Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4

Wang, Peng; Kammenzind, Bruce; Smith, Richard; Motta, Arthur; Aumand, Matthieu; Kaczorowski, Damien; Bachhav, Mukesh; Was, Gary

doi:10.1016/j.jnucmat.2024.155505

Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4

Journal Article · Fri Nov 08 00:00:00 EST 2024 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2024.155505· OSTI ID:2583951

^[1]; ^[2]; Smith, Richard ^[2]; ^[3]; Aumand, Matthieu ^[4]; Kaczorowski, Damien ^[5]; ^[6]; Was, Gary ^[1]

University of Michigan, Ann Arbor, MI (United States)
Naval Nuclear Laboratory, West Mifflin, PA (United States)
Pennsylvania State University, University Park, PA (United States)
Framatome Inc., Lynchburg, VA (United States)
Framatome, Lyon (France)
Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Using proton irradiation, this study investigates the individual influence of several factors on the corrosion kinetics of Zircaloy-4 in a hydrogenated water environment simulating a Pressurized Water Reactor (PWR). Using both simultaneous irradiation-corrosion and autoclave corrosion, we separately examine: (i) the effect of pre-irradiation on modifying the structure of the material, (ii) the impact of irradiation on creating defects in the growing oxide layer during corrosion, and (iii) the influence of irradiation on increasing the corrosion potential through radiolysis during corrosion. To replicate neutron-irradiated microstructure, two proton pre-irradiation schedules were employed: Schedule 1 (isothermal irradiation at 350 °C to 5 dpa) to simulate high-temperature PWR conditions, and Schedule 2 (two-step process: irradiation to 2.5 dpa at -10 °C followed by 2.5 dpa at 350 °C) to simulate lower temperature PWR and Boiling Water Reactor (BWR) conditions. Long-term autoclave corrosion testing over 360 days at 320 °C revealed no significant difference between unirradiated samples and those pre-irradiated according to either schedule, with all samples exhibiting sub-cubic kinetics within the pre-transition regime. Irradiated samples underwent Simultaneous Irradiation Corrosion (SIC) tests, corroding in 320 °C water while being irradiated with protons. Corrosion was found to accelerate in all SIC-tested samples relative to autoclave conditions, with the greatest increase observed in non-pre-irradiated samples. Pre-irradiation with either schedule resulted in a slower corrosion rate compared to non-pre-irradiated regions under SIC conditions. The degree of radiolysis observed in the SIC tests surpassed typical PWR conditions, approaching levels found in BWRs. Radiolysis products were identified as the primary contributors to accelerated corrosion, corroborated by radiolysis bar tests. Furthermore, these findings underscore the intricate interactions between irradiation, corrosion, and water chemistry in determining Zircaloy-4 corrosion kinetics within nuclear reactor environments.

View Accepted Manuscript (DOE)

Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Sponsoring Organization:: USDOE; USDOE Laboratory Directed Research and Development (LDRD) Program

Grant/Contract Number:: 89233018CNR000004; AC07-05ID14517

OSTI ID:: 2583951

Alternate ID(s):: OSTI ID: 2477471

Report Number(s):: INL/JOU--24-80040

Journal Information:: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: 1 Vol. 604; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (26)

