ML25258A153
| ML25258A153 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/2025 |
| From: | Bayssie M, Eric Focht, Amy Hull, Jeffrey Poehler, Madhumita Sircar, Robert Tregoning, Austin Young Office of Nuclear Regulatory Research |
| To: | |
| JEFF POEHLER 4158353 | |
| Shared Package | |
| ML25258A155 | List: |
| References | |
| RIL-2025-06 | |
| Download: ML25258A153 (191) | |
Text
RIL 2025-06 NRC Workshop on Structural Materials:
Research for 80 Years and Beyond, Summary Report October 1-3, 2024 Date Published: September 2025 Prepared by:
M. Bayssie E. Focht A. Hull J. Poehler M. Sircar R. Tregoning A. Young Nuclear Regulatory Commission Research Information Letter Office of Nuclear Regulatory Research
ii Disclaimer (required)
Legally binding regulatory requirements are stated only in laws, NRC regulations, licenses, including technical specifications, or orders; not in Research Information Letters (RILs). A RIL is not regulatory guidance, although NRCs regulatory offices may consider the information in a RIL to determine whether any regulatory actions are warranted.
iii TABLE OF CONTENTS LIST OF FIGURES.................................................................................................................... V LIST OF TABLES..................................................................................................................... VI ABBREVIATIONS AND ACRONYMS..................................................................................... VII EXECUTIVE
SUMMARY
........................................................................................................... X 1 INTRODUCTION.................................................................................................................1-1 1.1 Previous NRC Long-Term Operation Workshops........................................................1-1 1.2 2024 Workshop Motivation & Objectives......................................................................1-1 1.3 2024 Workshop Organization & Participants................................................................1-2 1.4 Organization of This Report..........................................................................................1-6 2
SUMMARY
OF PRESENTATIONS.....................................................................................2-1 2.1 Opening Session..........................................................................................................2-1 2.1.1 Welcome and Opening Remarks....................................................................2-1 2.1.2 Presentations.................................................................................................2-1 2.1.3 Key Themes of the Introductory Session........................................................2-2 2.2 Session 1Reactor Pressure Vessel...........................................................................2-2 2.2.1 Presentations.................................................................................................2-2 2.2.2 Panel Discussions..........................................................................................2-5 2.2.3 Session Summary..........................................................................................2-7 2.3 Session 2: Reactor Pressure Vessel Internals..............................................................2-7 2.3.1 Presentations.................................................................................................2-7 2.3.2 Panel Discussions........................................................................................ 2-13 2.3.3 Session Summary........................................................................................ 2-17 2.4 Session 3: Reactor Coolant Pressure Boundary Components.................................... 2-18 2.4.1 Presentations............................................................................................... 2-18 2.4.2 Panel Discussions........................................................................................ 2-21 2.4.3 Session Summary........................................................................................ 2-23 2.5 Session 4: Secondary Side Components.................................................................... 2-23 2.5.1 Presentations............................................................................................... 2-23 2.5.2 Panel Discussions........................................................................................ 2-26 2.5.3 Session Summary........................................................................................ 2-27 2.6 Session 5: Balance of Plant Systems......................................................................... 2-27 2.6.1 Presentations............................................................................................... 2-28 2.6.2 Panel Discussions........................................................................................ 2-29 2.6.3 Session Summary........................................................................................ 2-31 2.7 Session 6: Fatigue...................................................................................................... 2-32 2.7.1 Presentations............................................................................................... 2-32 2.7.2 Panel Discussions........................................................................................ 2-35 2.7.3 Session Summary........................................................................................ 2-36
iv 2.8 Session 7: Mitigation.................................................................................................. 2-37 2.8.1 Presentations............................................................................................... 2-37 2.8.2 Panel Discussion.......................................................................................... 2-39 2.8.3 Session Summary........................................................................................ 2-40 2.9 Session 8: Civil Structures, Concrete, and Components............................................. 2-40 2.9.1 Presentations............................................................................................... 2-40 2.9.2 Panel Discussion.......................................................................................... 2-55 2.9.3 Session Summary........................................................................................ 2-61 3 RESEARCH STRATEGY FOR LONG-TERM OPERATION...............................................3-1 3.1 Key Workshop Takeaways...........................................................................................3-1 3.2 Motivation and Objective..............................................................................................3-1 3.3 Approach......................................................................................................................3-1 3.3.1 PIRT Scoring..................................................................................................3-2 3.4 PIRT Evaluations for Individual Sessions.....................................................................3-5 3.4.1 Session 1, Reactor Pressure Vessel, PIRT Evaluation...................................3-5 3.4.2 Session 2, Reactor Pressure Vessel Internals, PIRT Evaluation................... 3-13 3.4.3 Session 3, Reactor Coolant Pressure Boundary (RCPB) Components, PIRT Evaluation......................................................................................................... 3-19 3.4.4 Session 4, Secondary Side Components, PIRT Evaluation.......................... 3-29 3.4.5 Session 5, Balance of Plant Systems........................................................... 3-35 3.4.6 Session 6, Fatigue, PIRT Evaluation............................................................ 3-42 3.4.7 Session 7: Mitigation, PIRT Evaluation......................................................... 3-51 3.4.8 Session 8, Civil Structures, Concrete, and Components, PIRT Evaluation.................................................................................................................. 3-56 3.4.9 Consolidation of PIRT Results...................................................................... 3-78 4
SUMMARY
AND CONCLUSIONS......................................................................................4-1 5 REFERENCES....................................................................................................................5-1 APPENDIX A WORKSHOP ATTENDEES........................................................................ A-1 APPENDIX B PRESENTER AND PANELIST BIOGRAPHIES.......................................... B-1 APPENDIX C
SUMMARY
OF PIRT SCORING RESULTS................................................ C-1
v LIST OF FIGURES Figure 3-1 Schematic illustrating the combinations of importance and knowledge scores suggesting various life management responses................................................3-5 Figure 3-2 PIRT Results for Session 1, RPV.......................................................................... 3-11 Figure 3-3 PIRT Results for Session 2, RVI........................................................................... 3-17 Figure 3-4 PIRT Results for Session 3, RCPB....................................................................... 3-27 Figure 3-5 PIRT Results from Session 4, Secondary Side Components............................... 3-35 Figure 3-6:PIRT results for Balance of Plant Session............................................................. 3-40 Figure 3-7: PIRT Results for Fatigue Session........................................................................ 3-49 Figure 3-8 PIRT Results for Session 7, Mitigation................................................................. 3-54 Figure 3-9 PIRT Results for Session 8 - Effects of Radiation on Structures........................... 3-65 Figure 3-10 PIRT Results, Session 8 - Aging of Post-tensioned Containments...................... 3-68 Figure 3-11 PIRT Results, Session 8 - Coupled Degradation Mechanisms (CDM)................. 3-70 Figure 3-12 PIRT Results, Session 8 - Crevice Corrosion Liner (CCL).................................. 3-72 Figure 3-13 PIRT Results, Session 8 - Enhance Inspection and Monitoring (EIM)................. 3-73 Figure 3-14 PIRT Results, Session 8 - Repair and Replacement Strategies (RRS)............... 3-75 Figure 3-15 PIRT Results, Session 8 - Effects of Climate Change (ECC).............................. 3-76
vi LIST OF TABLES Table 1-1 Workshop Final Agenda....................................................................................1-3 Table 3-1 PIRT Panel Members........................................................................................3-2 Table 3-2 PIRT Importance Scoring Criteria.....................................................................3-3 Table 3-3 PIRT Uncertainty Scoring Criteria.....................................................................3-4 Table 3-4 PIRT Knowledge Scoring Criteria......................................................................3-4 Table 3-5 PIRT Inputs for Session 1, Reactor Pressure Vessel........................................3-5 Table 3-6 PIRT Inputs for Session 2, Reactor Pressure Vessel Internals........................ 3-13 Table 3-7 PIRT Inputs for Session 3, Reactor Coolant Pressure Boundary Components................................................................................................... 3-19 Table 3-8 PIRT Inputs for Session 4, Secondary Side Components............................... 3-29 Table 3-9 PIRT Inputs, Session 5, Balance of Plant Systems......................................... 3-35 Table 3-10 PIRT Inputs, Session 6, Fatigue...................................................................... 3-42
vii ABBREVIATIONS AND ACRONYMS ACES Assessment or Advancement of Civil Engineering Structures ACI American Concrete Institute AI artificial intelligence ALARA as low as reasonably achievable AM additive manufacturing AMP aging management program APT atomic probe tomography ASCET Assessment of Structures subjected to Concrete Pathologies ASME American Society of Mechanical Engineers ASNR Nuclear Safety and Radiation Protection Authority ASR alkali-silica reaction ASTM American Society for Testing and Materials BIM Building Information Modeling BWR boiling-water reactor CANDU CANada Deuterium Uranium CEA French Alternative Energies and Atomic Energy Commission CEPCO Chugoku Electric Power Company CF corrosion fatigue CFD computational fluid dynamics CHUG CHECWORKS Users Group CHz carbohydrazide, CO(NHNH2)2 CIPP cured-inplace-pipe CNSC Canadian Nuclear Safety Commission CSA Canadian Standards Association CUFen Cumulative usage factor including environmental correction Fen DED direct energy deposition DEF delayed ettringite formation DEHA N,N-Diethylhydroxylamine, (C2H5)2NOH DO dissolved oxygen DOE Department of Energy (USA)
EA Erythorbic Acid, C6H8O6 EAF environmentally assisted fatigue EC environmental cracking
viii EDF
Électricité de France EMDA Expanded Materials Degradation Assessment EPRI Electric Power Research Institute ETA ethanolamine, HOCHCHNH FAC flow-accelerated corrosion FEM finite element method FIM field ion microscopy GA Gallic acid, C6H2(OH)3CO2H HAZ heat-affected zone HIP hot isostatic pressing IAEA International Atomic Energy Agency IASCC irradiation-assisted stress corrosion cracking IGALL International Generic Ageing Lessons Learned IGSCC intergranular stress corrosion cracking IHSI induction heating stress improvement IRSN Institut de Radioprotection et de Sûreté Nucléaire LAS low alloy steel LTO long-term operation LWR light-water reactor LWRS light-water reactor sustainability MIC microbially-induced corrosion MITI Ministry of International Trade and Industry ML machine learning MRP materials reliability program MSIP mechanical stress improvement process NEA Nuclear Energy Agency NRA Nuclear Regulation Authority (Japan)
NDE nondestructive evaluation NDT nondestructive testing NPP nuclear power plant NSAC Nuclear Safety Analysis Center OCP open-circuit potential ODOBA Observatory of the durability of reinforced concrete structures ODS oxide dispersion strengthened
ix OE operating experience OLNC OnLine Noble Chemistry PAS positron annihilation spectroscopy PBF powder bed fusion PIRT phenomenon identification & ranking table PNNL Pacific Northwest National Laboratory PT pressure temperature PTS pressurized thermal shock PSSP PWR supplement surveillance program PWR pressurized-water reactor QA/QC quality assurance/quality control RPV reactor pressure vessel ROV remotely operated vehicle RVI reactor vessel internals SCC stress corrosion cracking SCK CEN Belgian Nuclear Research Centre SCs structures & components SGR steam generator replacement SLR subsequent license renewal SSC system, structure, component SIPP sprayed-inplace-pipe SMR small modular reactor TLAA time-limited aging analysis TEM transmission electron microscopy THE Tsujikawa Hisamatsu Electrochemical VACS vessel aging calculation scheme VERCORS Verification Réaliste du COnfinement des RéacteurS
x EXECUTIVE
SUMMARY
The NRC staff conducted a workshop entitled NRC Workshop on Structural Materials:
Research for 80 Years and Beyond, October 1-3, 2024, at NRC Headquarters. The NRC staff designed the workshop to support the development of an NRC research strategy for long-term operation (LTO). This strategy assesses the state of knowledge related to aging management for plant operations beyond 80 years and provides recommendations on research needs to inform the existing aging management programs (AMPs) by identifying any needed modifications or additions. The workshop consisted of eight technical sessions, with seven sessions covering aging issues of metallic components and materials, and one session covering seven topics of aging issues for structural components and materials, particularly concrete structures.
Following the workshop, the NRC staff identified key research recommendations from the presentations and panel discussion for each session. The NRC staff then further investigated the research recommendations by conducting eight expert panel evaluations employing a modified phenomena identification and ranking table (PIRT) process. The expert panels were composed of subject matter experts from the Office of Nuclear Regulatory Research, Division of Engineering (RES/DE), Office of Nuclear Reactor Regulation Division of New and Renewed Licenses (NRR/DNRL), and Office of Nuclear Reactor Regulation Division of Engineering and External Hazards (NRR/DEX). The PIRT determined that the following areas should be considered for future NRC research to support LTO beyond 80 years:
Session 1: Reactor Pressure Vessel Perform research on harvested reactor pressure vessel support materials in order to improve embrittlement predictions for these materials.
Improve existing embrittlement trend curves, such as ASTM E 900, and incorporate improvements into regulatory guidance.
Session 2: Reactor Pressure Vessel Internals Accelerate guidance on qualification and licensing of advanced materials and repair methods.
Session 3: Reactor Coolant Pressure Boundary Perform research to improve accelerated testing methods (e.g., for thermal aging embrittlement, SCC initiation, etc.) to avoid excessive microstructural alterations while still obtaining useful data.
Session 5: Balance of Plant Systems Develop advanced methods to inspect cured-in-place-pipe and sprayed-in-place-pipe Session 8: Civil Structures, Concrete, and Components Harvest and analyze service-irradiated concrete to verify accelerated laboratory data (e.g.,
flux, specimen size) and reduce uncertainties. Investigate the formation of new phases in concrete as concrete ages).
Investigate long-term creep effects, structural modifications, steam generator replacement, loss of prestress, retensioning and interactions between original and replacement materials.
xi Collect actual inplant aged materials performance data through harvesting of inservice or decommissioned plant structures for validation and benchmarking. This research is highlighted for confirmation of aging effects using field data, but subject to the availability for harvesting from appropriate post-tensioned containment.
Design and perform experiments to understand coupled degradation mechanisms and validate findings with field observations and harvested material data.
Integrate advanced technologies like proven artificial intelligence, robotics, drones, building information models (BIM) for enhanced inspection efficiency, improved safety for personnel, increased accessibility, and streamlined data management. Also, improve visual inspections and leverage innovative NDE techniques to detect degradation and aging effects.
This Research Information Letter (RIL) also contains the rationale for highlighting the recommendations above. This RIL could be a foundational document for a broader strategic research assessment. This future work could more wholistically update the existing knowledge base and develop a more comprehensive research strategy for long-term operation.
1-1 1 INTRODUCTION 1.1 Previous NRC Long-Term Operation Workshops NRC staff designed the October 13, 2024, workshop to provide input to aid in development of an NRC strategy for long-term operation (LTO), which assesses the state of knowledge related to aging management for plant operation beyond 80 years and provides recommendations on any research needs to inform the existing aging management programs (AMPs) or identify any needed modifications or additions. In this way, it was meant to be a successor to and build upon the previous NRC LTO workshops and key activities.
In 2021, NRC completed an activity to assess the feasibility of extending the time period for license renewal of nuclear power plants from the current 20 year maximum to a maximum period of 40 years, and to identify options to implement this change. One of the three main recommendations of this activity was to consider an evaluation to identify ongoing research activities (related to concrete, cables, reactor vessel internals and reactor pressure vessels) that could be extended to greater exposure levels (e.g., higher fluence levels) to address the potential for reactor operations up to 100 years.
Earlier, in 2019, (RIL202012) NRC held an international workshop on age-related degradation of reactor vessels and internals. While the workshop presenters varied greatly among government, academic, and industry organizations, their experiences with materials aging management were remarkably similar.
In 2014, NRC published the five-volume Expanded Materials Degradation Assessment (EMDA), NUREG/CR7153, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr7153/v1/index.html [15] The EMDA uses the approach of the phenomena identification and ranking table (PIRT), wherein an expert panel was convened to rank potential degradation scenarios according to their judgment of susceptibility and current state of knowledge. The EMDA built upon NUREG/CR6923, "Expert Panel Report on Proactive Materials Degradation Assessment," referred to as the PMDA report, https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6923/index.html [68] NRC conducted a comprehensive evaluation of potential aging-related degradation modes for core internal components, as well as primary, secondary, and some tertiary piping systems, considering operation up to 40 years. This document has been a very valuable resource, supporting NRC staff evaluations of licensees' aging management programs and allowing for prioritization of research needs.
1.2 2024 Workshop Motivation & Objectives The purpose for the October 2024 workshop was to explore the state of knowledge (including operating experience, research activities, and associated engineering analyses) related to the performance of passive systems, structures and components (SSCs) during LTO of existing light-water reactors (i.e., > 80 years) and identify any research needs to inform the existing aging management programs. The workshop was focused on identifying new, or immature, research topics related to aging effects and mitigation that may become important at such long operating times. Such efforts are intended to be distinct from current and near-term research activities unless they have specific aspects that support extended operation.
1-2 1.3 2024 Workshop Organization & Participants During the four days, the workshop addressed both metallic and concrete degradation, and concrete harvesting.
The first two days of the workshop addressed metallic degradation through focused sessions on the following topics:
- 2. RPV internals
- 3. Reactor Coolant Pressure Boundary (RCPB) Components
- 4. Secondary Side Components
- 5. Balance of Plant Systems (e.g., Service Water, Buried Tanks, and Piping)
- 6. Fatigue
- 7. Mitigation In Days 1 and 2 on Metallic Structures, each session consisted of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of presentations from invited speakers, followed by a 45-minute panel session. The panel session discussed topics raised by the speakers and questions from the facilitator and audience. All panelists had ample opportunity to provide their unique perspectives.
Day 3 addressed Civil Structures Concrete and Components.
Day 4 was a separate workshop on harvesting of concrete structures [9].
The workshop final agenda is provided in Table 1-1. Appendix A contains the list of workshop participants.
1-3 Table 1-1 Workshop Final Agenda Tuesday, October 1 Introductory Session 8:00 Welcome and Introduction Rob Tregoning, NRC 8:05 Opening Remarks John Tappert, NRC Susan Lesica, DOE Jim Cirilli, EPRI 8:20 Research for Long Term Operation Jim Cirilli, EPRI 8:35 Overview of LWRS Materials Research Supporting LTO Frank Chen, ORNL 8:50 NRC Research: Readiness for Long Term Operation Rob Tregoning, NRC John Wise, NRC Session 1: Reactor Pressure Vessel (Coordinator: Jeff Poehler) 9:00 Irradiation Damage Behavior in RPV/Low Alloy Steels: Considerations for 80 years and beyond Grace Burke, INL 9:20 Thermal Annealing: Fresh Look at Old Issue Mikhail Sokolov, ORNL 9:40 EPRI MRP RPV Integrity LTO Research Elliot Long, EPRI 10:00:30 Break 10:30 Session 1 Panel Discussion Tim Griesbach, SIA Susan Ortner, NNL Brian Hall, Westinghouse Akiyoshi Nomoto, CRIEPI Session 2: Reactor Pressure Vessel Internals (Coordinator: Austin Young) 11:15 Connecting the dots to support Long Term OperationWhat out there and what is needed?
Pl Efsing, Vattenfall 11:35 Emerging material degradation issues in nuclear materials under extended inservice irradiation beyond 80 years Maxim Gussev, ORNL 11:55 Irradiation-Assisted Stress Corrosion Cracking:
From Mechanisms to Mitigation Steven Raiman, U. Michigan 12:03:30 Lunch Break 13:30 Session 2 Panel Discussion Yiren Chen, ANL Kyle Amberge, EPRI 14:15 14:45 Break Session 3: Reactor Coolant Pressure Boundary Components (Coordinator: Eric Focht) 14:45 Environmental Cracking During Extended LWR Operation Péter Andresen, Andresen Consulting
1-4 Tuesday, October 1 15:05 Addressing Gaps for Life Beyond 80: Experience based on SCC Initiation Research on RCPB Structural Materials Ziqing Zhai, PNNL 15:25 Long-Term Approach to Manage Stress Corrosion Cracking Thierry Couvant, EDF 15:45 Session 3 Panel Discussion Bogdan Alexandreanu, ANL Mychailo Toloczko, PNNL Takumi Terachi, INSS Concluding Remarks 16:30 Day One Wrap-Up Discussion Rob Tregoning, NRC 17:00 Adjourn Day 1 Wednesday, October 2 Introductory Session 8:00 Welcome and Introduction Jeff Poehler, NRC 8:05 Opening Remarks Christian Araguas, NRC Session 4: Secondary Side Components (Coordinator: Mekonen Bayssie) 8:10 Evaluation of Hydrazine Alternatives as Oxygen Scavenger in PWR Secondary Systems Yudai Yamamoto, CRIEPI 8:30 EPRI Research on Flow-Accelerated Corrosion and Erosion in Nuclear Power Plants Ryan Wolfe, EPRI 9:00 Session 4 Panel Discussion Jared Smith, EPRI Frank Gift, EPRI 9:45 10:15 Break Session 5: Service Water, Buried Tanks and Piping (Coordinator: Amy Hull) 10:15 Service Water Systems at Nuclear Plants George Licina, SIA 10:40 Technical advances in Cured-inPlace-Pipe (CIPP) linings John Cavallo 11:05 Session 5 Panel Discussion Dylan Cimock, EPRI Chris Burton, Structural Technol.
John Wise, NRC 11:53:20 Lunch Break Session 6: Fatigue (Coordinator: Rob Tregoning) 13:20 EPRI MRP Perspectives on Fatigue-Related Issues for 80 Years and Beyond Tom Damiani, EPRI
1-5 Wednesday, October 2 13:40 Fatigue Evaluation for Additive Manufacturing Materials Seiji Asada, Mitsubishi Heavy Industries 14:00 Uncomfortable with Uncertainty?
Andrew Morley, Rolls-Royce 14:20 Session 6 Panel Discussion Mark Gray, Westinghouse Tim Gillman, SIA Jonathan Mann, Amentum 15:05 15:30 Break Session 7: Mitigation (Coordinator: Dale Delisle) 15:30 Mitigation of Material Degradation Associated with Long Term Operation Darren Barborak, EPRI 15:55 Advanced Laser Welding Techniques for Repair of Irradiated Light-Water Reactor Components Jian Chen, ORNL 16:25 Session 7 Panel Discussion Taku Arai, CRIEPI Carol Moyer, NRC Darren Barborak, EPRI Concluding Remarks 17:00 Day Two Wrap-Up Discussion Robert Tregoning, NRC 17:30 Closing Remarks Steve Ruffin, NRC 17:35 Adjourn Day 2 18:30 Dinner at Pin Stripes https://www.pinstripes.com/bistro POC: Amy Hull Thursday, October 3, 2024 Introductory Session (Coordinator: Jose Pires) 9:00 Welcome and Introduction Laurel Bauer, NRC 9:05 Opening Remarks Marissa Bailey, NRC Sue Lesica, DOE 9:20 Overview of LWRS Materials Research Frank Chen, ORNL Session 8: Civil Structures Concrete and Components (Coordinators: Madhumita Sircar, Jose Pires) 9:35 Research for Beyond 80 YearsCivil Structures Mita Sircar, NRC 9:45 Operation Beyond 80: Knowledge Gaps and Synergistic Degradation Modes in Concrete Yann Le Pape, ORNL 10:1510:30 Break 10:30 EPRI Research on Long Term Operations:
Considerations for Life Beyond 80 Samuel Johnson, EPRI 11:00 AMPs and Future Research Need for LTO on Concrete Structures Masa Kojima, NRA Japan, Ippei Maruyama, U-Tokyo
1-6 Thursday, October 3, 2024 11:20 Research Results on Concrete Using the Decommissioning Plant of the Hamaoka Nuclear Power Plant Takashi Osaki, CEPCO 11:43:10 Lunch 13:10 NRC Workshop: What Research for Beyond 80 Years for Civil Structures - IRSN contribution Fabienne Ribeiro, IRSN 13:30 NRCEDF LTOEDF Approach Benoit Masson, EDF (R) 13:50 CNSC Perspective on Gaps in Knowledge for Concrete Containment Structures for LTO Cédric Androut, CNSC (R) 14:10 Aging Management of Nuclear Civil Structures -
Potential Areas for Enhancements Julia Tcherner, AtkinsRéalis 14:34:45 Break 14:45 Crevice Corrosion Assessment of Steel Liners &
Cement Interface in Nuclear Power Plants Valdir de Souza, SCKCEN (R) 15:00 Crevice Corrosion Modelling of Steel Liners &
Cement Interface in Nuclear Power Plants Anssi Laukkanen, VTT (R) 15:15 Creep and Creep-cracking in PCCV Randy James, SSC (R) 15:3515:50 Break 15:50 Panel Discussion Yann. LePape, ORNL Samuel Johnson, EPRI Ippei Maruyama, U Tokyo C. Jones, U-Kansas J. Tcherner, AtkinsRéalis Concluding Remarks 16:40 Day Three Wrap-Up: Key Takeaways J. Pires, M. Sircar, NRC 16:55 Closing Remarks L Bauer, J McKirgan, NRC 17:00 Adjourn Day 3 1.4 Organization of This Report Section 2 summarizes the presentations and panel discussions for each session and identifies research ideas and recommendations suggested by the presenters and panelists. Section 3 describes the Phenomenon Identification and Ranking Table (PIRT) process used to determine the high priority research recommendations to support operation beyond 80 years arising from the technical sessions and also describes the PIRT results for each session. Session 3.4.9 lists the highest priority research recommendations determined by the NRC staff based on the PIRT process. Section 4 contains a brief summary and conclusions. Session 5 contains references.
2-1 2
SUMMARY
OF PRESENTATIONS 2.1 Opening Session 2.1.1 Welcome and Opening Remarks Rob Tregoning, Senior Technical Adviser for Materials at NRCs Office of Nuclear Regulatory Research (RES), welcomed attendees, outlined the workshops purpose, and covered meeting logistics. Senior officials from NRC, DOE, and EPRI gave opening remarks. John Tappert, the Acting Director for RES, emphasized the importance of LTO for the existing nuclear fleet and the need for research to support safe operations beyond 80 years. Sue Lesica, Materials Engineer at the U.S. Department of Energy (DOE), highlighted DOEs interest in LTO and the need for research to address knowledge gaps. Jim Cirilli, Senior Technical Executive at the Electric Power Research Institute (EPRI), discussed EPRIs research focus on LTO and the importance of identifying research needs and knowledge gaps.
2.1.2 Presentations 2.1.2.1 Research for Long Term Operation (ML24274A176)
Presenter: Jim Cirilli, EPRI Jim Cirilli further discussed EPRIs research focus on LTO and the importance of identifying research needs and knowledge gaps. He emphasized that U.S. leads the timeline with respect to license renewal, and many of the international EPRI members and their regulators are keen observers of the NRC's actions related to license renewal. EPRI focuses on obtaining more carbon free megawatts from the existing international fleet through four main areas: (1) power up rates, (2) LTO (3) outage optimization, and (4) leveraging new technologies. The EPRI LTO relies on continuous aging management and issue assessment cycle, using the EPRI materials degradation matrix, and the issue management tables to identify high priority research needs.
2.1.2.2 Overview of LWRS Materials Research Supporting LTO (ML24282A772)
Presenter: Frank Chen, ORNL Frank Chen, as ORNL R&D staff member, also leading DOE's Light Water Reactor Sustainability (LWRS) Program, is particularly interested in the mechanical properties of structural materials under extreme environments, and the correlation of mechanical properties and materials microstructure. The main objective of the LWRS materials pathway is to provide data and methods to assess the performance damage, as well as mitigation options for systems, structure, and components (SSC)that are essential to safe and sustainable operation of existing nuclear fleets.
2.1.2.3 NRC Research: Readiness for Long Term Operation (ML24274A177)
Presenters: Rob Tregoning and John Wise, NRC As NRC representatives, they discussed the regulatory framework for license renewal and the importance of research in supporting regulatory decisions for operations beyond 80 years. They presented a timeline showing that of the oldest U.S. plants, there are two or three that are about 55 years old, and the average U.S. plant age right now is 42 years and thus an aging fleet.
Research is really a foundation of our regulatory framework along with current plant activities
2-2 (NUREG1412). From this meeting, they hope to identify research topics worth pursuing, gather opinions and knowledge gaps, and explore possible research strategies.
2.1.3 Key Themes of the Introductory Session This session set the stage for detailed discussions on the technical and regulatory aspects of extending the operational life of nuclear power plants, emphasizing the need for proactive research and collaboration. Speakers emphasized the following:
Research Needs: Identifying and addressing knowledge gaps related to materials degradation, aging management, and mitigation techniques.
Collaboration: Emphasis on collaboration between NRC, DOE, EPRI, and international partners to support LTO.
Regulatory Framework: Understanding the regulatory requirements and preparing for potential license renewals beyond 80 years.
Workshop Goals: Collecting ideas, prioritizing research topics, and developing strategies to address LTO challenges.
2.2 Session 1Reactor Pressure Vessel 2.2.1 Presentations 2.2.1.1 Irradiation Damage Behavior in RPV/Low Alloy SteelsConsiderations for 80 Years and Beyond (ML24282A715)
Presenter: M. Grace Burke, MFC/Idaho National Laboratory, University of Manchester (Emeritus)
Grace Burke discussed the history and advancements in the study of reactor pressure vessel (RPV) embrittlement, focusing on irradiation damage and the role of microstructure in materials performance and degradation. She highlighted the importance of surveillance programs that provided critical data on high-dose irradiations, especially for materials relevant to nuclear plants.
Burke covered the evolution of characterization techniques from the 1960s to the present, including Transmission Electron Microscopy (TEM), Field Ion Microscopy (FIM), Atom Probe Tomography (APT), and Positron Annihilation Spectroscopy (PAS). She emphasized the need for robust, reliable mechanical properties data and the importance of evaluating real materials through appropriate sampling and high-dose neutron irradiated materials.
The presentation also touched on the variability in material properties, the role of small angle neutron scattering, and the significance of manganese and nickel in developing stable neutron irradiation-induced hardening in high nickel steels. Burke concluded by stressing the importance of understanding material specifications and the potential impact of new production techniques on material properties.
The presentation identifies several key research needs in the study of RPV embrittlement and irradiation damage:
2-3
- 1. Surveillance Programs: The importance of surveillance programs that provide critical data on high-dose irradiations, especially for materials relevant to nuclear plants, is emphasized.
- 2. Characterization Techniques: and the evolution of characterization techniques such as TEM, FIM, APT, and PAS are highlighted.
- 3. Material Variability: The presentation stresses the importance of understanding material variability and ensuring valid, reproducible data, particularly with respect to microstructure and properties.
- 4. Low-Temperature Irradiation: The significance of low-temperature irradiation data for support structures and small modular reactors is mentioned, with reference to the Advanced Test Reactor (ATR2) program led by Bob Odette.
- 5. New Production Techniques: The potential impact of new production techniques on material properties and the importance of understanding material specifications are discussed.
- 6. Real Materials Evaluation: The need for evaluating real materials through appropriate sampling and high-dose neutron irradiated materials is emphasized. There is a need for robust and reliable mechanical properties data.
2.2.1.2 Thermal Annealing: A Fresh (80+) Look at Old (40+) Issue Presenter: Mikail Sokolov, Oak Ridge National Laboratory (ML24282A716)
The presentation provided an indepth discussion on thermal annealing as a technique to extend the life of nuclear RPVs. It started by highlighting a quiz to gauge the audience's awareness of thermal annealing in the United States, revealing that only two nuclear reactor pressure vessels have undergone this process: the unfinished Marble Hill in Indiana and the U.S. test reactor in Alaska, SM1 A.
Thermal annealing was described as a process distinct from metallurgical annealing, performed to recover and mitigate irradiation embrittlement. It involves heating the RPV to a temperature range between 340°C and 500°C for about a week. The effectiveness of recovery depends on factors such as annealing temperature, time, material composition, and level of damage.
However, it is noted that the microstructure of annealed material differs from its pre-irradiation state, affecting embrittlement differently.
The presentation outlined several engineering solutions to extend RPV life to 80 years or more, including:
Low-leakage core to reduce reactor flux Reevaluation of pressurized thermal shock (PTS) screening material Direct fracture toughness measurements Among these techniques, thermal annealing was highlighted as the only method that allows for the recovery of irradiated belt line transition temperature shift and recovery of outer shell toughness.
2-4 Sokolov also mentioned that various factors influence the effectiveness of annealing, such as temperature, time, material composition, and damage level. It concluded with an explanation of a visual representation showing the ductile to brittle transition temperature and PTS screening criteria, emphasizing the importance of these techniques in maintaining the longevity and safety of nuclear reactors.
The presentation outlined several key research recommendations related to the thermal annealing of nuclear RPVs. Here are the main points:
- 1. Effectiveness of Annealing: The effectiveness of annealing is not solely determined by how much the transition temperature can be recovered but by how long the reactor can operate afterward. The rate of reembrittlement is crucial.
- 2. Reembrittlement Studies: There is a need for systematic studies on the thermal stability of nickel manganese clusters and how they respond to annealing. Most of the previous research focused on copper precipitates, which are less relevant for materials that will operate for 80 years.
- 3. Direct Fracture Toughness Methods: The industry is moving toward direct fracture toughness methods, which could impact the annealing process. This approach allows for the expansion of surveillance programs and provides opportunities to assemble surveillance placements that follow embrittlement after annealing.
- 4. Technological Challenges: There are two types of annealing: wet and dry. Wet annealing is simpler but less effective, while dry annealing is more complex and requires the removal of core internal structures and water. Both methods have their own sets of challenges and benefits.
- 5. Regulatory Updates: Existing regulations and guidelines, such as Reg Guide 1.162, need to be updated to account for new findings related to nickel manganese clusters and other factors affecting embrittlement.
- 6. Future Research: To apply annealing for up to 80 years of operation, research needs to start now to ensure that the current fleet of reactors can benefit from these techniques.
2.2.1.3 EPRI MRP Low Alloy Steel Research for LTO Presenter: Elliot Long, Electric Power Research Institute (ML24274A178)
Elliott Longs presentation outlined the role of the Materials Reliability Program (MRP) and the ongoing research efforts to address potential gaps in RPV integrity for LTO. Long emphasized the importance of time-limited aging analyses (TLAAs) in ensuring the safety and reliability of nuclear plants as they seek license extensions from 40 to 60 years, and potentially up to 80 years or beyond.
Key areas of focus included the effects of neutron irradiation on RPV materials, the regulatory framework governing RPV integrity, and the methodologies for assessing and mitigating embrittlement. The presentation highlighted the pressurized-water reactor (PWR) Supplement Surveillance Program (PSSP), which aims to generate high-fluence data to validate
2-5 embrittlement trend correlations for long-term RPV operation. Long also discussed the potential need for additional surveillance capsules and the importance of updating material properties and regulatory guides to ensure continued safety.
The presentation concluded with a call for proactive data collection and research to support the safe extension of nuclear plant operations, emphasizing the collaborative efforts of industry groups, regulatory bodies, and research institutions.
The presentation identified several key research recommendations:
Enhance American Society for Testing and Materials (ASTM) E90015: Improve and enhance the existing ASTM E90015 model to better predict embrittlement trends at high fluence levels.
Generate High-Fluence Data: Continue the PSSP to gather high-fluence surveillance data from actual RPV materials irradiated in commercial PWRs.
Update Regulatory Guides: Revise regulatory guides, such as Reg. Guide 1.99, to incorporate new data and improve accuracy at higher fluence levels.
Additional Surveillance Capsules: Consider the need for additional surveillance capsules for plants expected to reach high fluence levels, ensuring comprehensive data collection for LTO.
Collaborative Research Efforts: Foster collaboration between industry groups, regulatory bodies, and research institutions to address potential gaps in RPV integrity and ensure the safe extension of nuclear plant operations.
2.2.2 Panel Discussions Panelists:
Tim Griesbach (Structural Integrity Associates, USA)
Susan Ortner (National Nuclear Laboratory, United Kingdom)
J. Brian Hall (Westinghouse Electric Company, LLC, USA)
Akiyoshi Nomoto (Central Research Institute of Electric Power Industry, Japan)
Summary of Discussion The discussion began with a question from the chat about parallel modeling efforts at the metal grain scale. Tim Griesbach explained that initial trend curve models were developed without the ability to see precipitate sizes, but newer models have confirmed the predictions of those earlier models. Brian Hall added that modeling at multiple scales is helpful but needs to be backed up with measurements, and Susan Ortner emphasized the need for better fracture modeling and more data to understand the scatter in trend lines.
The panel then moved on to pre-prepared questions, starting with the most likely mechanism that could limit the life of RPVs beyond 80 years. Tim Griesbach suggested that PT curves could be the most limiting factor, as they narrow the operating window for heat-up and cooldown. Brian Hall mentioned the need for improved methods of fracture toughness measurements and the potential use of annealing. Brian Hall also discussed reactor pressure
2-6 vessel supports as a potential issue that could benefit from additional data, since supports are made from a variety of steels, and most testing has been in high-flux test reactors. Brian Hall indicated that harvesting support materials from operating plants would be potentially useful.
Nomoto and Ortner discussed the importance of monitoring specimens and the potential issues with phosphorus segregation and thermal aging.
The second pre-prepared question addressed whether further knowledge refinement or mitigation techniques will be required to support operation beyond 80 years. Tim Griesbach emphasized the importance of credible and valid methods to predict embrittlement trends, while Hall and Nomoto discussed the potential need for flux reduction and thermal annealing. Susan Ortner highlighted the importance of understanding margins and avoiding double-accounting.
The discussion concluded with a question from the chat about the potential benefits of operational changes to improve the PTS situation. Tim Griesbach and Brian Hall explained that plant-specific differences can be incorporated into a probabilistic analysis framework to address the issue.
Overall, the panel discussion highlighted the complexity of modeling radiation-induced embrittlement and the need for continued research and data collection to support LTO of RPVs.
The following are the key research recommendations arising from the Session 1 panel discussion:
Parallel Modeling Efforts: There is a need for parallel modeling efforts at the metal grain scale to replicate radiation-induced embrittlement.
Fracture Toughness: Emphasis on the importance of fracture toughness backed by measurements to inform predictive embrittlement.
Empirical Radiation State: Use empirical radiation state between microstructure change and mechanical property change for future modeling.
Better Fracture Modeling: Need for better fracture modeling on many levels to understand scatter and correlations.
Data Collection: More data of an appropriate kind is always needed, especially for low-copper, low-phosphorous materials, which may not be well-represented by current embrittlement trend curves.
Reactor Pressure Vessel Supports: Harvesting of reactor pressure vessel support materials could be beneficial for improving embrittlement predictions for these materials.
PT Curves: Potential limitations beyond 80 years could be due to PT curves, which narrow the operating window for heat-up and cooldown.
Direct Fracture Toughness: Improved methods of fracture toughness measurements are helpful for LTO.
Monitoring Specimens: Lack of monitoring specimens for future LTO.
2-7 Phosphorus Segregation: Concerns about phosphorus segregation levels becoming dangerous for intergranular failure.
Mitigation Techniques: Greater use of mitigation techniques such as thermal annealing or flux reduction may be required.
2.2.3 Session Summary The presentations and the panel discussion highlighted several common themes for research on RPV integrity beyond 80 years. One common theme is surveillance programs and data collection, in particular the enhancement of surveillance programs to provide critical data on high-dose irradiation data, especially for materials relevant to nuclear plants, and proactive collection of data to support the safe extension of nuclear plant operations. In the area of materials characterization, the session highlighted the need for development of robust and reliable mechanical properties data through advanced characterization techniques such as TEM, FIM, APT, and PAS, to permit evaluation for real materials through appropriate sampling and high-dose neutron irradiated materials. Additionally, there is a need to understand material variability and ensure valid, reproducible data, particularly with respect to microstructure and properties.
Another area discussed in the session is thermal annealing and mitigation techniques. In particular, there is a need for systematic studies on the thermal stability of nickel manganese clusters and their response to annealing. Annealing may necessitate a move toward direct fracture toughness methods to expand surveillance programs and provide opportunities to assemble surveillance placements that follow embrittlement after annealing. Also, technological challenges associated with wet and dry annealing methods should be addressed.
The session pointed out that regulatory and methodological updates may be necessary, such as updates to Regulatory Guide 1.162, to account for new findings related to nickel manganese clusters and other factors affecting embrittlement, as well as improvements to the existing ASTM E90015 model to better predict embrittlement trends at high fluence levels.
Several presenters and panelists noted that direct fracture toughness methods may be needed to support LTO, and that better fracture modeling on many levels to understand scatter and correlations is needed.
Finally, the session highlighted the importance of collaboration between industry groups, regulatory bodies, and research institutions to address potential gaps in RPV integrity and ensure the safe extension of nuclear plant operations.
2.3 Session 2: Reactor Pressure Vessel Internals 2.3.1 Presentations 2.3.1.1 Connecting the dots to support Long Term OperationWhat out there and what is needed?
Pl EfsingRinghals AB/NUQ (ML24282A719)
Efsings presentation focused on the aging management of Reactor Vessel Internals (RVIs),
emphasizing the importance of early planning, continuous monitoring, and data-driven decision-making to ensure the safe and extended operation of nuclear power plants. He highlighted key
2-8 aging mechanisms, the role of material harvesting, and the need for improved documentation and benchmarking to enhance predictive modeling and regulatory compliance.
Efsing delivered a comprehensive perspective on aging management in nuclear power plants, emphasizing that aging must be addressed from the very inception of plant ownership. He articulated that aging management is integral to defense-indepth and vital to public and regulatory confidence, especially as facilities push beyond 60 years of operation and consider lifetimes exceeding 80 years. He reminded the audience that robust, reviewable, and evidence-based programs are essential for demonstrating fitness for service.
He traced the evolution of aging management practices from early inservice inspections and surveillance activities to todays more formalized aging management programs. Efsing stressed that these programs are most effective when grounded in operating experience (OE) and supported by ongoing R&D. He emphasized the universality of aging mechanisms across global fleets, noting that materials in Sweden, Japan, or the U.S. behave similarly under irradiation and thermal stress.
A core focus of his remarks was the need to reduce extrapolation in long-term predictions by improving the quality and completeness of the material knowledge base. This includes understanding both initial manufacturing conditions and degradation behaviors through advanced modeling and data derived from harvested materials. He pointed to reactor internals, especially welds and base metals, as areas of special interest, and advocated for benchmarking data from harvested materials against test reactor data to assess acceleration validity.
Efsing underscored the critical importance of documentation and knowledge retention, citing unknown knowns lost through attrition or lack of institutional memory. He cautioned against premature conclusions without adequate data and called for continual reassessment and validation of existing assumptions. He also highlighted the promise and limitations of modern tools like AI and machine learning, which, while useful for connecting known data points, cannot yet discover the unknown.
Ultimately, Efsing called for a strong OE exchange and continuous scientific rigor to support long-term safe operation. He advocated transparency, high-quality data collection, and better utilization of harvested materials to inform modeling and extend predictive capabilities.
The presentation identified several key research recommendations:
Improved characterization of initial manufacturing conditions: Many legacy components lack complete records of their fabrication, making it difficult to establish accurate baselines for aging analyses.
Advanced modeling tools to reduce extrapolation: Modeling should be used to extend the applicability of existing data and reduce reliance on long extrapolations for LTO.
Validation of accelerated testing using harvested materials: Comparative studies between test reactor-irradiated and service-irradiated materials are necessary to ensure conservatism (and avoid over-conservatism) in degradation predictions.
2-9 Focused research on reactor internals (e.g., weld metals, base metals): These materials, especially in high-flux zones, are primary candidates for detailed studies through material harvesting and analysis.
Expanded use and validation of degradation indicators: There is a need to benchmark and refine the indicators used to track aging, ensuring they truly represent the physical degradation taking place.
Understanding and managing wear and low-profile degradation mechanisms: Not all degradation is due to high-profile mechanismssome are basic, cumulative, and require better recognition in aging management.
Knowledge retention and documentation systems: Preventing knowledge loss through retirement or turnover is essential. Maintaining high-quality, accessible documentation is critical.
Development of better harvesting strategies and guidelines: Optimizing the use of harvested materials for research and model validation should be a coordinated industry priority.
Transparent use of machine learning and AI: These tools must be applied carefully, ensuring they complement human understanding without introducing false confidence in poorly understood areas.
2.3.1.2 Emerging material degradation issues in nuclear materials under extended inservice irradiation beyond 80 years Maxim N. GussevOak Ridge National Lab (ORNL)( ML24282A720)
Maxim Gussevs presentation focused on lesser-known, second-order material degradation phenomena in RVI, which may become critical issues under extended reactor operation. He highlighted several key concerns related to transmutation, phase instabilities, and radiation-induced transformations in austenitic stainless steels, particularly 304 and 316 stainless steelsthe primary materials used in RVI.
Gussev presented on emergent degradation mechanisms in austenitic steels under long-term irradiation in nuclear reactors, emphasizing phenomena that are currently considered secondary but may evolve into primary concerns as service durations extend beyond original design lifetimes. Unlike conventional degradation topics such as irradiation hardening or stress corrosion cracking, Gussev focused on transmutation effects, radiation-induced phase instability, helium/hydrogen accumulation, and nonlinear material transformations.
He highlighted that long-term neutron irradiation causes transmutation of key alloying elements (e.g., nickel, manganese, nitrogen), which destabilize the austenitic phase and promote ferritic and martensitic transformations. These changes are typically minor in short-term service but could accelerate with increasing fluence and flux, leading to nonlinear degradation and phase instability. Of particular concern is the formation of ferrite along grain boundaries and the potential for martensitic transformation at room temperature, which could degrade mechanical properties and corrosion resistance.
The presentation also underscored the inadequacy of existing models and data in predicting these emergent behaviors. Gussev cited helium accumulation and gas bubble evolution as
2-10 examples of poorly understood, nonlinear processes that may contribute to embrittlement or intergranular fracture. He called attention to neutron thermalization effects near water-metal interfaces, which can result in unexpected transmutation rates, and presented experimental observations suggesting abrupt increases in ferrite content and abnormal mechanical behavior at high doses and low fluxes.
Ultimately, Gussev emphasized the urgent need for detailed post-irradiation examinations (PIE) of materials from reactors undergoing license extensions, noting that current materials may contain legacy impurities and lack resilience to these evolving mechanisms.
The presentation identified several key research recommendations:
Improve understanding of transmutation-induced phase instability Enhance models and data to quantify the role of transmutation in destabilizing austenitic microstructuresparticularly the loss of Ni, Mn, and N, and the formation of ferrite-and martensite-promoting elements like V. Assess the implications for phase stability at extended lifetimes and higher fluences.
Clarify the role of helium and hydrogen in degradation mechanisms: Investigate helium/hydrogen accumulation, bubble formation, and their effects on grain boundary embrittlement. Establish mechanisms of bubble pressure evolution and interaction with radiation-induced defects under varying flux and temperature conditions.
