ML20357A006

RIL2020-12, International Workshop on Age-Related Degradation of Reactor Vessels and Internals

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Issue date:	01/31/2021
From:	Harris B, Amy Hull, Carol Moyer, Robert Tregoning Office of Nuclear Regulatory Research
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RIL 2020-12 INTERNATIONAL WORKSHOP ON AGE-RELATED DEGRADATION OF REACTOR VESSELS AND INTERNALS Date Published: January 2021 Prepared by:

A. Hull B. Harris C. Moyer R. Tregoning Research Information Letter Office of Nuclear Regulatory Research

Disclaimer This report was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third partys use, or the results of such use, of any information, apparatus, product, or process disclosed in this publication, or represents that its use by such third party complies with applicable law.

This report does not contain or imply legally binding requirements. Nor does this report establish or modify any regulatory guidance or positions of the U.S. Nuclear Regulatory Commission and is not binding on the Commission.

ABSTRACT The U.S. Nuclear Regulatory Commissions (NRCs) Office of Nuclear Regulatory Research and Office of Nuclear Reactor Regulation organized this International Workshop on Age-Related Degradation of Reactor Vessels and Internals, held May 23-24, 2019, at NRC Headquarters, 11545 Rockville Pike, Rockville, MD.

Approximately 70 attendees from 31 regulatory, research, and industry organizations representing 11 countries attended and presented information on their organizations state of knowledge, operating experience, and research activities related to age-related degradation of reactor pressure vessels and reactor vessel internals. The staff affirmed that aging management programs and research programs presented by international organizations are consistent and are evaluating similar issues as reactors enter periods of extended operation.

The objectives for this public workshop were to explore, via presentations representative of participating organizations from the United States, Japan, Czech Republic, Belgium, South Korea, United Kingdom, Hungary, France, and Switzerland, the following issues:

• state of knowledge, operating experience, and research activities related to reactor pressure vessel embrittlement at high fluence levels

• degradation of reactor vessel internals for operating periods beyond the original design up to 80 years

• degradation related to long-term operation of other safety-significant primary pressure boundary components During the workshop and in subsequent correspondence, the workshop participants addressed four questions:

(1) (a) What program or guidance forms the basis of your aging management approach?

(b) What do you believe are the most significant technical issues related to long-term operation?

(c) Does your country have any plans to update its regulatory guidance to address any of these aging management issues?

(2) Radiation-induced void swelling and creep are degradation mechanisms of potential concern during long-term operation. Summarize or reference any operating experience, research programs, or aging management programs related to void swelling or creep.

(3) (a) What is your current embrittlement trend curve (ETC) for reactor pressure vessel steels, and when was it implemented into your regulations?

(b) Does your country use ETCs for predictive purposes, or do you rely on surveillance results and then use ETCs for interpolating among surveillance results?

(4) Does your country have any plans to use additive manufacturing (or other advanced manufacturing techniques) for repair/replacement of components?

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iv FOREWORD The Atomic Energy Act of 1954, as amended, and U.S. Nuclear Regulatory Commission (NRC) regulations limit commercial power reactor licenses to an initial 40 years but also permit such licenses to be renewed. This original 40-year term for reactor licenses was based on economic and antitrust considerationsnot on limitations of nuclear technology. Due to this selected period, however, some structures and components may have been engineered on the basis of an expected 40-year service life. In 1982, the NRC established a comprehensive program for nuclear plant aging research. The results of this research indicated that most nuclear plant aging issues are manageable and do not pose technical impediments that would prevent them from operating for additional years beyond their original 40-year license period.

The NRC Office of Nuclear Regulatory Research provides specific research products to facilitate the evaluation of aging effects on passive long-lived systems, structures, and components (SSCs). The objective of this research is to generate independent technical data and confirmatory tools to enable development of regulatory guidance on the aging of SSCs and to support the regulatory review of future applications for license renewal. These products build on analysis methods, tools, and expertise developed as part of ongoing and new research activities, focused specifically on aging effects during long-term operation.

These new research activities include NRC and industry public workshops. These workshops are intended to address the state of knowledge on the technical issues identified in the staff requirements memorandum dated August 29, 2014, for SECY-14-0016, Ongoing Staff Activities to Assess Regulatory Considerations for Power Reactor Subsequent License Renewal, dated January 31, 2014, and discussed in NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, issued July 2017, and any new operating experience from the initial license renewal period.