Precipitates in zircaloy: Identification and the effects of irradiation and thermal treatment Yang, W. J. S.; Tucker, R. P.; Cheng, B. Journal of Nuclear Materials, Vol. 138, Issue 2-3 https://doi.org/10.1016/0022-3115(86)90005-X	journal	April 1986
Phase instability, decomposition and redistribution of intermetallic precipitates in Zircaloy-2 and -4 during neutron irradiation Griffiths, M.; Gilbert, R. W.; Carpenter, G. J. C. Journal of Nuclear Materials, Vol. 150, Issue 1 https://doi.org/10.1016/0022-3115(87)90093-6	journal	September 1987
A review of microstructure evolution in zirconium alloys during irradiation Griffiths, M. Journal of Nuclear Materials, Vol. 159 https://doi.org/10.1016/0022-3115(88)90093-1	journal	October 1988
Long-term corrosion of Zircaloy before and after irradiation Hillner, E.; Franklin, D. G.; Smee, J. D. Journal of Nuclear Materials, Vol. 278, Issue 2-3 https://doi.org/10.1016/S0022-3115(99)00230-5	journal	April 2000
How the crystallography and nanoscale chemistry of the metal/oxide interface develops during the aqueous oxidation of zirconium cladding alloys Ni, N.; Hudson, D.; Wei, J. Acta Materialia, Vol. 60, Issue 20 https://doi.org/10.1016/j.actamat.2012.09.021	journal	December 2012
Structure of zirconium alloy oxides formed in pure water studied with synchrotron radiation and optical microscopy: relation to corrosion rate Yilmazbayhan, Aylin; Motta, Arthur T.; Comstock, Robert J. Journal of Nuclear Materials, Vol. 324, Issue 1 https://doi.org/10.1016/j.jnucmat.2003.08.038	journal	January 2004
Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys Cox, B. Journal of Nuclear Materials, Vol. 336, Issue 2-3 https://doi.org/10.1016/j.jnucmat.2004.09.029	journal	February 2005
Transmission electron microscopy examination of oxide layers formed on Zr alloys Yilmazbayhan, Aylin; Breval, Else; Motta, Arthur T. Journal of Nuclear Materials, Vol. 349, Issue 3 https://doi.org/10.1016/j.jnucmat.2005.10.012	journal	March 2006
Origin and effect of lateral cracks in oxide scales formed on zirconium alloys Tejland, Pia; Andrén, Hans-Olof Journal of Nuclear Materials, Vol. 430, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2012.06.039	journal	November 2012
A facility for studying irradiation accelerated corrosion in high temperature water Raiman, Stephen S.; Flick, Alexander; Toader, Ovidiu Journal of Nuclear Materials, Vol. 451, Issue 1-3 https://doi.org/10.1016/j.jnucmat.2014.03.022	journal	August 2014
Transmission electron microscopy characterization of Zircaloy-4 and ZIRLO™ oxide layers de Gabory, Benoit; Motta, Arthur T.; Wang, Ke Journal of Nuclear Materials, Vol. 456 https://doi.org/10.1016/j.jnucmat.2014.09.073	journal	January 2015
Hydrothermal corrosion of SiC in LWR coolant environments in the absence of irradiation Terrani, K. A.; Yang, Y.; Kim, Y. -J. Journal of Nuclear Materials, Vol. 465 https://doi.org/10.1016/j.jnucmat.2015.06.019	journal	October 2015
Determination of dose rate effects on Zircaloy oxidation using proton irradiation and oxygen transport modeling Reyes, Michael; Wang, Peng; Was, Gary Journal of Nuclear Materials, Vol. 523 https://doi.org/10.1016/j.jnucmat.2019.05.039	journal	September 2019
Effect of radiation damage and water radiolysis on corrosion of FeCrAl alloys in hydrogenated water Wang, Peng; Grdanovska, Slavica; Bartels, David M. Journal of Nuclear Materials, Vol. 533 https://doi.org/10.1016/j.jnucmat.2020.152108	journal	May 2020
The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC Doyle, Peter; Sun, Kaichao; Snead, Lance Journal of Nuclear Materials, Vol. 536 https://doi.org/10.1016/j.jnucmat.2020.152190	journal	August 2020
Corrosion behavior of ferritic FeCrAl alloys in simulated BWR normal water chemistry Wang, Peng; Grdanovska, Slavica; Bartels, David M. Journal of Nuclear Materials, Vol. 545 https://doi.org/10.1016/j.jnucmat.2020.152744	journal	March 2021
Emulation of neutron damage with proton irradiation and its effects on microstructure and microchemistry of Zircaloy-4 Wang, Peng; Bowman, Josh; Bachhav, Mukesh Journal of Nuclear Materials, Vol. 557 https://doi.org/10.1016/j.jnucmat.2021.153281	journal	December 2021
Investigation of breakaway corrosion observed during oxide growth in pure and low alloying element content Zr exposed in water at 360°C Ensor, B.; Motta, A. T.; Lucente, A. Journal of Nuclear Materials, Vol. 558 https://doi.org/10.1016/j.jnucmat.2021.153358	journal	January 2022
Reproducing shadow corrosion on Zircaloy-2 using in-situ proton irradiation Wang, Peng; Nowotka, Karsten; Was, Gary S. Journal of Nuclear Materials, Vol. 558 https://doi.org/10.1016/j.jnucmat.2021.153406	journal	January 2022
Untangling competition between epitaxial strain and growth stress through examination of variations in local oxidation Yankova, Maria S.; Garner, Alistair; Baxter, Felicity Nature Communications, Vol. 14, Issue 1 https://doi.org/10.1038/s41467-022-35706-3	journal	January 2023
Current understanding of radiation-induced degradation in light water reactor structural materials Fukuya, Koji Journal of Nuclear Science and Technology, Vol. 50, Issue 3 https://doi.org/10.1080/00223131.2013.772448	journal	March 2013
Effects of Operational Parameters on Nodular Corrosion Characteristics Etoh, Yoshinori Journal of Nuclear Science and Technology, Vol. 26, Issue 8 https://doi.org/10.1080/18811248.1989.9734379	journal	August 1989
Microstructural Characterization of Oxides Formed on Model Zr Alloys Using Synchrotron Radiation Motta, A. T.; Gomes da Silva, M. J.; Yilmazbayhan, A. Journal of ASTM International, Vol. 5, Issue 3 https://doi.org/10.1520/JAI101257	journal	January 2008
Review of Corrosion and Dimensional Behavior of Zircaloy under Water Reactor Conditions Garzarolli, F.; Manzel, R.; Reschke, S. Zirconium in the Nuclear Industry https://doi.org/10.1520/STP36674S	book	January 1979
Oxidation of Zircaloy-4 during in situ proton irradiation and corrosion in PWR primary water Wang, Peng; Was, Gary S. Journal of Materials Research, Vol. 30, Issue 9 https://doi.org/10.1557/jmr.2014.408	journal	March 2015
Corrosion of Zirconium Alloys: Link between the Metal/Oxide Interface Roughness, the Degradation of the Protective Oxide Layer and the Corrosion Kinetics Bossis, Philippe; Lefebvre, Florence; Barbéris, Pierre Materials Science Forum, Vol. 369-372 https://doi.org/10.4028/www.scientific.net/MSF.369-372.255	journal	October 2001

Similar Records

SiC cladding corrosion and mitigation

Conference · Fri Jul 01 00:00:00 EDT 2016 · OSTI ID:22750105

SiC Composite for Fuel Structure Applications

Technical Report · Thu Dec 21 23:00:00 EST 2017 · OSTI ID:1415452

Irradiation Accelerated Corrosion of Stainless Steel and Ferritic-Martensitic Steel in Simulated Primary Water

Conference · Sun Jan 29 23:00:00 EST 2017 · Proceedings, CORROSION 2017 · OSTI ID:1474122

Related Subjects

Pressurized Water Reactor
Zircaloy-4
corrosion
proton irradiation
radiolysis
zirconium

Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4

Citation Formats

References (26)

Similar Records

Related Subjects