Assess thresholds for radiation-induced phase transformations: Determine critical irradiation conditions (e.g., dose, dose rate, flux) that result in ferritic and martensitic transformations in austenitic steels, with a focus on transformations along grain boundaries and at surface interfaces.
Develop predictive models for nonlinear material degradation: Address the limitations of existing linear degradation models by identifying mechanisms and thresholds that lead to abrupt changes in mechanical properties (e.g., ductility loss, strength anomalies) during long-term irradiation.
Evaluate localized transmutation effects near thermal interfaces: Examine the impact of thermal neutron spectrum shifts near water-metal interfaces on local transmutation rates and compositional changes. Improve spatial resolution of neutron transport models to better predict these localized phenomena.
Conduct post-irradiation examinations on service-exposed materials: Prioritize PIE on long-exposed reactor materials, especially those with legacy impurity profiles, to validate modeling assumptions and observe emergent degradation behavior under real operating conditions.
Investigate effects of radiation-induced microstructural evolution on corrosion behavior: Assess how phase transformations and transmutation products influence corrosion resistance and crack initiation in light-water reactor environments, particularly in relation to grain boundary chemistry and structure.
Characterize synergistic effects among multiple degradation mechanisms: Explore how transmutation, gas accumulation, phase transformation, and dislocation evolution
2-11 interact to produce accelerated or emergent degradation responses not predicted by single-mechanism models.
Conclusion & Future Considerations:
Many degradation mechanisms that were once considered minor may evolve into dominant failure mechanisms during extended reactor operation. As reactors age and materials are subjected to prolonged stress and radiation exposure, these seemingly insignificant issues can accumulate and pose serious threats to the structural integrity and reliability of critical components.
Nonlinear effects and complex material interactions have the potential to accelerate degradation in unexpected ways. These interactions can amplify the impact of environmental and operational stresses, leading to faster material deterioration than initially predicted, and posing challenges for accurate lifetime assessments and maintenance planning.
Post-irradiation evaluations of Reactor Vessel Internal (RVI) materials from reactors operating under license extension conditions are essential. These assessments provide critical insights into the actual condition of materials after prolonged exposure, enabling more accurate predictions of future performance and ensuring the continued safety and reliability of extended reactor operations.
Further studies are necessary to deepen our understanding of phase stability, transmutation effects, and material embrittlement in order to ensure the safe LTO of nuclear reactors. Advancing knowledge in these areas is critical for predicting material behavior under extended service conditions and for developing effective aging management strategies.
2.3.1.3 Irradiation-Assisted Stress Corrosion Cracking From Mechanisms to Mitigation Stephen Raiman University of Michigan (ML24282A721)
Steven Raimans presentation focused on Irradiation-Assisted Stress Corrosion Cracking (IASCC) in reactor vessel internals, outlining the mechanisms responsible for material degradation and potential mitigation strategies.
Raiman presented a comprehensive overview of recent advancements in understanding and mitigating Irradiation-Assisted Stress Corrosion Cracking (IASCC) in reactor core internals. He described the current state of IASCC research as being at an inflection point, where the mechanisms are now reasonably well understood, allowing researchers to shift focus toward practical mitigation strategies. The goal is to apply decades of mechanistic knowledge to engineer materials and microstructures that can resist IASCC and support plant life extensions to 80 years.
Raiman summarized three key mechanisms that contribute to IASCC:
Dislocation channeling, which leads to stress concentration at grain boundaries.
Grain boundary oxidation, which weakens those boundaries; and
2-12 Silicon segregation, which further reduces grain boundary integrity and contributes to crack growth.
He then outlined several research efforts aimed at mitigating these effects through microstructure engineering and thermal treatments, focusing on advanced manufacturing and material processing techniques such as ODS steels, nanograined alloys, additive manufacturing, and powder metallurgy with hot isostatic pressing (HIP).
The presentation identified several key research recommendations:
Mechanistic Understanding of IASCC: Refine and validate the role of three key mechanisms contributing to IASCC: (1) localized deformation via dislocation channeling, (2) grain boundary oxidation, and (3) silicon segregation and dissolution. Improve fundamental understanding of how these mechanisms interact and evolve under neutron irradiation and reactor coolant environments.
Advanced Materials with Engineered Microstructures: Investigate the IASCC resistance of novel materials engineered to interrupt or eliminate dislocation channeling, including:
Ultrafine/nanocrystalline alloys processed by equal channel angular extrusion (ECAE).
Oxide dispersion strengthened (ODS) steels.
Additively manufactured stainless steels produced via hot isostatic pressing (AM-HIP).
Conduct irradiation and environmental testing to evaluate crack initiation and growth in these materials under representative conditions.
Post-Irradiation Thermal Annealing Treatments: Evaluate the use of thermal annealing to reduce irradiation damage and improve IASCC resistance in stainless steel reactor internals. Determine optimal temperature and duration conditions to restore mechanical properties while maintaining corrosion resistance.
Optimization of Powder Metallurgy for Core Internals: Identify best practices for powder selection, degassing, forming, and HIP processing to produce robust, IASCC-resistant microstructures. Generate data on the long-term performance of powder metallurgy components under irradiation to support potential deployment in reactor systems.
Industry Integration and Deployment: Explore the practical application of thermal annealing and advanced manufacturing techniques for repair and replacement of reactor internals. Identify regulatory, technical, and operational considerations for qualification and licensing of materials with engineered microstructures.
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Conclusion:==
The understanding of irradiation-assisted stress corrosion cracking (IASCC) mechanisms has advanced significantly, allowing for a shift toward more proactive mitigation strategies. This improved knowledge has led to better predictive models and preventive measures, enabling the
2-13 industry to address IASCC risks before they lead to material failures, ultimately enhancing the safety and reliability of reactor components.
Novel materials and advanced manufacturing techniques are crucial for extending the life of reactor vessel internals to 80 years. By developing materials that can withstand prolonged radiation exposure and utilizing cutting-edge manufacturing processes, it is possible to enhance the durability and performance of reactor components, ensuring their continued reliability and safety over extended operational periods.
Collaboration with industry is essential for successfully implementing these solutions in operational nuclear reactors. By working closely with industry experts and stakeholders, it becomes possible to tailor solutions to real-world conditions, address practical challenges, and ensure that new technologies and materials can be effectively integrated into existing reactor systems to improve safety and performance.
2.3.2 Panel Discussions The panel discussion focused on the LTO of nuclear reactor pressure vessel internals, particularly the effects of aging, cracking mechanisms, and potential mitigation strategies.
Panel
Participants:
Kyle AmbergeElectric Power Research Institute (USA)
Yiren ChenArgonne National Laboratory (USA)
The panel session encompassed a technical discussion between experts from EPRI, Argonne National Laboratory, and other stakeholders on aging mechanisms in nuclear reactor materials, particularly in the context of LTO beyond 80 years. The conversation begins with panelist introductions, highlighting extensive expertise in materials degradation, corrosion, and irradiation effects. A major theme is the potential for changes in the dominance of known degradation mechanisms, rather than the emergence of entirely new ones. Panelists emphasized that synergistic effectswhere multiple degradation modes interact or accelerate one anotherwill likely become more prominent over extended lifetimes. Specific examples, such as irradiation-assisted stress corrosion cracking (IASCC), fatigue, and baffle-former bolt failures, were cited to illustrate the need for more integrated assessments of material behavior under combined stressors.
Another key topic was the impact of irradiation on material microstructures, including concerns about radiation-induced ordering in nickel-based alloys like Alloy 625. Although such phenomena have been observed in laboratory settings, their practical implications in light water reactor (LWR) environments remain uncertain and require further investigation. The discussion also addressed mitigation strategies, including alloy selection and heat treatments, though the feasibility of applying these to existing reactor components is limited. The panel concluded that while long-term degradation may not involve entirely new mechanisms, understanding the complex interactions among existing ones is crucial for ensuring safe and reliable reactor performance into the future.
The discussion continues to explore the challenges of managing materials degradation in nuclear reactors operating up to and potentially beyond 100 years. Concerns were raised about the unintended consequences of applying heat treatments to irradiated materials, which might introduce new or accelerated degradation phenomena due to changes in microstructure.
Panelists emphasized the need for caution when considering thermal aging remedies and noted
2-14 that, while mitigation strategies like reducing stress or modifying environmental conditions (e.g.,
silicon content) can help, applying these retroactively to irradiated reactor components is complex and often impractical.
The discussion then turned to the feasibility of component replacements, particularly smaller parts such as baffle-former bolts. While replacement is possible and has been done in some cases, practicality depends on radiation levels, component accessibility, and the business case for utilities. Heat-treating major components like core barrels was considered largely impractical due to ALARA (As Low as Reasonably Achievable) constraints and the logistics of high-radiation environments.
Leonard Bond raised the issue of detecting degradation earlybefore failures like bolt breakage occuremphasizing the need for practical, inservice inspection methods that move beyond lab-scale nanostructural analysis. Panelists agreed, citing the importance of timely inspections using proven techniques like ultrasonic testing. They also noted that national labs and industry must work together to better understand degradation mechanisms and translate that understanding into effective monitoring and mitigation strategies. Water chemistry control was cited as a successful example of this kind of targeted intervention.
In closing remarks, the panelists reaffirmed that achieving reactor lifetimes of 80 to 100 years is feasible, primarily because known degradation mechanisms are being actively managed. While perfect prediction is unrealistic, continuous learning from inspections and OE enables proactive aging management. There remains a need to better understand dose-dependent behaviors, particularly nonlinear effects such as swelling, and to determine whether these will emerge within extended lifespans. Despite the uncertainties, panelists expressed cautious optimism, stressing that vigilant monitoring and collaborative research are key to ensuring safe, long-term reactor operation.
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Conclusion:==
The discussion emphasized the need for continued research, monitoring, and adaptability in managing reactor internals. While new degradation mechanisms may not emerge, existing ones could evolve in significance. Collaboration between industry and national laboratories will be crucial in refining predictive models, improving inspection techniques, and ensuring that nuclear plants can safely operate for extended lifetimes.
Key Research Recommendations from the Question and Answer (Q&A) Session The panelists highlighted several critical research areas needed to support the LTO of nuclear reactors, particularly extending the lifespan of reactor internals up to 80-100 years. These research recommendations include:
Understanding Combined Degradation Mechanisms Further research is necessary to understand how various degradation mechanismssuch as stress corrosion cracking (SCC), fatigue, and thermal aginginteract and influence each other over extended operational periods. These mechanisms may not act independently; instead, their combined effects could accelerate material degradation in ways that are currently not well understood.
2-15 In particular, the phenomenon known as "unzipping," where localized degradation progressively propagates and leads to larger structural failures, highlights the need for more advanced and accurate predictive models. Improved modeling capabilities would allow engineers to better anticipate these cascading failures and implement preventive measures to maintain the integrity and safety of nuclear reactor components over the long term.
Investigating Effects of Irradiated Microstructures at Elevated Temperatures Research should thoroughly investigate how exposing irradiated materials to elevated temperatures influences their microstructure and mechanical properties. When materials that have already undergone radiation damage are subjected to thermal aging, complex and potentially unknown degradation mechanisms may emerge. These effects could include accelerated phase transformations, embrittlement, or unexpected changes in strength and ductility, which are not yet fully understood. As such, the nuclear industry must approach the use of heat treatments with caution. While thermal treatments may offer some potential to restore ductility or relieve residual stresses, it should not be assumed that they can fully reset or reverse the effects of irradiation. A clear understanding of the limits and consequences of post-irradiation heat treatment is essential to ensure long-term material performance and reactor safety.
Development of Inservice Inspection Techniques There is a critical need to develop and implement improved nondestructive evaluation (NDE) methods capable of detecting early-stage degradation phenomenasuch as cracking, embrittlement, and other structural failureswell before they reach critical levels. Current inspection techniques may not always provide the resolution or sensitivity required to identify such defects at their inception, particularly in complex and radiation-exposed reactor environments.
Therefore, the feasibility of advanced detection technologies, including nanometer-scale techniques, should be thoroughly explored. These ultra-sensitive methods have the potential to identify subtle microstructural changes and incipient damage, allowing for more proactive maintenance and risk management.
Additionally, new generations of ultrasonic or other advanced imaging technologies should be developed to provide accurate, high-resolution assessments of reactor internals while in service.
Such innovations must be robust enough to function under high-temperature, high-radiation conditions and provide reliable data to inform maintenance and life-extension decisions.
- Advancing Repair and Mitigation Techniques for Aging Components Ongoing research into welding and repair techniques for highly irradiated stainless steels is essential, particularly for applications involving reactor internals, which are exposed to extreme radiation and mechanical stresses over time. These efforts aim to develop reliable methods for restoring structural integrity without compromising safety or violating radiation exposure limits, such as those governed by ALARA principles.
In parallel, it is important to evaluate the technical feasibility and economic viability of replacing critical components, including baffle-former bolts, core barrels, and other reactor internals that may experience significant degradation. Such evaluations should consider not only the
2-16 engineering challenges but also the associated costs, operational downtime, and regulatory requirements.
Ultimately, the nuclear industry must prioritize the identification and implementation of cost-effective mitigation strategies that strike a practical balance between technical feasibility and the financial realities faced by utility operators, ensuring long-term safety and sustainability of nuclear power generation.
- Radiation Dose Accumulation and Long-Term Material Stability Studies should be conducted to quantify how prolonged and increased radiation exposure impacts the mechanical and structural properties of reactor components over time. This includes evaluating changes such as embrittlement, loss of ductility, reduction in fracture toughness, and other alterations that may compromise material performance under operational stresses.
Additionally, researchers must investigate the potential for dose-dependent nonlinear degradation effects, particularly in fast reactor environments where neutron flux is significantly higher. For example, phenomena like void swelling may not progress linearly with dose and could accelerate rapidly once a critical thresholdsuch as a specific displacements per atom (DPA) levelis reached. Identifying these thresholds is essential for developing accurate predictive models, informing inspection intervals, and ensuring the long-term integrity and safety of reactor structures.
Collaboration Between National Labs and Industry for Mechanism Understanding National laboratories should collaborate closely with the nuclear industry to refine the mechanistic understanding of material degradation processes and to develop targeted mitigation strategies. This partnership is crucial for translating advanced research findings into practical engineering solutions that can be implemented in real-world reactor environments.
To support long-term reactor operation, more comprehensive aging management programs should be established that integrate cutting-edge scientific research with day-today operational practices. These programs should not only monitor the condition of materials over time but also proactively incorporate new insights into degradation mechanisms and mitigation techniques.
Furthermore, successful examplessuch as the role of water chemistry control in mitigating stress corrosion cracking (SCC)should serve as models for future efforts. These case studies demonstrate the effectiveness of aligning science, engineering, and operations in addressing complex degradation phenomena and should guide the development of similar interdisciplinary strategies for other forms of aging and material degradation.
- Assessing Feasibility of 100Year Reactor Operation Research should focus on determining whether long-term material degradation can be effectively managed to enable the safe operation of nuclear reactors for up to 100 years. This includes assessing whether current aging management practices and mitigation strategies are sufficient to maintain structural integrity and performance over such extended periods.
Studies should also investigate how degradation mechanisms evolve beyond 60-80 years of service, identifying any new or unexpected failure modes that may emerge with prolonged exposure to radiation, temperature, stress, and environmental factors. Understanding these
2-17 late-stage degradation behaviors is critical for refining predictive models and ensuring the continued reliability of reactor components.
Additionally, the nuclear industry should actively collect and analyze OE from reactors worldwide to enhance aging models. Incorporating real-world data and lessons learned will help improve forecasting accuracy and support the development of robust strategies for long-term license extensions and reactor sustainability.
Conclusion The research recommendations emphasize the need for an integrated approach that combines fundamental material science, advanced inspection techniques, practical repair methods, and strategic collaboration between researchers and industry stakeholders. The overarching goal is to ensure that reactors can operate safely and reliably for extended lifetimes while minimizing unexpected degradation challenges.
2.3.3 Session Summary The session discussions emphasized the importance of proactive aging management, predictive modeling, and advanced mitigation strategies to ensure the safe, LTO of RVIs. Several key research themes emerged:
Understanding Combined Degradation Mechanisms Investigating how SCC, fatigue, irradiation effects, and thermal aging interact over extended operational lifetimes.
Addressing the potential for progressive degradation, such as "unzipping" failures in structural components.
Material Science and Long-Term Irradiation Effects Examining transmutation effects in austenitic stainless steels, including phase instabilities and helium accumulation.
Evaluating how prolonged neutron exposure impacts mechanical properties, particularly in high-dose environments.
Studying radiation-induced phase transformations and ferrite formation, which may contribute to unexpected material embrittlement.
Predictive Modeling and Benchmarking for Aging Assessments Enhancing predictive models by integrating empirical data from material harvesting and operational experience.
Improving documentation and benchmarking practices to account for variations in original manufacturing processes.
Developing more accurate life-extension projections beyond 60-80 years.
Advancements in Inspection and Monitoring Techniques Improving NDE methods to detect early-stage cracking, embrittlement, and other degradation effects.
2-18 Exploring nanometer-scale detection techniques and enhanced ultrasonic imaging for real-time assessment of reactor internals.
Ensuring data integrity and reliability for long-term monitoring.
Mitigation Strategies and Repair Technologies Investigating welding and repair techniques for highly irradiated reactor internals, particularly for baffle-former bolts and core barrels.
Exploring the potential of advanced manufacturing methods, such as Powder Metallurgy and HIP, to create more resilient materials.
Assessing the feasibility of thermal annealing as a method to restore material properties in aging reactor components.
Collaboration and Knowledge Transfer in Aging Management Strengthening partnerships between national laboratories, utilities, and regulatory agencies to refine best practices.
Addressing workforce challenges by developing structured knowledge retention programs to train the next generation of nuclear engineers.
Establishing a comprehensive aging management framework that incorporates scientific research, industry needs, and regulatory compliance.
Feasibility of 80-100 Year Reactor Operation Investigating whether long-term degradation can be sufficiently controlled to extend reactor lifespans up to 100 years.
Continuously reassessing safety margins as materials accumulate higher radiation doses.
Leveraging AI and machine learning to enhance predictive analytics, while recognizing their role in reinforcing existing knowledge rather than uncovering new mechanisms.
Ensuring the long-term safety and reliability of reactor vessel internals requires an integrated approach that combines advanced material science, predictive analytics, improved inspection methods, and proactive mitigation strategies. Collaboration among researchers, industry leaders, and regulatory bodies will be essential in adapting to evolving degradation challenges and supporting future license extensions.
2.4 Session 3: Reactor Coolant Pressure Boundary Components 2.4.1 Presentations 2.4.1.1 Environmental Cracking During Extended LWR Operation (ML24274A180)
Presenter: Peter Andresen Peter Andresen discussed the challenges and complexities of environmental cracking (EC) in the nuclear industry, emphasizing the importance of laboratory data, historical understanding, and the erosion of expertise. His presentation summarized some of the important environmental cracking issues for pressure boundary and related components exposed to high-temperature water, including a perspective on the origin of problems and their persistence. EC is a societal
2-19 concern that began with steam engine failures causing many deaths and injuries in the mid-1800s from SCC and corrosion fatigue (CF). The first American Society of Mechanical Engineers (ASME) code was published ~70 years later in 1914 and remains primarily a fatigue and fracture code, with offsets for the effect of environment, which was the dominant issue in early boiler operation. Andresen asserted that design codes are built on the concept of immunity, which is a major contributor to continued EC problems. Because immunity has proven inadequate, life evaluation curves were developed for CF and SCC in some environments.
There are rarely dependencies on anything but mechanical parameters, even though environmental and material factors often dominate. New problems surface on a roughly geometric time scale. EC is an extreme value phenomenon, and while cracking has occurred in some of the extreme value locations, there are many other somewhat less extreme locations. In Andresens opinion, the complex and multidisciplinary nature of EC makes relevant atomistic modeling many decades off. Attempts at employing artificial intelligence have been ineffectual because of the complexity of interdependencies and very uneven data quality. Thus, sustained expertise and laboratory capability are imperative, and have fallen to a critical point.
There continues to be a tendency not to believe that good laboratory data foretells plant problems.
Andresens presentation highlighted research needs as summarized below:
- 1. Continue research on reactor coolant pressure boundary (RCPB) materials: RCPB materials are not immune to EC, and there is no threshold effect (i.e., some parameter level below which EC does not occur). One area of concern includes the effect of using improved weld metals which might transfer the most susceptible microstructure to the heat-affected zone (HAZ). Also, attention is needed to be given to the effects of plant-relevant surface conditions and dynamic strain.
- 2. Low alloy steel: Research on low alloy steel (LAS) from temper bead welding which can cause an elevated yield strength and EC susceptibility.
- 3. Plant operating changes and upgrades: The effects of plant modifications and upgrades, including power uprates are potential areas of concern.
- 4. Maintaining expertise and capabilities: Continue research to help prevent loss of knowledge, expertise, and laboratory capability. The perception that EC issues have been solved by using immune materials and staying below thresholds may only push degradation further into the future. Thus, expertise in environmental cracking should be maintained to address emergent issues.
2.4.1.2 Gaps to Address for Life Beyond 80: Experience based on PNNLs LWRS SCC Initiation Research on Reactor Coolant Pressure Boundary (RCPB) Structural Materials (ML24282A726)
Presenter: Ziqing Zhai (Pacific Northwest National Laboratory)
Ziqing Zhai discussed the research on SCC initiation in nuclear power plant materials under the U.S. DOE's Light Water Reactor Sustainability (LWRS) Program. The research aims to understand the SCC initiation process, generate high-fidelity data, and evaluate mitigation measures for nuclear power plants, and includes nickel-based Alloy 600, nickel-based Alloy 690, and stainless steels, with ongoing research on Alloy 690 and potential future issues. Alloy 600 remains vulnerable to SCC initiation, with studies showing a consistent initiation process
2-20 involving intergranular oxide formation and crack nucleation. The research performed in the LWRS program found discrepancies in the relationship between temperature and SCC initiation time compared to other literature data, highlighting a gap in the mechanistic understanding.
Contrary to service experience, research found no obvious effect of grain boundary carbides on SCC initiation in Alloy 600. Early findings on Alloy 690 revealed that creep-induced grain boundary cavities can lead to microscopic cracking initiation in low cold work materials and macroscopic cracking in highly cold worked materials. For all levels of cold work, initiation time increased significantly as applied stress decreased. Further research on intergranular stress corrosion cracking (IGSCC) of 304L and 316L(N) stainless steel piping is motivated by SCC in French reactors.
The identified research needs focus on several key areas to address gaps in understanding and mitigating SCC initiation in nuclear power plant materials:
- 1. Mechanistic Understanding: There is a need for a better mechanistic understanding of the SCC initiation process in Alloy 600/82/182 and Alloy 690/52/152. This includes studying the early stages of intergranular oxide growth and the effects of stress, surface finish, and heat-toheat variability.
- 2. Predictive Models: Developing more accurate predictive models for SCC initiation is needed. Older models are mostly empirical and probabilistic, focusing on limited influencing factors. Mechanistic-based models, such as Électricité de France (EDF's) local model for Alloy 600/182, are promising but better data/knowledge is needed to better quantify/account for influencing factors.
- 3. Temperature Effects: There is a gap in understanding the relationship between temperature and SCC initiation time. Laboratory testing often uses higher temperatures to accelerate data generation, but this may not accurately reflect inservice conditions.
There is also some indication that temperature effects do not always follow Arrhenius behavior.
- 4. Grain Boundary Carbides: The role of grain boundary carbides in SCC initiation is somewhat unclear. Service experience suggests they improve crack propagation resistance, but experimental data shows no obvious effect.
- 5. Cold Work and Applied Stress: The effects of cold work and applied stress on SCC initiation time vary between different alloys. Understanding these differences may help in the development of unified models and determining the improvement factors for second-generation nickel-based alloys. This includes how cold work is applied (e.g., forging vs tensile strain).
- 6. Creep-Induced Cavities: Research on Alloy 690 has revealed that creep-induced grain boundary cavities can lead to microscopic cracking initiation in highly cold-worked materials tested at 360°C. Further studies are needed to understand this mechanism in lower cold work materials at lower temperatures.
- 7. Welding Defects: The impact of preexisting welding defects on SCC initiation in high-chrome weld metals needs to be better understood. These defects may act as high-stress regions that induce creep cavities and degrade materials over LTOs.
2-21
- 8. Stainless Steel Piping: This research aims to provide a better understanding of SCC initiation in stainless steel piping, focusing on factors such as thermal stratification, temperature, dissolved oxygen, surface condition, long-term aging, and welding residual stress/strain.
- 9. Mitigation Effects: Research may be needed to evaluate the effect surface treatments, such as peening, have on EC resistance.
2.4.1.3 Long-Term Approach to Manage Stress Corrosion Cracking (ML24274A181)
Presenter: Thierry Couvant (EDF)
Thierry Couvant discussed the challenges and strategies related to managing the aging of nuclear reactor components, with a focus on long-term research and operational solutions.
Some reactor components, such as the reactor pressure vessel, reactor containment, and internal cables, are impossible or difficult to replace, necessitating long-term research programs to address aging and temperature variations. The long-term research objectives include allowing safe operation beyond 60 years and preparing for ten-year indepth inspections. To address inservice degradation, it is important to understand the underlying mechanisms, adjust NDE strategies, and create NDE tools. Proactive funding of research is necessary to improve knowledge, develop tools and methodologies, and create good models that incorporate observed inservice data. Understanding inservice cracking requires knowledge of component fabrication, surface conditions, and operating conditions, and ready access to manufacturing records and onsite measurements are essential for understanding inservice conditions. SCC simulation is a valuable tool for validating assumed mechanisms and making predictions about crack initiation and growth.
The presentation outlined several research needs to address the aging of nuclear reactor components:
- 1. Testing Plant Materials: Research and testing on inservice materials and potential alternatives (e.g., XM19, A725) are necessary. This includes having knowledge of how components were fabricated and operating conditions. Access to plant records would be very beneficial. Evaluating representative mockups may also help understand component service performance.
- 2. Degradation Mechanisms: There is a need for a better understanding of the underlying mechanisms of inservice degradation. This includes adjusting NDE strategies and creating NDE tools to accurately assess cracking kinetics
- 3. Surface Finish and Crack Initiation: Research is needed to understand the effect of surface finish on crack initiation. This might involve testing both polished and realistic surface conditions to ensure accurate predictions.
- 4. Modeling and Simulation: There is a need for simulation and modeling to validate assumed mechanisms and make accurate predictions about crack initiation and growth.
This includes building comprehensive databases and benchmarking different models.
This also includes adjusting NDE strategies and creating NDE tools to accurately assess cracking kinetics.
2.4.2 Panel Discussions Panelists:
2-22 Bogdan Alexandreanu, Argonne National Laboratory Takumi Terachi, Institute of Nuclear Safety System, Inc.
Mychailo Toloczko, Pacific Northwest National Laboratory The panel discussion focused on several key themes related to LTO, particularly beyond 80 years. The discussion began with questions about whether current mitigation and aging management strategies are adequate and if the right mechanisms are being considered. The panelists explored various aspects of reactor coolant pressure boundary locations that might need attention, especially those not currently evaluated for aging.
Toloczko emphasized that many PWSCC mechanisms have already been considered, but there is still concern about the completeness of current understanding. He highlighted the importance of considering unanticipated issues that might arise due to incomplete knowledge.
Alexandreanu pointed out the significance of locations with welds or dissimilar metal welds, as these areas may show hardening and carbide coarsening due to aging. Terachi added that while the current maintenance program is well-documented, uncertainties still exist, particularly with SCC.
The discussion also touched on secondary effects that might become important later in the reactor's life. A question from the audience raised the issue of secondary effects, especially those that are thermally induced, and how they might impact LTO. Alexandreanu shared insights from his research on aged materials, highlighting the importance of understanding the effects of ordering and the behavior of different materials under long-term aging.
The panelists also discussed the challenges of accelerated testing (i.e., employing elevated temperatures, cold work, etc.) and its implications for understanding long-term aging. The panel discussed whether current acceleration techniques might introduce new mechanisms or obscure the ones being investigated. Alexandreanu and Terachi both emphasized the need for fundamental studies before applying data based on acceleration tests, as well as the importance of considering complex phenomena that might arise under accelerated conditions.
Toloczko recognized the challenges of accelerated testing but emphasized that valuable information can be learned by doing accelerated testing and it is important to interpret the results correctly. Andresen stressed the importance of having plant data to help validate laboratory testing, noting there may be differences in the magnitudes of effects such as the effect of dynamic strain, which is often employed in laboratory testing, but may not be occurring in plant components.
The panel discussion highlighted several key areas for future research to address aging-related degradation during long-term nuclear reactor operations beyond 80 years:
- 1. Validating Aging Programs with Plant Materials: There is a need to validate current aging programs using materials from actual plants. This involves comparing aged materials with those in service to ensure the aging programs are accurate and effective.
- 2. Studying Dissimilar Metal Welds and Interfaces: Research should focus on dissimilar metal weld interfaces and heat-affected zones, as these may exhibit hardening and carbide coarsening. Surface effects are also an important aspect to evaluate.
- 3. Understanding Long-Term Aging Mechanisms: Studying long-term aging mechanisms is important. This includes looking at secondary effects that may become important over time, such as thermally induced issues. The panel agreed that thermally
2-23 aging of materials of interest should begin as soon as practicable is wise and relatively inexpensive.
- 4. Acceleration Testing: There is a need to improve accelerated testing methods to avoid altering microstructures too much while still obtaining useful data. This involves finding a balance between accelerating tests and maintaining the integrity of the materials being tested.
- 5. Knowledge Transfer: It is essential to pass down knowledge to the next generation of researchers to ensure continuity in understanding EC and other aging-related issues.
- 6. Fatigue and Corrosion Fatigue: Emerging issues like fatigue and corrosion fatigue need to be addressed as reactors undergo more cycles over extended lifetimes.
2.4.3 Session Summary Session 3 focused on Reactor Coolant Pressure Boundary components and included presentations and discussions on various aspects of stress corrosion cracking (SCC) and environmental degradation of materials used in nuclear reactors. The presentations and discussions underscored the need for ongoing research, accurate modeling, and effective knowledge transfer. The experts stressed the importance of coupling research with practical plant operations and the need for a better understanding of inservice conditions to ensure the long-term safety and reliability of nuclear reactors. Some common themes that emerged included: the importance of high-fidelity laboratory testing data coupled with plant data, understanding inservice conditions, evaluating mitigation measures, developing accurate SCC models, and transferring knowledge to new staff.
2.5 Session 4: Secondary Side Components The objective of session 4 was to focus on existing knowledge gaps stemming from OE and research related to aging issues, such as thermal fatigue, flow-accelerated corrosion, wear, and vibration during LTO and then to also discuss, as appropriate, research ideas to address these gaps. Mekonen Bayssie organized and chaired this session; Yudai Yamamoto and Ryan Wolfe presented. Jared Smith and Frank Gift were panelists.
2.5.1 Presentations 2.5.1.1 Evaluation of Hydrazine Alternatives as Oxygen Scavenger in PWR Secondary Systems (ML24274A182)
Presenter: Yudai Yamamoto, CRIEPI, Japan The presentation outlined the research process, starting with a literature review to identify potential hydrazine (N2H4) alternatives. Promising candidates include Carbohydrazide (CHz, CO(NHNH2)2 and N,N-Diethylhydroxylamine (DEHA, (C2H5)2NOH), with CHz being the most promising despite its own toxicity. Other candidates like Erythorbic Acid (EA, C6H8O6) which is a stereoisomer of ascorbic acid and the naturally occurring Gallic Acid (GA, C6H2(OH)3CO2H) are safer but lack sufficient data.
Yamamotos presentation discussed hydrazine alternatives for water treatment in nuclear power plants. Hydrazine, commonly used in Japan as an oxygen scavenger, is highly harmful (toxic,
2-24 carcinogenic & environmentally toxic) and may be subject to future restrictions due to its carcinogenic properties. The research aims to find safer alternatives before hydrazine becomes unavailable.
The research involves measuring reaction rates, degradation products, and corrosion potential of these alternatives. Carbohydrazide and DEHA show potential, but further studies are needed to ensure their effectiveness and safety. The presentation concluded with a call for continued research to identify viable hydrazine alternatives.
The following research recommendations were made in Yamamotos presentation:
- 1. Prepare for hydrazine restrictions: Due to its high toxicity and potential future regulations, it is crucial to find alternatives to hydrazine for water treatment in nuclear power plants.
- 2. Conduct literature reviews: Identify potential hydrazine alternatives through comprehensive literature reviews.
- 3. Basic studies on alternatives: Perform basic studies to obtain data not available in literature, such as reaction rates, degradation products, and corrosion potential.
- 4. Promising candidates transition of oxygen scavenger: Focus on CHz in near term and DEHA future as the most promising alternatives, despite their own limitations.
- 5. Evaluate safer chemicals: Consider safer chemicals like EA and GA, which are used as food additives and have low toxicity.
- 6. Measure reaction rates: Conduct experiments to measure the reaction rates of hydrazine alternatives under defined conditions.
- 7. Analyze degradation products: Evaluate the degradation products of alternatives to understand their potential impact on corrosion.
- 8. Corrosion potential measurement: Measure the corrosion potential of alternatives using high-temperature, high-pressure water loops.
- 9. Consider Ethanolamine (ETA, HOCHCHNH): Explore the possibility of using Ethanolamine as a stand-alone water chemistry option due to its slight oxygen scavenging ability.
2.5.1.2 EPRI Research on Flow-Accelerated Corrosion and Erosion in Nuclear Power Plants.
Presenter: Ryan Wolfe, EPRI The presentation highlighted the significance of the EPRI Programmatic Guidance for Flow-Accelerated Corrosion Programs, which emerged from the 1986 incident at the Surry Nuclear Power Station. This incident led the NRC to require utilities to provide information on how they would address flow-accelerated corrosion (FAC) issues, resulting in the publication of NSAC202L by the Nuclear Safety Analysis Center. This document outlines six key elements of
2-25 an effective FAC program, including corporate commitment, analysis, OE, inspections, training, and long-term strategy.
Wolfes presentation focused on passive mechanical systems, specifically FAC and erosion.
Wolfe introduced the CHECWORKS Users Group (CHUG), which utilizes EPRI's CHECWORKS software to predict FAC rates, evaluate components, and plan future inspections. CHUG members, located throughout the U.S. and around the world, also engage in research and development projects and provide a forum for utilities to discuss operational experiences and training opportunities.
Wolfe discussed the importance of considering the OE of other plants and conducting inspections to measure remaining wall thickness and project service life. EPRI trains program owners and CHECWORKS software users to manage FAC programs effectively. The presentation also emphasized the need for a long-term strategy to reduce a plant's susceptibility to FAC through material upgrades and changes in chemistry.
The presentation also addressed the challenges of predicting FAC rates in small-bore piping and the need for different guidance when handling such piping. Wolfe stressed the importance of keeping guidance up to date by monitoring operational experiences and incorporating new technologies.
In addition to FAC, Wolfe covered erosion as a distinct degradation mechanism that has gained significance as FAC programs mature. Erosion includes cavitation erosion, liquid droplet impingement, flashing erosion, and solid particle erosion. EPRI has issued separate guidance on erosion and provides computer-based training to help utilities understand the differences between FAC and erosion.
Wolfe concluded by summarizing the importance of monitoring operational experiences, incorporating new technologies, and addressing research gaps to support continued license renewal. The presentation concluded with a Q&A session, during which Wolfe addressed questions regarding material changes and the use of artificial intelligence and machine learning in FAC prediction.
Wolfes presentation identified the following research recommendations:
Investigate erosion-resistant materials.
Develop better methods for selecting components for erosion inspections.
Perform state-ofthe-fleet assessments to identify the best practices and improvements to the guidance.
Incorporate OE and new technology into guidance updates.
Address the continued suitability of the existing exclusion criteria for not requiring monitoring of wall thickness in systems operating less than two percent of the time.
Expand data collection and collaboration with different plants and countries for better predictions.
2-26 Improve data curation before feeding it to machine-learning models.
2.5.2 Panel Discussions Panelists:
Jared Smith, Principal Technical Leader, the Electric Power Research Institute (EPRI).
Frank Gift, Senior Principal Technical Leader at the Electric Power Research Institute (EPRI).
A panel discussion featuring Jared Smith and Frank Gift examined various aspects of SCC and materials aging in nuclear reactors. Jared Smith, a data scientist with a Ph.D. in corrosion science, and Frank Gift, a senior principal technical leader at the EPRI, shared their insights on the importance of understanding and managing SCC in steam generator tubes, the impact of water chemistry, and the need for proactive asset management. They emphasized the significance of empirical data, inspection intervals, and continuing research to ensure the long-term reliability of nuclear components. The panelists addressed several questions, including whether new or previously unimportant stress corrosion mechanisms will gain significance in the future, what new research is necessary to support the separation of nuclear components on the secondary side for LTO, and whether flow-accelerated corrosion will remain manageable beyond 80 years with current strategies or if new approaches will be required. They were also asked if major piping or component replacements will be necessary for secondary side components to continue operating the current light-water reactor (LWR) fleet beyond 80 years and if further improvements in water chemistry control methods are needed to maintain the viability of secondary side components over this extended period.
Smith and Gift provided comprehensive responses to audience questions, highlighting advancements in secondary side chemistry control, stress management, and modern alloys that have significantly reduced the urgency of SCC as a primary degradation mode for steam generator tubes. Additionally, they discussed the need to understand the effects of low levels of impurities, the implication of multiple chemical cleanings, and the synergistic effects of different degradation mechanisms.
The panelists agreed on the necessity of ongoing research to support the LTO of nuclear components, particularly in the context of extended operational periods beyond 80 years. They stressed the importance of proactive asset management, understanding the flow, temperature, and steam quality in the steam generator's upper internals, and avoiding complacency in chemistry controls to ensure the longevity of secondary side components. The panelists identified the following research recommendations:
Proactively identify the bounds of susceptibility for SCC of modern steam generator tube alloys.
Investigate the effects of short-term excursions in secondary side water chemistry on long-term cracking.
Study the impact of long-term exposure to low levels of impurities on materials.
Assess the effects of multiple chemicals cleans on boiler materials over extended periods.
2-27 Examine the potential synergistic effects of different degradation mechanisms.
Explore chemistry solutions and sleeving technologies for mitigation and repair.
Improve detection and sizing capabilities for corrosion-resistant alloys.
Continue evolving and updating existing programs for managing FAC and erosion.
Investigate the impact of foreign material in steam generators and the formation and spallation mechanisms.
Monitor and adapt water chemistry control methods to ensure the longevity of secondary side components.
2.5.3 Session Summary Yudai Yamamoto's presentation emphasized the urgent need to prepare for potential hydrazine restrictions in nuclear power plant water treatment due to its toxicity. To address this, he recommended conducting thorough literature reviews to identify promising alternatives. For those alternatives lacking sufficient data, basic studies should be performed to determine crucial parameters like reaction rates, degradation products, and corrosion potential. While CHz and DEHA are currently considered promising, their limitations necessitate the evaluation of safer chemicals such as EA and GA, given their low toxicity and use as food additives. The research should involve measuring reaction rates of these alternatives under defined conditions, analyzing their degradation products to assess corrosion impact, and directly measuring their corrosion potential using high-temperature, high-pressure water loops. Finally, Yamamoto suggested exploring Ethanolamine (ETA) as a potential stand-alone water chemistry option due to its inherent oxygen scavenging properties.
Ryan Wolfe's presentation centered on enhancing the management of erosion in nuclear power plants. He advocated for the investigation of more erosion-resistant materials and the development of improved methodologies for selecting components that require erosion inspections. To further optimize current practices, Wolfe recommended performing state-ofthe-fleet assessments to identify the best practices and areas for guidance improvement. He also stressed the importance of incorporating both OE and new technological advancements into future guidance updates. A key issue to address is the two percent exclusion criteria related to continuing license renewal. To improve predictive capabilities, Wolfe proposed expanding data collection efforts and fostering greater collaboration among different plants and even international partners. Finally, he highlighted the necessity of enhancing data curation processes before utilizing the information in machine-learning models.
2.6 Session 5: Balance of Plant Systems The objective of session 5 was to focus on existing knowledge gaps stemming from OE and research related to aging issues, such as MIC, leaching, coating degradation, corrosion, new repair techniques, during LTO and then to also discuss, as appropriate, research ideas to address these gaps. While the scope of this section extended to all balance of plant systems, there was a focus on service water systems and buried tanks and piping due to their potential safety significance and OE with degradation in these systems. Rob Tregoning chaired this session which had been organized by Amy Hull; George Licina and John Cavallo presented.
Dylan Cimock, Chris Burton, and John Wise were panelists.
2-28 2.6.1 Presentations 2.6.1.1 Service Water Systems at Nuclear Plants (ML24274A183)
George Licina, retired, previously at Structural Integrity Associates The presentation addressed service water systems at nuclear power plants, focusing on their design and operation, degradation history, repair techniques, and long-term management strategies. Service water systems are large and complex, with varied water chemistries, materials, and operational conditions. Systems are often over-designed for accident scenarios which often lead to stagnant conditions. Such conditions promote corrosion that affects piping, heat exchangers and other components. Types of corrosion include general corrosion, localized corrosion, selective leaching, galvanic corrosion, microbially-induced corrosion (MIC), and erosion-corrosion. Cumulative corrosion damage, if unmitigated, can result in increased rates of leakage with continued operation.
Inspection and monitoring are important to identify and manage damage early. Effective inspection strategies employ initial screening inspections using, for example long range ultrasonics, followed by detailed inspections of suspect locations. Phased-array ultrasound is an advanced technique that can map thickness measurements with great fidelity. Real-time monitoring has the potential to replace inspections and can help avoid unplanned downtime, optimize outage time, and optimize mitigation treatments.
Mitigation and repair recommendations include cleaning, water treatments, cathodic protection, localized repairs, and upgrading materials to more corrosion-resistant alloys or non-metallic options to improve system longevity. Coating is also used to minimize galvanic coupling in heat exchangers. Licina concludes that service water systems cannot be entirely free of problems but can be effectively managed through better understanding, inspection, and proactive mitigation strategies to ensure their extended operational lifespan.
The presentation highlighted the following research needs to better manage the LTO of service water systems in nuclear power plants and ensure their structural integrity and efficient operation over extended lifespans.
Improved Predictive Models: Better methods to predict when, where, and how to inspect systems should be developed and enhance mapping and characterization of corrosion data should be pursued for more accurate forecasting Advanced Inspection and Monitoring Techniques: Degradation thresholds should be defined for interpreting inspection results. Flaw sizing techniques should be improved to allow for more precise damage evaluation. Real-time monitoring should be incorporated to complement traditional inspections Effective Mitigation Strategies: New materials and water treatments should be considered and evaluated for enhanced corrosion resistance. Localized repair techniques should be improved to address specific issues. The effectiveness of cathodic protection, coatings, and chemical treatments should be evaluated.
Evaluation of Repairs and Treatments: Methods to assess the long-term effectiveness of repairs and treatments should be developed. Solutions should be optimized to balance cost and performance.
2-29 2.6.1.2 Technical Advances in Cured-inPlace-Pipe (CIPP) Linings (ML24274A184)
Jon Cavallo, PE, FASTM The presentation discussed advancements in CIPP technology, its origins, and current applications. This trenchless method involves creating a "pipe within a pipe" to rehabilitate aging pipelines using materials like polyester, epoxy, or polyurea-polyisocyanate, without extensive excavation. These materials cure to form a durable, stand-alone or adhered lining, with lifespan estimates of 50-60 years. Limitations include inability to coat fittings or severely degraded pipes, and variability in cost depending on the extent of surface preparation required.
Spray-inplace-pipe (SIPP) technology is an emerging alternative that is currently in development. SIPP promises higher structural strength, easier surface preparation, and faster curing times compared to CIPP. The presentation concluded with a Q&A session, addressing inspection methods, degradation mechanisms, and temperature limitations of CIPP materials.
Cavallo emphasized tools like remote crawlers and ultrasonic testing for inspections and discussed external corrosion as a key concern for pipeline longevity.
The presentation highlighted the following research needs which aim to improve pipeline rehabilitation technologies and ensure greater structural strength, durability, and effective inspections:
Inspection Techniques for CIPP: Development of advanced methods to inspect CIPP and ensure long-term reliability. Current techniques like remote crawlers and ultrasonic testing are available, but further innovations are needed, particularly for inspecting gaps, fittings, and non-adhered linings.
Adhesion and Ultrasonic Inspection: Research into ultrasonic methods capable of assessing adhesion between the host pipe and the new lining is emphasized. This includes identifying the right frequencies to penetrate polymer thicknesses and enhance the credibility of inspections.
Degradation Mechanisms: Understanding external corrosion of host pipes and other degradation mechanisms that could affect the lifespan and performance of CIPP systems. Investigating soil conductivity and chemistry's impact on pipeline corrosion are essential considerations.
Standardization of SIPP: The emerging SIPP technology requires industrywide standards to qualify coatings and simplify adoption. Utilities are facing challenges with proprietary products, highlighting the need for uniform testing and qualification criteria.
2.6.2 Panel Discussions Panelists:
Chris Burton, Structural Technologies Dylan Cimock, Electric Power Research Institute John Wise, U.S. Nuclear Regulatory Commission The panel discussion addressed the management, inspection, and maintenance of buried and underground piping systems in nuclear facilities to support LTO. The emphasis was on buried piping and tank systems' vulnerabilities and addressing these aging infrastructure challenges by developing and implementing advanced strategic inspection and repair methodologies which
2-30 account for the safety significance of systems, structures, and components (SSCs). Key discussion topics included applicable degradation mechanisms, the role of non-metallic repair materials, and the importance of data sharing and standardization.
Critical systems consist of safety-related piping and tanks, which can include service water, fire protection, and circulating water systems. These systems employ a wide variety of materials such as carbon steel, reinforced concrete, prestressed concrete, and polymers. The systems can also be subjected to disparate environmental stressors. The numerous material and environment combinations can lead to various types of degradation mechanisms such as selective leaching, wall thinning, corrosion, sediment impact, and settlement.
These systems can be challenging to inspect as it is difficult to identify localized damage in buried infrastructure. There is subsequently a need for advanced NDE tools and improved data collection and sharing across the industry to support predictive maintenance and risk-based decision-making. Repair is also difficult due to access limitations. Economic and safety considerations should be weighed when considering whether to excavate and replace the damaged SSC or repair it. In situ repair technologies, such as polymer-based linings and carbon-fiber reinforcements should continue to be developed, with considerations for long-term performance and quality assurance/quality control (QA/QC) practices.
