In-service corrosion and grain boundary oxidation in neutron-irradiated 316 stainless steel baffle-former bolts

Lach, Timothy G.; Gussev, Maxim N.; Chen, Xiang (Frank)

doi:10.1016/j.corsci.2025.113106

Title: In-service corrosion and grain boundary oxidation in neutron-irradiated 316 stainless steel baffle-former bolts

Journal Article · Mon Jun 09 00:00:00 UTC 2025 · Corrosion Science

DOI: https://doi.org/10.1016/j.corsci.2025.113106 · OSTI ID:3002718

^[1];

^[2];

^[1]

Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Materials Science & Technology Division
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Reactor core internal components such as baffle-former bolts (BFBs) are subjected to significant mechanical stress, corrosive environment, and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and of the bolts. In this work, characterization was performed on the oxidized surface of stainless steel BFBs harvested from a commercial pressurized water reactor (PWR) after 40 years of operation. The analysis shows that a complex multilayered surface oxide with six identified layers formed that is different from 2-layer structure commonly observed in model experiments. The oxide varies by composition – predominantly Fe, Cr, and Ni, grain size, and phase, and has features resembling both unirradiated and radiation/ corrosion experiments likely due to the low radiation flux compared to ion-irradiation or the test reactor radiation. In addition, grain boundary oxidative attack featured a pathway for Fe and other elements to move from the metal matrix to the outermost oxide. In conclusion, the results help assess PWR lifetime extension, put into context previous experimental studies, and provide input for designing experiments combining radiation and corrosion effects.

View Accepted Manuscript (DOE)

Cite

Export

Save

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

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE); USDOE Office of Science (SC), Basic Energy Sciences (BES). Scientific User Facilities (SUF)

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 3002718

Journal Information:: Corrosion Science, Journal Name: Corrosion Science Vol. 255; ISSN 0010-938X

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (37)