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vi CONTENTS 1 Introduction ..........................................................................................................................1 1.1 Participating Organizations ..............................................................................................1 1.2 Workshop Focus and Agenda .........................................................................................3 2

SUMMARY

OF PRESENTATIONS ........................................................................................5 2.1 Aging Management and Subsequent License Renewal in the United States A. Hiser) .......................................................................................................................5 2.2 NRCs Aging and Materials Research Activities (R. Tregoning) ....................................5 2.3 U.S. Nuclear Electric Power Generation Industry Management of Age-Related Degradation (M. Burke) ................................................................................................6 2.4 Overview of Metals Research in LWRS Program MRP (F. Chen) .................................6 2.5 Overview of Safety Research on Metal Aging Due to Neutron Irradiation in S/NRA/R (K. Arai) .........................................................................................................7 2.6 CRIEPI Research Activities on Neutron Irradiation Embrittlement of RPV and Core Internals (T. Arai) .................................................................................................7 2.7 Czech Approach to Ageing Management of Reactor Pressure Vessel and Reactor Vessel InternalsState of Knowledge (J. Ertl) ................................................7 2.8 UJV Activities in International Research Projects in the Field of RPV and RVI (M. Zamboch) ...............................................................................................................8 2.9 Measurement of Core Shroud at NPP Temelin (M. Zamboch) ......................................8 2.10 Belgian R&D on Environmental Effects on Materials Degradation in LWRs (S. Gavrilov) .................................................................................................................8 2.11 Repair of Doel 1 NPP Reactor Vessel Head Penetrations (C. Dupuit) ..........................9 2.12 Belgian R&D Using the Enhanced Surveillance Strategy for RPV Embrittlement Assessment (M. Lambrecht) .........................................................................................9 2.13 Doel 1 & 2 Upper Plenum Injection Line Issue (M. De Smet) ........................................9 2.14 Inspection of Control Rod Guide Assemblies in Belgian NPPs (C. Dupuit) ...................9 2.15 Current Status of Aging Management on Reactor Vessels in Korea (focusing on surveillance test) (T-K Song) .................................................................................10 2.16 Ongoing Researches in Age-Related Degradation of Reactor Materials in Korea (B-S Lee)..........................................................................................................10 2.17 Operating Experience (OpE) on RV Internals, RV Head Penetrations, and RCS Small Bore Nozzles in Korea (J-S Yang) ....................................................................11 2.18 UK Regulatory Experience in Materials Ageing (G. Hopkin)........................................ 11 2.19 State of Knowledge and Research Activities on RPV Materials in UK (G. Burke) ....... 11 2.20 AM and LTO-Related Activities of RPV and Its Internals and Other Primary Pressure Boundary Components at the Paks NPP (S. Ratkai) ................................... 12 2.21 EDF Operating Experience RV Internals (R. Menand) ................................................ 12 2.22 Carbon Segregations in Heavy Forged Components (E. Viard) .................................. 12 2.23 Aging Management and LTO of NPPs in Switzerland: Status 2019 (R. Doering) ................................................................................................................13 2.24 Operating Experience of a Swiss BWR (J. Heldt) .......................................................13 3 DISCUSSION OF KEY ISSUES ...........................................................................................14 4 WORKSHOP

SUMMARY

....................................................................................................22 APPENDIX A WORKSHOP ATTENDEES.............................................................................1 vii

APPENDIX B PRESENTATION SLIDES ...............................................................................1 B.1 Aging Management and Subsequent License Renewal in the United States (A. Hiser) .....................................................................................................................1 B.2 NRCs Aging and Materials Research Activities (R. Tregoning)................................. 16 B.3 U.S. Nuclear Electric Power Generation Industry Management of Age-Related Degradation (M. Burke) .............................................................................................23 B.4 Overview of Metals Research in LWRS Program MRP (F. Chen) ............................. 41 B.5 Overview of Safety Research on Metal Aging Due to Neutron Irradiation in S/NRA/R (K. Arai) .....................................................................................................49 B.6 CRIEPI Research Activities on Neutron Irradiation Embrittlement of RPV and Core Internals (T. Arai) ..............................................................................................59 B.7 Czech Approach to Ageing Management of Reactor Pressure Vessel and Reactor Vessel InternalsState of Knowledge (J. Ertl) ............................................. 73 B.8 UJV Activities in International Research Projects in the Field of RPV and RVI (M. Zamboch)............................................................................................................88 B.9 Measurement of Core Shroud at NPP Temelin (M. Zamboch)................................... 97 B.10 Belgian R&D on Environmental Effects on Materials Degradation in LWRs (S. Gavrilov) ............................................................................................................113 B.11 Repair of Doel 1 NPP Reactor Vessel Head Penetrations (C. Dupuit) ..................... 118 B.12 Belgian R&D Using the Enhanced Surveillance Strategy for RPV Embrittlement Assessment (M. Lambrecht) ............................................................. 123 B.13 Doel 1 & 2 Upper Plenum Injection Line Issue (M. De Smet) .................................. 132 B.14 Inspection of Control Rod Guide Assemblies in Belgian NPPs (C. Dupuit) .............. 139 B.15 Current Status of Aging Management on Reactor Vessels in Korea (focusing on surveillance test) (T-K Song) ..............................................................................146 B.16 Ongoing Researches in Age-Related Degradation of Reactor Materials in Korea (B-S Lee) ......................................................................................................158 B.17 Operating Experience (OpE) on RV Internals, RV Head Penetrations, and RCS Small Bore Nozzles in Korea (J-S Yang) ........................................................ 167 B.18 UK Regulatory Experience in Materials Ageing (G. Hopkin) .................................... 181 B.19 State of Knowledge and Research Activities on RPV Materials in UK (G. Burke) ...............................................................................................................189 B.20 AM and LTO-Related Activities of RPV and Its Internals and Other Primary Pressure Boundary Components at the Paks NPP (S. Ratkai) ................................ 209 B.21 EDF Operating Experience RV Internals (R. Menand) ............................................ 238 B.22 Carbon Segregations in Heavy Forged Components (E. Viard)............................... 251 B.23 Aging Management and LTO of NPPs in Switzerland: Status 2019 (R. Doering) ............................................................................................................261 B.24 Operating Experience of a Swiss BWR (J. Heldt) .................................................... 297 viii