In particular, Burton highlighted circulating water systems as a critical issue due to their size, material variety, and potential for catastrophic failure (e.g., creating a "lake" at a station due to rupture). He emphasized the need for different inspection and repair technologies tailored to the diverse materials used in these systems. He also stressed the importance of non-safety-related, but risk significant, systems like circulating water and fire protection which are often overlooked despite their importance. He advocated for in situ repairs using technologies like carbon fiber-reinforced polymers and polymer-based linings.
Cimock agreed with Burton on the under-prioritization of non-safety-related systems like large-volume circulating water systems. He highlighted the challenge of detecting leaks and damage in deeply buried pipes. He also stressed that the variability in raw water systems complicates corrosion prediction. He encouraged optimizing inspection strategies, including using advanced NDE techniques for better data collection (e.g., high-resolution ultrasonic imaging) and called for a shift in inspection focus to longer pipe segments rather than localized regions to better assess system health. He also advocated improving data sharing within the industry to inform better risk-based inspection frameworks.
Wise questioned whether the nuclear industry collects and shares sufficient data on system conditions across facilities to make informed, industrywide decisions. He also highlighted the possibility for settlement impacts on buried pipes, particularly with materials that may experience microstructural changes over time (e.g., graphite formation in cast iron). He recommended expanding the scope of inspections to address emerging aging effects and degradation mechanisms and specifically highlighted the importance of developing far-field NDE techniques to simplify extent-ofcondition evaluations. He also emphasized the need for standardization in repair methodologies, particularly for new polymer-based materials.
The panelists identified the following research topics which emphasize advancing inspection tools, improving repair materials, and enhancing data sharing and standardization to address challenges to ensure effective aging management and maintenance of buried and underground piping systems in nuclear facilities.
2-31 Inspection Technologies: Advanced NDE techniques for large-diameter pipes and buried infrastructure should be developed to enable efficient, far-field, and aboveground inspections. Methods should be improved to inspect longer pipe segments instead of relying solely on localized inspections, to better assess system health comprehensively.
Research on high-resolution ultrasonic imaging and encoded data collection should be conducted to ensure accurate corrosion morphology evaluation.
Repair Methodologies: Novel in situ repair technologies that align with updated safety, economic, and environmental requirements should continue to be developed and standardized. Polymer-based and carbon-fiber-reinforced solutions (e.g., CIPP and SIPP), for large-diameter, deeply buried pipes that are challenging to excavate and replace are promising. However, the long-term mechanical and chemical stability (e.g.,
creep, water absorption) of these materials under operating conditions needs to be understood.
Material Performance: The integrity of existing SSCs should continue to be researched. The effects of transient pressures, load stresses, and environmental changes is an important topic. Research on the potential impacts of ground settlement on buried pipes is also needed to evaluate the effects of ovality (i.e., pipe deformation due to soil compaction and heavy equipment loads) and its impact on structural integrity over time. Methods to monitor and mitigate such effects in LTO scenarios should also be considered. Finally, Investigation of selective leaching should address fracture initiation and microstructural changes (e.g., graphite formation in cast iron).
Data and Predictive Maintenance: Data collection and sharing should be enhanced within the nuclear industry to support predictive maintenance strategies and informed decision-making. Revisiting and potentially reinstituting shared databases (e.g., the Buried Pipe Inspection Results Database) to analyze inspection results industrywide could help achieve this objective. Better data sharing would also support development of risk-informed inspection and maintenance frameworks to refine inspection prioritization and reduce unnecessary costs.
Standardization and QA/QC Practices: Standardized testing and QA/QC practices should be created for onsite repairs, particularly for polymer-based systems. Clear, standardized guidelines should be developed for defect characterization and acceptable defect sizes in inspection and repair processes. Finally, protocols should be established to address variability in installation methods (e.g., curing conditions) for non-metallic materials.
2.6.3 Session Summary The session provided key insights into service water systems, pipeline rehabilitation technologies, and buried piping aging management to support LTO at nuclear plants. Service water systems are large and complex and face multiple degradation challenges due to varied materials, service environments, and normally stagnant or low water flow. Corrosion mechanisms like MIC and galvanic corrosion can significantly degrade these systems and inspection is challenging to ensure appropriate coverage given the size and limited accessibility of much of the system.
There is a pressing need for implementing advanced NDE tools, such as phased-array ultrasonic imaging, to more accurately assess long pipe segments. Real-time monitoring is
2-32 another important tool for catching severe degradation early enough to avoid more costly repairs and reducing the more resource-intensive inspections during outages. Mitigation techniques include water treatments, localized repairs, and upgrading to more corrosion-resistant materials. Non-metallic materials are promising in this regard including CIPP, which forms a durable "pipe within a pipe" for aging pipelines. Emerging SIPP offers improved structural strength and faster curing times but requires standardization for broader adoption.
Research priorities include improving predictive models, developing real-time monitoring technologies, and continuing to improve ultrasonic inspection methods. Data sharing and standardization across the industry are needed to support risk-based inspection strategies and improve predictive maintenance. Research is required to continue to develop and standardize insitu repair technologies and ensure the long-term stability of polymer-based materials. A deeper understanding of degradation mechanisms like external corrosion are vital for assessing the remaining of the native metallic piping systems and also to improve the durability and reliability of non-metallic repair systems.
2.7 Session 6: Fatigue 2.7.1 Presentations 2.7.1.1 EPRI MRP Perspectives on Fatigue-Related Issues for 80 Years and Beyond (ML24274A185)
Presenter: Tom Damiani, Electric Power Research Institute The presentation focused on fatigue-related issues and projects related to LWRs for LTO.
Damiani initially discussed EPRIs MRP research in fatigue, including topics on Environmental Assisted Fatigue (EAF), vibration fatigue, thermal fatigue, and flaw tolerance approaches.
Damiani highlighted the distinction between OE driven vibration and high-cycle fatigue and regulatory-driven EAF and seismic fatigue. Computational fluid dynamics (CFD) is being used to model thermal fatigue in conjunction with flaw tolerance approaches to make more accurate lifetime predictions.
Component testing is underway to better link laboratory EAF results with component-level results. Refining the cumulative usage factor (CUF) including environmental correction factor (CUFen) calculations for components and using flaw tolerance predictions are two approaches that can support operation beyond 80 years. Efforts are underway to address seismic fatigue by using existing research and knowledge to revise conservativisms in ASME CC N900 Level D service requirements. The presentation emphasized the importance of continued fatigue evaluation and research for extended reactor operations beyond 80 years, including fatigue in additive manufactured materials and different water chemistries.
The following research ideas were identified to support continued long-term plant operation.
Develop Total Life Fatigue Methods: This approach accounts for all stages of crack initiation and growth in a probabilistic manner and may help bridge the gap between laboratory data and component life assessment.
Evaluate EAF in Different Water Chemistries: Study the impact of various water chemistries, particularly KOH, on EAF to improve fatigue assessments for LTO. The impact of the potential switch from LiOH to KOH on EAF in LWRs has not been
2-33 considered, and KOH should be investigated for potential use in small modular reactor (SMR) designs.
Understand Irradiation Effects on Fatigue: This issue is particularly important for reactor pressure vessel and internals components. The limited available data are inconclusive with regard to the impact of irradiation on the fatigue lives of materials exposed to LWR environments. Material harvesting provides opportunities to evaluate this effect.
Evaluate Fatigue of Additive Manufactured Materials: Understand the fatigue characteristics of these materials and develop appropriate codes and standards and assessment methods for their use in new designs and replacement parts. The equivalent structural strain range approach may provide a path forward.
Impact of Noble Metal Catalysts on Fatigue: Examine the effect of noble metal catalysts on the fatigue life of materials, especially under BWR conditions with optimized dissolved oxygen content. Boiling-water reactors (BWRs) have trouble maintaining or justifying dissolved oxygen (DO) content to optimize CUFen for carbon/low alloy steels (<
40 ppb) because these requirements can conflict with DO requirements for flow-accelerated corrosion (> 50 ppb).
2.7.1.2 Fatigue Evaluation for AM Materials ( ML24274A186)
Presenter: Seiji Asada, Mitsubishi Heavy Industries, Ltd.
This presentation discussed the fatigue evaluation of additive manufacturing (AM) materials for nuclear components. The motivation is that supply chain constraints may favor AM components to reduce lead time and avoid plant outages. Asada provided background on two prevalent AM technologies: Powder Bed Fusion (PBF) and Direct Energy Deposition (DED), highlighting their advantages and disadvantages. The ASME B&PV Code Committee and the BPV III Working Group are developing codes for AM materials in pressure-retaining equipment, and it is noted that while tensile properties of AM materials will be required to meet existing standards for conventional materials, fatigue evaluation requires further study. There is current consideration of applying the Master Curve Method but its applicability to Class 1 nuclear components is unknown, especially since it was developed for air environments.
Asada explained the basis of the Master S-N Curve methodology and EAF evaluation approach and suggests that these traditional methods may not be compatible with each other or applicable to AM materials. He proposed a new fatigue evaluation approach using the best-fit curve based on tensile strength and the Murakami root area method which potentially reduces the need for extensive fatigue testing. The proposal includes a three-step process for applying this method to AM components. The first step would develop fatigue curves for polished and asbuilt specimens. The second step would use the materials hardness and the near-surface defect area to determine the fatigue endurance limit and fatigue crack growth threshold. The third step is to apply the Fen method to account for environmental effects. There is limited data demonstrating the potential of the Master S-N approach but limited data on EAF effects of AM materials.
Asada highlighted the importance of further research and fatigue testing to develop reliable evaluation methods for AM materials, especially for Class 1 nuclear components to better investigate this approach. He concluded that AM materials hold promise for supporting LTO of LWRs, but more investigation is needed to ensure their reliability in fatigue-prone environments.
2-34 The following research ideas were identified to justify employing these materials to support continued long-term plant operation.
Understand Fatigue Phenomena in AM Materials: Investigate the more prominent phenomena expected to strongly affect component fatigue lives. These phenomena include effects of surface finish on fatigue endurance limits for AM materials, material orientation effects, the impact of different manufacturing processes, and the effect of specimen type (i.e., hollow vs. solid specimens).
Develop Fatigue Prediction Methods for AM Materials: There is a need to further develop these methods for AMmanufactured Class 1 nuclear components. This evaluation should assess the applicability of the Master S-N Curve Method, especially Murakami's root area approach. The applicability of EAF approaches for AM materials should also be assessed.
2.7.1.3 Uncomfortable with Uncertainty? (ML24274A187)
Presenter: Andy Morley, Rolls-Royce Submarines, Ltd.
The presentation by Andy Morley explored the emphasis on fatigue in engineering, particularly thermal and low-cycle fatigue, and challenges the conventional approach of placing undue importance on fatigue due to its quantifiable nature. He critiqued the deterministic design paradigm, which often relies on conservative assumptions about material resistance and loadings. He emphasized that fatigue failures often result from unaccounted-for loadings, such as branch line thermal stratification and mixing-tee turbulence. Morley next presented data from Ringhals 3 and 4 in Sweden that highlighted the significant variability and uncertainty in thermal loads, even in these nearly identical systems, due to small design and operational differences.
This experience underscores the importance of direct measurement and monitoring to replace assumptions with reality.
Morley advocated for a risk-informed approach to fatigue management, integrating direct measurements and analytics to enhance component reliability and safety. He also highlighted the importance of understanding both material resistance and loading assumptions to prevent unexpected failures and cautions against compromising reliability by eroding conservatism in fatigue assessments without a holistic understanding of all uncertainties. Such a holistic, risk-informed approach supports improved reliability and operational flexibility during operation beyond plants initial design life and helps managing uncertainties associated with flexible plant operations.
The following research ideas were identified to support continued long-term plant operation.
Improve Sensors and Data Gathering: There's a need for improved sensors to monitor equipment health and collect real-time data on loading and material resistance.
Enhanced data logging and sharing systems are also crucial for accurate fatigue analysis.
Tailor Data Analytics Tools: Developing advanced analytics to process large datasets and replace initial design assumptions with real-world data. This involves handling big datasets and performing detailed analytics to understand equipment behavior under different conditions.
2-35 Adapt Existing Equipment Health Monitoring Methods: Adapting methods from the aerospace industry, such as extending the time between maintenance intervals based on detailed monitoring and analysis of equipment health. This can help increase availability and reduce downtime.
Improve Modeling and Simulation Tools: Improving CFD and other modeling tools to better predict and understand the impact of small design variations on fatigue behavior.
This includes validating models against real-world data to ensure their accuracy.
Understand Effects of Flexible Operation: Researching the effects of flexible plant operation on fatigue and understanding the implications of changing power levels and other operational parameters on equipment longevity and safety.
Develop Risk-Informed Approach: Integrating direct measurements and analytics into a risk-informed approach to fatigue management. This approach helps in making more accurate assessments and managing uncertainties in flexible plant operations. Methods to integrate uncertainties into fatigue assessments holistically, rather than relying solely on conservative design assumptions also need to be incorporated within these methods.
2.7.2 Panel Discussions Panelists:
Mark Gray, Westinghouse Electric Company Tim Gilman, Structural Integrity Associates Jonathan Mann, Amentum Holdings, Inc.
The panel session focused on various aspects of fatigue evaluation in nuclear reactors and the challenges associated translating lab results to component applications to support LTO. There was consensus on the need for both design validation and plant monitoring to ensure safety and reliability as plants age. The key topics discussed included loading conditions, environmental fatigue, instrumentation and monitoring and unanticipated failures. The importance of understanding the accurate loading conditions in both design and operational phases was stressed. The transferability of laboratory-observed EAF phenomena of environmental fatigue for use in plant component life predictions has been a longstanding challenge and needs to be better understood to reduce potential computational conservatisms. Better instrumentation and monitoring in the plants would help address these issues but there are both logistical and regulatory challenges with implementing better instrumentation in plants and there is a need to practically balance the potential benefits with the drawbacks. The objectives of this improved knowledge and monitoring would be to reduce the frequency of unanticipated failures stemming from unforeseen issues, such as check valve malfunctions, which may significantly contribute to fatigue damage.
These conclusions were mutually supported by individual panelists opinions. Gray discussed the importance of understanding the actual load conditions during both the design and operational phases of nuclear power plants. He emphasized that these load conditions need to be managed as plants move past the 80-year mark and that robust, field-collected data be used inform both current evaluations and future design criteria. Gillman indicated that while original equipment designs anticipated certain loadings, unanticipated issues like safety injection lines leaking due to check valve malfunctions have occurred. He suggested that better instrumentation, despite its challenges, might help monitor these situations. Like Gray, Gilman
2-36 stressed the need to validate design loads using operating data. Mann led a fulsome debate on the influence of environmental factors, questioned how well the observed phenomena in controlled tests can be translated to real-world situations, and stressed the need for comprehensive evaluation.
The discussion among the panelists highlighted a common theme: the critical importance of aligning theoretical models with practical, real-world observations. Their insights suggest that the future of fatigue evaluation in nuclear reactors will involve a blend of advanced instrumentation, dynamic data analysis, and a reconsideration of traditional design assumptions based on operational experience. As reactors continue to age, these strategies will become more crucial to ensuring safe operations.
The following future research topics were identified during the panel discussion.
Field Data Collection and Validation: Research should focus on developing methods to gather dynamic, field-based data on loading conditions and validating and calibrating theoretical models with insitu measurements. This research would support safer operations by integrating field data into iterative design improvements and enabling predictive maintenance and proactive interventions.
Enhanced Instrumentation and Monitoring: Research in this area should explore novel sensors and monitoring techniques that can detect subtle changes in structural behavior and overcome logistical and regulatory challenges in retrofitting or integrating these technologies into existing plants. Objectives should include developing systems that can flag unanticipated mechanical anomalies and creating integrated diagnostic tools that couple sensor outputs with predictive analytics.
Environmental Fatigue - Bridging Laboratory and Field Observations: Research should aim to conduct comparative studies between lab-based environmental fatigue tests and real-world behavior to identify and quantify the factors that might affect performance. Since environmental factors (like temperature, radiation, and chemical exposure) can modulate fatigue behavior differently in the field, further research is needed to develop comprehensive models that incorporate these complex interactions and tailor fatigue assessments to reflect the unique conditions present in aging reactors.
Addressing Unanticipated Failures and System Variabilities: Research should focus on examining failure cases that were not anticipated during the design phase and developing diagnostic criteria and predictive frameworks to account for such unforeseen events.
Integration of Novel Materials and Techniques: With emerging materials (e.g., AM) entering the reactor environment, there is a concurrent research need to evaluate how these new materials behave under long-term cyclic loading and environmental exposure and revise existing fatigue models to accommodate the properties of innovative materials.
2.7.3 Session Summary The session provided perspectives on fatigue-related issues in nuclear reactors that are expected to be important for supporting LTO beyond 80 years. Tom Damiani discussed EPRIs ongoing and planned fatigue research, operating-experience and regulatory-based fatigue
2-37 challenges, and proposes research on developing total life fatigue methods, evaluating EAF in different water chemistries, understanding irradiation effects, and evaluating the fatigue of additive manufactured materials. Seiji Asada from Mitsubishi Heavy Industries addressed the fatigue evaluation of AM materials for nuclear components. He proposed studying the applicability of the Master S-N curve method based on tensile strength and the Murakami root area method in conjunction with the Fen method to account for environmental effects. Andy Morley from Rolls-Royce Submarines advocated for a risk-informed approach to fatigue management, integrating direct measurements and analytics to enhance component reliability and safety. He emphasized the need for improved sensors, data gathering, advanced analytics, and understanding the effects of flexible operation.
The panel discussion, featuring Mark Gray from Westinghouse Electric Company, Tim Gilman from Structural Integrity Associates, and Jonathan Mann Amentum Holdings, elaborated on the need to align theoretical models with practical, real-world observations. Research identified to support this goal includes collecting field data collection and using it to validate models; utilizing enhanced sensor and monitoring techniques; conducting comparative studies between laboratory-based environmental fatigue tests and real-world component behavior; addressing unanticipated failures and system variabilities; and integrating novel materials into the reactor fleet. The session promoted the need for continued research and development to better understand and predict fatigue degradation to ensure safe and reliable operation.
2.8 Session 7: Mitigation 2.8.1 Presentations 2.8.1.1 Mitigation of Material Degradation Associated with Long Term Operation (ML24274A188)
Darren Barborak, EPRI Darren Barboraks presentation outlined various initiatives and research efforts aimed at supporting the nuclear power generation industry by developing and optimizing repair, fabrication, and general requirements for both current and future nuclear fleets. Barborak emphasized the importance of advanced materials and joining repair technologies to ensure safe and effective repair or replacement fabrication technologies. The presentation covers four main initiatives: repair solutions for irradiated materials, spent fuel pool repairs, small-bore piping issues, and thermal aging of alloy 52/152.
For irradiated materials, the focus is on addressing the challenges posed by helium accumulation in grain boundaries during welded repairs, which can lead to cracking. The Welding and Repair Technology Center (WRTC) is working closely with Oak Ridge National Lab to collect weldability data, quantify helium concentrations, and develop field-deployable repair systems. In the area of spent fuel pool repairs, the presentation highlighted the need for operators to have options for addressing leaks, including non-metallic repairs that can be faster and more cost-effective than traditional welded repairs. The research includes reviewing detection and localization methods for fuel pool leaks and evaluating the advantages and limitations of non-metallic repair options.
Barborak also addressed small-bore piping issues, which are problematic due to stress concentration and fatigue. The WRTC is conducting high-cycle fatigue testing and analyzing structural support retrofitting to extend the life of small-bore piping configurations. Finally, the presentation discussed the thermal aging of alloy 52/152, which degrades over time due to
2-38 extended exposure to low temperatures. The research involves performing heat treatments and microscopy to understand the effects of thermal aging on the weld filler metal.
Overall, Barboraks presentation underscored the critical role of advanced materials, joining repair technologies, and peer engagement in ensuring the long-term safety and effectiveness of nuclear power plant operations.
Research Recommendations:
Development of New Repair MaterialsInvestigate alternative materials and coatings that minimize helium-induced cracking in irradiated components.
Enhancement of Non-Metallic Repair TechniquesFurther evaluate the long-term stability, chemical resistance, and mechanical properties of non-metallic repair solutions for spent fuel pools.
Advanced Monitoring for Small-Bore PipingImplement real-time monitoring systems using AI and sensors to predict and prevent fatigue-related failures.
Extended Testing of Thermal Aging EffectsConduct prolonged thermal aging studies under varying operational conditions to better understand degradation mechanisms.
Automation and Robotics for RepairsExplore robotic-assisted welding and repair technologies to improve precision and reduce radiation exposure for maintenance workers.
2.8.1.2 Advanced Laser Welding Techniques for Repair of Irradiated Light Water Reactor Components (ML24274A189)
Jian Chen, ORNL Jian Chens presentation discussed the challenges and advancements in welding techniques for light-water reactor components. Chen's presentation focused on the issues related to helium-induced cracking in reactor components, which becomes more pronounced with increased reactor operation time. The presentation explained that helium concentration in reactor cores increases over time, leading to weakened grain boundaries and tensile stress during welding, which can result in cracks. To address this, ORNL and the EPRI have developed advanced welding techniques, including the use of dual laser beams and friction welding, to mitigate these issues. The dual laser beam technique involves using one beam to melt the material and another to create a heated pattern that reduces tensile stress. Experimental validation has shown that this method significantly reduces tensile stress and the occurrence of helium-induced cracking. The presentation also detailed the materials used in the experiments, including various types of stainless steel and Alloy 182, and the results of these experiments, which indicate that the advanced welding techniques are effective in reducing helium-induced cracking. The research is ongoing, with efforts to refine welding parameters and further validate the techniques.
Research Recommendations:
Optimization of Dual Laser Beam ParametersFurther studies should focus on refining laser intensity, beam spacing, and heating duration to maximize effectiveness.
2-39 Long-Term Performance EvaluationConduct extended testing to assess the durability and reliability of the welds under operational conditions.
Expansion to Other AlloysInvestigate the effectiveness of these welding techniques on a broader range of reactor materials to ensure wider applicability.
Integration with Automated Welding SystemsExplore automation possibilities for consistent and large-scale implementation in nuclear plants.
Comparative Analysis with Traditional Welding MethodsConduct studies to compare cost, efficiency, and defect rates between advanced and conventional welding techniques.
2.8.2 Panel Discussion Session 7 Panelists Taku Arai, Central Research Institute of Electric Power Industry (CRIEPI)
Carol Moyer, Nuclear Regulatory Commission Darren Barborak, EPRI Panelists identified key challenges for long-term nuclear reactor operation mitigation, primarily focusing on Stress Corrosion Cracking (SCC) and the complexities of managing irradiated vessels and core internals. Taku Arai emphasized advanced water chemistry and SCC-resistant materials, while Carol Moyer highlighted water chemistry modifications for dissolved oxygen control and flux reduction, alongside the difficulties of remote inspection and repair. Barborak echoed the technical hurdles of remote operations in extreme environments.
The discussion covered mitigation techniques like water chemistry control, induction heating stress improvement (IHSI), and mechanical stress improvement process (MSIP), emphasizing core internal replacement. Regulatory differences between Japan (proactive replacement) and the U.S. were noted. The Q&A addressed repair feasibility, new radiation-resistant materials, and the effectiveness of repair welding. Jian Chen discussed research collaboration with EPRI on repair technologies for irradiated materials and the need for further testing and new, more resistant materials. Concerns about weld material behavior under irradiation were raised. The panel underscored the necessity of continuous evaluation, improvement of repair technologies, and industry-regulatory collaboration for the long-term safety and efficiency of nuclear reactors.
Key research recommendations included:
Advanced SCC Mitigation StrategiesInvestigate the long-term effectiveness of SCC-resistant alloys and optimize water chemistry to minimize degradation.
Improved Remote Inspection and Repair TechnologiesDevelop AIassisted robotic systems for precise, efficient repairs in highly irradiated, confined spaces.
Radiation-Resistant Materials DevelopmentDesign and test new materials with superior radiation tolerance for weld repairs and reactor internals.
2-40 Comparative Study of Regulatory ApproachesAssess the impact of proactive versus reactive maintenance strategies on long-term reactor safety and operational efficiency.
Enhanced Repair Welding TechniquesConduct further testing on weld materials and techniques to improve their performance in high-radiation environments.
Long-Term Performance MonitoringImplement real-time monitoring systems for early detection of material degradation and optimize predictive maintenance strategies.
2.8.3 Session Summary The key research recommendations from the presentations and panel discussion in Session 7 focused on addressing the challenges of material degradation in nuclear reactors. Darren Barborak's presentation emphasized the need for collecting and assessing weldability data for irradiated materials, quantifying helium concentrations, and developing field-deployable repair systems. Additionally, the research highlighted the importance of reviewing detection and localization methods for fuel pool leaks and evaluating non-metallic repair options. High-cycle fatigue testing and structural support retrofitting for small-bore piping, as well as understanding the effects of thermal aging on Alloy 52/152 through heat treatments and microscopy, are also recommended.
Jian Chen's presentation discussed advanced welding techniques for light-water reactor components, focusing on refining welding parameters to reduce helium-induced cracking and conducting additional experimental validations. The research also suggests investigating the use of other materials and alloys in welding experiments and exploring the long-term effects of helium concentration on reactor components.
The panel discussion underscored the development and application of water chemistry technology for mitigating SCC, implementing IHSI and MSIP for piping and nozzle welds, and exploring core internal replacement as a key technology for LTO. The focus on remote inspection and repair technologies for irradiated environments, achieving flux reduction at vulnerable structures, and considering new materials with better radiation resistance for LTO are also highlighted. Continuous evaluation and improvement of repair welding techniques for irradiated materials is emphasized, along with the need for collaboration between industry and regulatory bodies to address these challenges.
2.9 Session 8: Civil Structures, Concrete, and Components 2.9.1 Presentations 2.9.1.1 Research for Beyond 80 Years-Civil Structures (ML24282A963)
Presenters: Madhumita Sircar, George Wang (NRC)
The presentation delved into the challenges and opportunities associated with ensuring the long-term performance of passive, long-lived structures and components (SCs) in light-water reactors operating for more than 80 years. It provided an overview of existing knowledge, including OE and engineering analyses, while identifying research needs to support the adequacy of aging management programs. It emphasized the importance of focusing on emerging or evolving topics related to aging mechanisms, which may become critical for extended reactor operation.
2-41 Key brainstorming topics included the importance of aging management, improved inspection and monitoring technologies, and efficient methods to address aging effects. Advanced tools, such as demonstrated artificial intelligence (AI), machine learning (ML), drones, robotics, and sensors, were suggested for condition monitoring and inspections, particularly in inaccessible areas. Integrating Building Information Modeling (BIM) technology with aging data is recommended to modernize plant strategies and enhance safety measures.
The presentation also highlighted critical aging effects like irradiation on structures, post-tensioned containment creep, and corrosion. These concerns necessitate confirmatory testing under real-world conditions to reduce uncertainties. There was a strong focus on understanding synergistic effects, such as interactions between radiation and corrosion, and on evaluating their significance under various conditions. For post-tensioned containments, long-term impacts of creep, the effects of detensioning, retensioning of tendons, and the possibility of cracking due to internal strains were discussed, along with strategies for monitoring and management.
Contributions from the International Atomic Energy Agency (IAEA) International Generic Ageing Lessons Learned (IGALL) Working Group 3 identified potential research areas, such as new aging management programs (AMPs) for elastomers in civil structures and aggressive environments. Other areas include evaluating the functionality of concrete containment and irradiation effects on structures near the RPV. Overall, the presentation underscored the need for time-dependent aging and associated aging management activities to manage the effects of aging associated with the structural integrity of SSCs, newer approaches for monitoring and inspection, real-life data, and proven repair strategies to ensure safe and extended reactor operations beyond their original design lifetimes.
Key Research Recommendations/Emphasize:
Develop or Enhance AMPS: New AMPs for monitoring durability and assessing integrity of elastomers in civil structures. Enhance existing AMPs for aggressive coastal environments, long-term containment safety, and functionality.
Harvest: Conduct confirmatory testing to reduce uncertainties related to degradation due to radiation and other degradation mechanisms.
Inspection and monitoring using advanced technologies: Integrate advanced technologies like AI, robotics, BIM, and NDEs for efficient inspections and condition monitoring.
Aging of post-tensioned containments: Investigate long-term creep effects in post-tensioned containments, including the effects due to modification needed for steam generator replacement and other reasons, tendon retensioning, and mechanisms that causes radial strains. Also investigate long-term elastic and inelastic creep, poisons creep effect, tensile and bending creep, change of loading plane, difference of response on new and existing material (concrete, tendon) and tendon anchorage technologies.
Synergistic Effects: Explore synergistic effects of alkali-silica reaction (ASR),
carbonation, corrosion, radiation, and other aging mechanisms under various conditions.
Repair and replacement: Evaluate repair and replacement strategies for reactors operating beyond 80 years, ensuring feasibility and reliability.
2-42 2.9.1.2 Operation Beyond 80: Knowledge Gaps and Synergistic Degradation Modes in Concrete (ML24282A960)
Presenter: Yann Le Pape, ORNL In the United States, nuclear plants are aging while new projects like advanced nuclear reactors and small modular reactors are still in development. Extending the life of existing plants is crucial for continuous, affordable, and safe energy production, as well as for maintaining a well-trained and knowledgeable workforce. Despite the age of these plants, the concrete used is generally performing well. However, as with aging populations, the risk of issues increases over time. Over the past decade, research has focused on understanding the effects of irradiation on materials, but there is a need to update this knowledge to address new challenges.
Concrete in nuclear plants faces unique challenges due to varied designs and local material sourcing. Different concretes have different susceptibilities to degradation, influenced by factors such as operation, design, and materials. Understanding the structural integrity of the biological shield (or primary shield wall) under various conditions (service, accidental, seismic) and environments is crucial. Cracks in irradiated rocks/aggregates form in complex ways, both at the grain interface and within grains. The behavior of different concrete rock/aggregate types and their bonds is not fully understood, indicating the need for further research.
Stress relaxation in cement paste is essential, and there are two primary ways to dissipate stresses: through viscous mechanisms or by cracking. If expansion occurs very slowly, stresses may be relaxed. However, gaps in knowledge persist, particularly regarding neutron effects, which show that as neutron flux decreases, expansion also decreases. This makes understanding neutron effects a high research priority. Additionally, there is limited data on the interaction between irradiation and corrosion, and other synergistic degradation mechanisms.
Continued research is necessary to ensure the long-term integrity of nuclear plant construction materials.
Harvesting data from existing structures, materials, and components that have been inservice provides real-life data that is invaluable for understanding material degradation under actual operating conditions. This data helps validate accelerated testing methods, improve predictive models, identify new degradation mechanisms, and guide future research efforts. Researchers have found that gamma irradiation appears to increase the corrosion rate of steel, an important area for future research in structural concrete and steel embedments. Preliminary results on steel concrete bond strength and concrete creep relaxation mechanisms indicate the need for further research, especially at higher doses. The discovery of new phases in concrete from Hamaoka underscores the importance of ongoing research and data collection to confirm degradation scenarios inservice conditions. Addressing these challenges in future research efforts is essential for maintaining the structural integrity of aging nuclear plants.
Key Research Recommendations:
- 1. Enhance Knowledge on Irradiation Effects: Enhance existing knowledge on how irradiation affects materials in nuclear plants, focusing on the structural integrity of critical components under various conditions (service, accidental, seismic).
Neutron Flux Effects - Investigate the impact of neutron flux on material degradation. As preliminary data suggests, lower neutron flux may result in decreased expansion and degradation rates.
2-43 Stress Relaxation - Study the mechanisms of stress relaxation in cement paste, exploring the potential for acceleration of creep due to irradiation.
Gamma irradiation - Investigate how gamma irradiation affects the corrosion rate of steel and the overall properties of cementitious materials.
- 2. Synergistic Degradation Mechanisms: Research the interaction between irradiation and other degradation mechanisms, such as corrosion, carbonation and ASR. Assess how these mechanisms combine and affect concrete structures for LTO.
- 3. Harvesting Data from Inservice Structures: Collect and analyze real-life data from existing structures, materials, and components that have been inservice. This data is invaluable for validating accelerated testing methods, improving predictive models, identifying new degradation mechanisms, and guiding future research efforts. Formation of new phases in concrete as concrete age (lessons learned from harvesting in Japan).
Addressing these research needs will help ensure the long-term integrity, safety, and efficient operation of aging nuclear plants.
2.9.1.3 EPRI Research on Long-Term Operations: Considerations for Life Beyond 80 (ML24282A961)
Presenter: Samuel Johnson, Electric Power Research Institute Extending the life of the global nuclear energy landscape includes 438 operating nuclear reactors, with a significant portion having been in operation for 30 to 40 years and some are in operation for 50 years and more. In the United States, around 60 % of reactors have been operating for at least 40 years. This situation has turned license renewal and extended operations into a widespread priority, making LTO a global topic of interest. EPRI has been at the forefront of addressing these challenges through its LTO program, established in 2010.
EPRI's program, driven by the findings of the Expanded Materials Degradation Assessment (EMDA) report, aims to identify and address technical gaps to extend the life of nuclear power plants (NPPs). Significant milestones have been achieved, including the approval of subsequent license renewal applications in the United States. EPRI focuses on supporting its members in implementing aging management strategies for various components, including reactor vessels, core internals, primary components, and concrete structures.
The degradation of concrete structures in nuclear plants is likely and highly dependent on environmental conditions, loading conditions, and design. As time progresses, degradation becomes more likely, with key mechanisms being corrosion of reinforcement due to chloride ingress and carbonation, cracking due to ASR or delayed ettringite formation (DEF), and concrete irradiation. While carbonation takes a long time to occur, it is an important consideration as plants continue to age. Utilities operating beyond 80 years may exceed degradation thresholds, necessitating new mitigation strategies, retrofit, and modernizations.
Technological solutions, such as advanced inspection and NDE techniques, remote monitoring sensors, can provide better data for utilities, though these often involve business decisions regarding cost-benefit analysis. Longer operational periods allow for better recoupment of costs associated with these technologies. EPRI's path forward focuses on providing research, guidance, and practical solutions for LTO, relying on organizations like NRC and the Department of Energy (DOE) for applied research. Key areas include improved inspection and monitoring techniques, better aging modeling in analysis/simulation programs, and development of new repair materials and techniques.
2-44 Concrete aging management has been a significant focus for EPRI since 2010, encompassing research on irradiation, ASR, corrosion of reinforcement, and visual inspections. EPRI actively participates in industry working groups and provides support to utility members for aging management implementation. The research highlights the importance of ongoing studies to address existing degradation mechanisms and ensure structural functionality beyond 80 years of operation. Coordination and collaboration with organizations like the NRC and DOE are crucial in driving the industry forward.
Key Recommendations:
Potential for degradation to occur and progression of degradation can increase with time:
Corrosion of reinforcement due to chloride ingress, carbonation Cracking due to expansion (ASR, DEF)
Concrete Irradiation (More plants to exceed threshold)
Inspection and Monitoring Advanced Inspection and NDE Techniques Remote Monitoring Sensors Analysis / Evaluation: Degradation Modeling Guidance for Fitness for Service Mitigation / Modernization: Enhanced Mitigation Technologies Modernization Technologies Cathodic Protection Systems Enhanced Inspection and Monitoring (Drones, Remote Monitoring)
Repair / Replacement: New Repair Materials/ Advanced Repair methodologies Tools for Aging Management EPRI research moving forward will focus on tools to provide more reliable and efficient aging management of civil infrastructure 2.9.1.4 AMPs and Future Research Need for LTO on Concrete Structures - Part 1 (ML24282A962)
Presenter: Masa Kojima, NRA Japan The AMP for concrete strength degradation primarily focuses on pathologies such as carbonation, chloride penetration, alkali-aggregate reaction, elevated temperature, and irradiation. Routine visual inspections are essential, but various testing options are also employed, including measured testing, predictive formulas, numerical simulations, nondestructive testing, and destructive testing. Structural analysis is another valuable evaluation tool to assess structural integrity and age management of SSCs.
For carbonation, visual inspection is crucial, supported by measured carbonation depth and predictive formulas. Chloride penetration management also relies on visual inspection, with options for chloride ion content testing and predictive formulas. The alkali-aggregate reaction is monitored visually, and petrographic examination and reactivity testing are used.
2-45 Elevated temperatures are assessed through numerical methods, with a conservative threshold value of 66°C from NUREG-2192. Validity is checked by comparing numerical simulations with measured temperatures. Irradiation effects are managed using numerical methods and structural evaluations, including simulations and analyses.
Compressive strength can be tested non-destructively with methods like the rebound hammer or destructively by harvesting concrete samples. However, it is challenging to perform these tests during operation, particularly for elevated temperature and irradiation points, as direct access to the inner surface of the primary biological shield wall is limited.
The presentation emphasized the importance of visual inspections and identified future research needs, including improving NDE with demonstrated AI or remote/drone systems, and understanding synergistic degradation mechanisms.
Key Research Recommendations:
Inspection and Monitoring: Visual Inspection: detect indications of degradation such as cracks in a timely and appropriate manner through visual inspection. Improvement of NDEs using AI and drone systems. Enhancement of BIM and monitoring methods.
Measured Testing and Predictive Formulas: Detailed studies on carbonation, chloride penetration, and alkali-aggregate reaction, including predictive formulas and measured methods.
Numerical Simulation and Structural Evaluation: For elevated temperatures, validate of numerical simulations using measured data. For irradiation effects, validate numerical simulations using appropriate data and assess structural performance using plant-specific information in the structural analysis.
Synergetic Effects: Investigating the synergetic effects of degradation mechanisms.
2.9.1.5 AMPs and Future Research Need for LTO on Concrete Structures-Part 2 (ML24282A962)
Presenter: Ippei Maruyama, University of Tokyo Around 15 years ago, a research gap was identified regarding the evaluation of concrete soundness exposed to neutron and gamma ray irradiation. This led to a Nuclear Regulation Authority (NRA, Japan) project from 2008 to 2016, focusing on the degradation mechanisms of concrete due to irradiation, primarily caused by the volumetric mismatch between cement paste shrinkage and aggregate expansion. Despite progress, there remains a need for methodologies to assess the soundness of reinforced concrete members. Subsequent research by the Japanese Ministry of International Trade and Industry (MITI) developed several techniques and numerical calculation methods for this purpose.
To understand degradation and evaluation methods, it is necessary to address small-scale degradation at the mineral level, which is connected to aggregate-scale degradation and overall concrete degradation. The Rigid Body Spring Network was developed to evaluate aggregate expansion, crack propagation, and changes in physical properties. This technique helped assess the seismic performance reduction in cylindrical reinforced concrete structures. Real concrete behavior in reinforced structures, especially under realistic conditions, requires further investigation. Harvesting concrete from decommissioned plants revealed significant strength
2-46 contributions from reactions between rock-forming minerals and cement hydrates, emphasizing the importance of long-term reaction processes in concrete exposed to neutron irradiation.
Aggregates show varying expansion based on their mineral phases. For instance, quartz expands significantly under neutron irradiation, affected by temperature and healing mechanisms. Recent data confirms the flux impact on mineral expansion, indicating a need for detailed studies on the healing and annealing processes. Simple formulas have been proposed to account for these processes, but validation through harvested material investigation is essential. Concrete degradation due to heating and drying can reduce strength by up to 15 %.
Additionally, long-term changes in concrete, influenced by interactions between minerals and cement hydrates, require further understanding. Research is needed to confirm predictions made under accelerated conditions and to explore the real conditions in NPPs. The goal is to verify seismic performance, shielding performance, RPV supporting function for design-basis accident conditions, interference for other functions for example spalling of concrete in the cavity area.
Key Research Recommendations:
Enhance knowledge on Irradiation Effects: Study the behavior of rock-forming minerals under neutron and gamma irradiation and long-term interactions between minerals and cement hydrates in concrete. Research the mechanisms of healing and stress relaxation in irradiated concrete.
Testing and Numerical Simulation: Develop consensus on evaluating strength degradation due to heating, drying, and neutron irradiation. Create reliable techniques to evaluate the soundness of irradiated reinforced concrete members. Collaborate with industry and research organizations to align efforts and drive advancements.
Harvesting: Harvest and analyze concrete from decommissioned plants to collect real-life data and validate degradation and numerical methods.
2.9.1.6 Research Results on Concrete Using the Decommissioning Plant of the Hamaoka Nuclear Power Plant (ML24282A964)
Presenter: Takashi Osaki, Chugoku Electric Power Company (CEPCO)
The Hamaoka Nuclear Power Station has five units, with units 1 and 2 being decommissioned since 2009. Units 3, 4, and 5 are currently paused to implement new safety regulations. The decommissioning process, divided into four stages, is set to take around 30 years, with the research period from 2015 to 2022 focusing on dismantling equipment in the nuclear reactor area.
In Japan, in addition to visual inspections, large number of core samples are taken from the structures for destructive testing. Excessive core sampling, which can damage the concrete structures of NPPs, is a significant concern. Therefore, it is essential to minimize destructive testing while evaluating the soundness of the NPPs. The research aimed to develop a rational soundness evaluation method by collecting concrete core samples from reactors of units 1, 2, 3, and 5 at Hamaoka. The presentation primarily focused on the extensive examination of Unit 1.
Core samples extracted from the inner walls of Hamaoka Unit 1 showed that the center cores have higher compressive strengths compared to the surface cores. Chemical analysis revealed at the elevated temperature and high water content, reactions between portlandite and
2-47 dissolved feldspar, formed new hydrates which converted to Al-tobermorite resulting in an increase of concrete strength. Core samples from younger units (3 and 5) indicated increased reaction debris and aggregate content with material age. The research presentation concluded that understanding and evaluating these changes in concrete can be considered for assessing long-term NPP operations.
Key Research Recommendations:
Harvest Concrete Cores: Analyze concrete cores from decommissioned and paused plants to evaluate structural soundness. Collect samples from multiple sites to account for variations in environmental conditions and materials used, such as concrete containing fly ash. Gather comprehensive data on aged concrete to develop knowledge applicable across different NPPs.
Refine Evaluation Methods: Use findings to establish a rational soundness evaluation method tailored to LTO needs.
Facilitate Industry Collaboration: Promote knowledge-sharing among NPPs and research groups to improve understanding of aging concrete.
2.9.1.7 NRC Workshop: What Research for Beyond 80 Years for Civil Structures - IRSN Contribution (ML24282A965)
Presenter: Fabienne Ribeiro, Nuclear Safety and Radiation Protection Authority (ASNR)
In France, over 75 % of nuclear reactors were built between the 1980s and 1990s and are now reaching around 40 years of operation. License renewals for each reactor are decided after periodic inspections every 10 years. For the 900 MW reactors, the French regulatory authority, Institut de radioprotection et de sûreté nucléaire (IRSN), established rules and conditions for continued operation beyond the fourth periodic review in 2021. Similar rules will apply to the newer 1,300 MW reactors by 2025. For operations beyond 50 years, IRSN will decide during the fifth periodic review, and has asked the Électricité de France (EDF), the operator, to justify the continued operation of reactors for up to 60 years and beyond by 2025.
The current goal of IRSN, as the technical adviser to the French Safety Authority, is to investigate the aging and safety requirements of NPPs for 60 years and beyond. This includes ongoing research on containment buildings and individual structures, particularly focusing on predicting structural behavior under normal and extreme conditions such as external hazards and events such as earthquakes, storms, explosions/aggressions or internal events such as severe accidents. To achieve this, IRSN is developing simulation models and reference experimental data to understand the mechanisms of aging and validate the models. Aging mechanisms like corrosion, internal swelling reactions, delayed deformation (creep, shrinkage) and their impact on the mechanical properties of materials are also being studied. Scientific challenges are time effect, space or size effect, and coupling of degradation mechanisms.
If the difference in material properties and parameters remains negligible for up to 50 years of operation, the potential extension to 60 years, or even 80 years, will necessitate reconsidering the impact of irradiation on these materials. Much work has already been done by various groups, including data from open access programs and research conducted by organizations like the French Alternative Energies and Atomic Energy Commission (CEA), NRC, NRA (Japan),
and U.S. DOE. The first step involves reviewing existing knowledge and identifying needs and opportunities for further development.
2-48 Evaluating the differences in concrete compositions is important. The IRSN aims to extend the Vessel Aging Calculation Scheme (VACS), initially developed for vessel fluence, to evaluate vessel supports. This complex task requires addressing parts farther from the core than the vessel itself, necessitating advancements in radiation estimation methods like Monte Carlo and ongoing, dedicated validation of concrete vessel supports in irradiation states. Additionally, understanding the historical management of French PWRs, including the realistic irradiation history and core porosity over time, is essential.
Key Research Recommendations:
Modeling Tools and Experimental Data: Develop modeling tools and gather reference experimental data to understand the aging mechanisms and validate models. Develop durability indicators and multiscale, multiphysics simulation tools for evaluating concrete affected by pathologies.
Coupled/Synergistic Effects of Aging Effects: Evaluate the impact of different environmental conditions and material compositions on concrete behavior. Research the impact of corrosion, internal swelling reactions, and other aging mechanisms on the mechanical properties of materials. Address the scale effect and coupling of different aging phenomena such as corrosion and alkali-silica reactions.
Irradiation Impact: Assess the impact of irradiation on materials, especially for operations extending beyond 50, 60, and 80 years. Review existing knowledge from various organizations (e.g., CEA, NRC, NRA (Japan), U.S. DOE) to identify further development opportunities.
Radiation Transport Calculation: Extend the VACS Calculation Scheme initially developed for vessel fluence to concrete vessel supports, using advanced variance radiation methods and dedicated validation. Study the realistic irradiation history and core porosity of French PWRs over time.
Collaborative Efforts: Utilize collaborative R&D programs like the Observatory of the durability of reinforced concrete structures (ODOBA) Project and the MACUMBA Platform on containment leak tightness for comprehensive data collection and validation.
Nondestructive Examination: Develop NDE techniques to identify degradation, enhance inspection, and gather reference experimental data and validate simulation tools.
2.9.1.8 LTO-EDF Approach (ML24282A966)
Presenter: Benoit Masson, EDF The French approach to LTO of nuclear containment buildings focuses on structural integrity and safety enhancements. It emphasizes the French safety framework, the diversity of its fleetcomprising single-wall and double-wall containment structuresand attention to aging phenomena such as prestressing loss, drying, creep, and shrinkage. Key aspects of LTO include implementing accurate creep and ASR/DEF laws, performing nonlinear numerical simulation and calculations to account for aging factors, and analyzing deformation and stress to ensure containment strength during severe accidents. Research into creep, shrinkage, and coupled effects with accident conditions like temperature is vital.
2-49 The project "Verification Réaliste du COnfinement des RéacteurS" (VERCORS), a 1/3-scale mockup built with industrial-grade materials, aids in understanding and monitoring LTO.
Accelerated drying and creep tests (about nine times fasters than actual structure) are conducted alongside the examination of over 1,000 samples to evaluate modulus, shrinkage, and resistance. Advanced monitoring systems enhance insight into aging phenomena and leakage, such as pathway and cracking evolution, and refined finite element method (FEM) models for predictive accuracy.
