materials Conference, Clearwater, Florida. | |
| FY2015 | Shih, C., Katoh, Y., Kiggans, J., Koyanagi, T., Khalifa, H. E., Back, C. A., Hinoki, T., and Ferraris, M. (2015). Comparison of shear strength of ceramic joints determined by various test methods with small specimens. Ceramic Engineering and Science Proceedings, 35(7), 139-149. | Publication |
| FY2015 | Shih, C., Katoh, Y., Ozawa, K., Lara-Curzio, E., and Snead, L. (2015). Through thickness mechanical properties of chemical vapor infiltration and nano-infiltration and transient eutectic-phase processed SiC/SiC composites. International Journal of Applied Ceramic Technology, 12(3), 481-490. | Publication |
| FY2015 | Silva, C. M., Hunt, R. D., Snead, L. L., and Terrani, K. A. (2015). Synthesis of phase-pure U2N3 microspheres and its decomposition into UN. Inorganic Chemistry, 54(1), 293-298. | Publication |
| FY2015 | Silva, C. M., Katoh, Y., Voit, S. L., and Snead, L. L. (2015). Chemical reactivity of CVC and CVD SiC with UO2 at high temperatures. Journal of Nuclear Materials, 460, 52-59. | Publication |
| FY2015 | Silva, C. M., Lindemer, T. B., Voit, S. R., Hunt, R. D., Besmann, T. M., Terrani, K. A., and Snead, L. L. (2014). Characteristics of uranium carbonitride microparticles synthesized using different reaction conditions. Journal of Nuclear Materials, 454(1-3), 405-412. | Publication |
| FY2015 | Silva, C., Hunt, R., Snead, L., and Terrani, K. A. (2015). Synthesis of phase-pure U2N3 microspheres and its decomposition into UN. Inorganic Chemistry, 54(1), 293-298. | Publication |
| FY2015 | Snead, L. L., Katoh, Y., and Terrani, K. (2015). Discussion of minimum stress allowables for SiC composite cladding. Transactions of the American Nuclear Society, 112(1), 280-283. | Publication |
| FY2015 | Sooby Wood, E., Parker, S. S., Nelson, A. T., and Maloy, S. A. (2016). MoSi2 oxidation in 670-1498 K water vapor. Journal of the American Ceramic Society, 99(4), 1412-1419. | Publication |
| FY2015 | Terrani, K. A., and Silva, C. M. (2015). High temperature steam oxidation of SiC coating layer of TRISO fuel particles. Journal of Nuclear Materials, 460, 160-165. | Publication |
| FY2015 | Terrani, K. A., Kiggans, J. O., Silva, C. M., Shih, C., Katoh, Y., and Snead, L. L. (2015). Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form. Journal of Nuclear Materials, 457, 9-17. | Publication |
| FY2015 | Terrani, K. A., Yang, Y., Kim, Y.-J., Rebak, R., Meyer, H. M., and Gerczak, T. J. (2015). Hydrothermal corrosion of SiC in LWR coolant environments in the absence of irradiation. Journal of Nuclear Materials, 465, 488-498. | Publication |
| FY2015 | White, J. T., Nelson, A. T., Byler, D. D., Safarik, D. J., Dunwoody, J. T., and McClellan, K. J. (2015). Thermophysical properties of U3Si5 to 1773K. Journal of Nuclear Materials, 456, 442-448. | Publication |
| FY2015 | White, J. T., Nelson, A. T., Dunwoody, J. T., and McClellan, K. J. (2014). Oxidation resistance of uranium-silicide bearing composites for advanced nuclear reactor applications. Transactions of the American Nuclear Society, 110(1), 840-841. | Publication |
| FY2015 | White, J. T., Nelson, A. T., Dunwoody, J. T., Byler, D. D., Safarik, D. J., and McClellan, K. J. (2015). Thermophysical properties of U3Si2 to 1773K. Journal of Nuclear Materials, 464, 275-280. | Publication |
| FY2015 | Woolstenhulme, N. E. and D. M. Wachs, TREAT Water Loop Summary for IRP-NE-1, Task 2b, INL/EXT-14-33641, Rev 0, November 2014. | |
| FY2015 | Woolstenhulme, N. E., et al. (2015, August 25-27). ATF design for transient testing. AFC Integration Meeting, Brookhaven National Laboratory (BNL). | |
| FY2015 | Woolstenhulme, N. E., Wachs, D. M., and Beasley, A. A. (2014, November 9-13). Transient experiment design for accident tolerance fuels. Transactions of the American Nuclear Society, 111(1), 604-606, Anaheim CA. | Publication |
| FY2015 | Woolstenhulme, N., Baker, C. C., Bess, J. D., Davis, C., Housley, G. K., Jensen, C., O'Brien, R. C., and Snow, S. D. (2015, June 7-11). TREAT experiment vehicle design and future plans. Transactions of the American Nuclear Society, 112(1), 369-371. | Publication |
