Chromium-coated cladding analysis under simulated LOCA burst conditions

Sweet, Ryan; Mouche, Peter; Bell, Samuel; Kane, Kenneth; Capps, Nathan

doi:10.1016/j.anucene.2022.109275

Chromium-coated cladding analysis under simulated LOCA burst conditions

Journal Article · Fri Jun 17 00:00:00 EDT 2022 · Annals of Nuclear Energy

DOI:https://doi.org/10.1016/j.anucene.2022.109275· OSTI ID:1875359

Sweet, Ryan ^[1]; ^[1]; Bell, Samuel ^[2]; Kane, Kenneth ^[1]; Capps, Nathan ^[3]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Univ. of Tennessee, Knoxville, TN (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States)

Of the near-term accident-tolerant fuel concepts identified, chromium coatings have been shown to slow cladding oxidation without altering fuel system geometry or greatly affecting neutronic performance. To identify how coated cladding tubes perform under high-temperature accident conditions, pressurized-tube burst tests were conducted in the Severe Accident Test Station at Oak Ridge National Laboratory. To analyze how the coating affects cladding behavior, coated and uncoated cladding burst tests were simulated by using the BISON fuel performance code. Thermocouple data from these tests were fit into axial and azimuthal profiles and combined to generate 3D cladding surface temperatures, and pressure transducer data were compared until cladding failure. Additionally, the cladding temperatures at failure and the internal gas pressure evolution show relatively good agreement between the simulation and experiment results. Simulations were then extrapolated to demonstrate the coatings effectiveness to increasing the cladding burst temperature by using a parametric evaluation of the initial tube gas pressure. This work demonstrates the possibility of an increased cladding failure margin under transient conditions due to the addition of chromium coating, and more pertinently, an increased cladding failure margin from more accurate experimental characterization.

View Accepted Manuscript (DOE)

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

Sponsoring Organization:: USDOE

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1875359

Alternate ID(s):: OSTI ID: 1874056

Journal Information:: Annals of Nuclear Energy, Journal Name: Annals of Nuclear Energy Journal Issue: 1 Vol. 176; ISSN 0306-4549

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (14)

Oxidation of Zircaloy by steam Moalem, M.; Olander, D. R. Journal of Nuclear Materials, Vol. 182 https://doi.org/10.1016/0022-3115(91)90428-A	journal	May 1991
High-temperature creep of polycrystalline chromium Stephens, Joseph R.; Klopp, William D. Journal of the Less Common Metals, Vol. 27, Issue 1 https://doi.org/10.1016/0022-5088(72)90108-7	journal	April 1972
Current knowledge on core degradation phenomena, a review Hofmann, P. Journal of Nuclear Materials, Vol. 270, Issue 1-2 https://doi.org/10.1016/S0022-3115(98)00899-X	journal	April 1999
Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings Wagih, Malik; Spencer, Benjamin; Hales, Jason Annals of Nuclear Energy, Vol. 120 https://doi.org/10.1016/j.anucene.2018.06.001	journal	October 2018
Multidimensional multiphysics simulation of nuclear fuel behavior Williamson, R. L.; Hales, J. D.; Novascone, S. R. Journal of Nuclear Materials, Vol. 423, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2012.01.012	journal	April 2012
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
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
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
MOOSE: A parallel computational framework for coupled systems of nonlinear equations Gaston, Derek; Newman, Chris; Hansen, Glen Nuclear Engineering and Design, Vol. 239, Issue 10 https://doi.org/10.1016/j.nucengdes.2009.05.021	journal	October 2009
The TRANSURANUS mechanical model for large strain analysis Di Marcello, Valentino; Schubert, Arndt; van de Laar, Jacques Nuclear Engineering and Design, Vol. 276 https://doi.org/10.1016/j.nucengdes.2014.04.041	journal	September 2014
Thermal Creep of Zircaloy-4 Cladding under Internal Pressure Matsuo, Yutaka Journal of Nuclear Science and Technology, Vol. 24, Issue 2 https://doi.org/10.1080/18811248.1987.9735783	journal	February 1987
Zirconium Cladding Deformation in a Steam Environment with Transient Heating Chapman, R. H.; Crowley, J. L.; Longest, A. W. Zirconium in the Nuclear Industry https://doi.org/10.1520/STP36692S	book	January 1979
SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4 Hagrman, D. T.; Allison, C. M.; Berna, G. A. US Nuclear Regulatory Commission (NRC) https://doi.org/10.2172/100327	report	June 1995
Zirconium metal-water oxidation kinetics. IV. Reaction rate studies. [BWR:PWR] Cathcart, J. V.; Pawel, R. E.; McKee, R. A. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States) https://doi.org/10.2172/7317596	report	July 1977

Similar Records

Fuel Performance Analysis of Chromium-Coated Cladding under Burst Conditions

Technical Report · Wed Sep 01 00:00:00 EDT 2021 · OSTI ID:1874655

Fuel performance analysis of Cr-coated Zircaloy-4 cladding during a prototypical LOCA event using BISON

Journal Article · Fri Feb 09 19:00:00 EST 2024 · Annals of Nuclear Energy · OSTI ID:2405123

Behavior under LOCA conditions on enhanced accident tolerant chromium coated zircaloy-4 claddings

Conference · Fri Jul 01 00:00:00 EDT 2016 · OSTI ID:22765196

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS

Chromium-coated cladding analysis under simulated LOCA burst conditions

Citation Formats

References (14)

Similar Records

Related Subjects