Coating assessments are integral, particularly regarding seismic shutdown systems. Repairs to coatings, liner corrosion, gutters, and base raft thickening mitigate accident risks and strengthen seismic resilience. Monitoring systems, including optical fibers and innovative devices, improve detection of cracking and leakage.
Harvesting components like anchorage, concrete, and waterstop joints ensure long-term reliability. Prestressing management is crucial since it affects permeability and cracking.
Managing tertiary creep is essential as it significantly reduces concrete strength, especially under temperature impacts or accident tests. Periodic measurements confirm structural stability.
Collectively, the French strategy combines advanced monitoring, structural repairs, and enhanced modeling to ensure the sustainability and safety of nuclear infrastructure under LTO conditions.
Key Research Recommendations:
Containment Prestress Loss: Conduct comprehensive studies on aging phenomena such as prestressing loss, drying, creep, and shrinkage, incorporating nonlinear calculations and accurate laws (e.g., ASR/DEF) to ensure structural safety under accident conditions.
Advanced Monitoring: Utilize advanced monitoring systems like VERCORS mockup to gain insights into deformation, cracking evolution, leakage pathways, and structural behavior over time. Manage and monitor creep levels periodically to prevent the transition to tertiary creep, maintaining structural integrity and avoiding significant strength reductions under high stress or temperature conditions.
Modeling and Simulation: Enhance FEM models by integrating precise data on drying, creep, shrinkage, and aging to improve predictive accuracy and inform proactive structural interventions.
Repair and Mitigation: Implement systematic assessments and repairs, including coatings, liner corrosion, gutters, and base raft thickening, to mitigate risks related to severe accidents and improve seismic resilience.
NDE: Explore innovative methods for detecting cracks, leakage, and structural weaknesses, leveraging new NDE devices and measurement techniques.
2.9.1.9 CNSC Perspective on Gaps in Knowledge for Concrete Containment Structures for LTO (ML24282A967)
Presenter: Cedric Androuet, Canadian Nuclear Safety Commission (CNSC)
The CNSC has been focused on addressing the aging of their nuclear reactors, which are between 35 and 50 years old. Particular attention has been given to concrete structures,
2-50 prompted by the identification of ASR at the Gentilly-2 NPP during its license renewal 15 years ago. A research project was initiated to assess the consequences of ASR on containment structures, involving various material and structural tests and modeling, including notable tests at the University of Toronto. One key finding was that concrete with reactive aggregates had a higher ultimate capacity, but significantly reduced ductility compared to non-reactive concrete.
This research was further integrated into the Assessment of Structures subjected to Concrete Pathologies (ASCET) project by the OECD Nuclear Energy Agency (NEA). The ASCET project demonstrates that while the ultimate capacity of wall without ASR and walls with ASR is satisfactory, significant gaps remain in predicting displacements, hysteretic loops, failure modes, and loss of ductility in ASR-affected walls and structures. Further research and modeling are essential, particularly for understanding deformation limits, which dictate behavior under cyclic loading conditions like seismic events, given the reduced ductility.
Continued research efforts are ongoing through the ODOBA project and ACES project, focusing on the long-term safety performance of nuclear civil engineering structures.
The ODOBA project is studying large-scale testing, effects of reinforcement on the expansion and crack development, and monitoring by using nondestructive evaluation/ nondestructive testing (NDE/NDT). The ACES project underscored research on corrosion of embedded liners, internal swelling of concrete, delayed strain in containment buildings, and radiation effects on concrete.
The CNSC has recently launched a research project to evaluate the impact of climate change on NPP concrete structures. The project aims to assess the effects of increasing global temperatures and atmospheric humidity over the past 60 years and to project changes for the next 50 years, specifically in regions of Canada with nuclear plants. The research focuses on how climate change may impact the structural integrity, tightness, and durability of concrete structures, with particular attention to freeze-thaw cycles, carbonation, and corrosion.
Preliminary findings highlight concerns such as increased CO2 concentration leading to carbonation and corrosion. Higher temperatures affecting freeze-thaw cycles and overall concrete durability and potential increases in tornado occurrences and sea levels could be some of the factors impacting structural integrity and corrosion as well.
Key Research Recommendations:
Enhanced NDT techniques for structural evaluations.
Development of more accurate models and quantitative assessment methods to replicate the deformation behavior of ASR-affected structures.
Impact of climate change on concrete degradation mechanisms; examples below.
o Investigate effects of increased atmospheric CO2 levels on concrete structures, combined effects of multiple environmental stressors, such as temperature and humidity, on concrete durability.
o Study the impact of higher reservoir temperatures on the structural integrity and cooling systems of nuclear plants.
o Assess the resilience and safety of concrete structures against extreme precipitation and potential flooding.
o Evaluate the influence of rising sea levels on chloride diffusion and corrosion in concrete structures.
2-51 o
Explore the effects of increased tornado occurrences and tropical cyclones on structural integrity.
o Examine the kinetics of ASR under increased ambient temperatures and humidity.
2.9.1.10 Aging Management of Nuclear Civil Structures - Potential Areas for Enhancements (ML24282A968)
Presenter: Julia Tcherner, AtkinsRealis The presentation focused on the specifics of nuclear civil structures, emphasizing the importance of understanding aging, particularly in Canada. It highlighted the aging management activities pursued by AtkinsRéalis, evolved from Atomic Energy of Canada Limited. The discussion covered the general aspects of nuclear concrete structures and their role in mitigating extreme events. Detection and maintenance of degradation in these structures, along with areas for enhancement, were briefly touched upon. It was emphasized that nuclear concrete structures, although passive under normal conditions, undergo time-dependent changes that could alter their performance.
The presentation elaborated on aging management models and standards in Canada, referencing CNSC and the Canadian Standards Association (CSA) standards. The importance of understanding aging through R&D, OE, and condition assessments was discussed.
Identifying degradation mechanisms and performing inspection activities aim to detect and mitigate aging effects. Ongoing research collaborations, such as with the University of Toronto, investigating the interaction of various degradation mechanisms, were mentioned. Visual inspection methods were highlighted as a key tool for detecting signs of degradation in concrete structures.
Instances of corrosion of steel liners in steel-lined concrete structures were observed after nondestructive or destructive testing and laboratory analysis of samples, often due to objects left during construction. A project sponsored by the CANada Deuterium Uranium (CANDU)
Owners Group, in collaboration with Olson Engineering, tested different nondestructive techniques on large-scale specimens with various defects. Slab impulse response was effective in identifying delamination,. Recommendations were made to improve these previously identified methods, emphasizing the need for a bond between steel and concrete. Olson Engineering expanded these methods to remotely operated vehicles (ROVs) for faster scanning of structure surfaces. Digitized inservice inspection results allow more meaningful evaluations and decision-making regarding mitigative or monitoring strategies. A simple tool to document inspection results facilitates training and enhances aging management effectiveness.
The presentation also highlighted the importance of repairs, evaluating degradation, establishing the cause of distress, and selecting appropriate repair methodologies for long-lasting, economical solutions. Performance-based requirements and prequalification tests are crucial for successful repairs, tailored to specific cases. Examples of various repairs emphasized the importance of root cause assessment, material qualification, field support, and QA/QC. The increasing requirements for nuclear concrete structures due to plant life extensions, security, and evolving safety standards were acknowledged.
Key Research Recommendations:
Inspection and Monitoring - Develop novel techniques for detecting degradation mechanisms in inaccessible areas to enhance inspection accuracy. Create
2-52 comprehensive databases for inservice inspection results to enable tracking and trending of degradation over time.
Repair and Mitigation - Explore advancements in repair methods and materials tailored to specific aging mechanisms, ensuring durability and long-term performance. Take advantage of previously performed repairs with proven field success to refine future strategies.
Coupled/Synergistic Effects of Aging Effects - Investigate the interaction effects of coupled degradation mechanisms to predict long-term impacts on structural integrity and enhance structural resilience. Research the effects of advanced environmental stressors on concrete and steel components to strengthen design resilience to significant stressors and construction quality.
Numerical Simulation - Validate numerical simulations for elevated temperature assessments. Conduct structural evaluations incorporating numerical simulations for irradiation effects.
2.9.1.11 Crevice Corrosion Assessment of Steel Liners & Cement interface in Nuclear Power Plants (ML24282A748)
Presenter: Valdir De Souza, SCK CEN (Belgian Nuclear Research Centre)
The crevice assessment and experiments on steel liners and cement interfaces were part of the European commissioned project ACES (Assessment or Advancement of Civil Engineering Structures). The presentation covered background and objectives, experimental setup, techniques, conditions, results, and conclusions. Corrosion embedded in steel and concrete structures, particularly steel liners, was observed, despite the alkaline environment of concrete providing corrosion protection. The focus was on simulating and understanding crevice corrosion mechanisms at the contact between steel liners and concrete interfaces.
Experimental techniques included open-circuit potential (OCP) measurement, potentiodynamic polarization, and the Tsujikawa Hisamatsu Electrochemical (THE) method to determine crevice repassivation potential. These methods aimed to simulate crevice corrosion and understand the conditions under which it occurs. Different environments, such as high pH solutions with and without chloride, and low pH solutions, were tested under both oxic and anoxic conditions.
Results showed that high pH environments generally prevented corrosion except when chloride added crevice repassivation potential (CREV) and corrosion were noticed. In low pH environments corrosion was observed outside the crevice region conditions. The tests showed that in oxic conditions with high chloride, passive film breakdown and active corrosion occurred, while in anoxic conditions with oak, active behavior was observed. The method revealed no crevice repassivation potential in high pH, but chloride presence led to crevice repassivation potential. Low pH conditions showed corrosion outside the crevice region. These findings indicated the need for further research on accelerated techniques and crevice corrosion mechanisms.
Key Research Recommendations:
Investigating the impact of various environmental conditions on crevice corrosion mechanisms. Exploring the effects of different crevice formers on corrosion behavior.
2-53 Development of advanced experimental techniques for more accurate simulation and analysis of crevice corrosion.
Exploration of mitigation strategies to prevent crevice corrosion in steel-lined concrete structures.
2.9.1.12 Crevice Corrosion Modelling of Steel Liners & Cement Interface in Nuclear Power Plants (ML24284A244)
Presenter: Anssi Laukkanen, VTT
[Note: The presenter could not attend the workshop. Below is the summary from the presentation slides.]
The slides describe an extensive framework for modeling corrosion and electrochemistry processes, with applications spanning erosion-corrosion, crevice and pitting corrosion, tribocorrosion, elevated temperature oxidation and erosion, reactivity and slurry flow, and surface deposition techniques. Central to the modeling approach is its microstructural foundation, where a 3D representation of steel and concrete microstructures integrates multiple physical phenomena, such as chemical transport, electrochemistry, thermomechanics, and damage progression.
The VTT experiments involve specialized crevice corrosion setups, utilizing materials like steel plates and concrete slabs under varying conditions (e.g., pH levels and chloride concentrations).
Experimental setups include diverse crevice formersconcrete, carbonated concrete, and woodpaired with controlled environmental parameters. Models incorporate transport and conservation using Nernst-Planck equation. Furthermore, transport in concrete, corrosion product formation, and phenomenological modeling are aligned to mimic aging and degradation.
Key highlights include detailed modeling of crevice geometries, chemical species, and electrochemical reactions, achieved through finite element and finite volume methods.
Simulations predict behaviors like chloride enrichment, oxygen consumption, and nickel dissolution (as NiCl+), emphasizing the interdependence between crevice dimensions, chemistry, and corrosion dynamics.
The transport phenomena model extends to multiphysics scenarios, linking surface interactions, phase transformations, and thermomechanical impacts, including crack initiation and growth.
This comprehensive approach enables predictions on microstructural evolution, failure mechanisms, and the influence of operational conditions, offering critical insights for engineering materials and design.
Key Research Modelling Process:
Comprehensive corrosion modeling: The project focuses on developing and applying multiphysical 3D microstructural models to study corrosion behaviors, coupling chemical transport, electrochemistry, thermomechanics, and damage progression.
Experimental validation with diverse setups: Customized crevice corrosion experiments explore the effects of varying materials, geometries, and environmental conditions, such as pH levels and chloride concentrations, on corrosion dynamics.
2-54 Application of advanced simulation techniques: Models integrate finite element methods and Multiphysics solvers coupling with environmental behavior and corrosion, loss and damage of material, and mechanical behavior.
2.9.1.13 Creep and Creep-Cracking in PCCV (ML24284A243)
Presenter: Randy James, Structural Solutions Consulting (SSC)
This presentation discussed several key topics related to structural performance and creep in post-tensioned concrete containment (PCCV). Creep, which refers to time-dependent deformation under load, has significant implications for the structural integrity and longevity of PCCV. It was observed that as concrete ages, the creep rate decreases, but primary creep can reactivate under changing loads. Additionally, visco-elastic recovery with an irreversible component was noted.
Observations highlighted that creep is considered in the design of PCCV to maintain minimum design specifications but is generally not accounted for in thermal or mechanical load combination cases or short-lived accident conditions. However, creep becomes important in assessing long-term performance after significant load and temperature changes due to accidents.
Beneficial aspects of creep include its ability to reduce stress over time and redistribution of load paths, potentially improving the structural performance of concrete over long periods.
However, unfavorable aspects include increased deformation over time and the possibility of residual or "locked-in" creep strains contributing to cracking, especially under changing loads.
Creep-related structural issues emphasize that concrete is designed to carry compressive stress, and creep can exacerbate cracking, leading to maintenance and long-term performance issues. Structural modifications, such as modification for steam generator replacements PCCVs, require significant changes in loading over extended periods, which can affect creep performance.
Assessment of creep-related issues must consider that creep response is load, time, and temperature dependent, making the resulting structural effects path dependent. Tracking creep through elevated temperature gradients or cycles requires time-shifting between temperature dependent data. Environmental parameters, such as temperature gradients, cycles, humidity, and moisture transport, significantly affect creep rates, making it challenging to establish actual creep characteristics.
In terms of knowledge areas, most creep testing focuses on measuring creep rates under constant compressive load, often without considering the inelastic component, which involves changes in material structure. This inelastic component is crucial as it contributes to irrecoverable deformation. Additionally, data on the effects of differential creep across structural sections, mechanically induced cracking, and creep in three-dimensional stress states are necessary for a comprehensive understanding of structural performance.
Lastly, comments and discussions stressed the importance of pre-stressed containments for long-term performance, especially in containment accident scenarios. Creep due to elevated compressive stresses must be carefully monitored, particularly in significant structural modifications. Tendon relaxation, retensioning, local and regional de-tensioning, seismic event-induced cracking, cumulative refueling shutdowns, and post-accident condition assessments are all significant factors that need to be considered for maintaining structural integrity over time
2-55 Key Research Recommendations:
Assessment-Related Topics: Develop modeling techniques for time-path-dependent creep effects under variable stress states, temperatures, and environmental conditions.
Research should focus on examining creep behavior under varying load and temperature combinations, especially for long-term structural performance including seismic and design-basis accident conditions.
Knowledge Areas for Improvement: Study the effects of inelastic creep and its irrecoverable deformation, analyze drying and basic creep interactions, moisture transport and humidity, and investigate differential creep behaviors including thermal gradient, and Poissons Ratio variations in 3-dimensional stress states under pre-stress conditions considering tension and bending.
Creep-Related Structural Modifications and practical applications Examine the impact of major structural modifications (e.g., steam generator replacements) on creep performance, emphasizing load redistribution and risks of cracking. Research the role of creep in tendon relaxation and re-tensioning, structural integrity tests, and performance assessments in seismic events and post-accident containment conditions.
Enhance Monitoring and Maintenance Establish guidelines for monitoring deformations and protocols for restarting procedures after structural impacts or accident conditions.
2.9.2 Panel Discussion Moderators:
Madhumita Sircar, NRC Jose Pires, NRC Panelists:
Yann LePape, ORNL Samuel Johnson, EPRI Ippei Maruyama, U-Tokyo Christopher Jones, Kansas-State University Julia Tcherner, AtkinsRéalis Q1: What research should be prioritized considering very long-term operation (over 80 years) and the accumulation of fluence and gamma dose over that time?
The panel discussion focused on research priorities for very long-term operation (over 80 years) and the accumulation of fluence and gamma dose. Maruyama highlighted that current data is based on accelerated experiments and stressed the importance of obtaining real operational data to confirm existing accelerated data. Le Pape agreed and outlined five research priorities:
characterizing inservice irradiated concrete, understanding the flux effect, exploring chemical changes such as ASR, investigating creep and relaxation effects in cement paste, and studying the interaction between steel and concrete, particularly that of bonded irradiated rebar and concrete. Tcherner raised questions about the area of concrete affected by irradiation and the
2-56 conservativeness of accelerated irradiation, highlighting the difficulty in quantifying radiation exposure over time.
Maruyama and Le Pape discussed the accuracy of irradiation transport calculations used in the nuclear industry and the need for a consensus on modeling methodologies to assess radiation effects on concrete structures. Sircar emphasized the importance of free expansion data and its use in benchmarking modeling methodologies. The panel members agreed on the necessity of plant-specific evaluations and research objectives, and the importance of understanding the effects of radiation on both concrete and embedded reinforcement. The discussion concluded with a consensus on the need for continued research to address these priorities and enhance the safety and longevity of concrete structures in NPPs. Harvesting irradiated concrete and components from RPV supports and biological shield structure will provide data under operating environment, flux, for assessing the degradation, radiation attenuation profile, reducing uncertainty in using collected applicable data from accelerated testing. The real-life harvested data will be valuable for benchmarking modeling and simulation of structural performance.
Summary Key focus areas included understanding the effects of fluence and gamma radiation, characterizing irradiated concrete, exploring flux effects, size effects, structural confinement and restraining effects, addressing chemical changes like ASR, and investigating creep and relaxation in cement paste. There was also emphasis on refining modeling methodologies and irradiation transport calculations to better assess the impact of radiation on concrete and embedded reinforcements.
The importance of OE was highlighted, as it would help validate findings from accelerated tests, reduce uncertainties, and enhance simulations of structural performance. The discussion underscored the necessity of plant-specific evaluations and harvesting irradiated components from operating environments to provide valuable insights into degradation processes. Overall, the panel agreed on continued research efforts to improve safety and extend the longevity of nuclear concrete structures.
Key Research Recommendations:
Underscore the importance of obtaining operational data by harvesting irradiated concrete to confirm existing accelerated data (flux effects) and reduce uncertainty. Also compare data from free expansion vs. effects under structural confinement.
Explore chemical changes, such as ASR, primarily corrosion of steel due to disassociated cement water due to gamma radiation coupled potential with boric acid spillage in PWRs.
Investigate creep and relaxation effects in cement paste.
Study the interaction between steel and concrete.
Verify Irradiation transport through concrete calculations used in the nuclear industry.
Emphasize a consensus on modeling methodologies to assess radiation effects on concrete structures.
2-57 Q2: (a) What research prioritization is recommended for creep, shrinkage, coupled thermal stress (differential temperature inside and outside) effects on post-tensioned concrete containments considering retensioning, detensioning of tendons and plant modifications over time. What other synergistic effects should be considered?
(b) What monitoring methods in conjunction with creep studies will bring additional insights into the performance and integrity of post-tensioned concrete containments?
The panel discussion covered the research prioritization for addressing the effects of creep, shrinkage, and coupled thermal stress on post-tensioned concrete containments, considering the processes of retensioning and detensioning tendons, as well as plant modifications over time. Jones began by highlighting the importance of understanding stress reversals, particularly when shifting from compressive or mixed stress states to those involving tension. He emphasized the need for further research on creep under mixed stress states beyond just compression. Le Pape shared insights from his previous research on creep and shrinkage for containment buildings, noting that French containments are heavily instrumented, allowing for detailed monitoring of deformations over time. He also pointed out the challenge of transient (short lived) creep during accidental scenarios and the need for further research in this area.
Ms. Sircar noted the differences between French and other containments, specifically mentioning the ungrounded tendons and the impact of plant modifications on containment performance, primary creep reactivation under change in loading, and inelastic residual creep.
Pires emphasized the need to understand creep behavior under different load patterns, such as bending, and highlighted the importance of obtaining creep data specific to these conditions.
The panelists agreed on the necessity of continued research to address these challenges and improve the safety and longevity of post-tensioned concrete containments. The panel discussed the value of the monitoring methods in conjunction with creep studies that will bring additional insight for the performance and integrity of post-tensioned concrete containments.
Key Research Recommendations:
Investigating creep under mixed stress states other than just compression. Exploring the effects of temperature and humidity on creep behavior. Addressing transient creep during accidental scenarios.
Monitoring deformations in containments over time, particularly during integrity leak rate tests.
Understanding the impact of detensioning, retensioning of the ungrouted tendons and plant modifications on containment performance.
Investigating creep behavior under different load patterns, such as tension, bending.
Investigate primary creep reactivation under change in loading, and inelastic residual creep Q3:
Considering various degradation mechanisms such as corrosion / carbonation /
chloride ingression / thermal cycling and freeze-thaw, concrete cracking from internal chemical reactions, etc., how can the synergistic effects be addressed through research?
2-58 The combined panel discussion covered a wide range of research priorities for addressing concrete degradation mechanisms in NPP structures. Ms. Tcherner emphasized the importance of understanding coupled degradation mechanisms, such as freeze-thaw cycles combined with chloride-induced corrosion and exposure to demineralized water. She also highlighted carbonation as a degradation mechanism that is not well-researched, especially in combination with other mechanisms. Laboratory work is useful, but actual field testing when permitted would be equally important and beneficial to accurately assess LTO degradation.
Pires discussed the synergistic effects of ASR and cracking, which can facilitate the ingress of corrosive agents, and suggested small-scale tests to explore these interactions. Le Pape stressed the importance of designing experiments to cover a range of conditions to avoid misinterpretation and emphasized the need for field observations to validate laboratory findings.
Maruyama reiterated the challenges of accelerated experiments, while Johnson noted the extensive experimental testing required to develop generic models that can be adapted to specific plant needs. Sircar mentioned that synergistic mechanisms could be studied in operating plants without waiting for decommissioning.
In the continuation of the discussion, Le Pape emphasized the complexity of relating observed degradation in the field to specific mechanisms and the limitations of relying solely on OE to predict future conditions. He pointed out that OE provides a historical perspective but may not accurately forecast future degradation scenarios. Additionally, visible degradation does not always indicate the extent of internal damage, such as with ASR, which can occur within walls before becoming visible on the surface.
Johnson highlighted the importance of prioritizing research efforts and leveraging opportunities to gather data from both decommissioned and operating plants. Pires raised the topic of irradiation's potential impact on ASR and suggested experiments to test the stability of ASR gel under irradiation. The panelists agreed on the need for a deeper understanding of concrete chemistry and the importance of field data to validate laboratory findings. They also discussed the early stages of research into the interactions between irradiation and ASR, emphasizing the need to test hypotheses and expand knowledge to ensure safety.
Key Research Recommendations:
Investigating freeze-thaw cycles combined with chloride-induced corrosion and exposure to demineralized water.
Studying carbonation in combination with other degradation mechanisms (ASR, corrosion).
Conducting small-scale tests to explore the synergistic effects.
Designing experiments to understand coupled synergistic mechanisms and validate laboratory findings through field observations.
Developing generic models as guidance for plant-specific applications.
Studying synergistic mechanisms and verifying using field data from operating or decommissioned plants.
2-59 Investigating the synergistic effects of degradation mechanisms, such as irradiation combined with ASR, corrosion. Conducting experiments to test the stability of ASR gel under irradiation.
Expanding knowledge to predict future degradation scenarios and ensure the safety of nuclear structures.
Q4. How can the current inspection and monitoring practices be supplemented and enhanced by using newer technologies, e.g., drones, remote sensing, BIM, AI, displacements measurements, etc.?
The current inspection and monitoring practices in the nuclear industry, especially visual inspections, were deemed very effective as initial indicators of structural degradation. However, degradation mechanisms often precede visible signs, which makes the integration of newer technologies essential. Technologies like drones, remote sensing, BIM, demonstrated AI, and displacement measurements can significantly enhance visual inspections by providing comprehensive records and enabling comparisons over time. These technologies, already in use to some extent, offer potential improvements and more precise data collection. Additionally, structural health monitoring, which is prevalent in other civil infrastructures, can be adapted for nuclear structures, despite its current limited use in the industry.
The adoption of advanced technologies like BIM, digital twins, and structural health monitoring in the nuclear industry may be driven by the need to reduce personnel and maintenance costs, particularly for new reactors. The existing fleet might find it challenging to adopt these technologies due to the lack of historical data and the complexities of integrating new systems.
Different objectives, such as increasing inspection precision or reducing workforce reliance, could guide the application of these technologies. Training of personnel in new monitoring and inspection technologies coupled with using of robotics can make inspections more economical, efficient, and effective and help facilitate knowledge transfer from experienced inspectors to newer personnel.
Key Research Recommendations:
Development and Integration of New Technologies: Investigate how drones, sensors, augmented reality, and other technologies can supplement visual inspections and create complete inspection records for comparative analyses.
Data Management and Analysis: Establish systematic approaches for managing and analyzing large volumes of data collected through these technologies, potentially using AI for better aging management insights.
Sensor Reliability and Redundancy: Develop robust plans for sensor maintenance, replacement, and ensuring long-term reliability, especially for grouted tendon structures.
Benchmarking and Best Practices: Conduct studies to benchmark current practices and identify gaps or areas for improvement using newer technologies, ensuring that the added burden of data collection is justified by the benefits.
Cost-Effectiveness and Preventive Action: Evaluate the cost-effectiveness of integrating new technologies and their potential for early detection and preventive measures.
2-60 Adoption and Integration for New Reactors: Investigate how new nuclear reactors can integrate BIM, digital twins, and health monitoring technologies effectively to minimize costs and personnel needs.
Technological Objectives and Applications: Differentiate the objectives for applying advanced techniques, such as precision improvement, inaccessible location, workforce safety, to tailor the use of technologies like robotics and digital image correlation methods.
Standardization and Guidelines: Develop new methods and standards for NDE techniques to evaluate material properties and address specific issues such as concrete behind steel liners.
Early Detection and Preventive Measures: Establish effective early detection protocols for degradation using enhanced NDE methods, advanced technologies, ensuring timely preventive actions to manage aging infrastructure.
Q5. What approaches are you aware of that may be taken to develop repair and replacement strategies for structures by using proven or promising technology?
Research on repair materials shows significant potential for nuclear applications, but many are not yet implemented due to a lack of dedicated materials, engineering design changes, or code and standard limitations. There is a need for a focused effort within codes and standards organizations to identify and address research requirements for their adoption in NPPs.
The discussion covered several ideas related to repair techniques and materials in nuclear structures. Existing earthquake retrofitting experience and experimental data can serve as a basis for improving seismic performance, though adaptations for nuclear environments are needed. A forum or database to share long-term field performance and effective repair methods was proposed, focusing on nuclear safety-related structures requirements. Sharing repair data could enhance material usage and updating codes.
Environment friendly cement industry changes, including a shift from C 150 to C 595 cements, and future adoption of LC3 blended cement, require consideration for repair applications as older materials become unavailable. Repair methods are often plant-specific due to differences in structures and safety priorities, highlighting the need for advanced sensors, remote sensing, and NDE techniques to identify structural issues. Current standards on repairs are limited, necessitating updates and improved adaptability, including leveraging newer repair codes like the American Concrete Institute (ACI) repair code.
Key Research Recommendations:
Investigation of Repair Materials: Conduct research to understand and validate the use of new repair materials and technologies in NPPs and work toward their acceptance in industry codes and standards.
Forum for Sharing Best Practices: Establish a forum to share experiences, long-term performance data, and best practices regarding repair materials and methods used in nuclear facilities.
2-61 Database Creation: Develop a comprehensive database of repair techniques, materials, and their performance to inform future repair strategies and updates to codes and standards.
Adaptation to New Cement Standards: Research the implications of transitioning from C 150 to C 595 and LC 3 cement for repair strategies, ensuring compatibility with new cement types.
Plant-Specific Repair Approaches: Develop guidelines for creating plant-specific repair and mitigation strategies, considering unique conditions, structures, and safety requirements.
Modern Techniques for Monitoring: Investigate modern techniques like sensors, remote sensing, and NDE to detect unusual structural changes and aid in timely repairs.
Enhancement of Repair Codes: Improve and adapt existing repair codes (e.g., ACI repair code) to accommodate new materials, technologies, and industry needs 2.9.3 Session Summary Radiation Effects on Structures: Key focus areas included understanding the effects of fluence and gamma radiation, characterizing irradiated concrete, exploring flux effects, size effects, structural confinement and restraining effects, addressing chemical changes like ASR, and investigating creep and relaxation in cement paste. There was also emphasis on refining modeling methodologies and irradiation transport calculations to better assess the impact of radiation on concrete and interaction between concrete and embedded steel (e.g.,
reinforcements, structural steel, anchorages).
The importance of real operational data was highlighted, as it would help validate findings from accelerated tests, reduce uncertainties, and enhance simulations of structural performance. The discussion underscored the necessity of plant-specific evaluations and harvesting irradiated components from operating environments to provide valuable insights into degradation processes. Overall, the panel agreed on continued research efforts to improve safety and extend the longevity of nuclear concrete structures.
Aging of Post-tensioned Containments: The panel discussion focused on research priorities for understanding and addressing the effects of creep, shrinkage, and coupled thermal stress in post-tensioned concrete containments. Key considerations included the processes of retensioning and detensioning tendons, the impact of plant modifications over time, and stress reversalsparticularly transitions involving tension. Emphasis was placed on the need for research into creep under mixed stress states, transient creep during accidental scenarios, and the reactivation of primary and inelastic residual creep due to changes in loading.
The discussion also highlighted differences in containment structures, such as tendon configurations, and the importance of obtaining specific creep data for various load patterns like bending. Detailed monitoring and instrumentation were noted as valuable tools for understanding deformation and enhancing containment performance and integrity. The panel agreed on the necessity of continued research to ensure the safety and longevity of post-tensioned concrete containment structures.
2-62 Coupled Degradation Mechanisms: The panel emphasized the importance of understanding coupled degradation processes, such as freeze-thaw cycles paired with chloride-induced corrosion and demineralized water exposure. Carbonation was highlighted as an understudied degradation mechanism, especially in combination with other factors. While laboratory experiments are valuable, field testing was recognized as crucial for accurately assessing degradation.
The synergistic effects of ASR and cracking, which increase the ingress of corrosive agents, is another area for study. Accelerated experiments were acknowledged as challenging, and field observations were deemed essential for validating laboratory findings. Research priorities included adapting generic models to specific plant needs, studying mechanisms in operating plants, and leveraging OE alongside data from decommissioned plants.
Continued efforts to deepen knowledge of coupled concrete degradation and validate findings with real-world data were seen as critical for ensuring the safety and longevity of nuclear concrete structures.
Enhance Inspection and Monitoring: Current visual inspections in the nuclear industry are effective for detecting structural degradation, but newer technologies like drones, remote sensing, AI, and BIM are essential to address degradation mechanisms that precede visible signs. These technologies can enhance data collection, enable comparisons over time, and improve precision. Structural health monitoring, common in other infrastructures, could be adapted for nuclear plants despite limited use currently.
Adopting advanced tools like digital twins and robotics may help reduce costs and reliance on personnel, particularly in new reactors. However, the existing fleet faces challenges with historical data gaps and system integration. Robotics could also support knowledge transfer from experienced inspectors to newer staff.
Repair and Replacement Strategies: Research on repair materials shows promise for nuclear applications but faces limitations due to material availability, engineering changes, and code requirements. Earthquake retrofitting experience from Japan and experimental data can aid improvements for performance under seismic loading, with adaptations for nuclear environments. A proposed database could share repair performance data to refine material usage and update codes. Changes in cement standards and plant-specific needs require advanced monitoring techniques. Current repair codes remain limited, calling for updates and better adaptability.
3-1 3 RESEARCH STRATEGY FOR LONG-TERM OPERATION 3.1 Key Workshop Takeaways For each session, recommendations for future research were identified for each presentation and for the panel discussion. The numerous recommendations captured for each session emphasize that there remain many unanswered questions related to material performance during LTO. In this section of the report, the NRC staff attempts to identify the most significant of these recommendations to consider for future research to support LTO beyond 80 years.
3.2 Motivation and Objective The principal objective of the workshop was to collect ideas, from subject matter experts representing a diverse range of nuclear stakeholders, for research topics that should be considered to support the very LTO (i.e., > 80 years) of existing LWRs. Results from such research topics would augment existing technical knowledge and OE insights to inform the aging management programs during this timeframe. Research topics that study well-known phenomena out to conditions associated with LTO and the consideration of new phenomena not previously expected because of expected long incubation times are of interest.
The workshop was structured to solicit such potential research topics without prejudging or censoring them. However, it is recognized that the potential safety significance among the research topics will vary. The knowledge of the research topics and hence the uncertainty of their significance also varies. Unfortunately, practical resource constraints dictate that not all of the research topics identified in the workshop can be pursued. The NRC is interested in identifying research topics that have the greatest potential safety significance combined with the least amount of knowledge or, similarly, the greatest amount of uncertainty. These are the research topics that provide the most potential safety benefits if they are pursued. Some research topics may have lower safety significance but could be vitally important effective asset management. While such research is valuable, it is outside of the scope of NRCs regulatory framework.
The objective of this section is to systematically evaluate the research topics identified during the workshop. A systematic Phenomena Identification and Ranking Table (PIRT) process is used to uniformly evaluate research topics and identify ones that should be considered for further study. The NRC plans to work with industry and other government partners (e.g., DOE) to further explore such topics to determine the optimal implementation approach.
3.3 Approach Each workshop session identified research topics or ideas that are evaluated in this section.
Ideas were obtained from individual presentations and from the panel sessions. The session summaries in Chapter 2 delineate these topics. Many topics explore common or overlapping phenomena. The first evaluation step is to identify and consolidate like topics and ensure that the evaluation scope comprehensively captures the range of issues identified during the workshop. The rationale used to consolidate topics is explained, as appropriate.
Next, separate PIRT teams to evaluate the research topics for each session were assigned.
Unlike the workshop, each PIRT team were fully internal NRC. The team consisted of the session facilitator and at least three NRC subject matter experts. If appropriate, the facilitator
3-2 could also serve as a subject matter expert. Table 3-1 lists the PIRT panel members for each session. The experts collectively met to discuss each research topic to ensure a uniform understanding of the associated scope and objectives. Each expert then individually assessed each topic through a structured PIRT evaluation process (described in Section 3.3.1). The evaluation team then met again to collectively review their individual rankings and discuss the rationale supporting their evaluation. The primary goal of this meeting was to foster debate and exchange differing points of view. Each expert could then, if desired, adjust their assessments based on additional consideration stemming from the collective discussion. However, consensus evaluations were not sought. The PIRT scores for each topic reflect the average of the three individual evaluations, as buttressed by their supporting rationale. The difference, or standard deviation, associated with the scores is also indicated.
Table 3-1 PIRT Panel Members Session RES Coordinator PIRT Panel Members 1
Jeff Poehler Jeff Poehler, On Yee, Dave Rudland 2
Austin Young Jeff Poehler, John Tsao, Steven Levitus 3
Eric Focht Eric Focht, Jay Collins, Seung Min, Pat Purtscher, Andy Johnson 4
Mekonen Bayssie Andy Johnson, Rob Tregoning, Pat Purtscher 5
Rob Tregoning Rob Tregoning, Brian Alick, John Wise 6
Rob Tregoning Rob Tregoning, Yuken Wong, Chakrapani Basavaraju 7
Mekonen Bayssie Carol Moyer, Rob Tregoning, Amy Hull, John Wise 8
Madhumita Sircar Madhumita Sircar, George Wang, Andrew Prinaris 3.3.1 PIRT Scoring The traditional PIRT process is well established and has been used in many industries, including nuclear, for ranking and prioritizing issues. The PIRT process provides a systematic means of obtaining information from experts and involves generating lists of phenomena, where "phenomena" in this effort is associated with the research topics identified during the workshop, and then evaluating each phenomenon against a relevant figure-ofmerit related to reactor safety.
The PIRT scoring followed in this current evaluation is a modification of the process used to support the Proactive Material Degradation Assessment (PMDA) [68] and the Expanded Material Degradation Assessment (EMDA) [15.]. These processes were used to identify and evaluate phenomena associated with materials degradation beyond 40 years of operation (PMDA) and beyond 60 years of operation for an expanded numbers of systems, structures, and components (EMDA).
3-3 The PIRT scoring criteria consisted of evaluating the Importance, Knowledge, and Uncertainty scores associated with each research topic pertaining to 80-years or more of operation1. The Importance score rates the potential safety implications associated with each topic, on a scale from low (L) (minimal potential safety significance), to medium (M) (moderate potential safety significance) to high (H) (potential for high safety significance).
Table 3 details the importance criteria. The Uncertainty score measures each experts personal uncertainty in their Importance ranking, on a scale from low (L) (low uncertainty), to medium (M)
(moderate uncertainty), to high (H) (high uncertainty). Table 3-3 details the Uncertainty scoring criteria.
Finally, the Knowledge score rates each experts current belief about how thoroughly that research topic is understood through prior laboratory studies and/or OE, on a scale from unknown (UK) (poor understanding, little and/or low confidence in existing data/information), to partially known (PK) (partial knowledge of research topic), to known (K) (extensive, consistent data/information associated with that topic). Table 3 describes the Knowledge scoring criteria.
The L, M, and H scores were then converted to numerical scores with L=1, M=2, H=3. For the Knowledge scores, higher knowledge (K) was scored lower (K=1), partial knowledge (PK) =2, and unknown (UK) =3.
Table 3-2 PIRT Importance Scoring Criteria Rank Definition Points High (H)
Research topic has a controlling/strong potential safety significance that could be mitigated through knowledge gained by further research 3
Medium (M)
Research topic has a moderate potential safety significance that could be mitigated through knowledge gained by further research 2
Low (L)
Phenomenon has a minimal potential safety significance, and no research is needed to mitigate its effects 1
1 The PMDA and EMDA phenomena were degradation scenarios, and the scoring criteria were susceptibility (to the degradation scenario), confidence (in the susceptibility score), and knowledge. This evaluation changed the susceptibility criteria to risk significance because the research topics are broader than degradation scenarios.
Uncertainty was substituted for confidence so high uncertainty could more directly correlate with high-risk significance.
3-4 Table 3-3 PIRT Uncertainty Scoring Criteria Rank Definition Points High (H)
Potential safety significance ranking has high uncertainty due to insufficient expertise or understanding of potential effects of associated degradation 3
Medium (M)
Potential safety significance ranking has moderate uncertainty due to insufficient expertise or understanding of potential effects of associated degradation 2
Low (L)
Potential safety significance ranking has minimal uncertainty and is well established based on current understanding and operating experience 1
Table 3-4 PIRT Knowledge Scoring Criteria Rank Meaning Points Known (K)
Sufficient understanding of research topic to assess practical potential safety significance and associated ramifications 1
Partially Known (PK)
Partial knowledge and understanding of research topic to assess potential safety significance and associated ramifications 2 Unknown (UK)
Totally unknown or very limited knowledge of potential safety significance and associated ramifications 3
Using this process, the average Importance and average Knowledge scores can be plotted versus each other for each research topic. An example plot of Knowledge versus Importance is shown in Figure 3-1. The left side of the plot with the darker shading is indicative of low Knowledge, while the lighter shading on the right side of the plot is indicative of high Knowledge.
The labeled areas in the corners of the plot indicate the high Knowledge, low Safety Significance; high Knowledge, high Safety Significance; low Knowledge, high Safety Significance; and low Knowledge, low Safety Significance areas. Low Knowledge scores (>2) coupled with high importance scores (>2) are indicated by the red shading in and represent research topics with knowledge gaps that could be safety-significant during LTO. Conversely, High Knowledge scores (<1) coupled with low importance scores (<2) (green in Figure 1.1) are judged as relatively well understood and have low potential safety significance. Other combinations of Knowledge and Safety Significance fall between the cases listed above on the figure. Moving from upper right to lower left can be accomplished via additional research and development to better understand the research topic and its associated potential safety implications.
The Uncertainty scores are not directly plotted on the graphs but are represented by error bars. The corresponding recommendation number is indicated next to the marker in the plot. The average
3-5 uncertainty scores are converted to symmetrical error bars in the figures by assuming that the highest uncertainty (i.e., value of 3) corresponds to an importance uncertainty of 0.5 (i.e., 50% of a complete level). Also, the knowledge scores are slightly offset in some cases so that both the marker and the error bars associated with each recommendation number are clear. An example of a point representing a recommendation with error bars is included in the upper right quadrant of Figure 3-1.
The complete set of PIRT scoresheets for all sessions can be found in ADAMS Package No.
ML25183A340 or in APPENDIX C.
Figure 3-1 Schematic illustrating the combinations of importance and knowledge scores suggesting various life management responses 3.4 PIRT Evaluations for Individual Sessions 3.4.1 Session 1, Reactor Pressure Vessel, PIRT Evaluation 3.4.1.1 Initial Inputs to PIRT Scoring Twenty-seven recommendations were identified from the three presentations and the panel discussion for Session 1. After combining recommendations with similar themes, a final list of recommendations was generated for input to the PIRT process. To facilitate the combination process, the original recommendations were sorted into four groups:
material understanding, analysis techniques, mitigation, and monitoring.
The following table shows the original list of research recommendations and the consolidated set of recommendations.
Table 3-5 PIRT Inputs for Session 1, Reactor Pressure Vessel
3-6 Category Recommendation Combined Material Understanding Characterization techniques: The evolution of characterization techniques such as TEM, FIM, APT and PAS is highlighted.
Ensure that characterization techniques such as TEM, FIM, APT and PAS, continue to evolve.
Material Variability: Need to understand the importance of material variability and ensure valid, reproducible data, particularly wrt microstructure and properties.
Materials evaluated should be representative of operating reactor materials and should be evaluated for high-dose conditions.
New Production Techniques: The potential impact of new production techniques on material properties and the importance of understanding material specifications are discussed.
Real Materials Evaluation: The need for evaluating real materials through appropriate sampling and high-dose neutron irradiated materials is emphasized. There is a need for robust and reliable mechanical properties data, Low-temperature irradiation: The significance of low-temperature irradiation data for support structures and SMRs is mentioned, with reference to the ATR2 program led by Bob Odette.
Reactor Pressure Vessel Supports: Harvesting of reactor pressure vessel support materials could be beneficial for improving embrittlement predictions for these materials.
Reactor Pressure Vessel Supports: Harvesting of reactor pressure vessel support materials could be beneficial for improving embrittlement predictions for these materials.
Parallel Modeling Efforts: There is a need for parallel modeling efforts at the metal grain scale to replicate radiation-induced embrittlement.
Develop improved physics-based models for embrittlement.
Empirical Radiation State: Use empirical radiation state between microstructure change and
3-7 Category Recommendation Combined mechanical property change for future modeling.
Data Collection: More data of an appropriate kind is always needed, especially for low-copper, low-phosphorous materials, which may not be well-represented by current embrittlement trend curves.
Reworded 22: Generate additional embrittlement data on low-copper, low-phosphorus materials, which may not be well-represented by current embrittlement trend curves.
Analysis Techniques Enhance ASTM E90015: Improve and enhance the existing ASTM E90015 model to better predict embrittlement trends at high fluence levels.
Improve existing embrittlement trend curves, such as ASTM E 900, and incorporate improvements into regulatory guidance.
Update Regulatory Guides: Revise regulatory guides, such as Reg.
Guide 1.99, to incorporate new data and improved accuracy at higher fluence levels.
PT Curves: Potential limitations beyond 80 years could be due to PT curves, which narrow the operating window for heat-up and cooldown.
Direct fracture toughness methods: The industry is moving toward direct fracture toughness methods, which could impact the annealing process. This approach allows for the expansion of surveillance programs and provides opportunities to assemble surveillance placements that follow embrittlement after annealing.
Improve direct fracture toughness methodologies.
Fracture Toughness: Emphasis on the importance of fracture toughness backed by measurements to inform predictive embrittlement.
Better Fracture Modeling: Need for better fracture modeling on many
3-8 Category Recommendation Combined levels to understand scatter and correlations.
Direct Fracture Toughness:
Improved methods of fracture toughness measurements are helpful for LTO.
Regulatory Updates: Existing regulations and guidelines, such as Reg Guide 1.162, need to be updated to account for new findings related to nickel manganese clusters and other factors affecting embrittlement.
Update existing regulations and guidance, such as RG 1.162, to account for new findings related to nickel-manganese clusters and other factors affecting embrittlement.
Mitigation Effectiveness of Annealing: The effectiveness of annealing is not solely determined by how much the transition temperature can be recovered but by how long the reactor can operate afterward. The rate of reembrittlement is crucial.
Studies of reembrittlement after annealing to support LTO beyond 80 years is needed to support use of annealing for mitigation of embrittlement beyond 80 years.
Reembrittlement studies: There is a need for systematic studies on the thermal stability of nickel-manganese clusters and how they respond to annealing. Most of the previous research focused on copper precipitates, which are less relevant for material that will operate for 80 years.
Technological Challenges: There are two types of annealing: wet and dry. Wet annealing is simpler but less effective, while dry annealing is more complex and requires the removal of core internal structures and water. Both methods have their own set of challenges and benefits.
Additional research is recommended on the use of annealing to support LTO beyond 80 years, including determining whether annealing would be beneficial, and which type(s) of annealing should be pursued.
Future Research: To apply annealing for up to 80 years of operation, research needs to start now to ensure the current fleet of
3-9 Category Recommendation Combined reactors can benefit from these techniques.
Mitigation Techniques: Greater use of mitigation techniques such as thermal annealing or flux reduction may be required.
Monitoring Surveillance Programs: The importance of surveillance programs to provide critical data on high-dose irradiations, especially for materials relevant to nuclear plants, is emphasized.
Additional surveillance testing is needed to generate high-fluence embrittlement data. This could involve either testing existing high-fluence specimens, installing new capsules/specimens for high-fluence irradiation, or both.
Generate High-Fluence Data:
Continue PSSP to gather high-fluence surveillance data from actual RPV materials irradiated in commercial PWRs.
Additional Surveillance Capsules:
Consider the need for additional surveillance capsules for plants expected to reach high fluence levels, ensuring comprehensive data collection for LTO.
Monitoring Specimens: Lack of monitoring specimens for future LTO.
There is a need to ensure that there are sufficient surveillance specimens to monitor embrittlement during LTO.
3.4.1.2 Final List of Recommendations for Input to the PIRT Process RPV1 Ensure that characterization techniques such as TEM, FIM, APT and PAS, continue to evolve.
RPV2 Materials evaluated should be representative of operating reactor materials under high-dose conditions.