| FY2014 | Baek, J.-H., Byun, T. S., Maloy, S. A., and Toloczko, M. B. (2014). Investigation of temperature dependence of fracture toughness in high-dose HT9 steel using small-specimen reuse technique. Journal of Nuclear Materials, 444(1-3), 206-213. | Publication |
| FY2014 | Besmann, T. M., Ferber, M. K., Lin, H.-T., and Collin, B. P. (2014). Fission product release and survivability of UN-kernel LWR TRISO fuel. Journal of Nuclear Materials, 448(1-3), 412-419. | Publication |
| FY2014 | Bragg-Sitton, S. (2014). Development of advanced accident-tolerant fuels for commercial LWRs. Nuclear News, 57, 83-91. | Publication |
| FY2014 | Bragg-Sitton, S. (2014). Development of enhanced accident-tolerant fuels for light water reactors. In 2015 McGraw-Hill Yearbook of Science and Technology. | |
| FY2014 | Bragg-Sitton, S., Merrill, B., Teague, M., Ott, L., Robb, K., Farmer, M., Billone, M., Montgomery, R., Stanek, C., Todosow, M., and Brown, N. (2014). Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics (INL/EXT-13-29957 and FCRD-FUEL-2013-000264). Idaho National Laboratory. | Publication |
| FY2014 | Brown, N. R., Aronson, A., Todosow, M., Brito, R., and McClellan, K. J. (2014). Neutronic performance of uranium nitride composite fuels in a PWR. Nuclear Engineering and Design, 275, 393-407. | Publication |
| FY2014 | Brown, N. R., Todosow, M., and McClellan, K. J. (2014). Uranium nitride composite fuels in a pressurized water reactor: Exploration of multi-batch cycle length and UB4 admixture for reactivity control. In Proceedings of PHYSOR 2014, The Role of Reactor Physics Toward a Sustainable Future. Kyoto, Japan, September 28 â October 3, 2014. | Publication |
| FY2014 | Byun, T. S., Baek, J.-H., Anderoglu, O., Maloy, S. A., and Toloczko, M. B. (2014). Thermal annealing recovery of fracture toughness in HT9 steel after irradiation to high doses. Journal of Nuclear Materials, 449(1-3), 263-272. | Publication |
| FY2014 | Byun, T. S., Yoon, J. H., Hoelzer, D. T., Lee, Y. B., Kang, S. H., and Maloy, S. A. (2014). Process development for 9Cr nanostructured ferritic alloy (NFA) with high fracture toughness. Journal of Nuclear Materials, 449(1-3), 290-299. | Publication |
| FY2014 | Byun, T. S., Yoon, J. H., Wee, S. H., Hoelzer, D. T., and Maloy, S. A. (2014). Fracture behavior of 9Cr nanostructured ferritic alloy with improved fracture toughness. Journal of Nuclear Materials, 449(1-3), 39-48. | Publication |
| FY2014 | Carmack, J. (2014). Accident tolerant fuel development program. Nuclear Plant Journal, 46(1), January-February. | |
| FY2014 | Collin, B. P. (2014). Modeling and analysis of UN TRISO fuel for LWR application using the PARFUME code. Journal of Nuclear Materials, 451(1-3), 65-77. | Publication |
| FY2014 | Cunningham, N. J., Wu, Y., Etienne, A., Haney, E. M., Odette, G. R., Stergar, E., Hoelzer, D. T., Kim, Y. D., Wirth, B. D., and Maloy, S. A. (2014). Effect of bulk oxygen on 14YWT nanostructured ferritic alloys. Journal of Nuclear Materials, 444(1-3), 35-38. | Publication |
| FY2014 | Farmer, M. T., Leibowitz, L., Terrani, K. A., and Robb, K. R. (2014). Scoping assessments of ATF impact on late-stage accident progression including molten core-concrete interaction. Journal of Nuclear Materials, 448(1-3), 534-540. | Publication |
| FY2014 | George, N. M., Maldonado, I., Terrani, K., Godfrey, A., Gehin, J., and Powers, J. (2014). Neutronics studies of uranium-bearing fully ceramic microencapsulated fuel for pressurized water reactors. Nuclear Technology, 188(3), 238-251. | Publication |
| FY2014 | He, L. F., Pakarinen, J., Kirk, M. A., Gan, J., Nelson, A. T., Bai, X.-M., El-Azab, A., and Allen, T. R. (2014). Microstructure evolution in Xe-irradiated UO2 at room temperature. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 330, 55-60. | Publication |
| FY2014 | He, L.-F., Gupta, M., Yablinsky, C. A., Gan, J., Kirk, M. A., Bai, X.-M., Pakarinen, J., and Allen, T. R. (2013). In situ TEM observation of dislocation evolution in Kr-irradiated UO2 single crystal. Journal of Nuclear Materials, 443(1-3), 71-77. | Publication |
| FY2014 | Huang, Z., Harris, A., Maloy, S. A., and Hosemann, P. (2014). Nanoindentation creep study on an ion beam irradiated oxide dispersion strengthened alloy. Journal of Nuclear Materials, 451(1-3), 162-167. | Publication |