Influence of Irradiation on the Oxide Film Formed on 316 L Stainless Steel in PWR Primary Water Perrin, Stéphane; Marchetti, Loic; Duhamel, Cécilie Oxidation of Metals, Vol. 80, Issue 5-6 https://doi.org/10.1007/s11085-013-9401-3	journal	February 2013
Review on synergistic damage effect of irradiation and corrosion on reactor structural alloys Liu, Hui; Lei, Guan-Hong; Huang, He-Fei Nuclear Science and Techniques, Vol. 35, Issue 3 https://doi.org/10.1007/s41365-024-01415-3	journal	March 2024
Recent advances in the kinetics of normal/abnormal grain growth: a review Najafkhani, Fateme; Kheiri, Sara; Pourbahari, Bita Archives of Civil and Mechanical Engineering, Vol. 21, Issue 1 https://doi.org/10.1007/s43452-021-00185-8	journal	February 2021
A review of irradiation assisted stress corrosion cracking Scott, P. Journal of Nuclear Materials, Vol. 211, Issue 2 https://doi.org/10.1016/0022-3115(94)90360-3	journal	August 1994
Current status and future of IASCC research Shoji, Tetsuo; Suzuki, Shun-ichi; Raja, K. S. Journal of Nuclear Materials, Vol. 258-263 https://doi.org/10.1016/S0022-3115(98)00304-3	journal	October 1998
Atomistic simulations of grain boundary segregation in nanocrystalline yttria-stabilized zirconia and gadolinia-doped ceria solid oxide electrolytes Lee, Hark B.; Prinz, Friedrich B.; Cai, Wei Acta Materialia, Vol. 61, Issue 10 https://doi.org/10.1016/j.actamat.2013.03.027	journal	June 2013
Anisotropic radiation-induced segregation in 316L austenitic stainless steel with grain boundary character Barr, Christopher M.; Vetterick, Gregory A.; Unocic, Kinga A. Acta Materialia, Vol. 67 https://doi.org/10.1016/j.actamat.2013.11.060	journal	April 2014
Effect of machining on stress corrosion crack initiation in warm-forged type 304L stainless steel in high temperature water Chang, Litao; Volpe, Liberato; Wang, Yong Liang Acta Materialia, Vol. 165 https://doi.org/10.1016/j.actamat.2018.11.046	journal	February 2019
Clarifying the mitigation effect of proton irradiation on the intergranular oxidation of 316L stainless steel in high temperature water Wang, Shengkai; Zhang, Shihao; Xie, Jiayu Acta Materialia, Vol. 241 https://doi.org/10.1016/j.actamat.2022.118408	journal	December 2022
Clarifying the irradiation effect on the general oxidation of 316L stainless steel in high temperature hydrogenated water after 1000h immersion Wang, Shengkai; Wang, Qi; Zhang, Shihao Acta Materialia, Vol. 255 https://doi.org/10.1016/j.actamat.2023.119100	journal	August 2023
The time dependence of proton irradiation effect on the intergranular oxidation of 316 L stainless steel in high-temperature hydrogenated water Wang, Shengkai; Zhang, Shihao; Xie, Jiayu Acta Materialia, Vol. 260 https://doi.org/10.1016/j.actamat.2023.119340	journal	November 2023
Radiation induced segregation in titanium diboride Wei, Shuguang; Qureshi, Muhammad Waqas; Xi, Jianqi Acta Materialia, Vol. 267 https://doi.org/10.1016/j.actamat.2024.119739	journal	April 2024
Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary water Deng, Ping; Peng, Qunjia; Han, En-Hou Corrosion Science, Vol. 127 https://doi.org/10.1016/j.corsci.2017.08.010	journal	October 2017
Oxide growth and dissolution on 316L stainless steel during irradiation in high temperature water Hanbury, Rigel D.; Was, Gary S. Corrosion Science, Vol. 157 https://doi.org/10.1016/j.corsci.2019.06.006	journal	August 2019
Role of irradiation and irradiation defects on the oxidation first stages of a 316L austenitic stainless steel Boisson, M.; Legras, L.; Andrieu, E. Corrosion Science, Vol. 161 https://doi.org/10.1016/j.corsci.2019.108194	journal	December 2019
A high-resolution characterization of irradiation-assisted stress corrosion cracking of proton-irradiated 316L stainless steel in simulated pressurized water reactor primary water Kuang, Wenjun; Feng, Xingyu; Du, Donghai Corrosion Science, Vol. 199 https://doi.org/10.1016/j.corsci.2022.110187	journal	May 2022
A high-resolution study of the different surface state effects on the corrosion behaviors of a ferritic steel and an austenitic steel in supercritical water Chen, K.; Wang, J.; Zhang, L. Corrosion Science, Vol. 209 https://doi.org/10.1016/j.corsci.2022.110757	journal	December 2022
The effect of surface grinding on the stress corrosion cracking initiation of 316LN stainless steel in 600 °C supercritical CO2 Qian, Hongchen; Yang, Liujie; Feng, Xingyu Corrosion Science, Vol. 234 https://doi.org/10.1016/j.corsci.2024.112147	journal	July 2024
Irradiation-induced age-related degradation mechanisms effect on integrity of baffle former bolts Jyung, Jae-Min; Joo, Hwan-Jae; Chang, Yoon-Suk International Journal of Pressure Vessels and Piping, Vol. 208 https://doi.org/10.1016/j.ijpvp.2023.105116	journal	April 2024
Characterization of oxide layers grown on D9 austenitic stainless steel in lead bismuth eutectic Hosemann, P.; Hawley, M.; Koury, D. Journal of Nuclear Materials, Vol. 375, Issue 3 https://doi.org/10.1016/j.jnucmat.2007.12.005	journal	April 2008
Radiation-induced stress relaxation in high temperature water of type 316L stainless steel evaluated by neutron diffraction Ishiyama, Y.; Rogge, R. B.; Obata, M. Journal of Nuclear Materials, Vol. 408, Issue 2 https://doi.org/10.1016/j.jnucmat.2010.11.023	journal	January 2011
Deformation localization and dislocation channel dynamics in neutron-irradiated austenitic stainless steels Gussev, Maxim N.; Field, Kevin G.; Busby, Jeremy T. Journal of Nuclear Materials, Vol. 460 https://doi.org/10.1016/j.jnucmat.2015.02.008	journal	May 2015
Crack initiation testing of thimble tube material under PWR conditions to determine a stress threshold for IASCC Bosch, R. W.; Vankeerberghen, M.; Gérard, R. Journal of Nuclear Materials, Vol. 461 https://doi.org/10.1016/j.jnucmat.2015.02.038	journal	June 2015
Radiolysis driven changes to oxide stability during irradiation-corrosion of 316L stainless steel in high temperature water Raiman, Stephen S.; Bartels, David M.; Was, Gary S. Journal of Nuclear Materials, Vol. 493 https://doi.org/10.1016/j.jnucmat.2017.05.042	journal	September 2017
Accelerated corrosion and oxide dissolution in 316L stainless steel irradiated in situ in high temperature water Raiman, Stephen S.; Was, Gary S. Journal of Nuclear Materials, Vol. 493 https://doi.org/10.1016/j.jnucmat.2017.05.043	journal	September 2017
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
Irradiation-assisted stress corrosion cracking of Type 347 and Type 316 steels irradiated in commercial pressurized water reactors Ickes, Michael R.; McKinley, Joshua; Lee, Jung-Kun Journal of Nuclear Materials, Vol. 536 https://doi.org/10.1016/j.jnucmat.2020.152182	journal	August 2020
Effect of surface state of austenitic type 304L SS on oxide characteristics formed in high temperature high pressure water Das, Annesha; Roychowdhury, Supratik; Kain, Vivekanand Journal of Nuclear Materials, Vol. 564 https://doi.org/10.1016/j.jnucmat.2022.153672	journal	June 2022
Corrosion of 316L stainless steel in high temperature water and steam: Mechanisms of corrosion Hanbury, Rigel D.; Was, Gary S. Journal of Nuclear Materials, Vol. 599 https://doi.org/10.1016/j.jnucmat.2024.155180	journal	October 2024
Corrosion of 316L stainless steel in high temperature water and steam: effects of simultaneous proton irradiation Hanbury, Rigel D.; Was, Gary S. Journal of Nuclear Materials, Vol. 600 https://doi.org/10.1016/j.jnucmat.2024.155274	journal	November 2024
IASCC of neutron irradiated 316 stainless steel to 125 dpa Du, Donghai; Sun, Kai; Was, Gary S. Materials Characterization, Vol. 173 https://doi.org/10.1016/j.matchar.2021.110897	journal	March 2021
How irradiation promotes intergranular stress corrosion crack initiation Was, G. S.; Bahn, C. -B.; Busby, J. Progress in Materials Science, Vol. 143 https://doi.org/10.1016/j.pmatsci.2024.101255	journal	June 2024
Complexity of segregation behavior at localized deformation sites formed while in service in a 316 stainless steel baffle-former bolt Lach, Timothy G.; Gussev, Maxim N.; Chen, Xiang (Frank) Scripta Materialia, Vol. 255 https://doi.org/10.1016/j.scriptamat.2024.116371	journal	January 2025
General Corrosion of Materials under Simulated BWR Primary Water Conditions Hemmi, Yukio; Uruma, Yutaka; Ichikawa, Nagayoshi Journal of Nuclear Science and Technology, Vol. 31, Issue 5 https://doi.org/10.1080/18811248.1994.9735174	journal	May 1994
Microstructural Characterization of SCC Crack Tip and Oxide Film for SUS 316 Stainless Steel in Simulated PWR Primary Water at 320°C Terachi, Takumi; Fujii, Katsuhiko; Arioka, Koji Journal of Nuclear Science and Technology, Vol. 42, Issue 2 https://doi.org/10.1080/18811248.2005.9726383	journal	February 2005
Corrosion Behavior of Stainless Steels in Simulated PWR Primary Water—Effect of Chromium Content in Alloys and Dissolved Hydrogen— Terachi, Takumi; Yamada, Takuyo; Miyamoto, Tomoki Journal of Nuclear Science and Technology, Vol. 45, Issue 10 https://doi.org/10.1080/18811248.2008.9711883	journal	October 2008
Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components Lach, Timothy G.; Chen, Xiang (Frank) Volume 5: Materials & Fabrication https://doi.org/10.1115/PVP2023-107347	conference	July 2023