FIGURES Figure 1 - Allen Hiser, U.S.A., speaking about the NRCs aging and materials research activities (see presentation 2.1) ...................................................................................2 Figure 2 - Kensaku Arai, Japan, presenting a flow chart for RPV integrity evaluation (see presentation 2.5) .........................................................................................................2 ix

TABLES Table 1-1. Participating Organizations ........................................................................................1 Table 1-2. Focus Areas in Presentations Concerning Age-Related Degradation in Reactor Vessels and Internals ...................................................................................3 Table 1-3. Workshop Agenda......................................................................................................4 Table 3-1. Aging Management Approach ..................................................................................14 Table 3-2. Radiation-Induced Void Swelling and Creep ............................................................17 Table 3-3. Embrittlement Trend Curves ....................................................................................18 Table 3-4. Advanced Manufacturing Technologies....................................................................20 Table 3-5. Additional Japanese References ..............................................................................20 Table A-1 Workshop Attendees. A-1 x

ABBREVIATIONS AND ACRONYMS Term Description ADAMS Agencywide Documents Access and Management System AIC Ag-In-Cd surface treatment, in the context of ion-nitride treated RCCAs AM aging management AMP aging management program AMRC Advanced Manufacturing Research Centre, University of Sheffield (U.K.)

AMT advanced manufacturing technologies ASME American Society of Mechanical Engineers ASN Autorité de Sûreté Nucléaire (France)

ASTM ASTM International (formerly American Society for Testing and Materials)

Bel V technical support organization (Belgium)

BWR boiling-water reactor BWRVIP BWR Vessels and Internals Program CANDU Canada Deuterium Uranium (reactor)

CE Combustion Engineering CEZ eské Energetické Závody (Czech public electric utility)

CFR Code of Federal Regulations CGR crack growth rate CRDM control rod drive mechanism CRIEPI Central Research Institute of Electric Power Industry (Japan)

CT computed tension CVN Charpy V-notch DE Division of Engineering, in NRC/RES Department of the Environment, Transport, Energy, and Communications DETEC (Switzerland)

DMW dissimilar metal welds DNRL Division of New and Renewed Licenses, in NRC/NRR DOE U.S. Department of Energy EAC environmentally assisted cracking EAD engineering principles for aging and degradation EAF environmentally assisted fatigue EDF Électricité de France ENSI Swiss Federal Nuclear Safety Inspectorate ENSREG European Nuclear Safety Regulators Group EPRI Electric Power Research Institute ETC embrittlement trend curve FAC flow-accelerated corrosion GALL Generic Aging Lessons Learned, NUREG-1801 GCWM guide card wear measurement HCF high-cycle fatigue IAEA International Atomic Energy Agency IASCC irradiation-assisted stress-corrosion cracking IDOM Ingeniería y Dirección de Obras y Montaje (Spain) xi

Term Description IGALL International Generic Aging Lessons Learned IMT issues management tables (EPRI)

IRSN Institut de Radioprotection et de Sûreté Nucléaire (France)

ISI in-service inspection JAEA Japan Atomic Energy Agency JEPIC Japan Electric Power Information Center, U.S.A.

KAERI Korea Atomic Energy Research Institute KATAM catalogue of potential aging mechanisms KHNP Korea Hydro & Nuclear Power Co., Ltd KINS Korea Institute of Nuclear Safety KKL Kernkraftwerk Leibstadt Nuclear Power Plant (Switzerland)

KKM Muhleberg Nuclear Power Plant (Switzerland)

KSNP Korean Standard Nuclear Plant LBP late blooming phases LCF low-cycle fatigue LEA limited embrittlement area LRGD license renewal guidance document LTO long-term operation LWR light-water reactor LWRS Light Water Reactor Sustainability (Program)

MAPC Materials Action Plan Committee (EPRI)

MAWG Materials Assessment Working Group (EPRI)

MDM materials degradation matrix (EPRI)

MEM maintenance effectiveness monitoring Mn manganese MPC Materials Performance Centre, University of Manchester (U.K.)