RPV3 Reactor Pressure Vessel Supports: Harvesting of reactor pressure vessel support materials could be beneficial for improving embrittlement predictions for these materials, at low temperatures.
RPV4 Develop improved physics-based models for embrittlement.
3-10 RPV5 Generate additional embrittlement data on low-copper, low-phosphorus materials, which may not be well-represented by current embrittlement trend curves.
RPV6 Investigate phosphorus segregation levels with respect to intergranular failure.
RPV7 Improve existing embrittlement trend curves, such as ASTM E 900, and incorporate improvements into regulatory guidance.
RPV8 Improve direct fracture toughness methodologies.
RPV9 Update existing regulations and guidance, such as RG 1.162, to account for new findings related to nickel-manganese clusters and other factors affecting embrittlement.
RPV10 Studies of reembrittlement after annealing to support LTO beyond 80 years are needed to support use of annealing for mitigation of embrittlement beyond 80 years.
RPV11 Additional research is recommended on the use of annealing to support LTO beyond 80 years, including determining whether annealing would be beneficial, and which type(s) of annealing should be pursued.
RPV12 Additional surveillance testing is needed to generate high-fluence embrittlement data. This could involve either testing existing high-fluence specimens, installing new capsules/specimens for high-fluence irradiation, or both.
RPV13 There is a need to ensure that there are sufficient surveillance specimens to monitor embrittlement during LTO.
3.4.1.3 Summary of PIRT Results The complete scoring by the panel and the rationale for the scoring is contained in Session 1 PIRT Scoring - RPV. Figure 3-2 shows the PIRT results for Session 1.
3-11 Figure 3-2 PIRT Results for Session 1, RPV The recommendations fall into several distinct categories. The first category (higher importance, less knowledge), with points falling within the upper right quadrant, corresponds to those recommendations where research could be used to improve knowledge on a topic that the PIRT panelists expect to be important to safety during LTO. Recommendation RPV3, related to harvesting RPV support materials to assess embrittlement, is the only recommendation in this category. Panelists comments were somewhat split with two seeing the benefit of such research, and one questioning the practical application of the results. The uncertainty associated with RPV3 is also high, which could potentially increase the safety significance of this topic.
Recommendations RPV2, -7, -8, and -12, located in the upper left quadrant, are recommendations with relatively high importance (safety significance), but a higher level of knowledge, where further research will not be as beneficial in improving safety. RPV7 and RPV12, however, are located closest to the upper right corner of the graph, thus are considered slightly higher priority based on research being able to improve safety. For RPV7, related to improvements to ETCs, two panelists ranked high in importance and one medium; however, panelists perceived a relatively high knowledge level in this area as well. Two panelist rationales noted that they believe it is necessary/important to improve ETCs to reduce nonconservatism at high neutron fluences. For RPV12, related to generating additional RPV surveillance data at high fluences, two panelists also scored high in importance and one scored medium; however, panelists also rated the knowledge level fairly high. It was noted that more high-fluence data is needed to support the development of improved ETCs, so this recommendation is closely related to RPV7. Panelists did note existing industry programs to generate high-fluence data. Panelists also noted that RPV2, related to evaluating representative materials under high-dose conditions, was somewhat redundant with RPV12.
RPV8 is related to improving direct fracture toughness methodologies, and received varying importance scores from M/L to H. Also, panelists perceived a moderately high knowledge level in this area.
3-12 Recommendations RPV6, -10, and -11 fall in the lower right quadrant, indicating lower knowledge and also lower importance (safety significance). RPV10 and -11 both are related to additional studies of annealing of RPVs to support operating beyond 80 years. Panelist rationales generally indicated that the panelists do not believe annealing will be necessary or desirable to support operating beyond 80 years. It is also noted that for these three recommendations, the knowledge average score is only slightly greater than 2, so the knowledge level is not terribly low.
Recommendations 1, 4, 5, 9 and 13 fall into the lower left quadrant, indicating lower importance (safety significance) coupled with higher knowledge. Additional research on these topics would be less likely to significantly improve safety due to the lower safety significance and higher knowledge. RPV1 was related to improvement of laboratory characterization techniques, which the panelists noted was of low safety significance. RPV4 related to improved physics-based models, and panelists comments expressed skepticism about the need or likelihood of success of such models. RPV5 was to generate additional embrittlement data on low-copper and low-phosphorus materials; panelists questioned the importance to the U.S. fleet of such data. RPV9 was related to updating NRC guidance for annealing in RG 1.162 to account for new findings related to NiMn clusters (e.g. late-blooming phases). Panelists all rated this low in importance and rationales indicated that the panelists do not believe late-blooming phases will become significant factor in embrittlement of U.S. RPV steels. Also, as in the case of RPV10 and -11, panelists do not foresee that annealing would be necessary to support operation beyond 80 years. RPV13 was to ensure there are sufficient surveillance specimens to monitor embrittlement during LTO. Panelists were split on the importance of this recommendation and rated the knowledge level fairly high. Rationales also noted RPV13 may be redundant with RPV12.
3.4.1.4 Summary of Session 1 PIRT Scoring for RPV Components The PIRT panel scored 13 potential research topics identified during Session 1 on RPV components. Of the 13 topics scored, the panel identified three topics that could benefit from additional research to help improve our knowledge on safety-significant topics and reduce uncertainty to help clarify the importance to safety. These topics are:
RPV3 Harvesting of reactor pressure vessel support materials could be beneficial for improving embrittlement predictions for these materials.
RPV7 Improve existing embrittlement trend curves, such as ASTM E 900, and incorporate improvements into regulatory guidance.
RPV12 Additional surveillance testing is needed to generate high-fluence embrittlement data.
This could involve either testing existing high-fluence specimens, installing new capsules/specimens for high-fluence irradiation, or both.
3-13 3.4.2 Session 2, Reactor Pressure Vessel Internals, PIRT Evaluation Table 3-6 PIRT Inputs for Session 2, Reactor Pressure Vessel Internals Categories Research Topic Combined Recommendations Material Characterization and Data Integrity Improved Characterization of Initial Manufacturing Conditions Develop comprehensive baseline datasets for legacy and current reactor materials, integrating fabrication records, microstructural characterization, and degradation indicator benchmarking.
Expanded Use and Validation of Degradation Indicators Knowledge Retention and Documentation Systems Establish centralized knowledge retention systems and best practices for integrating expert judgment with machine learning tools in aging management.
Transparent Use of Machine Learning and AI Modeling and Simulation Advanced Modeling Tools to Reduce Extrapolation Develop integrated multimechanism degradation models that reduce reliance on extrapolation and simulate complex interactions among SCC, fatigue, thermal aging, and irradiation effects.
Understanding Combined Degradation Mechanisms Characterize Synergistic Effects Among Degradation Mechanisms Transmutation-Induced Phase Instability Advance mechanistic models and experimental studies to quantify the effects of irradiation, transmutation, and gas accumulation on phase stability and mechanical degradation in reactor materials.
Irradiated Material Behavior and Phase Stability Improve Understanding of Transmutation-Induced Phase Instability Clarify the Role of Helium and Hydrogen in Degradation Mechanisms Evaluate Localized Transmutation Effects Near Thermal Interfaces
3-14 Categories Research Topic Combined Recommendations Determine Effects of Irradiated Microstructures at Elevated Temperatures Identify thresholds and environmental conditions leading to nonlinear property changes, focusing on high fluence and long-term exposure environments Assess Thresholds for Radiation-Induced Phase Transformations Develop Predictive Models for Nonlinear Material Degradation Radiation Dose Accumulation and Long-Term Material Stability Material Harvesting, Post-Irradiation Examination (PIE) &
Validation Validation of Accelerated Testing Using Harvested Materials Design coordinated material harvesting and PIE campaigns on service-exposed components, especially reactor internals, to validate accelerated testing assumptions and improve degradation modeling.
Focused Research on Reactor Internals (e.g., Weld and Base Metals)
Post-Irradiation Examinations on Service-Exposed Materials Development of Better Harvesting Strategies and Guidelines Corrosion, Environmental, and Wear-Related Effects Understanding and Managing Wear and Low-Profile Degradation Mechanisms Quantify the effects of microstructural and environmental changes on corrosion initiation, propagation, and wear in light-water reactor conditions Investigate Effects of Radiation-Induced Microstructural Evolution and Corrosion Mechanistic Understanding of IASCC Conduct targeted experiments to evaluate and
3-15 Categories Research Topic Combined Recommendations Stress Corrosion Cracking and IASCC Mitigation Advanced Materials with Engineered Microstructures mitigate IASCC using post-irradiation annealing, advanced alloys, and powder metallurgy.
Post-Irradiation Thermal Annealing Treatments Optimization of Powder Metallurgy for Core Internals Industry Integration and Deployment Support regulatory and industry deployment through guidance on qualification and licensing of advanced materials and repair methods.
Inspection Monitoring and Component Repair Development of Inservice Inspection Techniques Develop next-generation NDE techniques for early-stage damage detection that are operable in high-radiation environments.
Advancing Repair and Mitigation Techniques Advance repair and replacement strategies for irradiated components that align with ALARA principles and long-term safety needs.
Strategic Research Integration for LTO Collaboration Between National Labs and Industry Establish collaborative mechanisms across national labs, regulators, and industry to prioritize and deploy impactful research outcomes Assessing Feasibility of 100Year Reactor Operation 3.4.2.1 Final List of Recommendations for Input to the PIRT Process RVI1 Develop comprehensive baseline datasets for legacy and current reactor materials, integrating fabrication records, microstructural characterization, and degradation indicator benchmarking.
RVI2 Establish centralized knowledge retention systems and best practices for integrating expert judgment with machine learning tools in aging management.
RVI3 Develop integrated multimechanism degradation models that reduce reliance on extrapolation and simulate complex interactions among SCC, fatigue, thermal aging, and irradiation effects.
3-16 RVI4 Design coordinated material harvesting and PIE campaigns on service-exposed componentsespecially reactor internalsto validate accelerated testing assumptions and improve degradation modeling.
RVI5 Quantify the effects of microstructural and environmental changes on corrosion initiation, propagation, and wear in light-water reactor conditions RVI6 Advance mechanistic models and experimental studies to quantify the effects of irradiation, transmutation, and gas accumulation on phase stability and mechanical degradation in reactor materials.
RVI7 Identify thresholds and environmental conditions leading to nonlinear property changes, focusing on high fluence and long-term exposure environments RVI8 Conduct targeted experiments to evaluate and mitigate IASCC using post-irradiation annealing, advanced alloys, and powder metallurgy.
RVI9 Support regulatory and industry deployment through guidance on qualification and licensing of advanced materials and repair methods.
RVI10 Develop next-generation NDE techniques for early-stage damage detection that are operable in high-radiation environments.
RVI11 Advance repair and replacement strategies for irradiated components that align with ALARA principles and long-term safety needs.
RVI12 Establish collaborative mechanisms across national labs, regulators, and industry to prioritize and deploy impactful research outcomes
3-17 3.4.2.2 Summary of PIRT Results for Session 2 Figure 3-3 PIRT Results for Session 2, RVI The complete scoring by the panel and the rationale for the scoring is contained in Session 2 -
PIRT scoring-RVI. The PIRT panel evaluated 12 research topics related to Reactor Vessel Internals (RVI) during Session 2. These topics were scored based on their importance to safety, the level of existing knowledge, and the degree of uncertainty associated with each topic. The recommendations were then categorized based on whether additional research could help reduce uncertainty, improve knowledge, or clarify the safety significance of the issues. The results are summarized below according to their research priority categories. Figure 3-3Figure 3-3 shows the PIRT results for Session 2.
The first group of recommendations, RVI3, 6, and 9, were voted to be of high importance and low knowledge representing topics where additional research could significantly reduce knowledge gaps in areas important to safety.RVI3 was viewed as important to enhancing our ability to forecast component behavior under extended operational periods. While individual degradation mechanisms have been studied, the integration of these mechanisms into a comprehensive model, especially under synergistic environmental and irradiation effects, is still underdeveloped. Additional research would help establish more robust predictive tools to support regulatory decisions. However, it was also stated that this will prove to be especially difficult to simulate and may not provide valuable return on investment in a timely manner. While modeling has improved, for RVI6 current methods may not fully account for spatial variability and spectral changes in neutron flux, especially in BWR and PWR environments. Specifically, the panel noted that transmutation due to irradiation has not been investigated much, and these
3-18 second-order effects could potentially become important during LTO. More refined modeling approaches, validated by inservice data, could help bridge current uncertainties and better support materials performance predictions. The panel acknowledged that efforts are already underway to prepare for the qualification and licensing of advanced materials particularly in response to the Section 401 provisions of the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy (ADVANCE) Act of 2024. The work for RVI9 continues to be an area that the panelists feel will be critical to both the licensing of new and advanced reactors, as well as operating reactors entering extended periods of operation.
RVI1, 4, 5, and 7 are considered important enough that further research could improve confidence and reduce uncertainty. The panelists determined that these topics all pose an above average level of importance and additional knowledge would prove valuable. While useful, RVI1 was not considered safety-significant on its own. The panelists acknowledged that efforts to collect and retain reactor vessel component information have been established and this may not need to be an effort taken up by research, however, ensuring a database is maintained can prove to be valuable. The panelists further suggested that industry may be better candidates to take this effort on. The work for RVI4 would provide real-world insight into the degradation of RVI components by examining materials affected by operational conditions that are difficult to replicate in laboratories. However, panelists were concerned that factors such as material availability and cost may create limitations in the feasibility of harvesting efforts.
RVI5 and RVI7 were similar in looking at the effects of environmental changes on degradation mechanisms. Both are considered to have relatively high importance with limited existing knowledge. While not as uncertain as some previous recommendations, given that property changes and crack formation/propagation have been observed and monitored over shorter operational durations, there is still a need to better understand potential degradation caused by higher fluences and long-term exposure. As operational time increases, it becomes increasingly important to determine whether more severe degradation develops from the likes of void swelling, IASCC, and other related mechanisms.
Acting on recommendations RVI8 and 10 could have a positive effect on safety significance, but the panel believes the level of required research is modest. RVI8 could have been included in the previous section due of the high degree of uncertainty. However, the importance was considered low because plant owners are more likely to forgo IASCC mitigation and internals component replacement for managing degradation through the same monitoring and evaluation methods that have been used in the past. For RVI10, there was agreement that advancing NDE tools capable of early flaw detection would prove valuable. However, there was uncertainty as to the relative importance compared to other recommendations considering NDE tools currently exist and there are some active efforts to improve NDE.
RVI2, 11, and 12 were all determined to either be sufficiently understood or of a low enough priority where they could be addressed through measures other than research. The topic of improved repair and replacement strategies was determined to be generally valuable but not a research topic to act on. The panel noted the value of knowledge management and strengthened NRC/industry/laboratory partnerships all have immense value but emphasized that this is more of a programmatic or administrative priority than a research need.
Key Research Recommendations Out of the 12 RVI topics scored, the panel identified the following as the highest priority areas where additional research could significantly improve safety understanding and reduce uncertainty:
3-19 RVI3: Develop degradation models that can better reduce reliance on extrapolation and better predict complex interactions between SCC, fatigue, thermal aging, and irradiation effects.
RVI4: Set up coordinated material harvesting efforts to collect and test materials that have been inside reactors for years. Focus on the most affected components to compare real-life changes with lab predictions.
RVI6: Create models and establish experimental studies to quantify the effects of irradiation, transmutation, and gas accumulation on phase stability and mechanical degradation.
RVI7: Identify the thresholds and environmental conditions which lead to nonlinear property changes, focusing on high fluence and long-term exposure.
RVI9: Accelerate guidance on qualification and licensing of advanced materials and repair methods.
These research activities are expected to provide high regulatory value, especially as the industry moves toward subsequent license renewal and faces increasing challenges related to extended service life of RVI components.
3.4.3 Session 3, Reactor Coolant Pressure Boundary (RCPB) Components, PIRT Evaluation 3.4.3.1 Initial Inputs to the PIRT Process Twenty-four recommendations were identified from the three presentations and the panel discussion for Session 3. After combining recommendations with similar themes, a final list of 15 recommendations was generated for input to the PIRT process. To facilitate the combination process, the original recommendations were sorted into four groups: mechanistic understanding and other mechanisms (i.e., mechanisms other than SCC), materials, modeling and effects parameters and maintaining expertise. After evaluating the two items under maintaining expertise, it was decided that maintaining expertise was a motivation for performing research, but not necessarily a research area. Thus, maintaining expertise was eliminated as a category.
The following table shows the original list of research recommendations and the combined set of recommendations.
Table 3-7 PIRT Inputs for Session 3, Reactor Coolant Pressure Boundary Components Category Initial Recommendations Combined Final Recommendations Mechanistic Understanding of SCC & Other Mechanisms Mechanistic Understanding:
There is a need for a better mechanistic understanding of the SCC initiation process in Alloy 600/82/182 and Alloy 690/52/152. This includes studying the early stages of Research on PWSCC initiation mechanisms should be conducted.
3-20 Category Initial Recommendations Combined Final Recommendations intergranular oxide growth and the effects of stress, surface finish, and heat-toheat variability These items apply generally to both Alloy 600 and Alloy 690 base and weld metals.
Items of particular interest include:
- The early stages of intergranular oxide growth
- The effects of stress
- The effect of surface finish
- Heat-toheat variability
- Creep cavity formation in lower-CW materials
- Creep cavity formation at lower temperatures
- The effect of grain boundary carbides.
These items most likely do not act independently toward contributing to PWSCC initiation. Thus, mechanistic research should involve understanding dependencies.
Degradation Mechanisms:
There is a need for a comprehensive understanding of the underlying mechanisms of inservice degradation.
Creep-Induced Cavities:
Research on Alloy 690 has revealed that creep-induced grain boundary cavities can lead to microscopic cracking initiation. Further studies are needed to understand this mechanism in lower cold work materials at lower temperatures Grain Boundary Carbides:
The role of grain boundary carbides in SCC initiation needs further investigation.
Service experience suggests they improve crack propagation resistance, but systematic data shows no obvious effect.
Surface Finish and Crack Initiation: Research is needed to understand the effect of surface finish on crack initiation. This involves testing on both polished and realistic surface conditions to ensure accurate predictions Fatigue and Corrosion Fatigue: Emerging issues like fatigue and corrosion fatigue need to be addressed as Degradation mechanisms such as fatigue and corrosions fatigue will likely continue to be of concern, especially as plants age.
Research may be warranted
3-21 Category Initial Recommendations Combined Final Recommendations reactors undergo more cycles over extended lifetimes.
if these mechanisms begin operating in unexpected areas or under unexpected conditions.
Materials Continue research on RCPB materials: RCPB materials are not immune to EC, and there is no threshold effect (i.e., some parameter level below which EC does not occur). One area of concern includes the effect of using improved weld metals which might transfer the most susceptible microstructure to the HAZ. Also, attention is needed on the effects of plant-relevant surface conditions and dynamic strain The research areas identified here are included in other parts of the table.
Testing Plant Materials:
Research and testing on inservice materials and potential alternatives (e.g.,
XM19, A725) are necessary.
This includes having knowledge of how components were fabricated and operating conditions.
Access to plant records would be very beneficial.
Evaluating representative mockups may also help understand component service performance Research on actual plant materials would be beneficial for helping to validate aging management programs, inspections intervals and predictive models. This would require committing to explant materials harvesting efforts.
Validating Aging Programs with Plant Materials: There is a need to validate current aging programs using materials from actual plants.
This involves comparing aged materials with those in service to ensure the aging programs are accurate and effective
3-22 Category Initial Recommendations Combined Final Recommendations Studying Dissimilar Metal Welds and Interfaces:
Research should focus on dissimilar metal weld interfaces and heat-affected zones, as these may exhibit hardening and carbide coarsening. Surface effects are also an important aspect to evaluate.
Research should focus on dissimilar metal weld interfaces and heat-affected zones, as these may exhibit hardening and carbide coarsening. Surface effects are also an important aspect to evaluate.
Welding Defects: The impact of preexisting welding defects on SCC initiation in high-chrome weld metals needs to be better understood. These defects may act as high-stress regions that induce creep cavities and degrade materials over LTO.
The impact of preexisting welding defects on SCC initiation in high-chrome weld metals needs to be better understood. These defects may act as high-stress regions that induce creep cavities and degrade materials over LTO.
Low alloy steel: Research on low alloy steel from temper bead welding and irradiation which can cause an elevated yield strength and EC susceptibility. Another potential area of concern is the effect of impurities on EC based on other research by other organizations where they reported a surprising sensitivity of LAS to chlorides. Also, the environmental effects have on fracture properties is a possible concern Research on degradation mechanisms associated with low alloy steel should be considered such as:
- temper bead welding and irradiation which can cause an elevated yield strength and EC susceptibility.
- environmental effects have on fracture properties Stainless Steel Piping: This research aims to provide a better understanding of SCC initiation in stainless steel piping, focusing on factors such as thermal stratification, temperature, dissolved oxygen, surface condition, Research on stainless steels to provide a better understanding of SCC initiation in stainless steel piping, focusing on factors such as:
- thermal stratification,
- temperature,
- dissolved oxygen,
3-23 Category Initial Recommendations Combined Final Recommendations long-term aging, and welding residual strength.
- surface condition, long-term aging, and
- welding residual strength.
Modeling & Effects Parameters (temp., stress, etc.)
Modeling and Simulation:
There is a need for simulation and modeling to validate assumed mechanisms and make accurate predictions about crack initiation and growth. This includes building comprehensive databases and benchmarking different models. This also includes adjusting NDE strategies and creating NDE tools to accurately assess cracking kinetics.
There is a need for modeling and simulation to validate assumed mechanisms and make accurate predictions about crack initiation and growth. This includes building comprehensive databases and benchmarking different models. This also includes adjusting NDE strategies and creating NDE tools to accurately assess cracking kinetics.
Predictive Models:
Developing more accurate predictive models for SCC initiation is needed. Older models are mostly empirical and probabilistic, focusing on limited influencing factors.
Mechanistic-based models, such as Électricité de France (EDF's) local model for Alloy 600/182, are promising but better data/knowledge is needed to better quantify/account for influencing factors Cold Work and Applied Stress: The effects of cold work and applied stress on SCC initiation time vary between different alloys.
Understanding these differences is important for developing unified models and determining the improvement factors for second-generation nickel-based alloys. This includes how cold work is applied The effects of cold work and applied stress on SCC initiation time vary between different alloys.
Understanding these differences is important for developing unified models and determining the improvement factors for second-generation nickel-based alloys. This includes how cold work is applied (e.g., forging vs tensile
3-24 Category Initial Recommendations Combined Final Recommendations (e.g., forging vs tensile strain).
strain). The effect of dynamic strain should also be evaluated.
Plant operating changes and upgrades: The effects of plant modifications and upgrades, including power uprates and low-leakage core (which may lead to SCC acceleration by thermal mixing) are potential areas of concern The effects of plant modifications and upgrades, including power uprates and low-leakage core (which may lead to SCC acceleration by thermal mixing) are potential areas of concern.
Temperature Effects: There is a gap in understanding the relationship between temperature and SCC initiation time. Laboratory testing often uses higher temperatures to accelerate data generation, but this may not accurately reflect inservice conditions. There is also some indication that temperature effects do not always follow Arrhenius behavior.
Research on how temperature affects SCC mechanisms and kinetics is needed to help develop models for both CGR and initiation. For CGR, emphasis is needed on potential microstructural changes that might occur during long-term thermal aging. For initiation, a more fundamental and broader understanding of temperature effects is needed focusing on validating kinetics and surface microstructure evolution (e.g., grain boundary oxidation, diffusion, etc.)
Understanding Long-Term Aging Mechanisms: Studying long-term aging mechanisms is important. This includes looking at secondary effects that may become important over time, such as thermally induced issue. The panel agreed that beginning thermally aging materials of interest as soon as practicable is wise and relatively inexpensive.
Acceleration Testing: There is a need to improve accelerated testing methods to avoid altering microstructures too much while still obtaining useful There is a need to improve accelerated testing methods to avoid altering microstructures too much while still obtaining useful data. This involves finding a
3-25 Category Initial Recommendations Combined Final Recommendations data. This involves finding a balance between accelerating tests and maintaining the integrity of the materials being tested balance between accelerating tests and maintaining the integrity of the materials being tested Water chemistry and mitigation: Research on the effects of oxygen in PWR make-up water and for optimal Pt distribution with OnLine NobleChem (OLNC) should be considered.
Research on the effects of oxygen in PWR make-up water and for optimal Pt distribution with OnLine NobleChem (OLNC) should be considered.
Mitigation Effects: Research may be needed to evaluate the effect surface treatments, such as peening, have on EC resistance.
Research may be needed to evaluate the effect surface treatments for mitigation, such as peening, have on EC resistance.
3.4.3.2 Final List of Potential Research Areas for Input to the PIRT Process RCPB1 Research should be continued on improving our understanding of PWSCC initiation mechanisms. These items apply generally to both Alloy 600 and Alloy 690 base and weld metals. Items of particular interest include: the early stages of intergranular oxide growth, the effects of stress on initiation mechanisms (this is separate from initiation time empirical modeling parameter effects), creep cavity formation in lower-CW materials, creep cavity formation at lower temperatures, and the effect of grain boundary carbides. These items most likely do not act independently toward contributing to PWSCC initiation. Thus, mechanistic research should involve understanding dependencies.
RCPB2 Conduct research on thermal fatigue and corrosion fatigue as these mechanisms will likely continue to operate as plants age and possibly operate in previously unexpected areas or under unexpected conditions.
RCPB3 Perform testing/research on actual plant materials to help validate aging management programs, inspections intervals and predictive models. This would require committing to explant materials harvesting efforts.
RCPB4 Research should focus on dissimilar metal weld interfaces and heat-affected zones, as these may exhibit hardening and carbide coarsening during thermal aging.
RCPB5 Investigate the impact of preexisting welding defects on SCC initiation in high-chrome weld metals. These defects may act as high-stress regions that induce creep cavities and degrade materials over LTO.
3-26 RCPB6 Research on degradation mechanisms associated with low alloy steel should be considered such as temper bead welding which can cause an elevated yield strength and environmental cracking (EC) susceptibility.
RCPB7 Research on stainless steels to provide a better understanding of SCC initiation in stainless steel piping, focusing on factors such as thermal stratification, temperature, dissolved oxygen, surface condition, long-term aging, and welding residual strain.
RCPB8 Conduct modeling and simulation to validate assumed mechanisms and make accurate predictions about PWSCC crack initiation and growth. This includes building comprehensive databases and benchmarking different models. This also includes adjusting NDE strategies and creating NDE tools to accurately assess cracking kinetics.
RCPB9 Conduct research to evaluate how cold work and applied stress affect SCC initiation time for different alloy groups (e.g., nickel-base alloys, stainless steels).
Understanding these differences is important for developing unified models and determining the improvement factors for second-generation nickel-based alloys.
This includes how cold work is applied (e.g., forging vs tensile strain). The effect of dynamic strain should also be evaluated.
RCPB10 Evaluate how plant modifications and upgrades, including power uprates and low-leakage core (which may lead to SCC acceleration by thermal mixing) affect degradation.
RCPB11 Perform research on how temperature affects SCC initiation mechanisms and kinetics to help develop models. This research would help develop a more fundamental and broader understanding of temperature effects focusing on validating kinetics and surface microstructure evolution (e.g., grain boundary oxidation, diffusion, etc.).
RCPB12 Perform research on how temperature affects SCC CGR mechanisms and kinetics to help develop models with emphasis on potential microstructural changes that might occur during long-term thermal aging.
RCPB13 Perform research to improve accelerated testing methods to avoid altering microstructures too much while still obtaining useful data. This involves finding a balance between accelerating tests and maintaining the integrity of the materials being tested.
RCPB14 Research may be needed to evaluate the EC resistance of components after the application of mitigation surface treatments such as peening.
3.4.3.3 Summary of PIRT Results for Session 3 The results of the PIRT panel scoring are shown in Session 3 - PIRT Scoring - RCPB. Figure 3-4, below.
3-27 Figure 3-4 PIRT Results for Session 3, RCPB The complete scoring by the panel and the rationale for the scoring is contained in Session 3 -
PIRT Scoring-RCPB. RCPB13 is related to evaluating the possibility of improving accelerated testing methods and fell into the category of having high safety importance and low knowledge.
In this category, research could help gain knowledge to help reduce uncertainty in our understanding of issues important to safety. The panelists were aligned on the importance of employing testing techniques that provide accurate data in a reasonable amount of time and that accelerated testing methods (e.g., elevated temperature, cold work) are a key approach to achieve this. Our knowledge of how accelerated testing methods might affect microstructures could be improved or verified to increase our confidence in the data. RCPB13 was one of the highest ranked research areas because accelerated testing methods are typically employed to obtain almost all PWSCC-related data and regulatory decisions are being made based on this data. Improving our knowledge in this area may decrease the uncertainty associated with our view of its importance.
RCPB11 on evaluating how temperature affects PWSCC initiation mechanisms was scored as being important for safety but having some knowledge to mitigate safety concerns. In this category, more research could improve our knowledge somewhat on issues important to safety.
For RCPB11, the panelists recognized the importance of understanding how temperature Importance Knowledge RCBP High Low High Low 1
9 10 8
2 7
6 5
4 14 11 3
12 13
3-28 influences PWSCC initiation mechanisms in general and that more knowledge in this area might be beneficial for improving our understanding. There is data on key materials such as nickel-based Alloy 600/82/182, which are susceptible to PWSCC, and it would not be too difficult to obtain data to improve our understanding of temperature on initiation mechanisms. The high-chrome nickel alloys Alloy 690/52/152 are very resistant to PWSCC and more research might improve our knowledge regarding initiation mechanisms, but given its level of resistance to PWSCC, such research would be a lower priority compared to other less resistant alloys such as Alloy 600/82/182 and stainless steels.
Three topics were scored in a category where the research topics were important to safety and with less knowledge available. The panel included RCPB3 in this category since it was on testing explant or service-aged materials to help validate laboratory testing. Based on our knowledge of operating aging mechanisms, research could be focused on high-value materials and components. RCPB12 is related to how temperature affects SCC crack growth rate mechanisms and kinetics to help develop models with an emphasis on thermal aging. For the high-chrome nickel alloys, thermal aging is not an immediate safety concern, but there is some data to suggest thermal aging can lead to increased hardness and higher susceptibility to SCC.
Additional research in this area might improve our knowledge and decrease the uncertainty associated with importance, which might decrease the safety importance. DOE is doing research in this area which will help supplement our knowledge. The panel felt that RCBP14 on the effect of mitigation techniques, such as peening, was an interesting research topic, but the information we have so far shows that mitigation is effective. More research might be required if cracks are discovered in mitigated components.
RCPB7 and RCPB8 were scored in the category of research topics that are expected to have safety significance but could likely be addressed with a modest level of research given the level of knowledge we already have. RCPB7 is related to research on stainless steels to provide a better understanding of SCC initiation in stainless steel piping, focusing on factors such as thermal stratification, temperature, dissolved oxygen, surface condition, long-term aging, and welding residual strain. The panelists generally felt the degradation mechanisms are fairly well known, but research could be done to identify the components that can be affected during the extended period of operation and obtain data on thermal aging. RCPB8 is on conducting modeling and simulation to validate PWSCC mechanisms and make accurate predictions of initiation times and crack growth rate. The panelists view is that it is important to have tools to help predict and assess crack initiation and growth, but the level of knowledge in this area is high and the difficulties of developing predictive models are well recognized. Additional research might help decrease uncertainty in the data used to develop models.
The remaining issues fall within the regime where significant knowledge exists to address the recommended topics through measures other than research, or the recommended topic is not expected to have safety significance. Some of the research topics that were scored into this regime seemed to be associated with high importance uncertainty, which might be reduced with more knowledge and lower their safety importance further. These topics include RCPB1 on research on PWSCC initiation mechanisms, RCPB2 on thermal fatigue and corrosion fatigue, RCPB4 focuses on dissimilar metal weld interfaces and heat-affected zones, RCPB5 investigates the impact of preexisting welding defects on SCC initiation in high-chrome weld metals, RCPB6 research on degradation mechanisms associated with low alloy steel, RCPB9 on conducting research to evaluate how cold work and applied stress affect SCC initiation time for different alloy groups, and RCPB10 is related to temperature and stress effects associated with power uprates and plant modifications.
3-29 For RCPB1, the importance scores were mixed with one panelist scoring it high citing the importance of understanding the factors that affect PWSCC initiation to help prevent, mitigate and manage PWSCC while most of the panelists scored RCBP1 as less important. The panelists that scored this as less important noted that current regulatory requirements typically assume crack initiation has occurred and, in the case of steam generator tubing that initiation is mitigated by stress relieving the tubes. RCPB2 is mitigated by industry-led programs to scope and identify thermal fatigue issues, but some research could be performed to help verify that the programs are bounding for the period of extended operations. Research on RCPB4 and RCPB5 is planned and ongoing and should sufficiently address these topics. RCPB6 was not considered a safety concern and research is not recommended at this time. For RCPB9, the panelists felt this was of low importance because the issue is well known, and good practices are in place to help minimize cold work. For RCPB10, the panelists felt that this research area was of low importance and sufficient knowledge exists regarding the effect of power uprates since they are reviewed regularly. Plant modifications typically have little effect on material degradation unless there are plant-specific issues.
3.4.4 Session 4, Secondary Side Components, PIRT Evaluation 3.4.4.1 Initial Inputs to PIRT Process Thirteen initial recommendations were identified from the two presentations and the panel discussion for Session 4. Although it should be noted that more than one potential research topic is identified in each recommendation. After combining recommendations with similar themes into broad categories, and considering each unique research topic within each recommendation, a final list of 14 recommendations was generated for input to the PIRT process. To facilitate the combination process, the original recommendations were sorted into categories:
The following table shows the original list of research recommendations and the combined final set of recommendations.
Table 3-8 PIRT Inputs for Session 4, Secondary Side Components Category Initial Recommendations Combined Final Recommendations Conduct literature reviews Identify potential hydrazine alternatives through comprehensive literature reviews.
Conduct comprehensive literature reviews and basic studies to identify and evaluate hydrazine alternatives, focusing on efficacy, safety, regulatory status, and key data such as reaction rates, degradation products, and corrosion potential.
3-30 Category Initial Recommendations Combined Final Recommendations Basic studies on alternatives Perform basic studies to obtain data not available in the literature, such as reaction rates, degradation products, and corrosion potential.
Conduct basic research to evaluate safer oxygen scavenger alternatives, focusing on reaction rates, degradation products, and corrosion potential, with a prioritized transition to CHz and future consideration of DEHA.
Promising candidates transition of oxygen scavenger Evaluate safer chemicals Focus on CHz in the near term and DEHA in the future as the most promising alternatives, despite their own limitations.
CHz and DEHA stand out as the most promising alternatives for near-and long-term use, respectively, while safer, low-toxicity compounds such as Ethyl Acetoacetate (EA) and Gluconic Acid (GA)both commonly used as food additivesalso warrant consideration.
Consider safer chemicals like EA and GA, which are used as food additives and have low toxicity.
Measure reaction rates Analyze degradation products, and Measure corrosion potential Conduct experiments to measure the reaction rates of hydrazine alternatives under defined conditions.
Conduct experiments to measure the reaction rates of hydrazine alternatives under defined conditions and analyze their degradation products to assess potential corrosion impact.
Evaluate the corrosion potential of these alternatives using high-temperature, high-pressure water loops for a comprehensive understanding of their performance and safety.
Evaluate the degradation products of alternatives to understand their potential impact on corrosion.
Measure the corrosion potential of alternatives using high-temperature, high-pressure water loops.
Corrosion potential measurement Consider Ethanolamine (ETA):
Explore the possibility of using Ethanolamine as a stand-alone water chemistry option due to its slight oxygen scavenging ability.
Explore the possibility of using Ethanolamine as a stand-alone water chemistry option due to its slight oxygen scavenging ability.
Focus on Alternatives and Safety Data and Research Identify safer alternatives to current materials and chemicals used in nuclear power plants.
Critical need to identify safer alternatives to the materials and chemicals currently used in nuclear power plants. The main research recommendations included conducting comprehensive literature reviews to assess existing knowledge, undertaking basic scientific Conducting comprehensive literature reviews, basic studies, and expanding data collection were common recommendations
3-31 Category Initial Recommendations Combined Final Recommendations across the presentations and panel discussion.
studies to explore novel materials, and expanding data collection to inform evidence-based decisions.
Corrosion and Erosion Management Ongoing research and improvements in managing corrosion and erosion in nuclear components.
Ongoing research and long-term strategies to enhance corrosion and erosion management in nuclear components.
Long-Term Strategy The importance of a long-term strategy for the continued operation and safety of nuclear components was a recurring theme.
SCC New mechanisms of SCC are unlikely to be discovered for modern steam generator tube alloys, as extensive research has already been conducted in this area.
Current strategies have significantly reduced the urgency surrounding SCC, mainly due to improvements in water chemistry control and the development of better alloys.
Discovering new mechanisms of SCC in modern steam generator tube alloys is unlikely due to extensive prior research.
Advances in water chemistry control and improved alloys have significantly mitigated SCC concerns.
LTO:
Future impact of secondary side water chemistry changes on SCC. Maintaining appropriate chemistry is crucial for long-term reactor operations exceeding 80 years.
Research topics proposed include the effects of brief excursions in secondary side water chemistry and the cumulative impact of low-level impurities.
Role of secondary side water chemistry in SCC and long-term reactor operations, with proposed research on brief excursions and low-level impurity effects.
Material Aging and Degradation Recommendations include studying the long-term effects of chemical treatments, including deposit management, and understanding potential synergistic effects caused by The recommendations stress the importance of studying the long-term effects of chemical treatments, particularly in deposit management. They also highlight the need to assess potential
3-32 Category Initial Recommendations Combined Final Recommendations flow-induced vibrations and foreign object wear.
Panelists indicate the importance of predictive maintenance strategies as operating conditions evolve.
interactions between flow-induced vibrations and foreign object wear.
Additionally, the role of predictive maintenance strategies in adapting to changing operating conditions is important.
FAC The current approach to managing FAC is viewed as flexible, capable of adapting to new OE and incorporating advancements in water chemistry.
Future research should continue to focus on the impacts of changing water chemistry, particularly the application of film-forming amines.
Future research should focus on the impacts of changing water chemistry, particularly the application of film-forming amines, while building on the flexible approach to managing FAC that adapts to new OE and advancements in water chemistry.
Component Replacement Needs The discussion touches on the necessity of monitoring major piping and components such as steam drums and moisture separators, as they can suffer from erosion and corrosion.
Replacement decisions for components will be based on plant-specific conditions and economic considerations, with inspections either justifying repairs or leading to replacements.
Monitoring key components is essential to prevent erosion and corrosion, with repair or replacement decisions guided by inspections, plant conditions, and economic factors.
Long-Term Management Strategies Effective management of steam generator chemistry and the implementation of diligent maintenance strategies are emphasized for sustaining plant integrity over extended operation periods.
Understanding flow, temperature, and steam quality as operational factors is critical Maintaining plant integrity over extended operation requires effective steam generator chemistry management, proactive maintenance, and a strong understanding of key operational factors to mitigate system deterioration.
3-33 Category Initial Recommendations Combined Final Recommendations for predicting and mitigating deterioration in aging systems.
SSC1 Conduct thorough literature reviews and research to identify and evaluate safer alternatives to hydrazine as oxygen scavengers. Assess candidates based on efficacy, safety, regulatory compliance, reaction kinetics, degradation products, and corrosion impact. Prioritize carbohydrazide (CHz) for initial evaluation, with diethylhydroxylamine (DEHA) as a secondary focus.
SSC2 CHz and DEHA stand out as the most promising alternatives for near-and long-term use, respectively, while safer, low-toxicity compounds such as Ethyl Acetoacetate (EA) and Gluconic Acid (GA)both commonly used as food additivesalso warrant consideration.
SSC3 Conduct experiments to measure the reaction rates of hydrazine alternatives under defined conditions and analyze their degradation products to assess potential corrosion impact. Evaluate the corrosion potential of these alternatives using high-temperature, high-pressure water loops for a comprehensive understanding of their performance and safety.
SSC4 Explore the possibility of using Ethanolamine as a stand-alone water chemistry option due to its slight oxygen scavenging ability.
SSC5 Ongoing research and long-term strategies to enhance corrosion and erosion management in nuclear components.
SSC6 Investigate how the flexible nature of current FAC management approaches can be enhanced to incorporate emerging OEsuch as the effects of brief excursions on long-term steam generator (SG) tube degradationand advancements in water chemistry.
SSC7 Prioritize research on the effects of evolving water chemistry on flow-accelerated corrosion (FAC), with a specific focus on the performance and implications of film-forming amines and the long-term impact of low levels of impurities on both steam generator (SG) tubes and other secondary side components.
SSC8 The recommendations stress the importance of studying the long-term effects of chemical treatments, particularly in deposit management. They also highlight the need to assess potential interactions between flow-induced vibrations and foreign object wear. Additionally, experts emphasize the role of predictive maintenance strategies in adapting to changing operating conditions.
SSC9 Future research should focus on the impacts of changing water chemistry, particularly the application of film-forming amines, while building on the flexible approach to managing FAC that adapts to new OE and advancements in water chemistry.
3-34 3.4.4.2 Summary of PIRT Results The complete scoring by the PIRT panel, along with the rationale for the scoring, can be found in Session 4 PIRT Scoring-SSC.
Figure 3-5 shows the PIRT results for Session 4. The average importance scores for the recommendations are generally in the low to medium range. Recommendations SSC2, -3, -4, and -5 fall in the lower right quadrant, indicating recommendations with relatively low knowledge, but for which additional research is considered less likely to significantly affect safety. SSC2, -3, and -4 are all related to research on alternative oxygen scavengers to hydrazine. The panel generally scored these relatively low in importance. One panelist commented that these recommendations are important from an asset management perspective for plant owners but are not critical to safety, due to robust programs for steam generator tube inspection and FAC management. SSC5 was related to investigating the effects of brief excursions on FAC management. One panelist commented that they doubted that brief excursions would have much effect on long-term S/G tube integrity, but another panelist commented that research in this area would be interesting, albeit not high in safety significance.
SSC6 was related to the long-term effects of film-forming amines and fell on the border of the lower left and lower right quadrants, indicating lower importance and moderate knowledge. One panelist scored SSC6 medium in importance and noted the potential for creating a SG tube integrity issue is low, due to the small amounts of material removed by FAC and erosion, but that corrosion products can result in deposit buildup over time. Another panelist rated SSC6 low in in importance due to perceived lower safety significance of secondary water chemistry issues.
SSC1 and SSC7 are on the border of the lower left and upper left quadrant, indicating moderate importance and fairly high knowledge. Therefore, additional research would not be expected to result in large gains in safety. SSC1 also involved hydrazine alternatives and was scored slightly higher for importance by the panelists. One panelist mentioned industry research on hydrazine alternatives (DEHA and carbohydrazide) but noted that there are still technical gaps. SSC7 recommended studying the long-term effects of chemical treatments on deposit management. One panelist commented that the value of this research is high because of the potential for film-forming amines to reduce deposit buildup in tube support plate crevices or on the top of the tubesheet, which is important with increased operating times between inspections of SGs, and that these deposits can lead to corrosion issues and changes to thermal hydraulic conditions if not addressed in a timely manner. However, another panelist commented that this research would be valuable from an asset management perspective but is not an area on which NRC should fund research.
SSC8 recommended assessing potential interactions between flow-induced vibrations and foreign object wear. SSC8 is located in the lower left quadrant indicating relatively high knowledge and relatively low importance. One panelist commented that flow-induced vibrations are not typically an issue in SGs, because they are addressed by the design of the tube support plates and antivibration bars, and because foreign object wear is readily and accurately sized by eddy current inspection. The panelist further noted that the generation of foreign objects is addressed by foreign material exclusion practices and cannot really be predicted, and that EPRI has developed a large database of foreign objects and has used computational fluid dynamics to create a model that does a good job evaluating the risk of foreign objects in SGs that cannot be retrieved. Another panelist agreed with this assessment.
To summarize the Session 4 PIRT results, none of the recommendations fell within the upper
3-35 right quadrant where knowledge is low and importance is high, which would indicate additional research could result in a significant improvement in safety. Due to the uncertainty indicated by the error bars, SSC3 falls the closest to the upper right quadrant. However, the staff determined that none of the recommendations from Session 4 should be highlighted as higher priority recommendations to support operation beyond 80 years.
Figure 3-5 PIRT Results from Session 4, Secondary Side Components 3.4.5 Session 5, Balance of Plant Systems 3.4.5.1 Initial Inputs to PIRT Process Thirteen initial recommendations were identified from the two presentations and the panel discussion for Session 5. Although it should be noted that more than one potential research topic is identified in each recommendation. After combining recommendations with similar themes into broad categories, and considering each unique research topic within each recommendation, a final list of 13 recommendations was generated for input to the PIRT process. To facilitate the combination process, the original recommendations were sorted into five categories: improved predictive modeling and data management supporting predictive maintenance, analysis techniques, mitigation, and monitoring; advanced inspection and monitoring techniques; repair and mitigation; long-term material performance; and standardization and QA/QC.
The following table shows the original list of research recommendations and the combined final set of recommendations.
Table 3-9 PIRT Inputs, Session 5, Balance of Plant Systems
3-36 Category Initial Recommendations Combined Final Recommendations Improved predictive modeling and data management supporting predictive maintenance Better methods to predict when, where, and how to inspect systems should be developed and enhance mapping and characterization of corrosion data should be pursued for more accurate forecasting.
Research supporting enhanced predictive maintenance through modeling, use of monitoring and inspection data, and forecasting.
Data collection and sharing should be enhanced within the nuclear industry to support predictive maintenance strategies and informed decision-making.
Revisiting and potentially reinstituting shared databases (e.g., the Buried Pipe Inspection Results Database) to analyze inspection results industrywide could help achieve this objective.
Better data sharing would also support development of risk-informed inspection and maintenance frameworks to refine inspection prioritization and reduce unnecessary costs.
Advanced Inspection and Monitoring Techniques Degradation thresholds should be defined for interpreting inspection results. Flaw sizing techniques should be improved to allow for more precise damage evaluation.
Real-time monitoring should be incorporated to complement traditional inspections.
Develop standardized guidelines for defect characterization and acceptable defect sizes in inspection and repair processes Develop real-time monitoring techniques Development of advanced methods to inspect CIPP and ensure long-term reliability.
Current techniques like remote crawlers and ultrasonic testing are available, but further innovations are needed, particularly for inspecting gaps, fittings, and non-adhered linings.
Development of advanced methods to inspect CIPP/SIPP and ensure long-term reliability Research into ultrasonic methods capable of assessing adhesion between the host pipe and the new
3-37 Category Initial Recommendations Combined Final Recommendations lining is emphasized. This includes identifying the right frequencies to penetrate polymer thicknesses and enhance the credibility of inspections.