| FY2014 | Kato, M., Murakami, T., Sunaoshi, T., Nelson, A. T., and McClellan, K. J. (2013). Property measurements of (U0.7Pu0.3)O2-x in PO2-controlled atmosphere. In Proceedings of the International Nuclear Fuel Cycle Conference: Nuclear Energy at a Crossroads, GLOBAL 2013 (pp. 852-856). Salt Lake City, UT, United States, September 29 - October 2, 2013. | Publication |
| FY2014 | Katoh, Y., Snead, L. L., Cheng, T., Shih, C., Lewis, W. D., Koyanagi, T., Hinoki, T., Henager, C. H., and Ferraris, M. (2014). Radiation-tolerant joining technologies for silicon carbide ceramics and composites. Journal of Nuclear Materials, 448(1-3), 497-511. | Publication |
| FY2014 | Katoh, Y., Terrani, K. A., and Snead, L. L. (2014). Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures. ORNL/TM-2014/210, Revision 1, Oak Ridge National Laboratory. | Publication |
| FY2014 | Koyanagi, T., Kiggans, J., Shih, C., and Katoh, Y. (2014). Pressureless joining of SiC by transient eutectic-phase method. Transactions of the American Nuclear Society, 110(1), 863-864. | Publication |
| FY2014 | Koyanagi, T., Kiggans, J., Shih, C., and Katoh, Y. (2014). Processing and characterization of diffusion-bonded silicon carbide joints using molybdenum and titanium interlayers. In Ceramic Materials for Energy Applications IV (pp. 151-160). | Publication |
| FY2014 | Mosbrucker, P. L., Brown, D. W., Anderoglu, O., Balogh, L., Maloy, S. A., Sisneros, T. A., Almer, J., Tulk, E. F., Morgenroth, W., and Dippel, A. C. (2013). Neutron and X-ray diffraction analysis of the effect of irradiation dose and temperature on microstructure of irradiated HT-9 steel. Journal of Nuclear Materials, 443(1-3), 522-530. | Publication |
| FY2014 | Nelson, A. T., Rittman, D. R., White, J. T., Dunwoody, J. T., Kato, M., and McClellan, K. J. (2014). An evaluation of the thermophysical properties of stoichiometric CeO2 in comparison to UO2 and PuO2. Journal of the American Ceramic Society, 97(11), 3652-3659. | Publication |
| FY2014 | Nelson, A. T., Sooby, E. S., Kim, Y.-J., Cheng, B., and Maloy, S. A. (2014). High temperature oxidation of molybdenum in water vapor environments. Journal of Nuclear Materials, 448(1-3), 441-447. | Publication |
| FY2014 | Ott, L. J., Robb, K. R., and Wang, D. (2014). Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions. Journal of Nuclear Materials, 448(1-3), 520-533. | Publication |
| FY2014 | Pint, B. A., Dryepondt, S., Unocic, K. A., and Hoelzer, D. T. (2014). Development of ODS FeCrAl for compatibility in fusion and fission energy applications. JOM, 66(12), 2458-2466. | Publication |
| FY2014 | Powers, J. J., George, N. M., Worrall, A., and Terrani, K. A. (2014). Reactor physics assessment of alternate cladding materials. In Proceedings of 2014 Water Reactor Fuel Performance Meeting/Top Fuel/LWR Fuel Performance Meeting (WRFPM 2014). Sendai, Miyagi, Japan, September 14-17, 2014. | Publication |
| FY2014 | Shih, C., Katoh, Y., Kiggans, J. O., Koyanagi, T., Khalifa, H. E., Back, C. A., Hinoki, T., and Ferraris, M. (2014). Comparison of shear strength of ceramic joints determined by various test methods with small specimens. In A. Gyekenyesi, M. Halbig, H.-T. Lin, Y. Katoh,; J. Mat (Eds.), Ceramic Materials for Energy Applications IV. | Publication |
| FY2014 | Teague, M., and Gorman, B. (2014). Utilization of dual-column focused ion beam and scanning electron microscope for three-dimensional characterization of high burn-up mixed oxide fuel. Progress in Nuclear Energy, 72, 67-71. | Publication |
| FY2014 | Teague, M., Gorman, B., King, J., Porter, D., and Hayes, S. (2013). Microstructural characterization of high burn-up mixed oxide fast reactor fuel. Journal of Nuclear Materials, 441(1-3), 267-273. | Publication |
| FY2014 | Teague, M., Gorman, B., Miller, B., and King, J. (2014). EBSD and TEM characterization of high burn-up mixed oxide fuel. Journal of Nuclear Materials, 444(1-3), 475-480. | Publication |
| FY2014 | Teague, M., Tonks, M., Novascone, S., and Hayes, S. (2014). Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel. Journal of Nuclear Materials, 444(1-3), 161-169. | Publication |