MRP Materials Research Pathway (part of LWRS Program)

MSIP Mechanical Stress Improvement Process MTR BR2 material test reactor BR2 (Belgium)

MVM Paks Magyar Villamos Mvek, NPP in town of Paks, Hungary NDE nondestructive examination NEI Nuclear Energy Institute Ni nickel NPP nuclear power plant NRA Nuclear Regulatory Authority (Japan)

NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation (in NRC)

NSSC Nuclear Safety and Security Commission (South Korea)

Normative Technical Documentation of Czech Association of Mechanical NTD ASI Engineers ODSCC outer diameter stress-corrosion cracking OE operating experience ONR Office for Nuclear Regulation (U.K.)

OpEx operating experience xii

Term Description ORNL Oak Ridge National Laboratory PCVN pre-cracked Charpy V-notch PFM probabilistic fracture mechanics PTS pressurized thermal shock PWR pressurized-water reactor PWROG Pressurized-Water Reactor Owners Group PWSCC primary water stress-corrosion cracking R&D research and development RCCA rod cluster control assembly RCS reactor cooling system RES Office of Nuclear Regulatory Research (in NRC)

RG regulatory guide RI reactor internal RPV reactor pressure vessel RR repair and replacement RTNDT Reference Temperature for Nil Ductility Transition.

RV reactor vessel RVI reactor vessel internal SALTO Safety Aspects of Long Term Operation Division of Research for Reactor System Safety, Regulatory Standard S/NRA/R and Research Department, Secretariat of Nuclear Regulation Authority (Japan)

SCC stress-corrosion cracking Studiecentrum voor Kernenergie; Centre dÉtude de lénergie Nucléaire SCK CEN (Belgium)

SLR subsequent license renewal SSC systems, structures, and components SUJB State Office for Nuclear Safety (Czech Republic)

TF thermal fatigue TLAA time-limited aging analysis TMF thermomechanical fatigue TTS transition temperature shift TWCF through-wall crack frequency UJV Rez Nuclear Research Institute Rez (Czech Republic)

U.K. United Kingdom UPI upper plenum injection UT ultrasonic technology VT visual testing water-water energetic reactor, series of pressurized-water reactor VVER (WWER) designs originally developed in the Soviet Union WEC Westinghouse Electric Corporation WOL weld overlay xiii

1 INTRODUCTION This research information letter summarizes the presentations and discussions at an international workshop held on May 23-24, 2019, at the U.S. Nuclear Regulatory Commission (NRC) Headquarters auditorium in Rockville, MD (Figures 1 and 2). Participants from 31 regulatory, research, and industry organizations representing 11 countries (Table 1) took part in presentations and discussions on the state of knowledge, operating experience (OE), and research activities related to reactor pressure vessel (RPV) embrittlement at high fluence levels, degradation of reactor vessel internals (RVIs) for operating periods beyond the original design up to 80 years, and degradation related to long-term operation (LTO) of other safety-significant primary pressure boundary components.

1.1 Participating Organizations Table 1-1 lists the organizations that participated in the workshop.

Table 1-1. Participating Organizations Organization Country Link Arizona Public U.S.A. http://aps.com Service ASN France http://www.french-nuclear-safety.fr/

Bel V Belgium https://www.belv.be/index.php/en/

CEZ Czech Republic https://www.cez.cz/en/home CRIEPI Japan https://criepi.denken.or.jp/en/

EDF France https://www.edfenergy.com/

ENSI Switzerland https://www.ensi.ch/en/

Entergy U.S.A. https://www.entergy.com/

EPRI U.S.A. https://www.epri.com/

Exelon U.S.A. https://www.exeloncorp.com/

Framatome France https://www.framatome.com/EN/home-57/index.html IAEA Austria https://www.iaea.org/

IDOM Spain https://www.idom.com/

IRSN France https://www.irsn.fr/EN/Pages/Home.aspx JEPIC-USA U.S.A. https://www.jepic-usa.org/

KAERI South Korea https://www.kaeri.re.kr/eng/

KHNP South Korea http://www.khnp.co.kr/eng/main.do KINS South Korea https://www.kins.re.kr/en/

KKL Switzerland https://www.kkl.ch MVM Paks Hungary http://www.atomeromu.hu/hu/Lapok/default.aspx NRA Japan https://www.nsr.go.jp/english/

NRC U.S.A. https://www.nrc.gov/

ONR U.K. http://www.onr.org.uk/

ORNL U.S.A. https://www.ornl.gov/

Ringhals AB Sweden https://group.vattenfall.com/se/var-verksamhet/ringhals SCK CEN Belgium https://www.sckcen.be/en 1

Organization Country Link Southern https://www.southerncompany.com/our-Nuclear U.S.A.

companies/southern-nuclear.html Company SUJB Czech Republic https://www.sujb.cz/en/

Tractebel Belgium https://tractebel-engie.com/en UJV Rez Czech Republic https://www.ujv.cz/en University of https://www.materials.manchester.ac.uk/materials-Manchester U.K.

performance-centre/

MPC This document includes the presentation slides, along with brief summaries of the workshop presentations. The workshop attendees are listed in Appendix A.

