POST IRRADIATION EXAMINATION OF PRESSURIZED WATER REACTOR STAINLESS STEEL INTERNAL COMPONENTS

Timothy G. Lach; Xiang Chen

doi:10.1115/PVP2023-107347

POST IRRADIATION EXAMINATION OF PRESSURIZED WATER REACTOR STAINLESS STEEL INTERNAL COMPONENTS

Timothy G. Lach, Xiang Chen

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution › peer-review

1 Scopus citations

Abstract

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Original language	English
Title of host publication	Materials and Fabrication
Publisher	American Society of Mechanical Engineers (ASME)
ISBN (Electronic)	9780791887486
DOIs	https://doi.org/10.1115/PVP2023-107347
State	Published - 2023
Event	ASME 2023 Pressure Vessels and Piping Conference, PVP 2023 - Atlanta, United States Duration: Jul 16 2023 → Jul 21 2023

Publication series

Name	American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP
Volume	5
ISSN (Print)	0277-027X

Conference

Conference	ASME 2023 Pressure Vessels and Piping Conference, PVP 2023
Country/Territory	United States
City	Atlanta
Period	07/16/23 → 07/21/23

Funding

This research was sponsored by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability Program, Materials Research Pathway, under contract DE-AC05-00OR22725 with UT-Battelle, LLC/Oak Ridge National Laboratory (ORNL). Atom probe tomography (APT) research was conducted as part of a user project at the Center for Nanophase Materials Sciences (CNMS), which is a US Department of Energy, Office of Science User Facility at Oak Ridge National Laboratory. We would like to thank Jonathan Poplawsky and James Burns for help with APT data collection. The authors extend their appreciation to Clay Morris, Jerid Metcalf, Mark Delph, and other colleagues at Irradiated Materials Examination and Testing Facility (IMET) at ORNL for their support during sample processing in the hot cell sample retrieval. In addition, we would like to thank Patricia Tedder and Travis Dixon at the Low Activation Materials Development and Analysis (LAMDA) Laboratory at ORNL, where mechanical behavior and TEM/STEM evaluation was conducted. We acknowledge Heather Malikowski and Robert Marcello of Exelon Corporation and Bernard Rudell formerly of Exelon Corporation for their assistance during the bolt harvesting process. Lastly, we would like to thank the late Mike Burke formerly with Electrical Power Research Institute who was involved in the bolt harvesting and characterization planning and led the sample preparation, machining, and shipping when he was working at Westinghouse Electric Company. This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).

Funders	Funder number
Bernard Rudell formerly of Exelon Corporation
LLC
Materials Research Pathway	DE-AC05-00OR22725
Office of Nuclear Energy, Light Water Reactor Sustainability Program
U.S. Department of Energy
Office of Science
Oak Ridge National Laboratory
Exelon Corporation

Keywords

Post irradiation examination
austenitic stainless steel
nuclear reactor internals
radiation effects

Access to Document

10.1115/PVP2023-107347

Cite this

Lach, T. G., & Chen, X. (2023). POST IRRADIATION EXAMINATION OF PRESSURIZED WATER REACTOR STAINLESS STEEL INTERNAL COMPONENTS. In Materials and Fabrication Article v005t06a063 (American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP; Vol. 5). American Society of Mechanical Engineers (ASME). https://doi.org/10.1115/PVP2023-107347

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title = "POST IRRADIATION EXAMINATION OF PRESSURIZED WATER REACTOR STAINLESS STEEL INTERNAL COMPONENTS",

abstract = "Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.",

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Lach, TG & Chen, X 2023, POST IRRADIATION EXAMINATION OF PRESSURIZED WATER REACTOR STAINLESS STEEL INTERNAL COMPONENTS. in Materials and Fabrication., v005t06a063, American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP, vol. 5, American Society of Mechanical Engineers (ASME), ASME 2023 Pressure Vessels and Piping Conference, PVP 2023, Atlanta, United States, 07/16/23. https://doi.org/10.1115/PVP2023-107347

POST IRRADIATION EXAMINATION OF PRESSURIZED WATER REACTOR STAINLESS STEEL INTERNAL COMPONENTS. / Lach, Timothy G.; Chen, Xiang.
Materials and Fabrication. American Society of Mechanical Engineers (ASME), 2023. v005t06a063 (American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP; Vol. 5).

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution › peer-review

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N2 - Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

AB - Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

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KW - austenitic stainless steel

KW - nuclear reactor internals

KW - radiation effects

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POST IRRADIATION EXAMINATION OF PRESSURIZED WATER REACTOR STAINLESS STEEL INTERNAL COMPONENTS

Abstract

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Funding

Keywords

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