| FY2014 | Toloczko, M. B., Garner, F. A., Voyevodin, V. N., Bryk, V. V., Borodin, O. V., Melnychenko, V. V., and Kalchenko, A. S. (2014). Ion-induced swelling of ODS ferritic alloy MA957 tubing to 500 dpa. Journal of Nuclear Materials, 453(1-3), 323-333. | Publication |
| FY2014 | Unocic, K. A., Hoelzer, D. T., and Pint, B. A. (2015). Microstructure and environmental resistance of low Cr ODS FeCrAl. Materials at High Temperatures, 32(1-2), 123-132. | Publication |
| FY2014 | Was, G. S., Jiao, Z., Getto, E., Sun, K., Monterrosa, A. M., Maloy, S. A., Anderoglu, O., Sencer, B. H., and Hackett, M. (2014). Emulation of reactor irradiation damage using ion beams. Scripta Materialia, 88, 33-36. | Publication |
| FY2014 | White, J. T., Nelson, A. T., Byler, D. D., Valdez, J. A., and McClellan, K. J. (2014). Thermophysical properties of U3Si to 1150K. Journal of Nuclear Materials, 452(1-3), 304-310. | Publication |
| FY2013 | Anderoglu, O., Byun, T. S., Toloczko, M., et al. (2013). Mechanical performance of ferritic martensitic steels for high dose applications in advanced nuclear reactors. Metallurgical and Materials Transactions A, 44(Suppl 1), 70-83. | Publication |
| FY2013 | Bhattacharyya, D., Dickerson, P., Odette, G. R., Maloy, S. A., Misra, A., and Nastasi, M. A. (2012). On the structure and chemistry of complex oxide nanofeatures in nanostructured ferritic alloy U14YWT. Philosophical Magazine, 92(16), 2089-2107. | Publication |
| FY2013 | Brown, N. R., Ludewig, H., Aronson, A., Raitses, G., and Todosow, M. (2013). Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part I: Lattice benchmarking, cycle length, and reactivity coefficients. Annals of Nuclear Energy, 62, 538-547. | Publication |
| FY2013 | Brown, N. R., Ludewig, H., Aronson, A., Raitses, G., and Todosow, M. (2013). Neutronic evaluation of a PWR with fully ceramic microencapsulated fuel. Part II: Nodal core calculations and preliminary study of thermal hydraulic feedback. Annals of Nuclear Energy, 62, 548-557. | Publication |
| FY2013 | Byun, T. S., Toloczko, M. B., Saleh, T. A., and Maloy, S. A. (2013). Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF. Journal of Nuclear Materials, 432(1-3), 1-8. | Publication |
| FY2013 | Crapps, J., DeCroix, D. S., Galloway, J. D., Korzekwa, D. A., Aikin, R., Fielding, R., Kennedy, R., and Unal, C. (2013). Separate effects identification via casting process modeling for experimental measurement of U-Pu-Zr alloys. Journal of Nuclear Materials, 443(1-3), 176-184. | Publication |
| FY2013 | Daw, J. E., Rempe, J. L., and Knudson, D. L. (2012). Hot wire needle probe for in-reactor thermal conductivity measurement. IEEE Sensors Journal, 12(8), 2554-2560. | Publication |
| FY2013 | Daw, J., Rempe, J., Palmer, J., Ramuhalli, P., Montgomery, R., Chien, H.-T., Tittmann, B., Reinhardt, B., and Kohse, G. (2013). Irradiation testing of ultrasonic transducers. In 2013 3rd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and their Applications (ANIMMA) (pp. 1-7). Marseille, France. Invited paper for ANIMMA 2013 Special Edition. | Publication |
| FY2013 | Egeland, G. W., Mariani, R. D., Hartmann, T., Porter, D. L., Hayes, S. L., and Kennedy, J. R. (2013). Reducing fuel-cladding chemical interaction: The effect of palladium on the reactivity of neodymium on iron in diffusion couples. Journal of Nuclear Materials, 432(1-3), 539-544. | Publication |
| FY2013 | Egeland, G. W., Valdez, J. A., Maloy, S. A., McClellan, K. J., Sickafus, K. E., and Bond, G. M. (2013). Heavy-ion irradiation defect accumulation in ZrN characterized by TEM, GIXRD, nanoindentation, and helium desorption. Journal of Nuclear Materials, 435(1-3), 77-87. | Publication |
| FY2013 | Hurley, D., Khafizov, M., Kennedy, R., and Burgett, E. (2013). Mechanical properties of nuclear fuel surrogates using picosecond laser ultrasonics. Proceedings of the 2013 International Congress on Ultrasonics, 686. Siong, G.W., Piang L.S., Cheong, K.B. eds., May 2-5, 2013. | Publication |
| FY2013 | Kim, B. G., Rempe, J. L., Knudson, D. L., Condie, K. G., and Sencer, B. H. (2012). In-situ creep testing capability for the Advanced Test Reactor. Nuclear Technology, 179(3), 417-428. | Publication |