The views and opinions presented in this report are those of the individual participants, and publication of this report does not constitute NRC approval or agreement with the information contained herein. As such, these proceedings are not a substitute for NRC regulations or guidance. Rather, the approaches and methods described in these proceedings and the recommendations from the discussions are provided for information only, and compliance is not required. Use of product or trade names in this report is for identification purposes only and does not constitute endorsement by the NRC.

Figure 1 - Allen Hiser, U.S.A., speaking about Figure 2 - Kensaku Arai, Japan, presenting a the NRCs aging and materials flow chart for RPV integrity research activities (see presentation evaluation (see presentation 2.5) 2.1) 2

1.2 Workshop Focus and Agenda For the readers reference, Table 1-2 indicates the represented countries, the presenters, their company or agency, and the general topic(s) on which they presented. The cross-hatch indicates presentations that were very focused and provided an in-depth discussion of the given topic. The solid grey shading indicates a more cursory discussion of that topic. Table 1-3 presents the formal workshop agenda.

Table 1-2. Focus Areas in Presentations Concerning Age-Related Degradation in Reactor Vessels and Internals Irradiation Aging Materials LTO RPV effects, Organization/ manage- aging concerns, embrittle-Country void Speaker ment research operating swelling, ment, ETC guidance activities experience curves creep U.S.A. NRC/NRR (A. Hiser)

NRC/RES U.S.A. (R. Tregoning)

U.S.A. EPRI (M. Burke)

U.S.A. ORNL (F. Chen)

Japan NRA (K. Arai)

Japan CRIEPI (T. Arai)

Czech Republic CEZ (J. Ertl)

Czech Republic UJV (M. Zamboch)

Belgium SCK CEN (S. Gavrilov)

Belgium Tractebel (C. Dupuit)

SCK CEN Belgium (M. Lambrecht)

Belgium Tractebel (M. De Smet)

Belgium Tractebel (C. Dupuit)

South Korea KINS (T-K Song)

South Korea KAERI (B-S Lee)

South Korea KHNP (J-S Yang)

U.K. ONR (G. Hopkin)

University of U.K. Manchester (G. Burke)

MVM Paks NPP Hungary (S. Ratkai)

France EDF (R. Menand)

France IRSN (E. Viard)

Switzerland ENSI (R. Doering)

Switzerland KKL (J. Heldt) 3

Table 1-3. Workshop Agenda Day 1 (May 23, 2019)

Time Presentation Topic 0800 Welcome/Introductions/Logistics 0815 Country Presentation, United States (NRC) 0900 Country Presentation, United States (EPRI) 1000 Break 1015 Country Presentation, Japan 1130 Lunch 1300 Country Presentation, Czech Republic 1415 Country Presentation, Belgium 1530 Break 1545 Panel Discussion 1645 Public Comments 1645 Adjourn Day 2 (May 24, 2019)

Time Presentation Topic 0800 Introduction 0815 Country Presentation, South Korea 0930 Country Presentation, United Kingdom 1045 Break 1100 Country Presentation, Hungary 1215 Lunch 1330 Country Presentation, France 1445 Country Presentation, Switzerland 1600 Break 1615 Panel Discussion 1700 Public Comments 1715 Adjourn 4

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SUMMARY

OF PRESENTATIONS On May 23-24, 2019, the NRC Office of Nuclear Regulatory Research (RES), Division of Engineering (DE), organized and hosted the International Workshop on Age-Related Degradation of Reactor Vessels and Internals. This workshop on the aging degradation of reactor vessels and internals included presentations (Figure 1) by senior staff at the NRC, as well as presentations by representatives from American and international industry, a U.S. Department of Energy (DOE) National Laboratory, and international nuclear regulators. All presentation materials are also publicly available in the NRCs Agencywide Documents Access and Management System (ADAMS) at Accession No. ML19150A174.

The audience included approximately 70 attendees representing 31 companies and organizations from 11 countries, including industry groups, government regulatory agencies, and both foreign and domestic research institutes (see Section 1.1). Table 1 of this report gives links to the organizations represented by the speakers.