Advanced NDE techniques for large-diameter pipes and buried infrastructure should be developed to enable efficient, far-field, and aboveground inspections. Methods should be improved to inspect longer pipe segments instead of relying solely on localized inspections, to better assess system health comprehensively.
Research on high-resolution ultrasonic imaging and encoded data collection should be conducted to ensure accurate corrosion morphology evaluation.
Develop advanced NDE techniques for large-diameter pipes and buried infrastructure should be developed to enable efficient, far-field, and above-ground inspections.
Research on high-resolution ultrasonic imaging and encoded data collection should be conducted to ensure accurate corrosion morphology evaluation, including identifying degradation thresholds and improving flaw sizing.
Repair and Mitigation New materials and water treatments should be considered and evaluated for enhanced corrosion resistance. Localized repair techniques should be improved to address specific issues. The effectiveness of cathodic protection, coatings, and chemical treatments should be evaluated.
Develop new materials and water treatments that align with updated safety, economic and environmental requirements that offer enhanced corrosion resistance.
Evaluate effectiveness of cathodic protection, coating, and chemical treatments Novel insitu repair technologies that align with updated safety, economic, and environmental requirements should continue to be developed and standardized.
Polymer-based and carbon-fiber-reinforced solutions (e.g., CIPP and SIPP), for large-diameter, deeply buried pipes that are challenging to excavate and replace are promising. However, the long-term mechanical and chemical stability (e.g., creep, water absorption) of these Develop methods to assess the long-term effectiveness of repairs and treatments, including mechanical and chemical stability
3-38 Category Initial Recommendations Combined Final Recommendations materials under operating conditions needs to be understood.
Methods to assess the long-term effectiveness of repairs and treatments should be developed.
Solutions should be optimized to balance cost and performance.
Long-term material performance Understanding external corrosion of host pipes and other degradation mechanisms that could affect the lifespan and performance of CIPP systems.
Investigating soil conductivity and chemistry's impact on pipeline corrosion are essential considerations.
Better understanding of attributes (e.g., soil conductivity/chemistry) driving external corrosion that could affect repair (e.g., CIPP) lifespan.
The integrity of existing SSCs should continue to be researched.
The effects of transient pressures, load stresses, and environmental changes is an important topic.
Research on the potential impacts of ground settlement on buried pipes is also needed to evaluate the effects of ovality (i.e., pipe deformation due to soil compaction and heavy equipment loads) and its impact on structural integrity over time. Methods to monitor and mitigate such effects in LTO scenarios should also be considered. Finally, Investigation of selective leaching should address fracture initiation and microstructural changes (e.g.,
graphite formation in cast iron).
Understand impacts of ground settlement on piping ovalization and its impact on structural integrity over time.
More investigation of selective leaching to address fracture initiation and microstructural changes and their impact on structural integrity.
Standardization and QA/QC Standardized testing and QA/QC practices should be created for onsite repairs, particularly for polymer-based systems. Clear, standardized guidelines should be developed for defect characterization and acceptable Develop uniform testing and qualification criteria and associated QA/QC practices for onsite repairs that address variability in installation methods
3-39 Category Initial Recommendations Combined Final Recommendations defect sizes in inspection and repair processes. Finally, protocols should be established to address variability in installation methods (e.g., curing conditions) for non-metallic materials.
The emerging SIPP technology requires industrywide standards to qualify coatings and simplify adoption. Utilities are facing challenges with proprietary products, highlighting the need for uniform testing and qualification criteria.
Develop standardized guidelines for defect characterization and acceptable defect sizes in inspection and repair processes 3.4.5.2 Final List of Recommendations for Input to the PIRT Process BOP-1. Research supporting enhanced predictive maintenance through modeling, use of monitoring and inspection data, and forecasting.
BOP-2. Research on high-resolution ultrasonic imaging and encoded data collection to ensure accurate corrosion morphology evaluation, including identifying degradation thresholds and improving flaw sizing.
BOP-3. Develop standardized guidelines for defect characterization and acceptable defect sizes in inspection and repair processes BOP-4. Develop real-time monitoring techniques BOP-5. Develop advanced NDE techniques for large-diameter pipes and buried infrastructure should be developed to enable efficient, far-field, and above-ground inspections BOP-6. Development of advanced methods to inspect CIPP/SIPP and ensure long-term reliability BOP-7. Develop new materials and water treatments that align with updated safety, economic and environmental requirements that offer enhanced corrosion resistance.
BOP-8. Evaluate effectiveness of cathodic protection, coating, and chemical treatments BOP-9. Develop methods to assess the long-term effectiveness of repairs and treatments, including mechanical and chemical stability BOP-10. Better understanding of attributes (e.g., soil conductivity/chemistry) driving external corrosion that could affect repair (e.g., CIPP) lifespan.
3-40 BOP-11. Understand impacts of ground settlement on piping ovalization and its impact on structural integrity over time.
BOP-12. Investigate selective leaching to address fracture initiation and microstructural changes and their impact on structural integrity.
BOP-13. Develop uniform testing and qualification criteria and associated QA/QC practices for onsite repairs that addresses variability in installation methods 3.4.5.3 Summary of PIRT Results The complete scoring by the PIRT panel, along with the rationale for the scoring, can be found in Session 5 - PIRT scoring - BOP. The average scores for the 13 balance of plant issues were calculated and they are depicted in Figure 3-6. The corresponding recommendation number is indicated next to the marker in the plot.
Figure 3-6:PIRT results for Balance of Plant Session Based on these results, the recommendations fall into several distinct categories. The first category (higher importance, less knowledge) corresponds to those recommendations where research could be used to improve knowledge on a topic that the PIRT panelists expect to be important to safety during LTO. Recommendation BOP6 (Develop Advanced Methods to Inspect CIPP/SIPP) falls into this category. This issue has importance because non-metallic repair techniques (e.g., CIPP/SIPP) are already in use and expected to grow in prominence as the existing fleet continues to operate. There is little explicit laboratory data or OE on the long-term performance of these repairs within nuclear power plant applications. This increases the importance of establishing inspection methods providing assurance of their long-term viability.
Current inspection methods are crude and rely on visual and acoustic tap-testing approaches.
3-41 Research in this area is underway, and it is recommended that this research continues to be prioritized so that high-throughput volumetric and surface inspection techniques can be developed that are effective in identifying the extent of potential degradation mechanisms.
Recommendations BOP5 (Develop Advanced NDE for Large-Diameter Pipes), 9 (Develop Methods to Assess Long-Term Effectiveness of Repairs and Treatments) and 11 (Understand Impacts of Ground Settlement) reflect topics with less knowledge that did not have as high as average importance score as BOP6. However, there is enough uncertainty in the score such that their importance could be elevated if the topics were better understood. This characteristic makes these topics good research candidates to either better understand their importance, develop or validate mitigation measures, or both. The topic on developing advanced NDE techniques (BOP5) is focused on developing efficient, far-field, or aboveground inspections that can be performed autonomously. Such characteristics could increase the frequency and coverage of current inspections. There are currently remote inspection methods that are implemented through access points as well as other less mature techniques.
Evaluating the long-term effectiveness of repairs and mitigation treatments (BOP9) addresses metallic, non-metallic, and surface treatment repairs. There is generally good knowledge and OE associated with metallic and historical surface treatments (e.g., coating, peening) and much less so with non-metallic repair techniques. Research in this area could also support development of acceptance criteria associated with inspection technique development (BOP6).
Effects of piping ovalization due to ground settlement (BOP11) is not an area that has received much, if any, prior study, and research could be used to determine the importance of this issue.
However, it is expected to be a secondary consideration without significant effects on safety given understanding of ovalizations effects on pressurized piping systems.
Recommendations BOP1 (Developing Enhanced Predictive Maintenance through Monitoring and Forecasting), and BOP13 (Develop Testing and Criteria for Onsite Repairs) are topics that are thought to be important to maintaining safety, but it is expected that current knowledge may be sufficient to address those issues without additional research. For BOP1, there are existing tools like BPWORKS and MAPPro that are available and, similar to CHECWORKS for secondary-side piping, can provide updated forecasting based on inspection results. Potential gaps are largely associated with the periodicity and coverage of the inspection results as it is essential that the forecasting models receive timely degradation-state information. Any such gaps could be improved by better information sharing among licensees and the development of more easily implementable inspection techniques (BOP5). Similarly, developing enhanced installation requirements, especially for non-metallic repairs (BOP13), is needed as these techniques become more prevalent. ASME Code Case N871 has qualification criteria for carbon-fiber-reinforced polymeric (CFRP) systems that may be appropriate and may serve as a model for other non-metallic repair techniques. There is also knowledge of installation best practices for CFRP and other non-metallics that provide a base. However, additional research may be helpful to gain a more systematic understanding of critical variables and development of acceptable ranges tied to the impacts on the repairs performance.
Recommendations BOP4 (Develop Real-Time Monitoring Techniques), and BOP12 (Investigate Selective Leaching to Address Fracture Initiation) are topics that could maintain or enhance safety although it is not as likely that research will contribute to significant safety benefits. Research would be needed to develop real-time monitoring techniques as they have a lower technical readiness level than other inspection methods. Robust and reliable sensors that could be implemented to provide acceptable system coverage are essential and monitoring strategies would need to be validated. More importantly, because the degradation rates are
3-42 relatively slow in these systems, periodic inspection is expected to be sufficient for timely assessment and a more cost-effective and mature alternative. Research to better understand margins associated with selective leaching could be useful to develop more accurate acceptance criteria. However, this issue only significantly affects a few materials (e.g., cast iron), and it is uncertain if such research will result in definitive and actionable results. It is more important to identify and monitor selective leaching locations and then likely more effective to just conservatively assume that the leaching region does not contribute to structural integrity and define replacement based on existing ASME Code requirements.
Recommendations BOP2 (Research on High-Resolution Ultrasonic Imaging), BOP3 (Develop Standardized Guidelines for Defect Characterization) and BOP7 (Develop New Materials and Water Treatments) are topics that the PIRT panelists expect have sufficient knowledge that will not be substantially improved through additional research. Existing ultrasonic techniques are thought to be sufficiently accurate to capture degradation mechanisms of concern. ASME Code acceptable wall thickness criteria are expected to be sufficient to maintain margins in these systems as it is material loss and bulk strength reduction, and not cracking, that are the principal concerns. Finally, currently available materials and water treatments are expected to be sufficient to manage degradation long term such that new material formulations or treatments to address nuclear plant challenges are not warranted.
Finally, BOP8 (Evaluate Effectiveness of Cathodic Protection, Coating, and Chemical Treatments) and BOP10 (Better Understanding of Attributes Driving External Corrosion that Affect Non-metallic Repair Integrity) are thought to be topics with low potential safety significance given that adequate knowledge exists to address the issue. Cathodic protection and chemical treatments are expected to remain effective if existing programs are properly implemented. Coatings may degrade but a focus on inspections to identify and characterize the extent of degradation is expected to be a more effective aging management strategy than correlating degradation to coating performance. Coatings can also simply not be credited as part of an aging management program as a conservative approach. External corrosion of the original systems (BOP10) is just one consideration as often the internal effects are limiting.
Principal external factors are generally known. Further, current non-metallic repair techniques (i.e., CFRP) do not credit the original piping within the repair other than the bonded region of the pipe.
3.4.6 Session 6, Fatigue, PIRT Evaluation 3.4.6.1 Initial Inputs to PIRT Process Eighteen initial recommendations were identified from the three presentations and the panel discussion for Session 6. Although it should be noted that more than one potential research topic is often identified in a recommendation. After combining recommendations with similar themes into broad categories, and considering each unique research topic within each recommendation, a final list of 14 recommendations was generated for input to the PIRT process. To facilitate the combination process, the original recommendations were sorted into five categories: Environmental fatigue attributes; fatigue in novel/AM materials; instrumentation, monitoring, and data collection; modeling and prediction; and implications for operation.
The following table shows the original list of research recommendations and the combined final set of recommendations.
Table 3-10 PIRT Inputs, Session 6, Fatigue
3-43 Category Initial Recommendations Combined Final Recommendations Environmental fatigue attributes Study the impact of various water chemistries, particularly KOH, on EAF to improve fatigue assessments for LTO.
The impact of the potential switch from LiOH to KOH on EAF in LWRs has not been considered, and KOH should be investigated for potential use in SMR designs.
Evaluate effects of water chemistry additions (e.g.,
KOH, NMCA, DO) on EAF Examine the effect of noble metal catalysts on the fatigue life of materials, especially under BWR conditions with optimized dissolved oxygen content. Boiling-water reactors (BWRs) have trouble maintaining or justifying Dissolved Oxygen (DO) content to optimize CUFen for Carbon/Low Alloy Steels (<
40 ppb) because these requirements can conflict with DO requirements for flow-accelerated corrosion (> 50 ppb).
Irradiation effects are particularly important for reactor pressure vessel and internals components. The limited available data are inconclusive with regard to the impact of irradiation on the fatigue lives of materials exposed to LWR environments. Material harvesting provides opportunities to evaluate this effect.
Evaluate the effects of irradiation on fatigue lives Fatigue in novel/AM materials With emerging materials, (e.g., AM) entering the reactor environment, there is a concurrent research need to evaluate how these new materials behave under long-Understand the fatigue characteristics of AM materials and effect of important attributes (e.g.,
3-44 Category Initial Recommendations Combined Final Recommendations term cyclic loading and environmental exposure and revise existing fatigue models to accommodate the properties of innovative materials surface finish, orientation, processing variables).
Understand the fatigue characteristics of AM materials and develop appropriate codes and standards and assessment methods for their use in new designs and replacement parts. The equivalent structural strain range approach may provide a path forward.
Investigate the more prominent fatigue phenomena expected to strongly affect AM component fatigue lives.
These phenomena include effects of surface finish on fatigue endurance limits for AM materials, material orientation effects, the impact of different manufacturing processes, and the effect of specimen type (i.e., hollow vs. solid specimens).
Develop appropriate codes and standards for the use of AM materials in new designs and replacement parts.
There is a need to further develop fatigue prediction methods for AMmanufactured Class 1 nuclear components. This evaluation should assess the applicability of the Master S-N Curve Method, especially Murakami's root area approach. The applicability of EAF approaches for AM materials should also be assessed.
Develop fatigue prediction and assessment methods for AMmanufactured safety-significant nuclear components
3-45 Category Initial Recommendations Combined Final Recommendations Instrumentation, monitoring, and data collection Field data collection and validation research should focus on developing methods to gather dynamic, field-based data on loading conditions and validating and calibrating theoretical models with insitu measurements.
This research would support safer operations by integrating field data into iterative design improvements and enabling predictive maintenance and proactive interventions.
Develop field-implementable methods to gather dynamic, field data on loading conditions.
Enhanced instrumentation and monitoring research should explore novel sensors and monitoring techniques that can detect subtle changes in structural behavior and overcoming logistical and regulatory challenges in retrofitting or integrating these technologies into existing plants. Objectives should include developing systems that can flag unanticipated mechanical anomalies and creating integrated diagnostic tools that couple sensor outputs with predictive analytics.
Explore novel sensors and monitoring techniques that can detect subtle changes in structural behavior affecting material resistance and are field-implementable Improve sensors and data gathering to monitor equipment health and collect real-time data on loading and material resistance.
Enhanced data logging and sharing systems are also crucial for accurate fatigue analysis.
Develop methods to validate and calibrate theoretical models with insitu measurement information.
3-46 Category Initial Recommendations Combined Final Recommendations Adapt existing equipment health monitoring methods from the aerospace industry, such as extending the time between maintenance intervals based on detailed monitoring and analysis of equipment health. This can help increase availability and reduce downtime.
Adapt existing equipment health monitoring methods from the aerospace industry.
Modeling and Prediction Tailor advanced data analytics tools to process large datasets and replace initial design assumptions with real-world data. This involves handling big datasets and performing detailed analytics to understand equipment behavior under different conditions.
Tailor advanced data analytics tools to process large datasets and understand/predict equipment behavior during operation Develop total life fatigue methods to account for all stages of crack initiation and growth in a probabilistic manner, which may help bridge the gap between laboratory data and component life assessment.
Develop total life fatigue methods to account for all stages of crack initiation and growth in a probabilistic manner Integrate direct measurements and analytics into a risk-informed approach to fatigue management. This approach helps in making more accurate assessments and managing uncertainties in flexible plant operations.
Methods to integrate uncertainties into fatigue assessments holistically, rather than relying solely on conservative design assumptions also need to be incorporated.
Develop a risk-informed approach to fatigue management which integrates direct measurements, analytics, and uncertainties.
3-47 Category Initial Recommendations Combined Final Recommendations Improve modeling and simulation tools such as CFD to better predict and understand the impact of small design variations on fatigue behavior. This includes validating models against real-world data to ensure their accuracy.
Improve modeling and simulation tools by examining real-world behavior, including unanticipated failure cases, to better predict and understand the impact of small design variations on fatigue behavior and identifying, quantifying and incorporating complex environmental and loading factors, and their interactions that most significantly affect performance and reflect the conditions present in aging reactors Implications for Operation Research should aim to conduct comparative studies between lab-based environmental fatigue tests and real-world behavior to identify and quantify the factors that might affect performance. Since environmental factors (like temperature, radiation, and chemical exposure) can modulate fatigue behavior differently in the field, further research is needed to develop comprehensive models that incorporate these complex interactions and tailor fatigue assessments to reflect the unique conditions present in aging reactors.
Research should focus on examining failure cases that were not anticipated during the design phase and developing diagnostic criteria and predictive frameworks to account for such unforeseen events and better understand the role of system variabilities in these failure cases.
Researching the effects of flexible plant operation on fatigue and understanding the implications of changing power levels and other operational parameters on Research the effects of flexible plant operation on fatigue and understanding implications of changing power levels and other operational parameters on
3-48 Category Initial Recommendations Combined Final Recommendations equipment longevity and safety equipment longevity and safety 3.4.6.2 Final List of Recommendations for Input to the PIRT Process FAT-1.
Evaluate effects of water chemistry additions (e.g., KOH, NMCA, DO) on EAF FAT-2.
Evaluate the effects of irradiation on fatigue lives FAT-3.
Understand the fatigue characteristics of AM materials and effect of important attributes (e.g., surface finish, orientation, processing variables).
FAT-4.
Develop appropriate codes and standards for the use of AM materials in new designs and replacement parts.
FAT-5.
Develop fatigue prediction and assessment methods for AMmanufactured safety-significant nuclear components FAT-6.
Develop field-implementable methods to gather dynamic, field data on loading conditions.
FAT-7.
Explore novel sensors and monitoring techniques that can detect subtle changes in structural behavior affecting material resistance and are field-implementable FAT-8.
Develop methods to validate and calibrate theoretical models with insitu measurement information.
FAT-9.
Adapt existing equipment health monitoring methods from the aerospace industry.
FAT-10.
Tailor advanced data analytics to process large datasets and understand/predict equipment behavior during operation FAT-11.
Improve modeling and simulation tools by examining real-world behavior, including unanticipated failure cases, to better predict and understand the impact of small design variations on fatigue behavior and identifying, quantifying and incorporating complex environmental and loading factors, and their interactions that most significantly affect performance and reflect the conditions present in aging reactors FAT-12.
Develop total life fatigue methods to account for all stages of crack initiation and growth in a probabilistic manner FAT-13.
Develop a risk-informed approach to fatigue management which integrates direct measurements, analytics, and uncertainties.
3-49 FAT-14.
Research the effects of flexible plant operation on fatigue and understand implications of changing power levels and other operational parameters on equipment longevity and safety 3.4.6.3 Summary of PIRT Results The complete scoring by the PIRT panel, along with the rationale for the scoring, can be found in Session 6 - PIRT scoring-Fatigue. The average scores for the 14 fatigue issues were calculated and they are depicted in Figure 3-. The corresponding recommendation number is indicated next to the marker in the plot.
T Figure 3-7: PIRT Results for Fatigue Session Based on these results, the recommendations fall into several distinct categories. The first category (higher importance, less knowledge) corresponds to those recommendations where research could be used to improve knowledge on a topic that the PIRT panelists expect to be important to safety during LTO. Recommendations FAT2 (Evaluating the Effects of Irradiation on Fatigue Lives) and FAT5 (Develop Fatigue Prediction and Assessment Methods for AMManufactured Safety-Significant Nuclear Components) fell into this category. For FAT2, the panelists noted that there is little data on irradiation effects and that there are some safety-significant internals components that are subjected to fatigue loadings. Its importance was tempered somewhat as one panelist indicated that the data that does exist near the fatigue endurance limit demonstrates better fatigue life due to irradiation because of the corresponding increase in tensile strength. For FAT5, it is recognized that there is very little fatigue data for
3-50 AM components in nuclear environments. AMproduced components have been shown to be more susceptible to fatigue than wrought or cast components in certain conditions. However, one of the panelists has slightly less concern because existing Code rules to address fatigue in these components should be sufficient assuming proper qualification and design practices are followed.
There are a trio of topics, FAT1 (Evaluate Water Chemistry Effects); FAT3 (Understand Fatigue Characteristics of AM Materials); and FAT11 (Improve Modeling and Simulation Tools by Examining Real World Behavior) that are associated with less knowledge but did not have as high as average importance score. However, there is enough uncertainty in that average score that their importance could be elevated. This makes these issues candidates for research to improve their underlying knowledge such that the importance of those topics to safety could be better understood. The consideration and modeling of real-world behavior (FAT11) is thought to be a potentially useful but very challenging topic as related small plant-specific differences and uncertainties in critical variables can dramatically affect fatigue performance. While quite a bit of research has been done on water chemistry effects (FAT1), studies on the optimization of dissolved oxygen limits for carbon/low-alloy steel materials could result in a better balance of fatigue and flow-accelerated corrosion degradation rates. Evaluation could also continue to validate the long-term effectiveness of hydrogenated water chemistry additions on stainless steel SCC in BWR plants. The FAT3 research recommendations can be addressed under FAT1 research, which is why its importance is tempered.
FAT6 (Develop Field-Implementable Methods to Gather Dynamic Field Data), FAT10 (Tailor Advanced Data Analytics Tools to predict Equipment Behavior During Operation), and FAT12 (Develop Probabilistic Total Life Methods) are all topics that have potentially high safety significance which are less likely to be addressed through additional research. Evaluating plant behavior (FAT6) is already possible and has been conducted at some plants, although some improvements may be possible through sensor development to improve reliability and reduce implementation cost. Interrogating (FAT10) plant instrumentation data to identify precursor field data is expected to be particularly challenging because the emblematic precursor signals are difficult to extract from the bulk instrumentation signals. Probabilistic tools (FAT12) are not expected to improve forecasting due to the importance of small changes in critical variables that are often plant-specific.
FAT4 (Develop Appropriate Codes and Standards for AM) and FAT7 (Explore Novel Sensor and Modeling Techniques) fall within a category of events having relatively low knowledge that could be improved through research, but the research outcomes associated with the topics are not expected to help maintain plant safety during LTO. For example, the fatigue requirements associated with existing design and operational codes are thought to be adequate to address AM components, and developing novel sensors would require extensive research and development with a low likelihood of success.
The remaining issues fall within the regime where significant knowledge exists to address the recommended topics through measures other than research, or the recommended topic is not expected to have safety significance. FAT8 (Develop Methods to Validate Models), FAT9 (Adapt Existing Methods from Aerospace Industry), FAT13 (Develop a Risk-informed Approach to Fatigue Management), and FAT14 (Research Effects of Flexible Plant Operation) all fell into these categories. FAT8, FAT9, and FAT13 are already well-established approaches such that additional research is not needed to implement such strategies. FAT13, in particular, represents a promising aging management approach with an established framework based on other risk-informed applications, which just needs to be tailored to fatigue applications. FAT14
3-51 could become important in the future if flexible plant operation, which will vary the operating conditions is allowed in the U.S.
3.4.7 Session 7: Mitigation, PIRT Evaluation 3.4.7.1 Initial Inputs to PIRT Scoring Thirteen initial recommendations were identified from the two presentations and the panel discussion for Session 7. Although it should be noted that more than one potential research topic is identified in each recommendation. After combining recommendations with similar themes into broad categories, and considering each unique research topic within each recommendation, a final list of eight recommendations was generated for input to the PIRT process. To facilitate the combination process, the original recommendations were sorted into categories:
Table 3-11 PIRT Inputs, Session 7, Mitigation of Material Degradation Associated with Long Term Operation.
The following table shows the original list of research recommendations and the combined final set of recommendations.
Category Initial Recommendations Combined Final Recommendations Development of New Repair Materials Investigate alternative materials and coatings that minimize helium-induced cracking in irradiated components.
Evaluate the effectiveness of alternative materials, coatings, and welding techniques in minimizing helium-induced cracking in irradiated components, with a focus on their applicability to a wide range of reactor materials.
Expansion to Other Alloys Investigate the effectiveness of these welding techniques on a broader range of reactor materials to ensure wider applicability.
Advanced Monitoring for Small-Bore Piping Implement real-time monitoring systems using AI and sensors to predict and prevent fatigue-related failures.
Implement real-time monitoring systems using AI and sensors to predict and prevent fatigue-related failures.
Extended Testing of Thermal Aging Effects Conduct prolonged thermal aging studies under varying operational conditions to better understand degradation mechanisms.
Conduct prolonged thermal aging studies under varying operational conditions to better understand degradation mechanisms.
Enhancement of Non-Metallic Repair Techniques Further evaluate the long-term stability, chemical resistance, and mechanical properties of non-metallic repair solutions for spent fuel pools.
Prioritize comprehensive long-term evaluation of non-metallic spent fuel pool repair solutions focusing on stability, chemical resistance, and mechanical
3-52 Category Initial Recommendations Combined Final Recommendations Enhanced Repair Welding Techniques Conduct further testing on weld materials and techniques to improve their performance in high-radiation environments.
integrity. Simultaneously, conduct targeted research on weld materials and techniques to enhance performance under high radiation. Investigate and develop automated systems for reliable and scalable implementation of these repair and welding technologies within nuclear facilities.
Integration with Automated Welding Systems Explore automation possibilities for consistent and large-scale implementation in nuclear plants.
Optimization of Dual Laser Beam Parameters Further studies should focus on refining laser intensity, beam spacing, and heating duration to maximize effectiveness.
Further studies should focus on refining laser intensity, beam spacing, and heating duration to maximize effectiveness.
Comparative Analysis of Traditional Welding Methods Conduct studies to compare cost, efficiency, and defect rates between advanced and conventional welding techniques.
Conduct studies to compare cost, efficiency, and defect rates between advanced and conventional welding techniques.
Advanced SCC Mitigation Strategies Investigate the long-term effectiveness of SCC-resistant alloys and optimize water chemistry to minimize degradation.
Investigate the long-term effectiveness of SCC-resistant alloys and optimize water chemistry to minimize degradation.
Improved Remote Inspection and Repair Technologies Develop AIassisted robotic systems for precise, efficient repairs in highly irradiated, confined spaces.
Develop AIassisted robotic systems for precise, efficient repairs in highly irradiated, confined spaces.
Radiation-Resistant Materials Development Design and test new materials with superior radiation tolerance for weld repairs and reactor internals.
Design and test new materials with superior radiation tolerance for weld repairs and reactor internals.
Comparative Study of Regulatory Approaches Assess the impact of proactive versus reactive maintenance strategies on long-term reactor safety and operational efficiency.
Assess the impact of proactive versus reactive maintenance strategies on long-term reactor safety and operational efficiency.
Long-Term Performance Monitoring Implement real-time monitoring systems for early detection of material degradation and optimize predictive maintenance strategies.
Implement real-time monitoring systems for early detection of material degradation and optimize predictive maintenance strategies.
3-53 3.4.7.2 Final List of Recommendations for Input to the PIRT Process MMD1 Evaluate the long-term stability, chemical resistance, and mechanical properties of non-metallic repair solutions for spent fuel pools.
MMD2 Explore automation possibilities (for welding) for consistent and large-scale implementation in nuclear plants.
MMD3 Evaluate and enhance alternative materials, coatings, and welding techniques to mitigate helium-induced cracking in irradiated reactor components. Design and test new materials with superior radiation tolerance for weld repairs and reactor internals.
Optimize welding parameterssuch as laser intensity, beam spacing, and heating durationfor improved performance in high-radiation environments, ensuring broad applicability across reactor materials.
MMD4 Conduct studies to compare defect rates between advanced and conventional welding techniques.
MMD5 Investigate the long-term effectiveness of SCC-resistant alloys and optimize water chemistry to minimize degradation.
MMD6 Develop AIassisted robotic systems for precise, efficient repairs in highly irradiated, confined spaces.
MMD7 Assess the impact of proactive versus reactive maintenance strategies on long-term reactor safety and operational efficiency.
MMD8 Implement real-time monitoring systems for early detection of material degradation and optimize predictive maintenance strategies.
3.4.7.3 Summary of PIRT Results The complete PIRT scoring, and the rationales for the scores, can be found in Session 7 -
Mitigation PIRT Scoring. Figure 38 presents the PIRT results for Session 7, showing that the average importance scores for the identified research recommendations generally fall within the low to medium range. Recommendations MMD6 and MMD1 are positioned in the lower right quadrant, representing areas where current knowledge is limited, but additional research is unlikely to result in significant safety improvements. MMD1, MMD2, and MMD4 focus on non-metallic repair solutions and welding technique evaluations for spent fuel pools, with MMD4 specifically addressing defect rate comparisons. These topics received relatively low importance scores from the expert panel. However, the panel emphasized the value of fostering innovation in technically demanding areas with long-term potential, such as MMD6, which involves AIassisted robotics for repairs in irradiated, confined spaces, and MMD2, which seeks to automate welding processes. MMD1 and MMD4, while less transformative, were acknowledged as foundational studies critical to sustaining material integrity and long-term system reliability.
MMD5 is situated in the upper left quadrant of the graph, indicating high importance and a medium level of existing knowledgesuggesting a moderate but timely research priority. This recommendation encompasses research into SCC-resistant alloys, proactive maintenance strategies, and real-time degradation monitoring, all of which present manageable technical
3-54 risks while offering significant benefits to nuclear safety, operational efficiency, and cost-effectiveness.
MMD3 stands out as a central and complex research challenge, located near the middle of the graph. It targets the mitigation of helium-induced cracking and the refinement of welding techniques for high-radiation environments. This area would be useful for the long-term sustainability of reactor components and would require a sustained and multidisciplinary research approach. MMD6, also in the lower right quadrant, highlights the development of AIassisted robotic systems for repair in highly irradiated, constrained environments. Though promising in terms of long-term impact, the substantial technological hurdles make it a future-oriented initiative best pursued through coordinated innovative efforts across multiple disciplines. One of the reasons panelists agreed to the ranking is because there had already Figure 3-8 PIRT Results for Session 7, Mitigation been much concern about work in this area, such as void swelling. Regulatory issues are being addressed in 50.55a via allowable fluence thresholds for underwater welding without approaching NRC.
MMD2 and MMD4, which explore the automation of welding for consistent, large-scale implementation, occupy a space on the graph that reflects medium impact and moderate development complexity. These areas are well suited to incremental research and pilot-scale demonstrations.
In contrast, MMD1, MMD2, and MMD4 reside in the lower-impact region of the graph. While these initiatives support system reliability and infrastructure maintenance, their expected returns are more limited, positioning them as complementary rather than central to the research agenda.
3-55 MMD7 and MMD8 appear near the upper middle of the graph, indicating higher importance and relatively strong existing knowledge. Though they approach the upper right quadrant where low knowledge and high importance would suggest strong justification for new research, the associated uncertainties, as reflected in the error bars, led the staff to conclude that these recommendations did not qualify as high-priority items. Panelists stated that an assessment of online monitoring for NPP monitoring, measurement, diagnostics, and prognostics should be integral to the adoption of proactive strategies.
In summary, none of the Session 7 recommendations occupy the upper right quadrant of the graph, which would represent areas where new research could significantly enhance safety.
3-56 3.4.8 Session 8, Civil Structures, Concrete, and Components, PIRT Evaluation Recommended PIRT Team (Madhumita Sircar, Andrew Prinaris (NRR\\DEX), George Wang (NRR\\DEX) 3.4.8.1 Initial Inputs to PIRT Process Session 8 covered seven different categories/topics : (1) Effects of Radiation on Structures, (2)
Aging of Post-tensioned Containments, (3) Coupled Degradation Mechanisms, (4) Crevice Corrosion, (5) Enhance Inspection and Monitoring, (6) Repair and Replacement Strategies, and (7) Effects of Climate Change. Of the seven different categories, 54 initial recommendations were identified from the 12 presentations and the panel discussion. Although it should be noted that more than one potential research topics is often identified in a recommendation. After combining recommendations with similar themes into broad categories, and considering each unique research topic within each recommendation, a final list of 35 recommendations was generated for input to the PIRT process.
The following table shows the initial list of research recommendations and the combined set of recommendations.
Category Initial Recommendations (IRs)
Combined Recommendations (CRs)
Effects of Radiation on Structures (RAD)
(12 IRs)
(7 CRs)
Data Collection:
- Harvest and characterize irradiated concrete for obtaining real-life data to verify accelerated lab test data, size effect, and reduce uncertainties in understanding degradation mechanisms.
- Compare radiation-induced free expansion data to verify effects under structural confinement.
Material Behavior:
- Investigate how gamma irradiation affects the corrosion rate of steel and the overall properties of cementitious materials.
- Study creep, stress relaxation, and healing mechanisms in irradiated cement paste and concrete.
- Study the potential for accelerated creep due to irradiation.
- Examine potential ASR and gel stability under irradiation.
Radiation Transport and Modeling:
- Refine methodologies for calculating radiation transport through concrete
- Harvest and analyze irradiated concrete to verify accelerated laboratory data (flux, size) and reduce uncertainties. Formation of new phases in concrete as concrete age.
- Compare radiation-induced free expansion data to effects under structural confinement. Explore damage depth
- Refine radiation transport models with realistic irradiation histories.
More plants to exceed thresholds with LTO.
- Verify attenuation of radiation through thickness of the structure and degradation.
- Study interactions of irradiation with ASR gel stability, creep in cement paste, and corrosion of rebars and embedded steel due to gamma radiation, further exacerbated by boric acid leakage often occurring at PWR RV cavities.
3-57 Category Initial Recommendations (IRs)
Combined Recommendations (CRs) incorporating realistic irradiation histories. More plants to exceed threshold with time.
Modeling and Simulation:
- Evaluate the impact of neutron and gamma irradiation on material degradation and structural integrity.
- Develop reliable methods to assess and numerical model soundness in irradiated reinforced concrete.
- Assess impact of secondary gamma and evaluate their effects on concrete.
Collaboration and Knowledge Expansion:
- Align research efforts with global organizations to address LTO beyond 80 years.
- Harvesting irradiated concrete and components
- Establish a consensus on testing and modeling methodologies.
- Collaborate globally to address LTO beyond 80 years.
- Standardize testing and improve predictive models and structural integrity assessment.
Aging of Post-tensioned Containments (APC)
(7 IRs)
(6 CRs)
Aging and Long-Term Behavior:
- Investigate long-term elastic and inelastic creep effects, including Poissons creep, tensile and bending creep.
- Study the impacts of structural modifications (e.g., steam generator replacement (SGR)), tendon retensioning, presence of small defects, and radial stain/cracking on containment integrity.
- Examine interactions and degradation between old and new materials (concrete, tendons),
loading.
- Investigate primary creep reactivation under change in loading, and inelastic residual creep Advanced Monitoring Techniques:
- Deploy advanced monitoring systems to track deformation,
- Investigate long-term creep effects, structural modifications, SGR, loss of prestress, retensioning and interactions between old and new materials.
- Study prestress loss, drying, creep, and shrinkage using reliable models to ensure structural safety under accident conditions.
- Analyze degradation mechanisms such as corrosion, ASR, freeze-thaw, carbonation etc., as plants age.
- Use advanced monitoring systems to track deformation, cracking, and structural behavior, preventing unrepairable damage.
- Enhance modeling with data on aging factors to improve
3-58 Category Initial Recommendations (IRs)
Combined Recommendations (CRs) evolution of cracking, and structural behavior.
- Periodically manage and monitor creep levels to prevent transitions to tertiary creep and maintain structural integrity under high stress or temperature.
Modeling and Simulation:
- Enhance FEM models with data on aging factors (drying, creep, shrinkage, temperature gradient),
change of loading, modification and repair to improve predictions and guide proactive interventions. Ensure structural safety under accident conditions.
predictions and guide interventions.
- Gather real-life data and information by harvesting.
Coupled Degradation Mechanisms (CDM)
(5 IRs)
(5 CRs)
Aging Effects:
Evaluate the impact of different environmental conditions and material compositions on concrete behavior.
Research the impact of coupled aging mechanisms on the mechanical properties of materials.
Address the scale effect and coupling of different aging phenomena such as radiation, alkali-silica reactions, corrosion, carbonation, and freeze-thaw.
Extreme environment e.g., marine or coastal area, extreme cold, extreme hot and dry environment.
Collaborative Efforts Utilize collaborative R&D programs like the ODOBA Project studying concrete pathologies and their consequences on nuclear structures and the MACUMBA Platform (test facilities) by ASNR for comprehensive data collection and validation.
Collect OE and conduct studies on corrosion, carbonation, chloride penetration, alkali-aggregate reactions including lessons learned.
Investigate freeze-thaw cycles combined with chloride-induced corrosion and exposure to demineralized water.
Design experiments to understand coupled degradation mechanisms and validate findings with field observations.
Develop generic models to serve as guidance for plant-specific applications.
Expand predictive knowledge and develop tools to address future degradation scenarios.
3-59 Category Initial Recommendations (IRs)
Combined Recommendations (CRs)
Crevice Corrosion Liner (CCL)
(6 IRs)
(3 CRs)
Environment conducive to crevice corrosion Investigate the impact of various environmental conditions on crevice corrosion mechanisms. Explore the effects of different crevice formers on corrosion behavior.
Experimental validation with diverse setups:
Customize corrosion experiments to explore the effects of varying materials, geometries, and environmental conditions, such as pH levels and chloride concentrations, on corrosion dynamics.
Develop advanced experimental techniques for more accurate simulation and analysis of crevice corrosion.
Comprehensive corrosion modeling (ongoing at Valtion Teknillinen Tutkimuskeskus - VTT)
The project focuses on developing and applying multiphysical 3D microstructural models to study corrosion behaviors, coupling chemical transport, electrochemistry, thermomechanics, and damage progression.
Application of advanced simulation techniques Models integrate finite element methods and Multiphysics solvers coupling with environmental behavior and corrosion, loss and damage of material, and mechanical behavior.
Mitigation Explore mitigation strategies to prevent crevice corrosion in steel-lined concrete structures.
Investigate crevice corrosion environment and crevice former and validate using OE and experiments Develop reliable comprehensive corrosion modeling Develop mitigation strategies to prevent crevice corrosion
3-60 Category Initial Recommendations (IRs)
Combined Recommendations (CRs)
Enhance Inspection and Monitoring (EIM)
(13 IRs)
(5 CRs)
Develop or Enhance AMPS New AMPs for elastomers in civil structures. Enhance existing AMPs for aggressive coastal environments and long-term containment functionality based on leakage tests.
Integrate advanced technologies incorporating demonstrated AI, robotics, BIM, sensors, and NDEs for efficient inspections and condition monitoring.
Inspection and Monitoring Detection of indications of degradation in a timely and appropriate manner through visual inspection, NDE. Not all degradations can be detected by visual inspection. Improvement of NDEs using AI and drone systems.
Enhancement of building information models (BIM) and monitoring methods.
Develop novel techniques for detecting degradation mechanisms in inaccessible areas to enhance inspection accuracy.
Create comprehensive databases for inservice inspection results to enable tracking and trending of degradation over time.
Measured Testing and Predictive Formulas Detailed studies on carbonation, chloride penetration, and alkali-aggregate reaction, including predictive formulas and methods for measuring.
NDE:
Leveraging new (NDE) devices and measurement techniques.
Improve visual inspections and leverage innovative NDE techniques to detect degradation mechanisms and aging effects.
Develop and enhance Aging Management Programs (AMPs) for elastomers in civil structures and for coastal environments.
Integrate advanced technologies such as AI, robotics, drones, building information models (BIM) for efficient inspections, improved safety for personnel, better access, and data management.
Implement enhanced mitigation technologies, such as cathodic protection systems, drones, and remote sensing and monitoring methods.
Create databases to track degradation trends over time and develop novel techniques for inspecting inaccessible areas.
3-61 Category Initial Recommendations (IRs)
Combined Recommendations (CRs)
Enhance nondestructive testing (NDT) techniques for structural evaluations.
Develop NDE techniques to gather reference experimental data and validate simulation tools.
Mitigation / Modernization Enhance Mitigation Technologies Cathodic Protection Systems Enhance Inspection and Monitoring (Drones, Remote Monitoring)
Tools for Aging Management - EPRI research moving forward will focus on tools to provide more reliable and efficient aging management of civil infrastructure Repair and Replacement Strategies (RRS)
(5 IRs)
(5 CRs)
Evaluate repair and replacement strategies for NPPs operating beyond 80 years, ensuring feasibility and reliability.
Explore new materials and advanced repair methodologies.
Implement systematic assessments and repairs, including coatings, liner corrosion, gutters, and base raft thickening, to mitigate risks related to severe accidents and improve seismic resilience.
Explore advancements in repair methods and materials tailored to specific aging mechanisms, ensuring durability and long-term performance.
Take advantage of previously performed repairs with proven field success to refine future strategies.
Evaluate repair strategies for nuclear plant structures beyond 80 years.
Research and validate new materials and methodologies.
Address implications of transitioning to new cement standards.
Develop databases for repair methods and performance data.
Share best practices through collaborative forums.
Implement systematic repairs to improve durability and resilience.
Leverage proven repair/rehabilitation techniques tailored to aging mechanisms.
Create owner group-specific repair guidelines.
Effects of Climate Change (ECC)
(6 IRs)
(4 CRs)
Environmental Stressors Effects of increased atmospheric CO2 levels on concrete structures, combined effects of multiple Assess impacts of higher reservoir temperatures and their effects (e.g., algae formation) on
3-62 Category Initial Recommendations (IRs)
Combined Recommendations (CRs) environmental stressors, such as temperature and humidity, on concrete durability.
Impact of higher reservoir temperatures on the structural integrity and cooling systems of nuclear plants.
Assess the resilience and safety of concrete structures against extreme precipitation and potential flooding.
Evaluate the influence of rising sea levels on chloride diffusion and corrosion in concrete structures.
Explore the effects of increased tornado and tropical cyclone frequencies and examine potential increased risk on structural integrity.
Examine the kinetics of ASR under increased ambient temperatures and humidity.
nuclear plant structures and cooling systems.
Evaluate concrete resilience Investigate chloride diffusion and corrosion due to sea level rise.
Analyze structural impacts of increased tornadoes and tropical cyclones, extreme precipitation, flooding, and rising sea levels.
3.4.8.2 Final List of Recommendations for Input to the PIRT Process Effects of Radiation on Structures (RAD) around RPV RAD-1 Harvest and analyze service-irradiated concrete to verify accelerated laboratory data (e.g., flux, specimen size) and reduce uncertainties. Investigate the formation of new phases in concrete as concrete ages.
RAD-2 Compare radiation-induced free expansion data from accelerated experiment to effects under structural confinement. Explore damage depth.
RAD-3 Refine radiation transport models using realistic irradiation histories, especially as more plants exceed radiation threshold during LTO.
RAD-4 Verify attenuation of radiation through thickness of the structure and study its impact on material degradation.
RAD-5 Study interactions of irradiation with ASR gel stability, creep in cement paste, and corrosion of rebar due to gamma.
3-63 RAD-6 Collaborate globally to address LTO beyond 80 years.
RAD-7 Standardize testing methods, improve predictive models and structural integrity assessment.
Aging of Post-tensioned Containments (APC)
APC-1 Investigate long-term creep effects, structural modifications, SGR, loss of prestress, retensioning and interactions between original and replacement materials.
APC-2 Study prestress loss, drying, creep, and shrinkage using reliable, validated models to ensure structural safety under design-basis loading including accident conditions.
APC-3 Evaluate degradation mechanisms such as corrosion, ASR, freeze-thaw, carbonation etc., on structural integrity, as plants age.
APC-4 Use advanced monitoring systems to track deformation, cracking, and structural behavior, preventing unrepairable damage.
APC-5 Enhance modeling with data on aging factors to improve predictions and guide interventions.
APC-6 Collect real-world performance data through harvesting of inservice or decommissioned plant structures for validation and benchmarking.
Coupled Degradation Mechanisms (CDM)
CDM-1 Collect OE and conduct targeted studies on degradation mechanisms in reinforced concrete, including corrosion, leaching of calcium hydroxide, carbonation, chloride/sulfates penetration, alkali-aggregate reactions, and boric acid attack.
Incorporate lessons learned to inform future practices.
CDM-2 Investigate the synergistic effects of freeze-thaw cycles, chloride-induced corrosion, and exposure to demineralized water on concrete durability.
CDM-3 Design and perform experiments to understand coupled degradation mechanisms and validate findings with field observations and harvested material data.
CDM-4 Develop generic models to serve as guidance for plant-specific applications.
CDM-5 Expand predictive knowledge and develop tools to address future degradation scenarios.
Crevice Corrosion Liner (CCL)
CCL-1 Investigate crevice corrosion environment and crevice former and validate findings using OE and experiments CCL-2 Develop reliable comprehensive corrosion modeling CCL-3 Develop mitigation strategies to prevent crevice corrosion
3-64 Enhance Inspection and Monitoring (EIM)
EIM-1 Improve visual inspections and leverage innovative NDE techniques to detect degradation mechanisms and aging effects.
EIM-2 Develop and enhance Aging Management Programs (AMPs) for elastomeric materials used in civil structures and for coastal environments.
EIM-3 Integrate advanced technologies as AI, robotics, drones, building information models (BIM) for enhanced inspection efficiency, improved safety for personnel, increased accessibility, and streamlined data management.
EIM-4 Implement enhanced mitigation technologies, such as cathodic protection systems, drones, and remote monitoring methods.
EIM-5 Create databases to track degradation trends over time and develop novel inspection techniques for hard-toaccess or embedded structural areas.
.Repair and Replacement Strategies (RRS)
EIM-1 Evaluate repair strategies to support safe operation of nuclear plants beyond 80 years.
EIM-2 Research and validate new materials and methodologies. Address implications of transitioning to new cement standards.
EIM-3 Develop databases for repair methods and performance data. Share best practices through collaborative forums.
EIM-4 Implement systematic repair program to enhance structural durability and resilience.
EIM-5 Leverage proven repair techniques tailored to specific aging mechanisms. Develop plant-specific repair guidelines.