| FY2013 | Kim, J. H., Byun, T. S., Hoelzer, D. T., Kim, S.-W., and Lee, B. H. (2013). Temperature dependence of strengthening mechanisms in the nanostructured ferritic alloy 14YWT: Part I-Mechanical and microstructural observations. Materials Science and Engineering: A, 559, 101-110. | Publication |
| FY2013 | Kim, J. H., Byun, T. S., Hoelzer, D. T., Park, C. H., Yeom, J. T., and Hong, J. K. (2013). Temperature dependence of strengthening mechanisms in the nanostructured ferritic alloy 14YWT: Part II- Mechanistic models and predictions. Materials Science and Engineering: A, 559, 111-118. | Publication |
| FY2013 | Koenig, T. W., Olson, D. L., Mishra, B., King, J. C., Fletcher, J., Gerstenberger, L., Lawrence, S., Martin, A., Mejia, C., Meyer, M. K., Kennedy, R., Hu, L., Kohse, G., and Terry, J. (2011). Advanced non-destructive assessment technology to determine the aging of silicon containing materials for Generation IV nuclear reactors. AIP Conference Proceedings, 1335, 1200-1207. Melville, NY, 2012. | Publication |
| FY2013 | Lim, H. C., Rudman, K., Krishnan, K., McDonald, R., Dickerson, P., Byler, D., and McClellan, K. (2013). Microstructurally explicit simulation of intergranular mass transport in oxide nuclear fuels. Nuclear Technology, 182(2), 155-163. | Publication |
| FY2013 | Mariani, R. D., Porter, D. L., Hayes, S. L., and Kennedy, J. R. (2012). Metallic fuels: The EBR-II legacy and recent advances. Procedia Chemistry, 7, 513-520. | Publication |
| FY2013 | McMurray, J. W., Shin, D., Slone, B. W., and Besmann, T. M. (2013). Thermochemical modeling of the U1yGdyO2±x phase. Journal of Nuclear Materials, 443(1-3), 588-595. | Publication |
| FY2013 | Morris, C., Bourke, M., Byler, D., Chen, C., Hogan, G., Hunter, J., Kwiatkowski, K., Mariam, F., McClellan, K. J., Merrill, F., Morley, D., and Saunders, A. (2013). Qualitative comparison of bremsstrahlung X-rays and 800 MeV protons for tomography of urania fuel pellets. Review of Scientific Instruments, 84(2), 023902-1-7. | Publication |
| FY2013 | Nelson, A. T., Giachino, M. M., Nino, J. C., and McClellan, K. J. (2014). Effect of composition on thermal conductivity of MgO-Nd2Zr2O7 composites for inert matrix materials. Journal of Nuclear Materials, 444(1-3), 385-392. | Publication |
| FY2013 | Park, Y., Huang, K., Paz y Puente, A., et al. (2015). Diffusional interaction between U-10 wt pct Zr and Fe at 903 K, 923 K, and 953 K (630°C, 650°C, and 680°C). Metallurgical and Materials Transactions A, 46(1), 72-82. | Publication |
| FY2013 | Rudman, K., Dickerson, P., Byler, D., McDonald, R., Lim, H., Peralta, P., and McClellan, K. (2013). Three-dimensional characterization of sintered UO2+x: Effects of oxygen content on microstructure and its evolution. Nuclear Technology, 182(2), 145-154. | Publication |
| FY2013 | Shin, D., and Besmann, T. M. (2013). Thermodynamic modeling of the (U,La)O2 ±x solid solution phase. Journal of Nuclear Materials, 433(1-3), 227-232. | Publication |
| FY2013 | Toloczko, M. B., Garner, F. A., and Maloy, S. A. (2012). Irradiation creep and density changes observed in MA957 pressurized tubes irradiated to doses of 40-110 dpa at 400-750°C in FFTF. Journal of Nuclear Materials, 428(1-3), 170-175. | Publication |
| FY2013 | Usov, I. O., Dickerson, R. M., Dickerson, P. O., Hawley, M. E., Byler, D. D., and McClellan, K. J. (2013). Thin uranium dioxide films with embedded xenon. Journal of Nuclear Materials, 437(1-3), 1-5. | Publication |
| FY2013 | Wei, C.-C., Aitkaliyeva, A., Luo, Z., Ewh, A., Sohn, Y. H., Kennedy, J. R., Sencer, B. H., Myers, M. T., Martin, M., Wallace, J., General, M. J., and Shao, L. (2013). Understanding the phase equilibrium and irradiation effects in FeZr diffusion couples. Journal of Nuclear Materials, 432(1-3), 205-211. | Publication |
| FY2013 | White, J. T., and Nelson, A. T. (2013). Thermal conductivity of UO2+x and U4O9-y. Journal of Nuclear Materials, 443(1-3), 342-350. | Publication |
| FY2013 | Xing, C., Jensen, C., Hua, Z., Ban, H., Hurley, D. H., Khafizov, M., and Kennedy, J. R. (2012). Parametric study of the frequency-domain thermoreflectance technique. Journal of Applied Physics, 112(10), 103105. | Publication |