In the weeks following the workshop, interaction continued between the NRC staff and speakers to further clarify their perspective on key issues (see Section 3). In this section, presentation summaries are based on the authors post-meeting contribution, if any, and NRC staff notes from the workshop.

May 23, 2019Morning Session United States 2.1 Aging Management and Subsequent License Renewal in the United States (A. Hiser)

Allen Hiser, Senior Technical Advisor for License Renewal Aging Management in the NRC Office of Nuclear Reactor Regulation (NRR), Division of New and Renewed Licenses, spoke on aging management and subsequent license renewal (SLR) in the United States (ADAMS Accession No. ML19143A269) (Figure 1a). He gave an overview of the U.S. experience with license renewal and SLR and stressed that in the United States, the license renewal safety principles are that plant safety is assured by the regulatory process and requires additional actions for aging management of passive, long-lived structures and components during the license renewal period. He emphasized neutron embrittlement of the RPV and high-fluence effects on RVIs as being SLR technical concerns, and he included neutron embrittlement for RPV support steel elements as a new LTO technical issue arising from the review of SLR applications. In general, his concerns for aging were related to the ability to identify potential new aging phenomena.

2.2 NRCs Aging and Materials Research Activities (R. Tregoning)

Rob Tregoning, NRC Technical Advisor for Materials Engineering (RES/DE), next presented an overview of the NRCs work in materials and aging research (ADAMS Accession No. ML19150A186). He explained that research objectives in this area typically focus on supporting regulatory decision making related to the use of new materials, manufacturing technologies, and in-service inspection (ISI) techniques; address knowledge gaps related to materials degradation during long-term plant operation to 80 years; or inform and enhance the use of risk information in regulatory decision making. His overview summarized the NRCs research activities on primary water stress-corrosion cracking (PWSCC); irradiation-assisted 5

stress-corrosion cracking (IASCC); steam generator tube integrity; aging management; spent fuel dry storage; neutron absorber materials; advanced manufacturing technologies; RPV integrity; piping integrity; probabilistic component integrity evaluation; and nondestructive examination (NDE). For each topic, he summarized the objective, motivation, and intended regulatory application; listed collaborative partners; identified recent accomplishments and deliverables; and discussed the next steps in the research. More information on each of these programs is available in NUREG-1925, Revision 4, Research Activities: FY 2018-2020, issued March 2018 (ADAMS Accession No. ML18071A139).

2.3 U.S. Nuclear Electric Power Generation Industry Management of Age-Related Degradation (M. Burke)

Michael Burke, technical executive at the Electric Power Research Institute (EPRI), presented EPRIs systematic approach to the management of reactor components aging degradation (ADAMS Accession No. ML19150A187). This approach has been employed by the U.S. nuclear power generation industry and supported by the formal processes of EPRIs Boiling-Water Reactor (BWR) Vessels and Internals Program (BWRVIP) and Materials Reliability Program since 2004. The presentation described how, building on the recognition of emerging aging issues from around 2000, the industry undertook a systematic approach to manage the mechanisms of materials aging degradation. The development of this systematic approach, based on the integrated issues management initiative outlined in Nuclear Energy Institute (NEI) 03-08, 1 led to the systematic categorization of aging degradation issues as a framework 0F for aging management of the fleet. Dr. Burke described how the BWRVIP and Materials Reliability Program currently categorize aging management issues as assessment, inspection and evaluation, mitigation, repair and replacement, and regulatory-driven to identify, analyze, and assess the importance of, provide resources to, and resolve the key aging management issues on a timely basis. The process initially focused on U.S. industry issues under the aegis of NEI 03-08 but is now used for international members as well. The approach of collecting OE and identifying the most urgent issues in each category as the basis for systematic plant aging management via EPRIs Materials Degradation Matrix was described. The use of the issues management tables (IMTs) and their Knowledge Gap tables was described and discussed.

Dr. Burke presented examples of how the IMT process has been employed to address and close knowledge gaps and to support successful management of age-related degradation of reactor vessels and internals.

2.4 Overview of Metals Research in LWRS Program MRP (F. Chen)

Frank Chen, senior scientist at Oak Ridge National Laboratory (ORNL), presented an overview of research on metals degradation in the Light Water Reactor Sustainability (LWRS) Program (http://lwrs.inl.gov) Materials Research Pathway (MRP) (ADAMS Accession No. ML19150A188).

ORNLs research involves collaborations with partners including EPRI, Westinghouse, the Pressurized-Water Reactor Owners Group (PWROG), Central Research Institute of Electric Power Industry (CRIEPI), Rolls Royce, Exelon, and the NRC. In describing the key benefit of the MRP research, Dr. Chen stated, Understanding which components are susceptible to certain forms of degradation, and their predictive behavior, will permit more focused component inspections, component replacements, and more detailed regulatory guidelines. The LWRS MRP was illustrated by a Venn diagram including experimental testing, harvested materials, and modeling. Dr. Chen described elements of the MRP metal-related project portfolio in detail, 1

NEI 03-08, Revision 3, Guideline for the Management of Materials Issues, February 2017, ADAMS Accession No. ML19079A256.

