The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts

Kane, K.; Bell, Sam; Capps, Nathan; Garrison, Ben; Shapovalov, K.; Jacobsen, G.; Deck, C.; Graening, Tim Seibert; Koyanagi, Takaaki; Massey, Caleb

doi:10.1016/j.jnucmat.2022.154152

Title: The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts

Journal Article · Tue Nov 15 00:00:00 EST 2022 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2022.154152· OSTI ID:1997815

^[1]; Bell, Sam ^[2]; Capps, Nathan ^[3]; Garrison, Ben ^[3]; Shapovalov, K. ^[4]; Jacobsen, G. ^[4];

^[4]; Graening, Tim Seibert ^[3];

^[3]; Massey, Caleb ^[3]

Johns Hopkins Univ., Baltimore, MD (United States); Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States)
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
General Atomics, San Diego, CA (United States)

Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. Here, in this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO₂ formation via Cr₂O₃ formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al₂O₃ regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H₂ generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H₂ generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H₂ generation by orders of magnitude relative to bare Zry-4.

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Research Organization:: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE), Nuclear Fuel Cycle and Supply Chain. Advanced Fuel Campaign

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1997815

Journal Information:: Journal of Nuclear Materials, Vol. 574, Issue 1; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (37)

Accident tolerant fuels for LWRs: A perspective Zinkle, S. J.; Terrani, K. A.; Gehin, J. C. Journal of Nuclear Materials, Vol. 448, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2013.12.005	journal	May 2014
Design, properties, and weldability of advanced oxidation-resistant FeCrAl alloys Gussev, M. N.; Field, K. G.; Yamamoto, Y. Materials & Design, Vol. 129 https://doi.org/10.1016/j.matdes.2017.05.009	journal	September 2017
Handbook of SiC properties for fuel performance modeling Snead, Lance L.; Nozawa, Takashi; Katoh, Yutai Journal of Nuclear Materials, Vol. 371, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2007.05.016	journal	September 2007
Development of Cr cold spray–coated fuel cladding with enhanced accident tolerance Ševeček, Martin; Gurgen, Anil; Seshadri, Arunkumar Nuclear Engineering and Technology, Vol. 50, Issue 2 https://doi.org/10.1016/j.net.2017.12.011	journal	March 2018
Characterization of SiC–SiC composites for accident tolerant fuel cladding Deck, C. P.; Jacobsen, G. M.; Sheeder, J. Journal of Nuclear Materials, Vol. 466 https://doi.org/10.1016/j.jnucmat.2015.08.020	journal	November 2015
Evaluating steam oxidation kinetics of environmental barrier coatings Kane, Kenneth; Garcia, Eugenio; Stack, Padraig Journal of the American Ceramic Society, Vol. 105, Issue 1 https://doi.org/10.1111/jace.18093	journal	September 2021
Qualitative and quantitative analysis of neutron irradiation effects in SiC/SiC composites using X-ray computed tomography Arregui-Mena, José David; Koyanagi, Takaaki; Cakmak, Ercan Composites Part B: Engineering, Vol. 238 https://doi.org/10.1016/j.compositesb.2022.109896	journal	June 2022
Burst characteristics of advanced accident-tolerant FeCrAl cladding under temperature transient testing Garrison, B.; Cinbiz, M. N.; Gussev, M. Journal of Nuclear Materials, Vol. 560 https://doi.org/10.1016/j.jnucmat.2021.153488	journal	March 2022
Measurement of Zircaloy-4 cladding tube deformation using a three-dimensional digital image correlation system with internal transient heating and pressurization Kim, Dong-Hyun; Choi, Gyeong-Ha; Kim, Hyochan Nuclear Engineering and Design, Vol. 363 https://doi.org/10.1016/j.nucengdes.2020.110662	journal	July 2020
Burst behavior of nuclear grade FeCrAl and Zircaloy-2 fuel cladding under simulated cyclic dryout conditions Kane, K. A.; Lee, S. K.; Bell, S. B. Journal of Nuclear Materials, Vol. 539 https://doi.org/10.1016/j.jnucmat.2020.152256	journal	October 2020
The Stressing for Test Purposes of Materials in Tubular Form Using Elastomeric Inserts—Experimental and Theoretical Development Mosley, K. Proceedings of the Institution of Mechanical Engineers, Vol. 196, Issue 1 https://doi.org/10.1243/PIME_PROC_1982_196_014_02	journal	June 1982
Silicon carbide and its composites for nuclear applications – Historical overview Katoh, Yutai; Snead, Lance L. Journal of Nuclear Materials, Vol. 526 https://doi.org/10.1016/j.jnucmat.2019.151849	journal	December 2019
Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors Yamamoto, Y.; Pint, B. A.; Terrani, K. A. Journal of Nuclear Materials, Vol. 467 https://doi.org/10.1016/j.jnucmat.2015.10.019	journal	December 2015
Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors Brachet, Jean-Christophe; Idarraga-Trujillo, Isabel; Flem, Marion Le Journal of Nuclear Materials, Vol. 517 https://doi.org/10.1016/j.jnucmat.2019.02.018	journal	April 2019
Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions Bell, S. B.; Kane, K. A.; Massey, C. P. Journal of Nuclear Materials, Vol. 557 https://doi.org/10.1016/j.jnucmat.2021.153242	journal	December 2021
Post-transient examination of performance of uranium silicide fuel and silicon-carbide composite cladding under reactivity-initiated accident conditions Schulthess, Jason; Kamerman, David; Winston, Alexander Journal of Nuclear Materials, Vol. 560 https://doi.org/10.1016/j.jnucmat.2022.153520	journal	March 2022
C-ring testing of nuclear grade silicon carbide composites at temperatures up to 1900 °C Shapovalov, Kirill; Jacobsen, George M.; Shih, Chunghao Journal of Nuclear Materials, Vol. 522 https://doi.org/10.1016/j.jnucmat.2019.04.047	journal	August 2019
Technology readiness levels for advanced nuclear fuels and materials development Carmack, W. J.; Braase, L. A.; Wigeland, R. A. Nuclear Engineering and Design, Vol. 313 https://doi.org/10.1016/j.nucengdes.2016.11.024	journal	March 2017
Modeling Environmental Degradation of SiC-Based Fibers Parthasarathy, Triplicane A.; Przybyla, Craig P.; Hay, Randall S. Journal of the American Ceramic Society, Vol. 99, Issue 5 https://doi.org/10.1111/jace.14086	journal	February 2016
Cladding burst behavior of Fe-based alloys under LOCA Massey, Caleb P.; Terrani, Kurt A.; Dryepondt, Sebastien N. Journal of Nuclear Materials, Vol. 470 https://doi.org/10.1016/j.jnucmat.2015.12.018	journal	March 2016
Current status and recent research achievements in SiC/SiC composites Katoh, Y.; Snead, L. L.; Henager, C. H. Journal of Nuclear Materials, Vol. 455, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2014.06.003	journal	December 2014
Handbook on the Material Properties of FeCrAl Alloys for Nuclear Power Production Applications (FY18 Version: Revision 1) Field, Kevin G.; Snead, Mary A.; Yamamoto, Yukinori https://doi.org/10.2172/1474581	report	August 2018
Chemical Vapor Infiltrated SiC/SiC Composites (CVI SiC/SiC) Lamon, Jacques Handbook of Ceramic Composites https://doi.org/10.1007/0-387-23986-3_3	book	January 2005
Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors Maier, Benjamin; Yeom, Hwasung; Johnson, Greg JOM, Vol. 70, Issue 2 https://doi.org/10.1007/s11837-017-2643-9	journal	November 2017
Investigation of the C-ring test for measuring hoop tensile strength of nuclear grade ceramic composites Jacobsen, G. M.; Stone, J. D.; Khalifa, H. E. Journal of Nuclear Materials, Vol. 452, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2014.05.002	journal	September 2014
HRTEM and chemical study of an ion-irradiated chromium/zircaloy-4 interface Wu, A.; Ribis, J.; Brachet, J. -C. Journal of Nuclear Materials, Vol. 504 https://doi.org/10.1016/j.jnucmat.2018.01.029	journal	June 2018
High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments Pint, B. A.; Terrani, K. A.; Brady, M. P. Journal of Nuclear Materials, Vol. 440, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2013.05.047	journal	September 2013
Silicon Carbide Oxidation in Steam up to 2 MPa Terrani, Kurt A.; Pint, Bruce A.; Parish, Chad M. Journal of the American Ceramic Society, Vol. 97, Issue 8 https://doi.org/10.1111/jace.13094	journal	July 2014
Physics and technology of magnetron sputtering discharges Gudmundsson, J. T. Plasma Sources Science and Technology, Vol. 29, Issue 11 https://doi.org/10.1088/1361-6595/abb7bd	journal	November 2020
Accident tolerant fuel cladding development: Promise, status, and challenges Terrani, Kurt A. Journal of Nuclear Materials, Vol. 501 https://doi.org/10.1016/j.jnucmat.2017.12.043	journal	April 2018
Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux Petrie, Christian M.; Koyanagi, Takaaki; McDuffee, Joel L. Journal of Nuclear Materials, Vol. 491 https://doi.org/10.1016/j.jnucmat.2017.04.058	journal	August 2017
Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding Hazan, J.; Gauthier, A.; Pouillier, E. Journal of Nuclear Materials, Vol. 550 https://doi.org/10.1016/j.jnucmat.2021.152940	journal	July 2021
Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures Mao, Keyou S.; Massey, Caleb P.; Yamamoto, Yukinori Acta Materialia, Vol. 231 https://doi.org/10.1016/j.actamat.2022.117843	journal	June 2022
Thermal diffusivity and thermal conductivity of SiC composite tubes: the effects of microstructure and irradiation Koyanagi, Takaaki; Wang, Hsin; Mena, Jose' D. Arregui Journal of Nuclear Materials, Vol. 557 https://doi.org/10.1016/j.jnucmat.2021.153217	journal	December 2021
Burst and oxidation behavior of Cr-coated Zirlo during simulated LOCA testing Bell, S. B.; Graening, T.; Evans, A. Journal of Nuclear Materials, Vol. 564 https://doi.org/10.1016/j.jnucmat.2022.153679	journal	June 2022
Ceramic Gas Turbine Development: Need for a 10Year Plan van Roode, Mark Journal of Engineering for Gas Turbines and Power, Vol. 132, Issue 1 https://doi.org/10.1115/1.3124669	journal	September 2009
Quantifying deformation during Zry-4 burst testing: a comparison of BISON and a combined in-situ digital image correlation and infrared thermography method Kane, K.; Bell, S.; Garrison, B. Journal of Nuclear Materials, Vol. 572 https://doi.org/10.1016/j.jnucmat.2022.154063	journal	December 2022

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