| FY2012 | Anderoglu, O., Van den Bosch, J., Hosemann, P., Stergar, E., Sencer, B. H., Bhattacharyya, D., Dickerson, R., Dickerson, P., Hartl, M., and Maloy, S. A. (2012). Phase stability of an HT-9 duct irradiated in FFTF. Journal of Nuclear Materials, 430(1-3), 194-204. | Publication |
| FY2012 | Bell, G. L., McDuffee, J. L., Ellis, R. J., Glasgow, D. C., Sitterson, R. G., Snead, L. L., and Schmidlin, J. E. (2012). Summary of FY12 HFIR irradiation testing activities (ORNL/TM-2012/454). Oak Ridge National Laboratory. | |
| FY2012 | Carmack, W. J., Porter, D. L., Chichester, H. J., and Hayes, S. L. (2012, September 24-27). Irradiation performance comparison of experimental and prototypic length metallic (U-10Zr) fuel. In 12th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, Prague, Czech Republic. | Publication |
| FY2012 | Chao-Chen Wei, Assel Aitkaliyeva, Zhiping Luo, Ashley Ewh, Y.H. Sohn, J.R. Kennedy, | |
| FY2012 | Chichester, H. J. M., Mariani, R. D., Hayes, S. L., Kennedy, J. R., Wright, A. E., Kim, Y. S., Yacout, A. M., and Hofman, G. L. (2012). Advanced metallic fuel for ultra-high burnup: Irradiation tests in ATR. Transactions, 106(1), 1349-1351. In Embedded Topical Meeting: Nuclear Fuel and Structural Materials for the Next Generation Nuclear Reactors (NFSM 2012) | Poster Session. Chicago, IL, 24-28 June 2012. | Publication |
| FY2012 | Chichester, H. J. M., Porter, D. L., and Hayes, S. L. (2012, September 24-27). Postirradiation examination of high burnup metallic fuels for transmutation. In HOTLAB 2012. The 49th Conference on Hot Laboratories and Remote Handling, Marcoule, France. 24-27 September 2012. | Publication |
| FY2012 | Farzbod, F., and Hurley, D. H. (2012). Resonant ultrasound spectroscopy: Using mode shapes. IEEE Transactions on Ultrasonics, Ferroelectrics, and Frequency Control, 59(11), 2413-2421. | Publication |
| FY2012 | Hua, Z., Ban, H., Khafizov, M., Schley, R., Kennedy, J. R., and Hurley, D. H. (2012). Spatially localized measurement of thermal conductivity using a hybrid photothermal technique. Journal of Applied Physics, 111(11), 113505. | Publication |
| FY2012 | Huang, K., Park, Y., Ewh, A., Sencer, B. H., Kennedy, J. R., Coffey, K. R., and Sohn, Y. H. (2012). Interdiffusion and reaction between uranium and iron. Journal of Nuclear Materials, 424(1-3), 82-88. | Publication |
| FY2012 | Hurley, D. H., Reese, S. J., and Farzbod, F. (2012). Application of laser-based resonant ultrasound spectroscopy to study texture in copper. Journal of Applied Physics, 111(5), 053527. | Publication |
| FY2012 | Mariani, R. D., Porter, D. L., O'Holleran, T. P., Hayes, S. L., and Kennedy, J. R. (2011). Lanthanides in metallic nuclear fuels: Their behavior and methods for their control. Journal of Nuclear Materials, 419(1-3), 263-271. | Publication |
| FY2012 | McDonald, R., Rudman, K., Luther, E., Peralta, P., Stanek, C., and McClellan, K. (2012). Porosity characterization of surrogates for oxide nuclear fuels: A statistical analysis of correlations among grain boundary misorientation and pore character and location. Poster presentation at the TMS Annual Meeting, Orlando, FL. 2012. Poster presentation. | |
| FY2012 | Pint, B. A., Brady, M. P., Keiser, J. R., Cheng, T., and Terrani, K. A. (2012, May). High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments. In Proceedings of the 8th International Symposium on High Temperature Corrosion and Protection of Materials, Les Embiez, France (Paper #89). | |
| FY2012 | Teague, M. M. (2012). Post irradiation examination of legacy FFTF oxide fuel (INL/LTD-1226386). | |
| FY2012 | Usov, I. O., Won, J., Devlin, D. J., Jiang, Y.-B., Valdez, J. A., and Sickafus, K. E. (2011). A novel method for incorporating fission gas elements into solids. Journal of Nuclear Materials, 408(2), 205-208. | Publication |
| FY2012 | Wright, A. E., Hayes, S. L., Bauer, T. H., Chichester, H. J., Hofman, G. L., Kennedy, J. R., Kim, T. K., Kim, Y. S., Mariani, R. D., Pointer, W. D., Yacout, A. M., and Yun, D. (2012). Development of advanced ultra-high burnup SFR metallic fuel concept - Project overview. Transactions, 106(1), 1102-1105. In Embedded Topical Meeting: Nuclear Fuel and Structural Materials for the Next Generation Nuclear Reactors (NFSM 2012) | Advanced Fuel - I. Chicago, IL, 24-28 June 2012. | Publication |