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including RPV studies, mitigation technologies, harvesting and integrated research, and studies on core internals and piping. For example, ORNLs RPV studies have developed a testing technology for determining master curve fracture toughness for RPV steels using mini compact tension specimens machined from Charpy impact specimens.

Japan 2.5 Overview of Safety Research on Metal Aging Due to Neutron Irradiation in S/NRA/R (K. Arai)

Kensaku Arai, in the Division of Research for Reactor System Safety, Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R), presented an overview of safety research on metal aging due to neutron irradiation (Figure 1b) (ADAMS Accession No. ML19150A183). After summarizing previous research programs on this topic, he discussed current work on RPV embrittlement and probabilistic fracture mechanics (PFM). A PFM guidebook for RPVs has been developed that supports the calculation of probabilistic numerical indices such as through-wall crack frequency by PFM analyses. This presentation introduced the status of the safety research programs for neutron irradiation embrittlement and PFM that are currently being conducted by S/NRA/R.

2.6 CRIEPI Research Activities on Neutron Irradiation Embrittlement of RPV and Core Internals (T. Arai)

Taku Arai, CRIEPI, continued with another discussion of research activities on neutron irradiation embrittlement of RPV and core internals (ADAMS Accession No. ML19150A184).

CRIEPIs research has focused on understanding the mechanism of embrittlement through correlating changes in microstructure and mechanical properties. CRIEPI proposes that Japan adopt a new embrittlement trend curve (ETC) based on high-fluence surveillance data and extensive atom probe tomography data. An ongoing confirmatory study on embrittlement of irradiated steels uses material harvested from the decommissioned Zion Unit 1. Another part of the research is investigating the combined effect of initial toughness distribution and fluence attenuation in RPV steel to evaluate conservatism in integrity assessment.

May 23, 2019Afternoon Session Czech Republic 2.7 Czech Approach to Ageing Management of Reactor Pressure Vessel and Reactor Vessel InternalsState of Knowledge (J. Ertl)

Jakub Ertl, CEZ, gave an overview of the Czech approach to the aging management of RPVs and RVIs (ADAMS Accession No. ML19150A180). He explained how the Czech Republic has two different types of aging management programs (AMPs)component-based AMPs and specific AMPs (focused on the degradation mechanisms). Specific AMPs were created according to generic attributes prescribed by International Atomic Energy Agency (IAEA) recommendations (International Generic Aging Lessons Learned (IGALL 2)). He presented a 1F 2 International Atomic Energy Agency, Ageing Management for Nuclear Power Plants: International Generic Ageing Lessons Learned (IGALL), IAEA Safety Reports Series No. 82, Vienna, Austria, 2015.

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new specific AMP, now in preparation, for radiation damage of internals. Another interesting document in preparation concerns RVI lifetime assessment.

2.8 UJV Activities in International Research Projects in the Field of RPV and RVI (M. Zamboch)

Miroslav Zamboch, UJV Rez, gave an overview of UJV activities in international RPV and reactor pressure internals research programs (ADAMS Accession No. ML19150A181). One such European research project is SOTERIA. The projects name is an abbreviation derived from Safe lOng-TERm operation of light water reactors based on Improved understanding of radiation effects in nuclear structural mAterials. Other collaborative activities discussed included various European and international materials research programs. Each of these multinational research programs is focused on a different aspect of the aging, degradation, and integrity of RPV and RVI materials, such as environmental fatigue, radiation embrittlement, and probabilistic assessment of RPV integrity. The Czech Republic is performing a deterministic assessment of RPV embrittlement based on the results of the surveillance program Charpy impact tests and normative trend curves. The probabilistic approach to the assessment of RPV embrittlement using a master curve approach is performed as a voluntary, supplementary assessment.

2.9 Measurement of Core Shroud at NPP Temelin (M. Zamboch)

Miroslav Zamboch, UJV Rez, next discussed measurement of the core shroud at NPP Temelin, a VVER 1000 type, in operation since 2001 (ADAMS Accession No. ML19150A182). He emphasized that RPV internals of VVER 1000 are, because of operating conditions, sensitive to the development of irradiation-induced creep and swelling (resulting from high fluences, high doses, and high temperature due to gamma heating). At this moment, the evaluation of this degradation mechanism in RVIs is only computational in the Czech Republic. The development of the swelling may result in macroscopic change of the core shroud. For LTO (60+ years), it is useful to confirm and monitor the development of the swelling by measurement at the NPP directly. The measuring device able to detect specific changes of core shroud geometry was designed and developed at UJV Rez, and the methodology for measurement was prepared and approved by the NPP. The expected degradation mechanisms include radiation swelling and radiation creep, as well as fatigue, IASCC, wear, loss of fracture toughness, and mechanical damage. The presentation extensively discussed radiation swelling and radiation creep, including a basic explanation of the degradation mechanisms and methods to calculate their effects. UJVs work on measuring dimensional changes in the core shroud, beginning in 2021, is intended to provide real phenomenological information about any actual radiation swelling.