Effects of Climate Change (ECC)
EIM-1 ECC1. Assess the impacts of elevated reservoir temperatures on nuclear plant structures and cooling systems.
EIM-2 Evaluate the resilience of concrete structures under changing climate EIM-3 Investigate chloride diffusion and corrosion due to sea level rise.
EIM-4 Analyze structural impacts of increased tornadoes and tropical cyclones, extreme precipitation, flooding, and rising sea levels.
3.4.8.3 Summary of PIRT Results The structural materials concrete and components have been divided into seven categories; (1)
Effects of Radiation on Structures, (2) Aging of Post-tensioned Containments, (3) Coupled Degradation Mechanisms, (4) Crevice Corrosion, (5) Enhance Inspection and Monitoring, (6)
3-65 Repair and Replacement Strategies, and (7) Effects of Climate Change. The links for the PIRT scoring results are provided in Appendix C. The error bars associated with the point for each recommendation are a function of the average uncertainty scores, with longer error bars associated with higher uncertainty. The corresponding recommendation number is indicated next to the marker in the plot. Note that the average uncertainty scores have been converted to symmetrical error bars in this figure by assuming that the highest uncertainty (i.e., value of 3) corresponds to an importance uncertainty of 0.5 (i.e., 50 % of a complete level). Also, the knowledge scores have been slightly offset in some cases so that both the marker and the error bars associated with each recommendation number are clear.
Long-term high priority research recommendations from Session 8 were determined based on a combination of relatively high importance and relatively low knowledge scores thus placing them in the upper right quadrant of Figure 3.-9 to 3-15.
The average scores for the seven different recommendations related to effects of radiation on structures were calculated and they are depicted in Figure 3.
Figure 3-9 PIRT Results for Session 8 - Effects of Radiation on Structures SummaryEffects of Radiation on Structures The complete PIRT scoring and rationale for the effects of radiation on structures can be found in Session 81 PIRT Scoring-RAD. On Figure 3-9, it is noted that four recommendations, RAD1, RAD2, RAD5, and RAD6, fell in the upper right corner(red) higher importance and lower knowledge category. RAD3 fell in the red-orange transition zone of the upper right corner (a little better knowledge) but it remains in the higher importance and lower knowledge. RAD4 and RAD7 are also related to RAD1 which fell on moderate knowledge and high importance.
RAD1 is harvesting whereas RAD2 to RAD7 are the confirmatory research that can be done after RAD1.
3-66 For RAD1 (Harvest and analyze service-irradiated concrete to verify accelerated laboratory data and reduce uncertainties. Investigate the formation of new phases in concrete as concrete ages) the PIRT panelists expect research could advance knowledge to improve safety for LTO.
The panelists noted, verifying irradiated concrete data is important to reduce uncertainties in understanding the long-term radiation effects of irradiation on concrete structures. There are significant uncertainties in correlating accelerated test data with real-world behavior, and validation against harvested inservice materials is largely missing. While the size effect in concrete is understood, the behavior of small test specimens irradiated in test reactors may not accurately reflect the behavior of structural concrete under radiation. Harvested specimens from actual reactors are needed for true structural behavior insights. It is essential to reduce uncertainty by verifying various factors such as flux effect, internal strain relaxation, size effect, radiation-induced volumetric expansion (RIVE) under structural confinement, radiation transport, attenuation profile, degradation depth, and validating numerical approaches.
RAD2 (Compare radiation-induced free expansion data from accelerated experiment to effects under structural confinement. Explore damage depth) and RAD1 above are related. For RAD-2, the PIRT panelists expect research could advance knowledge to improve safety for LTO. The panelists noted, understanding radiation-induced expansion under confinement is important for structural integrity, but high uncertainty remains due to limited data for free expansion available from accelerated experiments. While some studies address free expansion and use numerical simulation for confinement effects, a comprehensive understanding of inservice performance of the structures is still lacking. Evaluating how irradiation-induced aggregate expansion is counterbalanced by paste shrinkage, quantifying the effects of radiation on triaxiality with respect to concrete depth, and incorporating new knowledge from Hamaoka concrete are essential. Additionally, damage under confinement, which includes intrinsic (residual, internal) strain and mechanical strain, extends deeper than the expansion zone according to numerical modeling and simulation which should be verified by harvesting materials from plants with appropriate pedigree.
For RAD5 (Study interactions of irradiation with ASR gel stability, creep in cement paste, and corrosion of rebar due to gamma) PIRT panelists expect research could advance knowledge to improve safety for LTO. The panelists noted the interaction of irradiation with ASR, creep, and corrosion of rebar is important, however, it is noted that there is significant uncertainty in these coupled effects, and there is limited data and lack of understanding. The extent to which radiation can precondition concrete biological shield (CBS) to alkali-silica reaction (ASR) is relatively unknown, and it is unclear whether radiation with liner-trapped moisture can initiate ASR. For thick concrete sections of the CBS wall covered with liners, the presence of trapped moisture within the concrete has not been explored. In addition, the interaction of creep of cement paste can accommodate some radiation-induced volumetric expansion (RIVE).
Panelists also noted that the effects of creep under radiation and corrosion of steel under gamma radiation are not well studied, especially considering CBS parameters and environment.
For RAD6 (Collaborate globally to address LTO beyond 80 years), the panelists noted the need to develop a consensus knowledge by sharing and leveraging experimental data, numerical simulations, predictive models, based on harvesting and testing. Global collaboration on LTO beyond 80 years can enhance efficiency of data sharing, testing standards, and research on radiation effects, while also advancing and implementing novel aging management strategies, harmonizing regulations, and supporting design of future nuclear power plants.
However, achieving meaningful results could be challenging due to variations in design and materials, construction methods, and accessibility, making it essential to focus on developing bounding and generic results. The PIRT panelists expect that global collaboration could
3-67 advance knowledge by avoiding duplicative efforts and making it cost and time efficient to improve safety for LTO.
For RAD3 (Refine radiation transport models using realistic irradiation histories, especially as more plants exceed radiation threshold during LTO), the panelists noted, as neutron and gamma exposures increase over time, verified and validated radiation transport methods are important to predict the levels of accumulated radiation (neutron, gamma, heating) for assessing potential degradation. Recently, NRC has published a research report NUREG/CR7281, Radiation Evaluation Methodology for Concrete Structures, for providing insights into the effect of material and geometrical variations on the magnitude of the neutron flux, heating rate, and gamma dose rates incident on the bioshield as well as their attenuation through the reinforced concrete structures. Such models must account for multiple parameters and be verified under realistic operating conditions and materials. Refining radiation transport models using realistic irradiation histories will improve the prediction of radiation-induced degradation in reactor structures, especially as more nuclear power plants exceed irradiation thresholds during LTO.
Improved definitions of transport effects are needed for material degradation and time-dependent loading, including thermal effects. This research can be implemented along with RAD1.
For RAD4 (Verify attenuation of radiation through thickness of the structure and study its impact on material degradation), PIRT panelists noted that while radiation attenuation is well understood, it is important to verify modeling data using data from inservice irradiated structures to reduce uncertainty and address knowledge gaps in assessing radiation-induced degradation.
Verifying numerical models with irradiated concrete cores from service environments can help reduce uncertainty and assess the structural implications of radiation exposure. One panelist noted numerous recent publications on the subject led to moderate scoring. However, this research is also related to harvesting in RAD1.
For RAD7 (Standardize testing methods, improve predictive models and structural integrity assessment), the panelists noted testing methods and multiscale modeling approaches have been developed, using the knowledge gained through accelerated experiments. After verifying these methods and models with real-world service-irradiated data, the understanding can be extended to develop user-friendly guidance and applied methods. Standardized testing and improved predictive models help reduce uncertainty, build shared understanding, and support sound safety and regulatory decisions. Additionally, standardizing coring techniques is important as they can affect concrete strength and modulus values and correlating the data from past small test specimens to standard harvested test specimens.
Research on the effects of radiation on concrete structures and harvesting concrete has been recognized as a high priority by several countries. Since the release of NUREG/CR7153 Volume 4 (commonly referred to as the EMDA Report), knowledge in this area has advanced using data from accelerated testing of small specimens. Nonetheless, harvesting is expected to offer valuable real-world insights and help reduce remaining uncertainties. The seven
3-68 recommendations outlined above are interconnected and should be pursued collectively to ensure a comprehensive understanding.
Figure 3-10 PIRT Results, Session 8 - Aging of Post-tensioned Containments SummaryAging of Post-tensioned Containments For aging of post-tensioned containments, the complete PIRT scoring and rationale can be found in Session 82 PIRT Scoring - APC. It is observed that two recommendations, APC1 and APC6, fall on the upper right quadrant on the red-orange transition zone indicating medium-low knowledge and high importance.
For APC1 (Investigate long-term creep effects, structural modifications, SGR, loss of prestress, retensioning and interactions between original and replacement materials), PIRT panelists expect research could advance knowledge to improve safety for LTO. The panelists noted, structural integrity in aging of post-tensioned containment could be influenced by long-term phenomena such as prestress loss, creep behavior, and interactions between original and replacement materials. While basic understanding of creep exists, practical application for LTO of post-tensioned containment (PCC) under real-world conditions remains limited and uncertain.
Long term creep including differential temperature, inelastic creep, reactivation of primary creep, creep under mixed stress state, and transient creep during accidental scenarios is not commonly considered in design, specifically in the context of retensioning and detensioning tendons, presence of small defects, the impact of plant modifications over time, creating a large opening for SGR, stress reversals transitioning to tension, Poissons creep effect, and radial strain. As a lesson learned from Crystal River NPP Unit 3 (CR3) containment, in 2019, the ASME Code has added a provision for radial reinforcement to resist cracking due to radial strain and preventing delamination, which will benefit future PCCs. On the other hand, the operating PCCs are likely to benefit from research.
3-69 For APC6 (Collect real-world performance data through harvesting of inservice or decommissioned plant structures for validation and benchmarking) the panelists noted that collecting real-world data by harvesting is valuable to validate models. Availability of decommissioned plants and plant owners willingness to share information could induce limitations. Before harvesting from an international site, applicability of the data to domestic fleets and NRC for LTO and future licensing should be carefully assessed.
On Figure 3.10, APC2, APC3, APC4 and APC5 fall on the upper left quadrant orange zone indicating better than moderate knowledge and high importance. All four of these recommendations can be considered as slightly lower importance as to some extent knowledge exists but structural significance is high because PCCs are relied on for safety.
For APC2 (Study prestress loss, drying, creep, and shrinkage using reliable models to ensure structural safety under design-basis loading including accident conditions), panelists noted that creep data commonly available is from short-term compression creep experiments. Reliable models should include combinations of various phenomena as noted above for APC1. Reliable models of prestress loss and creep will advance the understanding to ensure structural safety of the PCCs under design-basis loading including accident conditions.
For APC4 (Use advanced monitoring systems to track deformation, cracking, and structural behavior, preventing unrepairable damage), the panelists suggested that advanced monitoring systems enable early detection of the issues, prevent unrepairable damage, support informed decisions, and extend the safe, cost-effective operation of nuclear plants beyond 80 years.
Advanced monitoring systems may inform or alert when deformations, cracking, and overall behavior would exceed acceptance criteria. For bonded tendon plants monitoring effectiveness relies on Integrated Leakage Rate Testing (ILRT), which mostly now is performed every 15 years. Considering LTO >80 years and advancements of technology, modernized monitoring systems will enable more efficient and safer work environment, and improved structural monitoring. This topic is related to EIM4 (Implement enhanced mitigation technologies, such as cathodic protection systems, drones, and remote monitoring methods) which is presented.)
For APC3 (Analyze degradation mechanisms such as corrosion, ASR, freeze-thaw, carbonation etc., as plants age), the panelists noted that these aging mechanisms create additional complex conditions for post-tensioning structures with limited data. The effects of the degradation mechanisms listed have been examined, but the importance of the topic remains.
The susceptibility of a posttensioned containment to ASR is limited to aggregates used for its construction. An understanding of plant-specific and/or combined aging effects can be gained and issues resolved for the most part through computational mechanics (modeling and simulation) efforts incorporating field data. The aging mechanisms are reasonably understood.
However, the synergistic degradation effects for post-tensioned structures are more complex than RC structures. Early detection and preventive measures can avoid unacceptable degradation.
For (APC5 Enhance modeling with data on aging factors to improve predictions and guide interventions), the panelists noted that this research coupled with advanced monitoring systems can improve predictability, reduce uncertainty, and help identify when and where interventions are needed. However, developing criteria that would alter constitutive laws would be a challenge. Reliable monitoring data should be used for improving predictive modeling and simulation for enabling timely prevention or mitigation.
3-70 Figure 3-11 PIRT Results, Session 8 - Coupled Degradation Mechanisms (CDM)
SummaryCoupled Degradation Mechanisms The complete PIRT scoring, and the rationale for the scores, can be found in Session 83 PIRT Scoring - CDM. It is observed that recommendation CDM3 fell on the upper right quadrant on the red-orange transition zone indicating medium-low knowledge and high importance. CDM5 fell on the orange zone aligned on the medium knowledge and high importance. CDM1 and CDM4 fell on the left upper quadrant yellow zone. CDM1 and CDM4 are in the range of moderate-high importance and moderate-high knowledge. CDM2 fell into yellow-green zone, indicating moderate importance and to some extent better than moderate knowledge, like CDM4.
CDM3 and CDM5 are linked and based on Figure 311, grouped in the higher safety significance and lower knowledge category.
For CDM3 (Design and perform experiments to understand coupled degradation mechanisms and validate findings with field observations and harvested material data.), the panelists expect that analyzing OE to identify potential coupled degradation mechanisms, using them for improving knowledge, setting up experiments, and validating field data and trends can enhance aging management strategies to ensure safe LTO. Although, it could be a complex and demanding endeavor and site specific.
For CDM5 (Expand predictive knowledge and develop tools to address future degradation scenarios; provided CDM3 can deliver advancement of knowledge), the panelists noted that knowledge gained from CDM3 and developing tools will contribute directly to assessing and monitoring structural performance. This research is important and likely to be accomplished without major complexity.
3-71 CDM1 and CDM4 have high safety significance but are not recognized as high priority as discussed below.
For CDM1 (Collect OE and conduct targeted studies on degradation mechanisms in reinforced concrete, including corrosion, leaching of calcium hydroxide, carbonation, chloride/sulfates penetration, alkali-aggregate reactions, and boric acid attack. Incorporate lessons learned to inform future practices.), the panelists considered that reasonable knowledge on individual degradation mechanism exists. They suggested analyzing the insights gained from operating experience at the plant-specific level and advancing the knowledge for managing and/or mitigating coupled degradation modes. This knowledge has been developed and implemented in part. The proposed research could be viewed as a plant-specific activity.
For CDM4 (Develop generic models to serve as guidance for plant-specific applications.),
the panelists suggested that generic models provide a valuable foundation for plant-specific applications, but their development and implementation come with notable challenges. Due to limited access to site-specific data, researchers typically produce broader, generalized insights.
While these models offer practical guidance across various contexts, adapting them to individual plant conditions requires careful evaluation and modification to account for unique design and environmental factors. Although the creation of comprehensive generic models would be ideal, such an effort is likely to require substantial resources.
For CDM2 (Investigate the synergistic effects of freeze-thaw cycles, chloride-induced corrosion, and exposure to demineralized water on concrete durability.), the panelists noted that while the combination presents a potential degradation concern, existing knowledge and aging management practices have already been developed and partially implemented which provide a reasonable foundation for addressing these risks. As such, the proposed research is considered relatively accessible and additional research may not be of high priority.
3-72 Figure 3-12 PIRT Results, Session 8 - Crevice Corrosion Liner (CCL)
SummaryCrevice Corrosion Liner The complete PIRT scoring, and the rationales for the scores, can be found in Session 84 PIRT Scoring - CCL. All three recommendations CCL1, CCL2, and CCL3 fell on the upper left quadrant. CCL1 on the orange zone indicates better than medium knowledge and high significance. CCL2 fell on the yellow zone, indicating better than medium knowledge and medium-high significance. CDM3 fell on the left upper quadrant greenish-yellow zone indicating a trend toward high significance, which is well understood. This topic was explored previously in the USA and recently ACES Project in Europe investigated this topic further. In the workshop the lead researchers presented the study that was part of ACES project.
For CCL1 (Investigate crevice corrosion environment and crevice former and validate findings using OE and experiments), the panelists noted that crevice corrosion is a form of localized corrosion that frequently develops in occluded metallic structures, especially those containing iron. Although it is a well-documented and common degradation mechanism, identifying the specific microenvironment that drives its occurrence can be challenging. The significance of crevice corrosion is ranked based on factors such as its importance, safety implications, and the degree of uncertainty, which is closely tied to how well its microenvironment is understood.
Gaining a clear understanding of the crevice environment as well as the materials that contribute to corrosion is essential to prevent accelerated degradation, particularly under LTO conditions. Hence CCL1 could benefit from research to improve the understating and developing preventive or mitigation methods.
CCL2 (Develop reliable comprehensive corrosion modeling)
3-73 Corrosion imposes significant costs, necessitating the development of reliable models, predictive tools to assess material degradation in specific electrochemical environments, and proven inspection methods especially for LTO >80. This topic is of moderate-high significance and some modeling approaches exist. However, additional research can improve the reliability of comprehensive corrosion modeling.
For CCL3 (Develop mitigation strategies to prevent crevice corrosion) falls under the category that adequate knowledge is available, additional research may not be of high priority.
Figure 3-13 PIRT Results, Session 8 - Enhance Inspection and Monitoring (EIM)
Summary-Enhance Inspection and Monitoring (EIM)
The complete PIRT scoring, and the rationale for the scores, is contained in Session 85 PIRT Scoring - EIM. It is observed that two recommendations, EIM4 and EIM5, fall on the upper middle orange zone indicating medium knowledge and high importance or safety significance, however, uncertainty associated with EIM5 is lower than EIM4. Recommendations EIM1, EIM2, and EIM3 also fall in the upper left quadrant indicating knowledge better than medium.
However, the safety significance for EIM3 (lighter orange) is little higher than EIM1 (lighter orange) and safety significance for EIM1 is higher than EIM2 (yellow zone). These items are all considered high significance and relatively low knowledge that would benefit from additional research to potentially improve aging management.
For EIM4 (Implement enhanced mitigation technologies, such as cathodic protection systems, drones, and remote monitoring methods), the panelists expected that research can improve inspection, monitoring and aging management and enhance safety. Use of proven methods, including newer technologies, will make monitoring safer and more cost efficient. This in part has been developed and implemented in non-nuclear industry and for nuclear plant structures
3-74 abroad. Cathodic protection is an example of an established method. Use and availability of drones, sensors and other remote monitoring are becoming common. The proposed research could be viewed as relatively less complex but will enhance aging management greatly.
For EIM5 (Create databases to track degradation trends over time and develop novel inspection techniques for hard-toaccess or embedded structural areas), the panelists considered that Integrated databases within fleets and specific industry organizations exist that enable improvements in industry performance. Research can improve inspection for hard-toaccess areas including high elevation such as containment dome, areas blocked by equipment/structure, location of unsafe environment (corrosive chemical, radiation etc.) except for those that are buried.
EIM -3 and EIM1 are related and both are of high safety significance. It is likely that they can improve aging management directly as discussed below.
For EIM3, (Integrate advanced technologies like AI, robotics, drones, building information models (BIM) for enhanced inspection efficiency, improved safety for personnel, increased accessibility, and streamlined data management), the panelists recommended that this research can be considered as plant modernization that can improve personnel safety and increase efficiency. Inspection cost may increase initially but it may save expensive repair, plant outage, or even shut down. It is important to note that AI needs to be tailored to specific or group of plants, just as plant specific programs apply to specific plants. Algorithmic bias, not applicable OE, and poor-quality data could skew results at the end creating more issues at hand.
For EIM1 (Improve visual inspections and leverage innovative NDE techniques to detect degradation mechanisms and aging effects), the panelists suggested that in many situations, visual inspection may not be effective in timely detection of degradation. Integrating NDE techniques in high-risk areas through a risk-informed approach enhances detection while balancing cost and effort. Robotic machine vision in challenging environments has been accomplished in space exploration. Application of that technology to the nuclear fleet is possible. Based on OE, critical plant locations for machine visual inspections could be identified and programmed for NDE detection and repairs.
For EIM2 (Develop and enhance Aging Management Programs (AMPs) for elastomeric materials used in civil structures and for coastal environments.), some research on material chemistry could be beneficial. This research has been proposed to be addressed by IAEA, and thus it was concluded not to be a high priority for the NRC research strategy.
3-75 Figure 3-14 PIRT Results, Session 8 - Repair and Replacement Strategies (RRS)
Summary - Repair and Replacement Strategies (RRS)
The complete PIRT scoring, and the rationales for the scores, can be found in Session 86 PIRT Scoring - RRS. Recommendation RRS2 falls on the upper middle orange zone, indicating knowledge is medium and importance is high. RRS1, RRS3, and RRS5 are on the upper left quadrant orange zone, indicating close to medium knowledge and high importance. RRS4 falls on yellow-green zone, indicating relatively higher knowledge and of medium importance.
RRS2, RRS1, RRS3, and RRS5 are related and identified as relatively high safety significance and lower knowledge (medium-high priority). Although knowledge is medium or better, repair and replacement strategies are an important topic for very LTO (LTO>80 years).
For RRS2 (Research and validate new materials and methodologies. Address implications of transitioning to new cement standards, high strength steel rebars and concrete.), panelists noted that researching and validating new materials and methods is important for ensuring durable and safe structures while modifying existing aged structures. While not the highest priority, this work is worthwhile given the possible widespread use of these materials and their significant impact on long-term performance. Materials and methods may become obsolete or unavailable with time. Pros and cons for newer materials and methods should be evaluated.
This research could be a good candidate for collaboration through specialized research organizations including industry engagement and input.
For RRS1 (Evaluate repair strategies to support safe operation of nuclear plants beyond 80 years.), the panelists thought that repair strategies proven in both nuclear and non-nuclear industries should be explored for timely maintenance actions, especially as plants age beyond 80 years and structural repairs become more common.
3-76 For RRS3 (Develop databases for repair methods and performance data. Share best practices through collaborative forums.), the panelists noted that information sharing is often hindered by policies, confidentiality and regional variability, resulting in inconsistent approaches. However, valuable insights may still be leveraged by sharing best practices through collaborative forums.
For RRS5 (Leverage proven repair techniques tailored to specific aging mechanisms. Develop plant-specific repair guidelines.), panelists suggested that drawing on OE and repair strategies that are proven in both nuclear and non-nuclear sectors should be actively explored, especially as cumulative aging effects occur, repairs will become increasingly necessary in plants operating beyond 80 years.
For RRS4 (Implement systematic repair program to enhance structural durability and resilience.), the panelists commented that ideally a standardized, non-deferrable maintenance strategy can extend plant life, enhance durability, and prepare for unexpected issues. However, its implementation will depend on corporate norms and plant-specific challenges. This is a topic for implementation for which there is time for plants to reach 80 years; at this point this activity is not urgent.
Figure 3-15 PIRT Results, Session 8 - Effects of Climate Change (ECC)
Summary - Effects of Climate Change (ECC)
The complete PIRT scoring, and the rationales for the scores, can be found in Session 87 PIRT Scoring - ECC. Recommendation ECC2 falls on the upper left yellow-orange zone, indicating knowledge is better than medium and importance is to some extent higher than medium.
ECC4, ECC1, ECC3 are on the lower left quadrant green zone, indicating higher than medium knowledge and lower importance.
3-77 CNSC presented this topic of their ongoing research activity. This is an important topic for awareness and tracking the gradual change of climate. The PIRT panelists expect that it could be part of a structural monitoring program by default. Also, the outcome of the recommendations for Repair and Replacement Strategies (RRS) would be useful for mitigating gradual climate change. The remarks from the PIRT panelists are noted below.
For ECC2 (Evaluate the resilience of concrete structures under changing climate), the panelists noted that concrete is generally durable and can withstand extreme conditions, often surpassing building code expectations and proving repairable in many cases. However, when degradation does occur, repairs can be complex and challenging. For structures operating beyond 80 years, depending on the climate, deterioration may accelerate, making the assessment of resilience of concrete structures critical for ensuring continued safe operation.
For ECC1 (Assess the impacts of elevated reservoir temperatures on nuclear plant structures and cooling systems.), panelists agreed to the importance of tracking climate change and taking additional measures if applicable; awareness is important. This may have a direct impact on water-control structures only.
For ECC4 (Analyze structural impacts of increased tornadoes and tropical cyclones, extreme precipitation, flooding, and rising sea levels), the panelists noted that additional measures may be needed if they exceed the original design criteria.
For ECC3 (Investigate chloride diffusion and corrosion due to sea level rise.) the panelists noted that this topic is important for the plants near the coasts; in general, chloride diffusion and corrosion are part of the aging management.
3-78 3.4.9 Consolidation of PIRT Results This subsection highlights certain research recommendations that the staff determined would be most beneficial for the NRC to consider conducting research in the future. These recommendations are a subset of the recommendations determined to be higher priority by the PIRT process for each session. This section discusses additional considerations which led to the staff highlighting these recommendations. This subsection also discusses research recommendations from some sessions that were scored relatively high by the PIRT process, but are not being highlighted by the NRC staff for future research, with explanation provided for these decisions. The recommendations highlighted for future research consideration are in bold text.
The following recommendations are highlighted for consideration of future research by NRC:
Session 1, RPV. The PIRT panel identified RPV3, -7, and -12 as research topics that could benefit from additional research to help improve our knowledge of safety-significant topics and reduce uncertainty to help clarify the importance to safety. RPV3 and RPV7 are highlighted as topics for consideration for future research by NRC. RPV12, the other higher priority recommendation from Session 1, is not recommended for future research by NRC.
o RPV3 Harvesting of reactor pressure vessel support materials could be beneficial for improving embrittlement predictions for these materials. This recommendation is highlighted because there has been little or no research on embrittlement of RPV supports, and actual inplant aged materials are the best way to obtain representative data on embrittlement of structural steels under low-temperature and low-fluence conditions.
o RPV7 Improve existing embrittlement trend curves, such as ASTM E 900, and incorporate improvements into regulatory guidance. This recommendation is highlighted because it had the highest importance among the Session 1 recommendations and it aligns with the NRC concern with non-conservatism of the current NRC-approved embrittlement trend curve in RG 1.99, Rev. 2 at high neutron fluence.
o RPV12 Additional surveillance testing is needed to generate high-fluence embrittlement data. This recommendation was one of the higher priority recommendations identified by the PIRT. While the staff considers this an important area of research, and agrees such data is needed, the industry already had existing programs to generate more high-fluence data, so this is of lower priority for NRC to sponsor research.
Session 2, RVI. The PIRT panel identified RVI3, -4, -6, -7, and -9 as research topics that could benefit from additional research to help improve our knowledge of safety-significant topics and reduce uncertainty to help clarify the importance to safety. Of these topics RVI3,
-6, and -9 were determined to be the highest priority topics for further consideration due to both lack of knowledge and determined level of importance. RVI9 was considered the highlighted topic to be considered for future research by NRC.
o RVI9: Accelerate guidance on qualification and licensing of advanced materials and repair methods. The reason for this recommendation being highlighted as the highest priority for research is threefold: 1. advanced materials and repair methods are of
3-79 rising interest throughout industry; 2. significant knowledge gaps need to be addressed; and 3. there are current time-sensitive efforts throughout the NRC to develop guidance on licensing activities for advanced manufacturing and materials.
o RVI3: Develop degradation models that can better reduce reliance on extrapolation and better predict complex interactions between SCC, fatigue, thermal aging, and irradiation effects; and RVI6: Create models and establish experimental studies to quantify the effects of irradiation, transmutation, and gas accumulation on phase stability and mechanical degradation; were both topics of high importance to the panelists, however it was determined that focusing on RVI9 would be the most valuable in the near-term.
Session 3RCPB. The PIRT panel identified RCPB11, -12, and -13 as three research topics that could benefit from additional research to help improve our knowledge of safety-significant topics and reduce uncertainty to help clarify the importance to safety.
o RCPB13 Perform research to improve accelerated testing methods to avoid altering microstructures too much while still obtaining useful data. This involves finding a balance between accelerating tests and maintaining the integrity of the materials being tested. This recommendation was determined to be the most important for consideration for future research because we employ accelerated testing methods to obtain almost all PWSCC-related data and regulatory decisions are being made based on this data. To the best of our knowledge, no systematic research on accelerated testing effects on key PWSCC parameters has been performed on the materials of interest.
o RCPB11 and RCPB12, while being topics of interest, are being addressed by current or planned NRC, DOE, and/or EPRI research projects.
Session 4Secondary Side. The staff determined that none of the recommendations should be highlighted as higher priority research recommendations to support operation beyond 80 years.
Session 5BOP. The staff highlighted one recommendation for consideration for future NRC research.
o BOP6 Develop advanced methods to inspect CIPP/SIPP: Even though research has started in this area, it is only at the beginning stages and more R&D funding and exploration is needed. There is an expectation that CIPP/SIPP use will become predominant as the plants age and there is still great uncertainty about the long-term performance of these systems in nuclear environments. Lacking this information makes it more important to develop techniques that can identify and interrogate the different degradation modes that are associated with CIPP/SIPP materials.
Session 6Fatigue. The PIRT panel identified two higher priority recommendations:
FAT2 (Evaluating the effects of irradiation on fatigue lives) and FAT5 (Develop fatigue prediction and assessment methods for AMmanufactured safety-significant components. Although they are both important topics, the staff does not believe either of these warrant elevation to the summary section for the following reasons:
3-80 o FAT2: Irradiation affects only a subset of safety-significant components that may be subject to fatigue. There is information on CGRs in these environments for established cracks, so the biggest gap is with respect to irradiation-induced crack initiation in these environments. While more information in this area would be helpful, it is not clear that the information will really be valuable to impact either the design of replacement components or the current inspection regimes for existing components.
Many of the safety-significant components in these locations have both redundancy and/or significant design margins against failure.
o FAT5: This issue only affects potential AM components. While it is expected that AM will be used more frequently as plants age, initial efforts have focused on qualifying AM components consistent with code qualification requirements for non-AM components. Fatigue qualification is part of the qualification requirements. Also, many of the fatigue issues are related to surface roughness issues, material anisotropy, and near-surface porosity. These issues can be mitigated, at least somewhat or entirely, by post-fabrication machining and heat treatment and using appropriate process parameters and other QA/QC measures to control porosity.
Ultimately, development of more generic fatigue prediction and assessment methods are topics more appropriate for the industry and DOE if some of these processing and fabrication requirements are to be relaxed.
Session 7, Mitigation - The staff determined that none of the recommendations should be highlighted as higher priority research recommendations to support operation beyond 80 years.
Session 8, Civil Structures Concrete and Components consists of seven categoriesRAD, APC, CDM, CCL, EIM, RRS, ECC. The PIRT panel identified RAD1, APC1, CDM3, and EIM3 & EIM1 as higher priority research topics that could bring valuable insights, reduce uncertainty, and improve knowledge for LTO. These topics are recommended to discuss with our partners and stakeholders for further discussion and collaboration.
o RAD1 Harvest and analyze service-irradiated concrete to verify accelerated laboratory data (e.g., flux, specimen size) and reconciling calculated vs actual neutron fluence, gamma dose for reducing uncertainties. Investigate the formation of new phases in concrete as concrete ages). RAD2, RAD3, RAD4, and RAD7 are the various prongs of the research recommendations using harvested materials. This research is highlighted because there is no relevant real-world data available which can be used to reduce uncertainty by verifying various factors such as flux effect, size effect, radiation-induced volumetric expansion (RIVE) under structural confinement, radiation transport, attenuation profile, degradation depth, and validating numerical approaches. This research should be collaborated through OECD/NEA and ICIC in the international arena and with domestic partners EPRI, DOE/ORNL, DOE/NSUF.
o APC1 Investigate long-term creep effects, structural modifications, SGR, loss of prestress, retensioning and interactions between original and replacement materials.
This research is likely to improve knowledge on long-term creep including differential temperature, inelastic creep, reactivation of primary creep, creep under mixed stress state, transient creep during accidental scenarios which are not commonly considered in design, specifically in the context of retensioning and detensioning tendons, presence of small defect, the impact of plant modifications over time, creating large opening for SGR, stress reversals transitioning to tension, Poissons creep effect, and radial strain.
3-81 o APC6 Collect actual inplant aged materials performance data through harvesting of inservice or decommissioned plant structures for validation and benchmarking. This research is highlighted for confirmation of APC1 using field data, but subject to the availability of materials for harvesting from appropriate post-tensioned containment.
o CDM3 Design and perform experiments to understand coupled degradation mechanisms and validate findings with field observations and harvested material data.
This research has been highlighted because it is likely that analyzing OE to identify potential coupled degradation mechanisms, using them for improving knowledge, setting up experiments, and validating field data and trending can enhance aging management strategies to ensure safe LTO. The research could be a complex and demanding endeavor and site specific.
o EIM3 Integrate advanced technologies that include demonstrated AI, robotics, drones, building information models (BIM) for enhanced inspection efficiency, improved safety for personnel, increased accessibility, and streamlined data management, and EIM1 Improve visual inspections and leverage innovative NDE techniques to detect degradation mechanisms and aging effects, are grouped together. In many situations, visual inspection may not be effective in timely detection of degradation. Integrating NDE techniques in high-risk areas through a risk-informed approach enhances detection while balancing cost and effort. Use of proven methods and including newer technologies will make monitoring safer and more cost efficient. The proposed research could be viewed as relatively less complex but will enhance aging management greatly. Evaluate worldwide methodologies used to help improve aging management activities.
o RRS2 Research and validate new materials and methodologies. Address implications of transitioning to new cement standards, high strength steel rebars and concrete. This research is important, however, not the highest priority. This research work is worthwhile given the possible widespread use of these materials and their significant impact on long-term performance. Materials and methods may become obsolete or unavailable with time. Pros and cons for newer materials and methods should be evaluated. This research could be a good candidate for collaboration through specialized research organizations including industry engagement and input where NRC may join as observer.
4-1 4
SUMMARY
The NRC staff conducted a workshop entitled NRC Workshop on Structural Materials:
Research for 80 Years and Beyond, October 13, 2024, at NRC Headquarters and virtually.
The NRC staff designed the workshop to provide input to aid in development of an NRC strategy for LTO. This strategy assesses the state of knowledge related to aging management for plant operation beyond 80 years and provides recommendations on any research needs to inform the existing AMPs or identify any needed modifications or additions.
Participants in the workshop were from diverse organizations including the nuclear industry, academia, regulators, and national laboratories, from approximately 20 countries, all of which are engaged in research on materials aging to support LTO of LWRs.
The workshop consisted of eight technical sessions, with seven sessions covering aging issues of metallic components and materials, and one session covering seven topics of aging issues for structural components and materials, particularly concrete structures. Each session featured several short presentations by researchers, followed by a discussion by a panel of technical experts. The technical session presentations and panel discussions are summarized in this report.
Following the workshop, the NRC staff identified key research recommendations from the presentations and panel discussion for each session. The NRC staff then conducted a PIRT process to determine which of the recommendations should be considered for future research by the NRC. This report describes the PIRT process which was implemented by an individual team consisting of NRC SMEs for each session. The report also details consolidated research recommendations that the NRC staff determined should be considered for future research to support LTO beyond 80 years. These highlighted recommendations are a relatively high-priority subset of all research recommendations arising from the workshop as determined by the PIRT.
5-1 5 REFERENCES
- 1. NUREG/CR7153 Vol 1, "Expanded Materials Degradation Assessment (EMDA): Executive Summary of EMDA Process and Results." October 31, 2014 (ADAMS Accession No. ML14279A321)
- 2. NUREG/CR7153 Vol 2, "Expanded Materials Degradation Assessment (EMDA): Aging of Core Internals and Piping Systems." October 31, 2014 (ADAMS Accession No. ML14279A331)
- 3. NUREG/CR7153 Vol 3, "Expanded Materials Degradation Assessment (EMDA): Aging of Reactor Pressure Vessels." October 31, 2014 (ADAMS Accession No. ML14279A349)
- 4. NUREG/CR7153 Vol 4, "Expanded Materials Degradation Assessment (EMDA): Aging of Concrete and Civil Structures." October 31, 2014 (ADAMS Accession No. ML14279A430)
- 5. NUREG/CR7153 Vol 5, "Expanded Materials Degradation Assessment (EMDA): Aging of Cables and Cable Systems." October 31, 2014 (ADAMS Accession No. ML14279A461)
- 6. NUREG/CR6923, "Expert Panel Report on Proactive Materials Degradation Assessment."
February 28, 2007 (ADAMS Accession No. ML070710257)
- 7. NUREG/CR6923, "Expert Panel Report on Proactive Materials Degradation Assessment,"
Appendix A, Materials Degradation Modes and Their Prediction - Appendix B, Background Papers. February 28, 2007 (ADAMS Accession No. ML070710258)
- 8. NUREG/CR6923, "Expert Panel Report on Proactive Materials Degradation Assessment,"
B.8, "Stress Corrosion Cracking of Carbon and Low Alloy Steels" - Appendix C, Panel Members. February 28, 2007 (ADAMS Accession No. ML070710260)
- 9. Summary Report of the International Concrete Harvesting Workshop, June 6, 2025 (ADAMS Accession No. ML25161A138)
A-1 APPENDIX A WORKSHOP ATTENDEES Last Name First Name Organization Abdallah Maha CNSC-CCSN Abdul Habib Mohamed National Nuclear Regulator Adia Jean-Luc EDF Adjirackor Theophilus Narh-Korli Nuclear Regulatory Authority Alexandreanu Bogdan Argonne National Laboratory Alinejad Majid Iran Nuclear Regulatory Authority Allik Brian U.S. NRC Alnaggar Mohammed ORNL Alvarado-Guilloty Lydiana U.S. NRC/NRR/DNRL/NCSG Amberge Kyle EPRI Andresen Péter Andresen Consulting Androut Cédric CNSC-CCSN Annett James Constellation Anor-nyarko Michael Nuclear Regulatory Authority Ghana Araguas Christian U.S. NRC Arai Taku CRIEPI Asada Seiji Mitsubishi Awudi-Dekpeh Emefa Ghana AEC Secretariat Badenhorst Anton ESKOM Koeberg Nuclear Power Station Bahr Ludwig GRS Bailey Marissa U.S. NRC Baker Tod Westinghouse Barbier Gauzelin IRSN Barborak Darren EPRI Barr Christopher DOE-NE Bastawros Ashraf Iowa State University Bauer Laurel U.S. NRC Bayssie Mekonen U.S. NRC Becker Donald U.S. NRC Behardien Naa-ilah Eskom Holdings SOC Ltd Benson Michael U.S. NRC Blankenship William Dominion Engineering
A-2 Last Name First Name Organization BlomstrÃm Johan unknown Bloom Steven U.S. NRC Bond Leonard Iowa State University Bonsu Kwabena Ghana AEC Bouhjiti David IRSN Bouhrizi Sofia EDF Bourguigne G.
unknown Bowers Crystal Duke Energy Boyan Dallas AEP Bozga John U.S. NRC Braden Michael Sargent & Lundy Bran Anleu Paula ORNL Brooks Adam ORNL Bruiners Rodger National Nuclear Regulator Buford Angie U.S. NRC Burgos Brian EPRI Burkardt Markus Dominion Engineering Burke Grace INL Burton Christopher Structural Technologies Byrnes Nicholas unknown Candelario-Quintana Luissette U.S. NRC Carlin Daniel Westinghouse Electric Co Cavallo Jon Jon R. Cavallo, PE Cervenka Petr MIT Chen Xiang (Frank)
ORNL Chen Jian ORNL Chen Yiren Argonne National Laboratory Chitose Keiko OECD NEA Choi Jinbok KAERI Cholakian A. Egon Harvard/ NIH/Oxford (Retired)
Chu Yi-Lun U.S. NRC Cicero Roman Innomerics Cimock Dylan EPRI Cinson Tony Framatome Cirilli Jim EPRI Coetzee Ubert National Nuclear Regulator Collins Jay U.S. NRC Cooper Paula U.S. NRC Region II
A-3 Last Name First Name Organization Cordes James court reporter Costa Oriol JSI, Jozef Stefan Institute Cotte Anne-Francoise EDF Couvant Thierry EDF R&D Craft Nathaniel Dominion Engineering Cumblidge Stephen U.S. NRC Cunningham Kelly Nuclear Science User Facilities Dale E
unknown Dallas J
unknown Damiani Tom EPRI Daniel Jason U.S. NRC Daniels Darrel Eskom de Araujo Jenna Eskom de Graaf Brandon Constellation De Souza Valdir SCK CEN Delisle Dale U.S. NRC Dewynter Veronique CEA Dijamco David U.S. NRC Domke Matt U.S. NRC Doorene S.
unknown Dring Ralph ENSI Downey Steven U.S. NRC Region II Dumaine Mario AkzoNobel Coating Eastman John unknown Ebrahim Gadija Eskom Efsing Pl Vattenfall/Ringhals AB Elen Muthu PNNL Fairbanks Carolyn U.S. NRC Ferreira Miguel VTT Technical Research Centre Fifield Leo PNNL Files Mary unknown Florencia Renteria del Toro WiN Global Young Generation Focht Eric U.S. NRC Franco Ingrassia Nucleoelectrica Argentina S.A.
Gabor Petofi IAEA Gallman Aaron U.S. NRC NRR/DEX/ESEB Ganjigatte Umapathy Indian Inst. Technology Delhi
A-4 Last Name First Name Organization Garg Krishan PSEG Nuclear Garner Frank Texas A&M University Gasparrini Claudia Jensen Hughes/Imperial College Gavula James U.S. NRC Gee Petra Innomerics Geza Siyabulela National Nuclear Regulator Ghosh Amitava U.S. NRC/NRR/DEX/ESEB Gift Frank EPRI Gilman Timothy Structural Integrity Associates Glass Samuel (Bill)
PNNL Goedjen Jackson Johns Hopkins University Goetz Sujata DOE-NE Golliet Matthew Westinghouse Gotou Mineo Japan NUS Co., Ltd.
Govender Sasha Eskom Gray Mark Westinghouse Griesbach Timothy Structural Integrity Associates Gubela Wonderboy National Nuclear Regulator (RSA)
Guillodo Michael Framatome Gunter Paul Beyond Nuclear Gussev Maxim ORNL Hahn Matthew DOE-NE Hakii Junichi TEPCO Hall James Brian Westinghouse Hannon Jacob unknown Hata Kuniki Japan Atomic Energy Agency Haywood Emma U.S. NRC Heffner Andrew Westinghouse Heidrich Brenden INL Heller Kevin U.S. NRC Henkerman Zharea Koeberg (Eskom)
Heo YeongAe Sandia National Laboratories Hiser Allen Self Hojo Kiminobu Mitsubishi Heavy Industries Holden-Remick Jonathan Radiation Safety and Control Services Honcharik John U.S. NRC/NRR/DNRL/NPHP Hull Amy U.S. NRC
A-5 Last Name First Name Organization Hurley Monica EPRI Huxmann William Portsmouth STEM Academy Huxmann Jessica Alabama Connections Academy Huxmann Jeffrey Huxy, Inc.
Ingrassia Franco Nucleoelectrica Argentina S.A.
Ista Ata U.S. NRC Ito Kiichi Japan NUS Iwamatsu Fuminori Hitachi, Ltd.
Jacobus Cornelius Talen Energy Jacquemain Didier OECD NEA Jaffer Shahzma CNL James Randy Structural Solutions Consulting, LLC Jatav Hemant Devi Ahilya University Jenkins Joel U.S. NRC Jenssen Anders Studsvik Nuclear AB Johnson Samuel EPRI Johnson Andy U.S. NRC/NRR/DNRL/NCSG Jones Christopher Kansas State University Joshua Morton Vistra Jun Jiheon ORNL Junji Eto Mitsubishi Research Institute, Inc.
Kalikian Varoujan U.S. NRC/NRR Kalnas Ronald Fluor Marine Propulsion Kanno Masanori CRIEPI Karkkainen Ryan UT Battelle Ke Jia-Hong INL Kearns Kieran Constellation Energy Kehler Beth Dominion Energy Kim Heejeong University of Arizona Kim Sungwoo KAERI Kim Jongmin KAERI Kim Yongsang KINS Kim Hongdeok KAERI King Christine EPRI Kirk Mark PEAI Klein Paul U.S. NRC Kojima Masayoshi Nuclear Regulation Authority (NRA)
A-6 Last Name First Name Organization Kontani Osamu Kajima Corporation Koopman Magrieta Eskom Kordina Tomas State office for nuclear safety Koutecka Gabriela Czech Technical University Kriz Antonin State Office for Nuclear Safety -
CZ Kumar Srijan NRG Kuutti Juha VTT Technical Research Centre Lai Shao U.S. NRC Lauferts Ulrike GRS; Gesellschaft für Anlagen-und Reaktorsicherheit Laukkanen Anssi VTT Technical Research Centre Lehman Bryce IAEA LePape Yann ORNL Lesica Susan DOE-NE Levitus Steven U.S. NRC Li Huan U.S. NRC Li Meimei Argonne National Laboratory Licina George Structural Integrity Associates Lindqvist Sebastian VTT Technical Research Centre Lisová Dana State office for nuclear safety Liu X
Westinghouse Long Elliot EPRI Lorenzo Stefanini NRG Luissette Candelario-Quintana U.S. NRC Lunceford Wayne EPRI Lundqvist Péter Vattenfall AB / Energiforsk Magnuson Eric U.S. NRC Makar Greg U.S. NRC Malaka Sammy NECSA - SAFARI-1 Malikowski Heather EPRI Manenc Christelle IRSN Mann Jonathan Amentum Manoly Kamal U.S. NRC/NRR Maples-Abidi Cindy TVA Marquie Christophe IRSN Maruyama Ippei The University of Tokyo Masson Benoit EDF
A-7 Last Name First Name Organization Mathebula Muhluri National Nuclear Regulator Matjee Mapula Eskom Matthews Samantha U.S. NRC Mays Ben Westinghouse McCormick Kevin U.S. NRC McGuire Francis Constellation McKirgan John U.S. NRC McNeil Péter RSCS Mdluli Garetshose Eskom Medoff James U.S. NRC Mehmood Tariq PAEC Melchionna Jim PSEG - Salem/Hope Creek Min Seung U.S. NRC Mitsuhashi Tadahiro Toshiba Miyazaki Masato Mitsubishi Heavy Industries Momoti Songo Eskom Mondal Kunal ORNL Mori Atsushi Toshiba Morito Ryo CRIEPI Morley Andrew Rolls-Royce Submarines Limited Morris Derek unknown Moyer Carol U.S. NRC Mphahlele Koketso Koeberg Nuclear Power Station Muggleston Casey Constellation Energy Mura Michelle EPRI Nagai Masaki CRIEPI Nahas Georges IRSN Naicker Balin Nakoski John NEA Nelson Jade GXO Logistics Nhleko Sifiso National Nuclear Regulator Nicolson Haco Eskom Nie JS U.S. NRC Nikaeen Ali unknown Nio Daisuke Japan Atomic Energy Agency Nomnnganga Loyiso Eskom Nomoto Akiyoshi CRIEPI Nove Carol U.S. NRC
A-8 Last Name First Name Organization Oakman Jamie Westinghouse Obeng Asiedu Godfred Ghana AEC Ogawa Takuya Toshiba Orita Shuichi TEPCO Ortner Susan UK National Nuclear Laboratory Osaki Takashi Chubu Electric Power Co., Inc.