| FY2011 | Beeler, B., Good, B., Rashkeev, S., Deo, C., Baskes, M., and Okuniewski, M. (2010). First principles calculations for defects in U. Journal of Physics: Condensed Matter, 22(50), 505703. | Publication |
| FY2011 | Beeler, B., Good, B., Rashkeev, S., Deo, C., Baskes, M., and Okuniewski, M. (2012). First-principles calculations of the stability and incorporation of helium, xenon and krypton in uranium. Journal of Nuclear Materials, 425(1-3), 2-7. | Publication |
| FY2011 | Bell, G. L., McDuffee, J., Hobbs, R. W., Ott, L. J., Ellis, R. J., Okuniewski, M. A., and Hayes, S. L. (2010, November). Preparations for low fluence fuels testing in the High Flux Isotope Reactor hydraulic rabbit facility. In OECD Nuclear Energy Agency Eleventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, San Francisco, CA. | |
| FY2011 | Cole, J. I., O'Holleran, T. P., Keiser, D. D., Jr., and Kennedy, J. R. (2011). Out-of-pile effects of lanthanides on fuel-cladding compatibility. Journal of Nuclear Materials. | |
| FY2011 | Daw, J. E., Rempe, J. L., and Wilkins, S. C. and Idaho National Laboratory (United States) (2002). Ultrasonic thermometry for in-pile temperature detection. In Proceedings of the 7th International Topical Meeting on Nuclear Plant Instrumentation, Control and Human-Machine Interface Technologies (NPICandHMIT 2002), Las Vegas, NV, United States. | Publication |
| FY2011 | Hurley, D. H., Khafizov, M., Shinde, S., Hurley, D. (2011). Measurement of Kapitza resistance across a bicrystal interface. Journal of Applied Physics, 109(8), 083504. | Publication |
| FY2011 | Janney, D. E., Kennedy, J. R. (2010). As-cast microstructures in U-Pu-Zr alloy fuel pins with 5-8 wt.% minor actinides and 0-1.5 wt% rare-earth elements. Materials Characterization, 61(11), 1194-1202 | Publication |
| FY2011 | Knudson, D. L. (2012). Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing. Measurement Science and Technology, 23(2), 025604. | Publication |
| FY2011 | Knudson, D., and Rempe, J. and Idaho National Laboratory (United States) (2001). Recommendations for use of LVDTs in ATR high temperature irradiation testing. In Proceedings of the 7th International Topical Meeting on Nuclear Plant Instrumentation, Control and Human-Machine Interface Technologies (NPICandHMIT 2001), Las Vegas, NV, United States. | Publication |
| FY2011 | Mariani, R. D. (2011). Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys and related methods (U.S. Patent Application No. 13/021,480). U.S. Patent and Trademark Office. | |
| FY2011 | Mohanty, R. R., Bush, J., Okuniewski, M. A., Sohn, Y. H. (2011). Thermotransport in γ(bcc) U-Zr alloys: A phase-field model study. Journal of Nuclear Materials, 414(2), 211-216. | Publication |
| FY2011 | Myers, M. T., Sencer, B. H., and Shao, L. (2012). Multi-scale modeling of localized heating caused by ion bombardment. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 272, 165-168. | Publication |
| FY2011 | Rempe, J., Knudson, D. L., Daw, J., Condie, K. G., Palmer, J. R., Skerjanc, W. F., Wilkins, S. C., and Davis, K. L. (2012). Enhanced in-pile instrumentation at the Advanced Test Reactor. IEEE Transactions on Nuclear Science, 59(4), 1214-1223. | Publication |
| FY2011 | Xing, C., Hua, Z., Ban, H., Hurley, D., and Kennedy, J. R. (2011). Evaluation of uncertainties of one-directional analytical model for thermoreflectance technique. Proceedings of the ASME 2011 International Technical Conference and Exhibition on Packaging and Integration of Electronic and Photonic Microsystems, AJTEC2011-44539, T10057. | Publication |
| FY2011 | Xing, C., Jensen, C., Ban, H., Mariani, R., and Kennedy, J. R. (2010). An electromotive force measurement system for alloy fuels. In Proceedings of the ASME 2010 International Mechanical Engineering Congress and Exposition, Volume 7: Fluid Flow, Heat Transfer and Thermal Systems, Parts A and B (pp. 403-408). Vancouver, British Columbia, Canada. American Society of Mechanical Engineers. ASME. | Publication |
| FY2010 | Choi, K., Tong, W., Maiani, R. D., Burkes, D. E., and Munir, Z. A. (2010). Densification of nano-CeO2 ceramics as nuclear oxide surrogate by spark plasma sintering. Journal of Nuclear Materials, 404(3), 210-216. | Publication |