Similar Records

In-service Oxidation and IASCC in High Fluence Baffle-Former Bolts Retrieved from a Westinghouse PWR

Technical Report · Thu Aug 01 04:00:00 UTC 2024 · OSTI ID:2441052

Lach, Timothy G.; Gussev, Maxim N.; Chen, Xiang (Frank)

Post-Irradiation Examination of High Fluence Baffle-Former Bolts Retrieved from a Westinghouse Two-Loop Downflow Type PWR

Technical Report · Wed Jul 31 04:00:00 UTC 2019 · OSTI ID:1557483

Chen, Xiang (Frank); Chen, Tianyi; Parish, Chad M.; +3 more

Microstructural Characterizations of Two High Fluence Baffle-Former Bolts Retrieved from a Westinghouse Two-loop Downflow Type PWR

Technical Report · Fri Sep 01 04:00:00 UTC 2023 · OSTI ID:2434390

Lach, Timothy G.; Gussev, Maxim N.; Chen, Xiang (Frank)

Related Subjects

Baffle-former bolt
Crack initiation
In-service irradiation and corrosion
Irradiation assisted stress corrosion cracking (IASCC)
Multilayered oxide structure
Radiation-induced segregation (RIS)

Title: In-service corrosion and grain boundary oxidation in neutron-irradiated 316 stainless steel baffle-former bolts

Citation Formats

References (37)

Similar Records

Related Subjects