Oumaya Toru Kansai Electric Power Co.
Palm Nathan EPRI Palmer Eric U.S. NRC Palou Martin Slovak Acad Scie, Inst. Construct
& Architec.
Park Si Hwan U.S. NRC Park Joon U.S. NRC, Region 3 Parker Cory U.S. NRC Payne Ryan U.S. NRC Peinado Javier Consejo de Seguridad Nuclear Peng Yan China Inst. Atomic Energy Persoz Matthieu EDF Petofi Gabor IAEA Pires Jose U.S. NRC Pitora Vladislav UJV Rez Poehler Jeffrey U.S. NRC Pope Steven ISL Prinaris Andrew U.S. NRC/NRR/DEX/ESEB Purtscher Patrick U.S. NRC Raiman Stephen University of Michigan Rama Neelan Eskom Renteria Florencia Women in Nuclear Young Generation Retfalvi Eszter Hungarian Atomic Energy Authority Rezai Ali U.S. NRC Ribeiro Fabienne IRSN Rivera Isella unknown Rosario Eucherius U.S. NRC Ross Miranda U.S. NRC Rosseel Thomas Imtech Subcontract to ORNL Rova Bob DOE-NE Rudland David U.S. NRC
A-9 Last Name First Name Organization Ruffin Steve U.S. NRC Rydlova Jolana State Office for Nuclear Safety Sabatino Samantha ORNL Sagals Genadijs CNSC-CCSN Saito Toshiyuki Toshiba Sakaguchi Shohei The Kansai Electric Power Co.,
Inc Sampson Michelle U.S. NRC Sanchez Javier IETcc-CSIC Sarrafian Joshua Constellation Sato Toshihiko Mitsubishi Heavy Industries Saueressig Patrick PNPP Schwartz Zach Constellation Seber Dogan Self Employed Seppnen Tommi VTT Technical Research Centre Shah Mansi Radiation Safety & Control Services Shaikh Atif U.S. NRC Sham Ting-Leung U.S. NRC Sharida Ullah Xcel Energy Sheehan Neil U.S. NRC Shim Do Jun EPRI MRP Sida Karen U.S. NRC Siddiqui Muzzammil U.S. NRC Silva Juan U.S. NRC Sims Paul unknown Sircar Madhumita U.S. NRC Smith Janine U.S. NRC Smith Jared EPRI Smith Steven Fluor Marine Propulsion Smith Jean EPRI Sobek Kamil Research Center Rez Sock Frederick U.S. NRC Sokolov Mikhail ORNL Soliman Alix Nature Magazine Song Rongjie INL Song Tae-Kwang KINS Spls Johanna Ringhals NPP / Vattenfall Sperbeck Silvio GRS
A-10 Last Name First Name Organization Steinfeldt Thomas U.S. NRC Stemberk Petr Czech Technical University Stevens Gary Self Strongbow Richmond Ghana AEC Sun Hongbin ORNL Suty Martina ENSI Suzuki Yoshitaka Chubu Electric Power Co., Inc.
Swartz Jasmine National Nuclear Regulator Sydnor Christopher U.S. NRC, NMSS/DFM/MSB Számel Péter Hungarian Atomic Energy Authority Tajuelo Elena ORNL Takaya Shigeru Japan Atomic Energy Agency Taller Stephen ORNL Tapani Eurajoki Fortum Power and Heat Tappert John U.S. NRC Tcherner Julia AtkinsRéalis Teague Melissa DOE-NNSA Terachi Takumi Institute of Nuclear Safety System Terry Leslie U.S. NRC Thabethe Sibongiseni National Nuclear Regulator Thomas Damiani EPRI Tokey Jason U.S. NRC Toloczko Mychailo PNNL Tregoning Robert U.S. NRC Tregoures Nicolas IRSN Tsao John U.S. NRC Tshetlwane Moila Eskom Ullah Sharida Constellation Ulmer Christopher U.S. NRC Uytdenhouwen Inge SCK CEN Vare Christophe EDF Venter Emuel Eskom Wall James EPRI Wallace Jay U.S. NRC Wang George U.S. NRC Wasiluk Bogdan CNSC-CCSN Whaley Tyler PWROG
A-11 Last Name First Name Organization White Megan Jacobs Widrevitz Dan U.S. NRC Willson Theresa AEP - DC Cook Nuclear Plant Wise Brandon U.S. NRC Wise John U.S. NRC Wolfe Ryan EPRI Wudi Vaclav Czech Technical University Xi Jianqi University of Illinois, Urbana-Champaign Xi Zuhan U.S. NRC NRR Xi Liu unknown Xinjian Duan CANDU Energy Inc.
Xuejun Wei CNSC-CCSN Yamamoto Masato CRIEPI Yamamoto Yasunori JAPAN NUS CO., LTD.
Yee On U.S. NRC Young Garry EPRI Young Austin U.S. NRC Youngblood Robert INL Yudai Yamamoto CRIEPI Zbynek Hlavac Research Center Rez Zhai Ziqing PNNL Zhao Yunfei University of Maryland Zhu Jinying University of Nebraska-Lincoln Zimmerman Julia unknown
B-1 APPENDIX B PRESENTER AND PANELIST BIOGRAPHIES B.1 Opening Session Jim Cirilli Jim Cirilli is a Senior Technical Executive in the Nuclear Materials Division at EPRI. Mr. Cirilli has a broad background in materials and welding engineering, materials testing, nondestructive examination and failure analysis, with specific experience in evaluation of material degradation concerns, technologies for degradation mitigation and oversight of fabrication activities for new or replacement components.
Prior to joining EPRI, Mr. Cirilli was Senior Manager, Engineering Programs at Exelon responsible for the following Programs: BWR and PWR Vessel and Internals Integrity, Materials Degradation Management, BACC, ISI, Repair-Replacement, NDE and Welding. He held similar positions at PSEG Nuclear, Northeast Utilities and LILCO. From 1997 to 2004 he managed the MMR Group, Inc.'s Lehigh Testing Laboratories, an independent materials testing and failure analysis laboratory providing materials consulting and testing services to a diverse industrial, legal and insurance customer base.
Over the years Mr. Cirilli has been active in various industry materials groups and initiatives, including chairing some of those groups.
He graduated from The Pennsylvania State University with a bachelors degree in Metallurgy.
He also has a MS in Metallurgy from the Rensselaer Polytechnic Institute at the Hartford Graduate Center.
Frank Chen Dr. Xiang (Frank) Chenis a Senior R&D staff member within the Materials Science &
Technology Division at ORNL. In his role, he leads the Materials Research Pathway for the LWRS Program. Additionally, Dr. Chen acts as the principal investigator for various U.S. DOE initiatives, including those focused on Fossil Energy and Carbon Management, Light Water Reactor Sustainability, and Fusion Energy Science.
His research focuses on understanding the mechanical behavior of structural materials in extreme environments, as well as establishing connections between mechanical properties and materials microstructure. With a wealth of expertise, Dr. Chen has contributed significantly to the field, boasting over 140 publications in peer-reviewed journals, book chapters, conference proceedings, and technical reports.
Dr. Chen's contributions extend beyond publications, as he holds two patents for advanced high-strength steels designed for use in automobile applications. His academic background includes a Ph.D. degree and an M.S. degree in Nuclear Engineering from the University of Illinois at Urbana-Champaign, complemented by a B.E. degree in Nuclear Engineering from Shanghai Jiao Tong University.
Rob Tregoning
B-2 Dr. Tregoning is the Senior Technical Advisor for Materials Engineering in the Office of Nuclear Regulatory Research (RES) at the NRC. Rob is currently working on several projects within RES the Division of Engineering including research supporting the increased enrichment rulemaking effort, advanced non-light water reactor material issues, research associated with advanced manufacturing techniques, degradation of reactor vessel internals, reactor pressure vessel embrittlement, environmentally assisted fatigue, stress corrosion cracking, and probabilistic and deterministic structural integrity assessment. Rob has previously supported the resolution of issues related to GSI-191, the development of loss-of-coolant-accident frequencies currently used in most U.S. PRAs, and the structural integrity assessment of the degraded Davis-Besse reactor pressure vessel head.
Before joining the NRC, Rob worked for the U.S. Navy as a materials engineer at the Naval Surface Warfare Center (NSWC). At NSWC, Rob performed research, testing, evaluation, and analysis pertaining to structural integrity issues for ship and submarine applications.
Rob received a B.S. degree in Mechanical Engineering from the University of Maryland, and M.S. and Ph.D. degrees in Mechanical Engineering from Johns Hopkins University.
John Wise Dr. Wise is the Senior Technical Advisor for License Renewal Aging Management in the Office of Nuclear Reactor Regulation (NRR) at the NRC. John supports the resolution of technical and regulatory issues related to materials degradation and aging management of nuclear power plant components - and previously served a similar role for spent fuel dry storage systems. He also participates in international efforts to develop guidance for long term operations, including the IAEA IGALL Program. Prior to joining the NRC in 2010, he was a research assistant professor at Colorado School of Mines, research engineer at Bethlehem Steel Corporation, and failure analysis consultant at Exponent, Inc. He received his bachelors degree in metallurgical engineering and his doctorate in materials science and engineering from Michigan Tech and Northwestern Universities.
B.2 Session 1 Session 1 - Presenters M. Grace Burke Dr. Burke is currently a Laboratory Fellow at the Materials & Fuels Complex at Idaho National Laboratory and the Scientific Lead for Reactor Structural Material, based in the Characterization and Advanced PIE Division. Prior to joining INL, Grace was a Corporate Fellow at ONRL. She is presently an Emeritus Professor at the University of Manchester, where she was a Professor and the Director of the Materials Performance Centre from 20112021, and is an adjunct Professor at The Ohio State University. She also spent nearly 30 years in industry (U.S. Steel Research Laboratory, Westinghouse STC, and the Bettis Atomic Power Laboratory). Grace is a physical metallurgist who has spent her career studying environment-sensitive behavior of materials including irradiation damage of steels and Nibase alloys and SCC. She is an expert in advanced materials characterization with an emphasis on understanding the role of microstructure in materials performance and degradation.
Mikhail Sokolov
B-3 Dr. Sokolov is the RPV Embrittlement Task Leader for the DOE Light Water Reactor Sustainability Program. He joined the Oak Ridge National Laboratory in 1992. His area of expertise is the effects of radiation on mechanical properties, especially fracture toughness, of various structural materials for nuclear application with emphasis on RPV materials.
Elliot J. Long Elliot J. Long is a Senior Principal Technical Leader in the MRP at the EPRI. Mr. Long specializes in RPV Integrity and Low Alloy Steel (LAS) Research. Prior to joining EPRI in 2017, he worked in this area of expertise at Westinghouse, where his responsibilities included all major evaluations and contract offerings within the RPV Integrity product area for customers in the U.S., Europe, and Asia. Mr. Long received a George Westinghouse Signature Award for his work developing the Material-Orientation Toughness Assessment (MOTA), which was an innovative approach to address the potential nonconservatism of methods used to estimate initial fracture toughness. In 2018, he also received a Chauncey, EPRIs highest award for technical achievement, for his work on the Review of the Japanese Nuclear Operators Aging Management Plan for Prolonged Shutdown Periods. This work, publicly available in MRP435, reviewed the proposed Japanese aging management plans to safely recover plant operating life after the imposed shutdowns due to the Great Japan Earthquake of 2011. He has 15 years experience in the nuclear power industry, as an engineering consultant and researcher, all focused on RPV Integrity. Prior to Westinghouse, he worked at Powerex Inc., a semiconductor manufacturer, and Siemens Energy, in fuel cells. He graduated from The Pennsylvania State University in 2005 with a Bachelor of Science degree in Chemical Engineering and a minor in Chemistry. In 2012, he completed a Master of Science degree in Materials Science and Engineering from the University of Pittsburgh.
Session 1 - Panelists Susan Ortner Susan Ortner is a Principal Materials Scientist at the UKs National Nuclear Laboratory. She has worked in the nuclear industry for more than 30 years focusing on the interplay between microstructure and bulk properties such as yield, toughness and corrosion resistance. Her RPV experience covers mechanical testing and fracture modelling, and microstructural analyses from optical microscopy through scanning electron microscopy (SEM) and SIMS to TEM and APT.
She has used this basis to derive an understanding of RPV embrittlement mechanisms and to develop and underpin dose-damage relationships, including RPV embrittlement trend curves.
Susan has carried out projects on RPV embrittlement for customers in the UK, the U.S. and Japan and participated in European collaborative programs within the 5th, 6th and 7th Framework programs and the Horizon 2020 program. She is a major contributor to the UK Radiation Embrittlement Forum, the UKs knowledge management program on RPV embrittlement and internals degradation and is Vice-Chair of the International Group on Radiation Damage Mechanisms. She is also the UK representative on the Materials Working Group of the Jules Horowitz Materials Test Reactor and on the Metals Subgroup of the Generation IV Forum for the Very High Temperature Reactor design.
Akiyoshi Nomoto Dr. Akiyoshi Nomoto is a senior research scientist in the Materials Science Division at the Energy Transformation Research Laboratory, CRIEPI. He joined CRIEPI in 1997. His work primarily focuses on the embrittlement of nuclear structural materials and the development of
B-4 the Japanese embrittlement trend curve for reactor pressure vessel steels. With a background in metallurgy, he holds a Doctor of Engineering from Yokohama National University. He has served as the head of his division since 2024.
J. Brian Hall Brian has gained unique experience in the nuclear power industry over the last 30 years enabling him to understand the interrelationship between material mechanical behavior, fracture mechanics and aging/embrittlement. Brians experience includes material specifications for component replacement and new plants, failure analysis, materials testing, fracture mechanics, and material aging evaluations for uprates, repairs and license renewal. His contributions have been in the areas of reactor vessel (RV) integrity, RV internals aging, and other reactor coolant system materials. He has supported individual utility, the Pressurized Water Reactor Owners Group (PWROG) Materials Subcommittee, and EPRI MRP. He is chair of the ASTM E10 Committee, Behavior and Use of Nuclear Structural Materials which is responsible for standards related to RV surveillance. Through publications and leadership in industry activities he is recognized internationally. He has advanced the use of the master curve method (direct fracture toughness) for the benefit of the utilities owning PWRs, enabling operation beyond 40 years and improved operation. Brian holds a B.S. in Engineering Science (Honors) and an M.S.
in Engineering Mechanics from Pennsylvania State University.
Timothy Griesbach Tim Griesbach received his Bachelors and Masters degrees in Metallurgy and Materials Science from Case Western Reserve University. Tim is internationally known for his expertise in the areas of reactor vessel embrittlement and vessel integrity management. Early in his career he worked for Combustion Engineering during the time when reactor vessel pressurized thermal shock became a major safety concern. Tim was a project manager with EPRI from 1983 to 1993 before entering the consulting world for the past 30 years with ATI Consulting and Structural Integrity Associates. He has authored or coauthored numerous papers on reactor vessel embrittlement and vessel integrity. Tim was chairman of the ASME Section XI Working Group on Operating Plant Criteria for over 30 years, and he has been a member of the ASME Section XI, Standards Committee since 2015.
B-5 B.3 Session 2 Session 2Presenters Pal Efsing Dr. Pl Efsing is a Senior Specialist in Aging and Degradation of Metallic Materials at Vattenfall Ringhals. He began his career at Barsebck nuclear power plant in 1995, focusing on solid mechanics and SCC. After Barsebck 1 was shut down in 1999, he transitioned to Ringhals, where he worked on reactor internals and LTO. Efsing holds a Ph.D. from the Royal Institute of Technology, Stockholm, and is an International Welding Engineer. He has been an adjunct professor there since 2010.Efsing chairs the EPRI Materials Reliability Project, represents Sweden in the IAEA LifeTime Management group, and is involved in repurposing materials from decommissioned NPPs for industry use.
Maxim Gussev Dr. Maxim N. Gussev is a research scientist at ORNL with 29 years of experience in nuclear science. His expertise encompasses the effects of radiation on materials, including long-term irradiation, material properties and behavior at high damage doses, as well as radiation-induced hardening and embrittlement. He specializes in plasticity, deformation mechanisms in irradiated metals and alloys, phase instability, and twinning in austenitic steels. Dr. Gussev is also skilled in advanced mechanical testing methods, such as Digital Image Correlation (DIC) and insitu mechanical testing, including electron backscatter diffraction (EBSD) analysis.
Steven Raiman Dr. Stephen Raiman is an assistant professor of nuclear engineering and radiological sciences and materials science and engineering at the University of Michigan. He is interested in corrosion and degradation of materials in extreme environments, with a focus on nuclear systems. Before coming to The University of Michigan, he was an assistant professor of Nuclear Engineering and materials science and engineering at Texas A&M University. He also spent 4 years as a staff researcher in the Materials Science and Technology Division of Oak Ridge National Laboratory. He graduated from The University of Michigan in 2016 with a Ph.D. in Nuclear Engineering and Radiological Sciences. He also holds a B.S. in Physics from the University at Buffalo.
Session 2-Panelists Yiren Chen Dr. Yiren Chen is a principal metallurgical engineer in the Nuclear Science and Engineering Division at ANL. He received his Ph.D. from the University of Illinois at Urbana-Champaign and joined ANL in 2005. His research focuses on the degradation of structural materials in light-water reactors and sodium-cooled fast reactors due to irradiation damage and corrosive environments. He has extensive experience in conducting mechanical tests on radioactive materials under simulated reactor conditions. He has led numerous research projects on irradiation-assisted SCC, radiation damage and effects, sodium-material interactions, as well as welding and additive manufacturing technologies. In addition, he actively contributes to nuclear materials research initiatives, aiming to advance the understanding of degradation mechanisms in structural materials in nuclear power systems.
B-6 Kyle Amberge Kyle Amberge is a Technical Executive with EPRI Materials Reliability Program and has been with EPRI since 2012. Prior to 2012, Kyle was a Materials Engineer and Manager at Bettis Atomic Power Lab (Naval Nuclear) in Pittsburgh, from 19922008, and with PSEG-Nuclear in New Jersey from 20082012. Kyle has an M.S. in Materials Engineering (1992) and B.S. in Mechanical Engineering (1990) both from Rensselaer Polytechnic Institute in Troy, NY.
B.4 Session 3 Session 3-Presenters Peter L. Andresen Dr. Péter Andresen retired after 38 years at GE Global Research Center as a Principal Scientist in the Corrosion & Electrochemistry Lab. His expertise spans 54 years in the area of corrosion and environmental effects on mechanical properties and integrity of materials. His research has focused on corrosion and environmental fracture of iron-and nickel-base alloys under conditions of interest to the energy industries. He is a member of the U.S. National Academy of Engineering, and an Advisory Professor at the Shanghai Jiao Tong University in China. Dr.
Andresen is a Fellow of the American Society for Metals and the National Association of Corrosion Engineers (NACE) and serves on the Board of Editors for Corrosion Journal. He earned a B.S. in Materials Engineering (Cum Laude), 1972, Rensselaer Polytechnic Institute and Ph.D. and M.S. degrees in Materials Science, 1978 and 1974, Rensselaer Polytechnic Institute. He has over 500 publications and 27 patents.
Ziqing Zhai Dr. Ziqing Zhai is a materials scientist at Pacific Northwest National Laboratory. She received a B.S. in Mechanical Engineering and an M.S. in Reliability Engineering from Shanghai Jiao Tong University in China and a Ph.D. from Tohoku University in Japan. She also has an Engineers diploma on Reliability Engineering from Ecole des Mines de Nantes in France. Since joining PNNL in 2015, her research has focused on utilizing advanced testing methods and microscopy tools to elucidate the mechanisms of environmental degradation of metal alloys in harsh environments, especially on structural materials for LWR applications. She is currently leading the crack initiation in metal alloys research under the U.S. DOE LWRS program at PNNL.
Thierry Couvant Dr. Thierry Couvant defended his PhD thesis in 2003. He was employed at EDF as a research engineer, then as a project manager. He shares his time between experiments, modeling and the development of an SCC simulation code. His activity is focused on the improvement of the bridging between Research and Engineering, promoting the use of simulation. He is in charge of a project fully dedicated to SCC of auxiliary lines since 2023.
Session 3Panelists Bogdan Alexandreanu Dr. Bogdan Alexandreanu is a Principal Nuclear Materials Engineer at Argonne National Laboratory. Research interests center on Environmentally Assisted Cracking (EAC) in nuclear
B-7 reactor environments, particularly on environmental effects on fatigue crack initiation and growth, SCC of nickel alloys and weldments, including the effect of the welding parameters on nickel-based weld SCC susceptibility. He currently leads research programs evaluating LTO 80+ of Nibased alloys and weldments for the DOE LWRS and the NRC. Bogdan received MSc and PhD degrees in Nuclear Engineering and Radiological Sciences from The University of Michigan at Ann Arbor.
Takumi Terachi Dr. Takumi Terachi currently works in the Nuclear Power Plant Aging Research Group at the Institute of Nuclear Safety System, Inc. (INSS). He has 20 years of experience performing research related to aging management and SCC in nuclear power plants. His work has involved indepth studies of SCC behavior in stainless steel within PWR primary system environments, contributing to proactive research efforts aimed at enhancing the long-term safety and reliability of nuclear facilities. He has also worked for Kansai Electric Power Company (KEPCO) at various times throughout his career. Dr. Terachi earned his B.S. in Industrial Chemistry from the Osaka Prefectural College of Technology in 1996, and a PhD in Nuclear Power and Energy Safety Engineering from the University of Fukui in 2009.
Mychailo Toloczko Dr. Mychailo Toloczko has 30 years of experience evaluating radiation effects on structural materials, covering mechanical testing, microstructure, theory, and development of improved alloys. He has 20 years of experience investigating SCC of materials for LWR systems with a focus on generating quantitative initiation and growth rate data on a wide range base metals, welds, and interface regions.
B.5 Session 4 Session 4Presenters Yudai Yamamoto Dr. Yudai Yamamoto is a Research Scientist in CRIEPI. He finished Doctor course in Graduate School of Chemical Sciences and Engineering of Hokkaido University and got a Ph.D. in 2019.
His specialty is corrosion and water chemistry. He has been responsible for a research project on hydrazine alternatives since he joined CRIEPI in 2019. He is also in charge of a research project on metal corrosion under disposal conditions. He is a member of the Review Committee of PWR Water Chemistry Guidelines issued by the Atomic Energy Society of Japan.
Ryan Wolfe Dr. Ryan Wolfe is a Technical Executive at the EPRI. He is leading and managing research on FAC, mechanical erosion, and high-density polyethylene (HDPE) piping for balance of plant applications.
Prior to his current role, Dr. Wolfe previously managed the Steam Generator Management Program committee responsible for addressing the materials, chemistry, and thermal hydraulic issues affecting steam generators. His work included assessing corrosion issues affecting steam generator life, advancing deposit buildup maintenance strategies, and investigating advanced water chemistry technologies. Before joining EPRI, Dr. Wolfe worked at the Bettis
B-8 Atomic Power Laboratory supporting the design, construction, and operation of Naval nuclear propulsion plants.
Dr. Wolfe holds a Bachelor of Science degree in Materials Science, a Master of Science degree in Metallurgy, and a Ph.D. in Engineering Science from The Pennsylvania State University.
Session 4Panelists Jared Smith Dr. Jared Smith received a PhD in Chemistry, specializing in corrosion science, from The University of Western Ontario (Canada) in 2007. From 2007 to 2010, he worked as a Corrosion Scientist for Kinectrics Inc. specializing in metallurgical failure analyses and buried pipe inspections for the Canadian nuclear fleet. From 2010 until 2022, Dr. Smith was employed by Atomic Energy of Canada Ltd. Chalk River Laboratories, where he worked as an R&D Scientist, Corrosion Section Head, and eventually Manager of the Reactor Chemistry & Corrosion Branch.
During his time at Chalk River Labs, he served several roles for the CANDU Owners Group, including Chair of the Steam Generator Integrity Working Group and author of the COG Strategic R&D Materials Roadmaps for LTO of the CANDU fleet. Dr. Smith joined the International Materials Research Program at EPRI in 2022 as a Principal Technical Leader, where he manages research programs associated with nuclear materials aging, including SCC and IASCC of stainless steels and SG tube alloys.
Frank Gift Frank Gift is a Senior Principal Technical Leader at the Electric Power Research Institute (EPRI). Frank earned his Bachelor of Science and Master of Science Degrees in Materials Science & Engineering from Lehigh University. Frank joined EPRI in September 2021 as a member of the International Materials Research (IMR) Program, responsible for research in materials performance and aging degradation phenomena associated with light-water nuclear reactors. He leads EPRIs Nuclear Sector materials harvesting and testing projects, including research projects focused on materials aging degradation in support of reactor sustainability.
From 2003 to 2021, Frank worked for Westinghouse Electric Company in a variety of roles, where he led work in materials testing, research, specifications, procurement, fabrication, repair, and performance assessmentssupporting operating plants, new plant design, and corporate R&D. Frank served as an engineering manager at Westinghouses hot cell and research facility, as well as in the role of business manager for its global testing services.
B.6 Session 5 Session 5Presenters George Licina George Licina has a B.S.in Metallurgical Engineering from the University of Illinois. He worked in the power industry from 1972 through 2022, first for General Electric, and, from 1986 through 2022, for Structural Integrity Associates. He was the Chief Materials Consultant with Structural Integrity Associates, specializing in problems related to environmental degradation of materials including corrosion, embrittlement, and irradiation effects. He is a recognized expert in the area of microbiologically influenced corrosion (MIC), is the author of the two EPRI Sourcebooks for MIC in Nuclear Power Plants, as well as many other EPRI publications including Service Water
B-9 Piping Guidelines, Guidelines for Replacement Materials for Service Water Systems, Open Cooling Water System Chemistry Guidelines, and Guideline for Preventing Galvanic Corrosion in Aboveground Dissimilar Metals Piping Systems. He is an inventor of the BIGEORGE' electrochemical system for online monitoring of biofilm formation. He has authored more than 125 publications in open literature and four patents.
Jon R. Cavallo Jon Cavallo is a Registered Professional Engineer in three states and holds a Bachelor of Science in Engineering Technology from Northeastern University in Boston, Massachusetts. He is active on a number of national technical societies including SSPC, NACE and ASTM. Mr.
Cavallo received the ASTM Award of Merit in 2010 and is an ASTM Fellow. During his 50 years of work in the coatings and corrosion mitigation field, he has gained a reputation as a world-renowned expert. His recent work has included assignments in USA, Canada, Slovenia, India, Peoples Republic of China, Japan and South Korea.
Session 5Panelists Chris Burton Mr. Chris Burton is the Director, Nuclear Infrastructure and Underground Piping Solutions, at Structural Technologies. His duties focus on condition assessment of nuclear underground and aboveground piping, piping repairs, and finite element analysis of piping. He has 39 years of nuclear experience which encompasses 22-years as the stations High Pressure and Raw Water Piping Owner and 5 years as the Corporate High Pressure and Raw Water Piping Owner for three stations. He is the Nuclear Industry Chairperson of the EPRI Buried Piping Integrity Group (BPIG), the Task Force Chairperson for the Underground Piping & Tank Integrity Initiative (UPTIINEI 0914), and the coauthor of the UPTII. He is also a member of NEI and the License Renewal Task Force Mechanical Group Dylan Cimock Mr. Dylan Cimock is a Senior Technical Leader with EPRI. He joined EPRI in 2016 and oversees research and development related to buried piping and tanks, cathodic protection, selective leaching, and raw water piping corrosion. Prior to joining EPRI, he worked for Constellation for nearly 7 years, primarily as part of their license renewal team. Dylan has been involved with buried piping and cathodic protection since 2009.
John Wise See Section B.1 for biography for Dr. Wise.
B.7 Session 6 Session 6Presenters Tom Damiani Dr. Thomas M. (Tom) Damiani is a Principal Technical Leader within the MRP at the EPRI. Dr.
Damianis research focus within MRP is related to structural fatigue analysis and the impact of
B-10 LWR environments on both the fatigue and fracture characteristics of primary pressure boundary and related internal materials.
Prior to joining EPRI in December 2021, Dr. Damiani was the Manager of Structural Methods Development and Implementation at the Naval Nuclear Laboratory (NNL). Dr. Damiani received his B.S. degree in Engineering Physics from West Virginia Wesleyan College, and a M.S. and Ph.D. in Mechanical Engineering from West Virginia University. Dr. Damiani is a member of the ASME Boiler and Pressure Vessel Code Committees Working Group on Fatigue Strength and Environmental Fatigue Evaluation Methods within Section III.
Seiji Asada Dr. Asada joined Mitsubishi Heavy Industries in April 1989. At Mitsubishi Heavy Industries, Dr.
Asada is in charge of designing and development of Reactor Pressure Vessels and his career has focused on stress analysis, fatigue analysis, irradiation embrittlement issue of RPVs, and fracture mechanics. He obtained a PhD from Kobe University in March 2009. He is the current Chair of SDO (Standard Development Organization) Convergence Board and an active member of ASME BPV III, WG on Environmental Fatigue Evaluation Method, JSME Main Committee, SG on Fatigue Evaluation, to name a few.
Andy Morley Dr. Morley is a Chartered Engineer and Member of the IMechE with over 15 years of experience in the nuclear industry with Rolls-Royce. During his time with Rolls-Royce Andy has held roles in pressure vessel and core design, but more recently in development of structural integrity methods with a specific focus on methods to appropriately account for the impact of a pressurized-water reactor environment on the fatigue of nuclear plant materials. Dr. Morleys background is in materials science, in which he holds a first-class masters and a doctorate from the University of Oxford.
Session 6Panelists Mark Gray Mr. Gray is a Fellow Engineer at Westinghouse Electric Company near Pittsburgh, Pennsylvania. He has worked for Westinghouse since 1981 in areas related to design, analysis, and management of nuclear power plant components and piping systems. Related to these activities, he has also contributed significantly to the development of fatigue aging management solutions, led Westinghouse Owners Group programs, participated in EPRIs Fatigue Issues Task Group and Environmental Fatigue Expert Panel, and coordinated industry Nuclear Plant Fatigue Applications workshops. Mr. Gray is a member of the ASME Section III Working Group on Piping Design and Subgroup on Design Methods and currently chairs the Working Group on Environmental Fatigue Evaluation Methods. His degrees from the University of Pittsburgh include Bachelors and Masters Degrees in Mechanical Engineering and a Nuclear Engineering certificate. He is a Registered Professional Engineer.
Tim Gilman Mr. Gilman is an Associate at Structural Integrity Associates with over 25 years of experience in engineering related to structural mechanics, fatigue and fracture in the energy industry. Mr.
Gilman is the technical lead for work related to the evaluation and management of fatigue and
B-11 environmentally assisted fatigue for nuclear plant components and has been involved in a majority of the license renewal and SLR applications todate in the United States. Mr. Gilman has a Masters Degree in Civil/Structural Engineering from the University of California at Berkeley.
Jonathan Mann Dr. Mann has been working on nuclear structural integrity, with a primary focus on environmental fatigue and methods development, since 2012. Dr. Mann completed an engineering doctorate at the University of Manchester, before spending 5 years at Rolls-Royce supporting environmental fatigue research programs. He is currently working at Amentum, continuing to support UK R&T testing programs. Jonathans background is in mechanical engineering and materials science, and he has been an active member of Europes INCEFA research programs and supported several ASME code case activities.
B-12 B.8 Session 7 Darren Barborak Dr. Darren Barborak is a Technical Executive at the Electric Power Research Institute (EPRI).
He holds strategic leadership responsibilities within the Welding & Repair Technology Center (WRTC), a world class research and development group supporting the nuclear power generation industry. WRTC supports both the current and future nuclear fleet by providing technical input to code and regulatory entities to inform the development and optimization of relevant repair, fabrication, and joining requirements. The program also develops and tests advanced materials, joining, and repair technologies for operating and new nuclear plant applications, supporting the implementation of safe, effective repair/replacement/fabrication technologies. WRTC supports peer engagement and technology transfer through guidance documents, onsite assistance, training and education, and information exchange.
Dr. Barborak is a subject matter expert in arc welding processes, welding technology, and welding codes & standards with over 35 years experience. His current focus is on welding process research and development, codes and standards engagement and associated training, and repair assistance. Prior to joining EPRI, he was Director of Materials and Joining Technology for WSI Specialty Welding, a global leader in developing and delivering specialized installation and maintenance solutions through automated welding, weld cladding, and weld overlay technologies throughout the energy industry. Dr. Barborak is a member of ASME Boiler and Pressure Vessel standards committees on welding, brazing, and fusing (BPV IX), subgroup on brazing (BPV IX SG-BRAZ), working group on welding and special repair processes (BPV XI WG-W&SRP), task group on temper bead welding (BPV XI TG-TB), and task group on weld overlay (BPV XI TG-WOL). He is also a life member of the American Welding Society (AWS).
Dr. Barborak received Bachelor of Science, Masters of Science, and Doctorate degrees in Welding Engineering from The Ohio State University, where he currently serves as adjunct professor and lecturer in the Department of Materials Science and Engineering.
Jian Chen Dr. Jian Chen has been working at ORNL for more than 10 years since graduation from The Ohio State University. He is currently Senior Research Staff in the Materials Joining Group, leading and supporting a variety of fundamental and applied research and technology innovations sponsored by government agencies and industries. He is the task lead on welding mitigation of Light Water Reactor Sustainability (LWRS) Program.
Carol Moyer Mrs. Carol Moyer is a Senior Materials Engineer and Technical Reviewer in NRRs Division of New and Renewed Licenses, Vessels and Internals Branch. She joined the NRC staff in 2001 in the Office of Nuclear Regulatory Research, directing research on stress corrosion cracking, NDE capabilities, and development of consensus standards. Carol earned a bachelors degree in Materials Science & Engineering from the University of Michigan, and a masters degree in Materials Science & Engineering from Carnegie Mellon University.
B.9 Session 8 Session 8 Presenters
B-13 Madhumita Sircar Madhumita Sircar joined U.S. NRC, Office of Nuclear Reactor Regulation in 2009, transitioning to the Office of Regulatory Research in 2010. In her role, she leads various research projects that support licensing and regulatory activities. An active contributor to industry codes and standards, she serves on American Society of Mechanical Engineers and chairs the American Concrete Institute's nuclear code committee 349.
Prior to NRC, Ms. Sircar was an engineering group supervisor and technical expert at Bechtel, where she specialized in the design and construction of power plants. She holds a Master of Science degree in Civil Structural Engineering from the University of Maryland, USA.
George Wang Mr. George Wang is a Civil Engineer (Structural) at NRC/NRR, holding B.S. degree in civil engineering and M.S. degree in structural engineering with 35 years' professional experiences in structural engineering. One of his duties has been performing aging management reviews for the Nuclear Power Plants license renewal applications.
Yann Le Pape Dr. Le Pape is a Distinguished Scientist specializing in concrete aging and structural assessment, currently serving as the Nuclear Structures and Construction Group Leader at Oak Ridge National Laboratory (ORNL). Since 2020, he has been instrumental in advancing research on concrete aging, characterization, modeling, and the development of low carbon concrete solutions. His leadership at ORNL emphasizes fostering innovation and collaboration in these critical areas.
Dr. Le Pape has earned significant recognition for his contributions to the field. He received the Japan Concrete Institute 2022 Award for his groundbreaking work on the MOSAIC numerical method for assessing concrete aging. His research has also been honored with RILEM Outstanding Papers Awards in 2015 and 2020, highlighting his impactful studies on concrete carbonation and irradiation-induced damage. In 2014, he was awarded the Significant Contribution Award by the Materials Science and Technology Division of the American Nuclear Society. Additionally, he served as Chair of the International Committee on Irradiated Concrete (ICIC) from 2017 to 2022.
Samuel Johnson Mr. Samuel Johnson is a Sr. Team Leader in the Nondestructive Evaluation Group within the Nuclear Sector at the Electric Power Research Institute (EPRI). In his role, Mr. Johnson leads and manages research related to Balance of Plant and Non-Nuclear NDE, including the Concrete Research Program.
Mr. Johnson has worked with the EPRI Nuclear Nondestructive Evaluation group since 2014.
During that time, he has conducted research in the areas of visual inspections and containment monitoring for both U.S. and international nuclear power plants. He served on an expert panel for repair and mitigation strategies for reinforcement corrosion of a containment structure, and he has led EPRIs research on aging management of civil structures and infrastructure. Mr.
Johnson participates in the IAEA IGALL Working Group 3 for Civil Structures and the NEI License Renewal Task Force. Mr. Johnson has also been involved in research on the use of unmanned aerial systems for inspections and monitoring of nuclear power plants since 2016.
B-14 Mr. Johnson received a Bachelors degree in civil engineering and a Masters degree in civil engineering from Clemson University. Both degrees were with an emphasis on civil structures.
Masayoshi Kojima Dr. Masa Kojima is a chief researcher in the Nuclear Regulation Authority Japan (NRA Japan).
He joined the NRA Japan in 2014. He works primarily on research on degradation of metal and concrete materials. In addition, he is involved in the inspection and audit on the aging evaluation of the nuclear power plants. He has a Ph.D. in nuclear engineering and masters degrees in mechanical engineering.
Prior to joining the NRA Japan, he worked at the Japan Nuclear Energy Safety Organization (JNES) for 10 years, where he organized safety research on the nuclear facilities for 4 years and was a regulatory inspector on the nuclear facilities for 6 years. In addition, he worked at the Toshiba Corporation for 9 years before the JNES, where he held roles as a technical engineer on the manufacturing and repairing of the BWR primary system components.
Ippei Maruyama Dr. Ippei Maruyama is a Professor at the Department of Architecture at the University of Tokyo, and he is a material scientist and concrete engineer whose main research interests are the mechanism development of cement-based materials and their durability under general and extreme environments.
Takashi Osaki Mr. Osaki works as a manager at Chubu Electric Power Company.
He was researching spectral analysis of seismic ground motion and fragility assessment of buildings at the Department of Architecture, Graduate School of Engineering, the University of Tokyo.
Since joining the company, He conducted probabilistic seismic hazard assessments and probabilistic building fragility assessments in nuclear power. He was also in charge of the maintenance and management of the concrete buildings of the Hamaoka Nuclear Power Plant.
He is currently coordinating research on the maintenance and management of aging concrete for Japan's nuclear power companies. He is the secretary of the Nuclear Building Maintenance and Management Subcommittee of the Architectural Institute of Japan.
Fabienne Ribeiro After more than 20 years of research in the field of modeling nuclear materials (fuel and metal) at the atomic scale, under multiscale approaches, Dr. Fabienne Ribeiro led a team of modeling researchers, notably on the issue of material weathering (steel, fuel, concrete). She is now a Senior Expert on materials weathering and aging at IRSN.
Benoit MASSON Benoit Masson started his career in a nuclear power plant, then he went on to work in the field of structural monitoring, before specializing in containment building. He is currently the EDF
B-15 expert in containment building; the third containment barrier, and supervise research, modification and maintenance activities for the French nuclear fleet.
Cédric Androut After obtaining his Masters degree from University of La Rochelle (France), Cédric Androut managed the concrete laboratory of Polytechnique Montréal for 13 years, where he developed and characterized various innovative concrete mix designs along with innovative structural systems using the concrete mixes developed. Cédric later served as a Materials Specialist at WSP for two years, where he was in charge of materials Quality Assurance and Quality Control for several high-profile projects in the Canadian National Capital Region. Cédric joined the Canadian Nuclear Safety Commission in March 2021, where he is a Civil Engineering Specialist, ensuring existing and new nuclear projects in Canada meet applicable requirements from a civil engineering standpoint.
Julia Tcherner Julia Tcherner is a Specialist Civil Engineer and an SME (subject matter expert) at AtkinsRéalis with over 20 years of experience in the nuclear industry leading projects on condition assessment, aging management, and LTO of nuclear structures; performing inspections and assessments; consulting on containment capabilities and leak tightness; troubleshooting problems with new and existing construction, repair, and modification projects.
Ms. Tcherner is a member of the CSA N287/291 Executive Committee and a Chair of several subcommittees developing standards for concrete containment and safety-related structures.
She participates in IAEA expert missions including SALTO (Safety Aspects of LTO). She contributes to the development of national and international technical documents, and is the author of many papers on assessing, improving, and maintaining the performance of structures in nuclear industry.
Valdir De Souza Dr. Valdir DeSouza is a chemist with a Ph.D. in Materials and corrosion engineering. He has 25 years of experience as a scientific researcher and engineer in electrochemistry, materials, and corrosion. As a member of the Electrochemistry and Metallic Materials research team at Belgian Nuclear Research Center SCK CEN he is involved in corrosion studies of steel in alkaline environment, characterization and leaching behavior of the radionuclides in nuclear graphite and metals remelting for metal recycling purposes.
Anssi Laukkanen Dr. Anssi Laukkanen is a Research Professor on Computational Materials and Data Sciences at VTT with a D.Sc. in Material Science, Adj. Professor at Univ. of Oulu. At VTT, responsible for multiscale materials modeling and AI for materials research.
Randy J. James Mr. James received a Master of Science degree in Engineering Mechanics from the University of Texas at Austin, Texas, USA, in 1977, and spent his career as a practitioner applying mechanics principles and advanced computational methods in addressing real-world structural engineering problems. Mr. James had a 33-year career with ANATECH Corp in San Diego CA,
B-16 USA, retiring as President, before opening a consulting firm, Structural Solutions Consulting, LLC, in 2016. Over the years, Mr. James has worked on a broad range of applications, including nonlinear dynamics for high-energy impact loads, pressure fragility assessments for nuclear containments, assessing effects of creep and property degradation at elevated temperatures under load cycling, and evaluating long-term performance of civil structures affected with AAR deterioration.
Session 8 Panelists The biographies of fellow panelists Yann LePape, Samuel Johnson, Ippei Maruyama, Julia Tcherner, and Madhumita Sircar, who also served as presenters, are included in the Presenters' Section above.
Jose Pires Dr. Jose A. Pires served as Senior Technical Advisor for Civil Structural Engineering in the Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission (NRC) from June 2013 to March 2025. He joined the NRC in 2008 as a Sr. Structural Engineer in the NRCs Office of Nuclear Regulatory Research, Division of Engineering. He received the NRC Meritorious Service Award in 2017. At the NRC, he worked on a variety of civil and structural engineering issues related to design basis and severe accidents conditions for nuclear power plants. He led structural and seismic performance aspects of NRC integrated seismic, structural, geotechnical and risk studies related to design basis, severe accidents and other beyond design basis events as well as research on benchmarking of structural analysis codes using available experimental data. This research included development of technical bases for regulatory guidance for structural design of safety-related structures for nuclear power plants including seismic safety, evaluation of containment capacity and performance, and structural and seismic aspects of probabilistic risk assessment. He worked for several years on structural engineering issues related to aging of nuclear power plant structures for LTO of nuclear power plants as well as on the development of technology inclusive, risk-informed and performance-based approaches for seismic safety primarily for new and non-light reactors. He was also the civil structural engineering lead for several agencywide multidisciplinary studies dealing with severe accident consequences and probabilistic risk assessment. In addition, he worked on the analysis of full-scale regulatory transportation accidents tests for spent nuclear fuel transportation casks as part of an international cooperative research agreement program. He was also the NRC representative for several years to the Concrete and Seismic subgroups of the Working Group on Integrity and Ageing of Components and Structures (WGIAGE) of the OECD/NEA/CSNI.
Prior to joining the NRC, Dr. Pires was a distinguished member of the technical staff of Applied Research Associates (ARA) where he worked on research and other projects for the Department of Defense, the NRC and private organizations. Early in his career, he was a member of the scientific staff of Brookhaven National Laboratory (BNL) in Long Island, New York, where he worked on research and safety review projects for the NRC. Following his work at BNL, he was Assistant Professor of Civil Engineering at Washington University in St. Louis and at the University of California, Irvine where he taught and prepared undergraduate and graduate courses in soil mechanics, foundation engineering, structural dynamics, finite elements methods for structural engineering and continuum mechanics, and structural reliability. He also conducted research in topics such as structural reliability, earthquake engineering, reliability of electric power transmission systems under earthquakes, and seismic design criteria for reinforced concrete structures accounting for reliability and optimal life-cycle cost; and directed
B-17 MS and PhD students. He has published in peer-reviewed journals and conferences and co-authored numerous NUREG research reports. Dr. Pires obtained MS (1980) and PhD (1983) degrees in civil engineering from the University of Illinois at Urbana-Champaign and his undergraduate degree in civil engineering (Licenciatura) from the Universidade to Porto, Portugal.
Chris Jones Dr. Chris Jones received his doctorate in civil engineering from Texas A&M University in 2011, where he measured and modeled fundamental material properties of calcium silicate hydrate the major component of hardened Portland cement. After graduation, he accepted a technical staff position at Sandia National Laboratories within Sandias Nuclear Energy and Fuel Cycle Programs Center studying reactor containment behavior under severe accident conditions.
Jones achieved several project leadership roles while working at Sandia on Department of Energy and Nuclear Regulatory Commission funded R&D efforts. He also worked at the United States Nuclear Regulatory Commissions Office of Research for one year under an Intergovernmental Agency Personnel Agreement. In 2022, Jones was promoted to his current position as a professor in the department of civil engineering at K-State.
C-1 APPENDIX C
SUMMARY
OF PIRT SCORING RESULTS PIRT SCORING RESULTS ADAMS ML#
ML25183A354 Session 2 - PIRT scoring-RVI.
ML25183A353 Session 3 - PIRT Scoring - RCPB ML25183A352 Session 4 PIRT Scoring-SSC.
ML25183A351 Session 5 - PIRT scoring - BOP.
ML25183A350 Session 6 - PIRT scoring-Fatigue.
ML25183A349 Session 7 - Mitigation PIRT Scoring.
ML25183A343 Session 81 PIRT Scoring-RAD.
ML25183A348 Session 82 PIRT Scoring - APC ML25183A347 Session 83 PIRT Scoring - CDM ML25183A346 Session 84 PIRT Scoring - CCL ML25183A345 Session 85 PIRT Scoring - EIM ML25183A344 Session 86 PIRT Scoring - RRS ML25183A342 Session 87 PIRT Scoring - ECC ML25183A341