| FY2010 | de Almeida, V. F., Hunt, R. D., and Collins, J. L. (2010). Pneumatic drop-on-demand generation for production of metal oxide microspheres by internal gelation. Journal of Nuclear Materials, 404(1), 44-49. | Publication |
| FY2010 | Hunt, R. D., Hunn, J. D., Birdwell, J. F., Lindemer, T. B., and Collins, J. L. (2010). The addition of silicon carbide to surrogate nuclear fuel kernels made by the internal gelation process. Journal of Nuclear Materials, 401(1-3), 55-59. | Publication |
| FY2010 | Hunt, R. D., Montgomery, F. C., and Collins, J. L. (2010). Treatment techniques to prevent cracking of amorphous microspheres made by the internal gelation process. Journal of Nuclear Materials, 405(2), 160-164. | Publication |
| FY2010 | Hurley, D. H., Reese, S. J., Park, S. K., Utegulov, Z., Kennedy, J. R., and Telschow, K. L. (2010). In situ laser-based resonant ultrasound measurements of microstructure mediated mechanical property evolution. Journal of Applied Physics, 107(6), 063510. | Publication |
| FY2010 | Mariani, R. (2010). Dopants for high burnup in metallic nuclear fuels. U.S. Patent No. 12/702,077 | |
| FY2010 | Mariani, R. (2010). Nuclear fuel bodies having shell and core regions, nuclear reactors including such nuclear fuel bodies, and related methods. U.S. Patent No. 12/893,503. | |
| FY2010 | Mohammadian, M. A., Allen, T. R., Sridharan, K., Cole, J. I., Fielding, R. F., and Young, C. (n.d.). Characterization of vanadium-lined fuel cladding fabricated with various process parameters. | |
| FY2010 | Nerikar, P. V., Rudman, K., Desai, T. G., Byler, D., Unal, C., McClellan, K. J., Phillpot, S. R., Sinnott, S. B., Peralta, P., Uberuaga, B. P., and Stanek, C. R. (2010). Grain boundaries in uranium dioxide: Scanning electron microscopy experiments and atomistic simulations. Journal of the American Ceramic Society, 94(6), 1893-1900. | Publication |
| FY2010 | Park, S. K., Baik, S. H., Cha, H. K., Reese, S. J., and Hurley, D. H. (2010). Characteristics of laser resonant ultrasonic spectroscopy system for measuring elastic constants of materials. Journal of the Korean Physical Society, 57, 375-379. | Publication |
| FY2010 | Rudman, K., Peralta, P., Stanek, C., Wheeler, K., Parra, M., Byler, D., and McClellan, K. (2010). Quantification of microstructure variability in surrogates for oxide nuclear fuels. In TMS Annual Meeting, Seattle, WA. | |
| FY2026 | Idaho National Laboratory. (2026, February). Advanced Fuels Campaign Newsletter (INL/MIS-26-91368). | |
| FY2026 | Chang, T. Y., Trowbridge, T. L., Burns, J., Swearingen, A. L., Yao, T., Wang, Y., … Jensen, C. B. (2026). Fuel-cladding chemical interaction in HT9/U-10Zr fuels: correlation with cladding temperature and fuel burnup from FFTF MFF series data. Journal of Nuclear Science and Technology. | Publication |
| FY2026 | Armstrong, R., Vermaak, J., Folsom, C., Seo, S., Pacheco Duarte, J., & Jensen, C. (2026). Advanced Multiphysics Code Coupling for Cladding Surface Thermocouples During Two-Phase Heat Transfer from Nuclear Fuel. Nuclear Technology, 1–15. | Publication |
| FY2026 | Yang, G., Mauseth, T. J., Pradhan, A., Shah, S., Howard, C. B., Yao, T., Matos, M. D., Schulthess, J. L., Teng, F., Murray, D. J., Bachhav, M. N., Gonderman, S., Gazza, J., Feltus, M. A., Xu, F., & Xu, P. (2026). Multi-scale, multi-modal characterization of silicon carbide cladding following safety testing in the Transient Reactor Test facility (TREAT). Materials Characterization, 237, 116416. | Publication |
| FY2025 | Wachs, D. (2025). The U.S. Next Generation Fuels (NGF) Program: Unleashing Advanced Nuclear Fuels: Unleashing Advanced Nuclear Fuels. In TopFuel 2025: Nuclear Reactor Fuel Performance Conference: Nuclear Reactor Fuel Performance Conference (pp. 1158-1164). (Proceedings of the TopFuel 2025: Nuclear Reactor Fuel Performance Conference). American Nuclear Society. https://doi.org/10.13182/TOPFUEL25-48730 | Publication |
| FY2026 | American Nuclear Society. (2025). Proceedings of the TopFuel 2025: Nuclear Reactor Fuel Performance Conference, October 5-9, 2025, Nashville, Tennessee. American Nuclear Society. https://ans.org/meetings | Publication |