Belgium 2.10 Belgian R&D on Environmental Effects on Materials Degradation in LWRs (S. Gavrilov)

Serguei Gavrilov, SCK CEN, discussed Belgian research and development (R&D) on the effects of the environment on materials degradation in LWRs (ADAMS Accession No. ML19150A175).

The presenter summarized results from IASCC studies and addressed hydrogen effects on RPVs in depth. Work continues on corrosion fatigue to address effects of surface conditions, hold time, and mean stress/strain.

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2.11 Repair of Doel 1 NPP Reactor Vessel Head Penetrations (C. Dupuit)

Charles Dupuit, Tractebel, gave a detailed presentation of work done to repair the Belgian Doel 1 NPP reactor vessel head penetrations (ADAMS Accession No. ML19150A176). The Doel 1 plant is a two-loop pressurized-water reactor (PWR) connected to the grid in 1974. He emphasized that before Doel LTO, PWSCC indications on reactor vessel head penetrations were detected and dispositioned through a Justification of Continuous Operation. After the LTO decision, plans for a vessel head replacement were deemed not feasible, and a repair of vessel head penetrations was therefore mandatory. This presentation covered the repair scope, repair process (Inside Diameter Temper Bead), repair qualification and issues encountered, as well as the ISI program for later outages.

2.12 Belgian R&D Using the Enhanced Surveillance Strategy for RPV Embrittlement Assessment (M. Lambrecht)

Marlies Lambrecht, SCK CEN, presented Belgian research to assess RPV embrittlement via an enhanced surveillance strategy (ADAMS Accession No. ML19150A177). The enhanced surveillance supplements the mandatory conventional surveillance, based on Charpy impact tests, and includes tensile testing with sub-sized specimens, facilitating radiation damage modeling. Researchers at SCK CEN are continuously developing and improving the ETC model with the support of data from the high-performance material test reactor BR2 and reliable databases. This work should provide increased reliability in the surveillance database in the context of LTO.

2.13 Doel 1 & 2 Upper Plenum Injection Line Issue (M. De Smet)

Michel De Smet, Tractebel, presented OE associated with the Doel 1 and 2 upper plenum injection (UPI) lines, part of the safety injection system and typical for a Westinghouse two-loop PWR (ADAMS Accession No. ML19150A178). He noted that in April 2018, a leak occurred in one of the two UPI lines of Doel 1 NPP. NDE of the UPI lines revealed degradation in the longer UPI-A lines of Doel 1 and 2 NPP: cracking in the bottom part of the straight pipe, upstream of the weld between straight pipe and elbow, and circumferential cracking in the weld between straight pipe and elbow. No cracking was found in the shorter UPI-B lines. Destructive examinations performed in two independent laboratories confirmed that the degradation was due to thermal fatigue. The affected straight parts of the UPI-A lines were replaced. A safety demonstration and justification for safe restart of both units, relying on repair, comprehensive monitoring, and future inspections, were submitted and were approved by the Belgian safety authorities. The installed monitoring confirms the presence of thermal cycles in the UPI lines at full-power conditions. Structural integrity evaluations (stress, fatigue, and fatigue crack growth) are ongoing.

2.14 Inspection of Control Rod Guide Assemblies in Belgian NPPs (C. Dupuit)

Charles Dupuit, Tractebel, described his experience inspecting control rod guide assemblies in Belgian PWRs, including Tihange 3 and Doel 4 (ADAMS Accession No. ML19150A179). He noted that following inspections in accordance with international guidelines and recommendations, control rod guide card wear was observed in most Belgian units. Excessive wear was detected at Tihange 3. As this unit shares many similarities with Doel 4, where no excessive wear was present, the wear mechanism was studied. Tractebel hypothesized that the difference in observed wear is possibly related to the different length of ion nitride (AIC) coating 9

present in the rod cluster control assemblies (RCCAs) at Tihange 3 versus those at Doel 4.

Mr. Dupuit also presented an overview of similar inspections of other units.

May 24, 2019Morning Session South Korea 2.15 Current Status of Aging Management on Reactor Vessels in Korea (focusing on surveillance test) (T-K Song)

Tae-Kwang Song, KINS, discussed the current status of aging management of reactor vessels in South Korea, focusing on surveillance test requirements (ADAMS Accession No. ML19150A193). He noted that a surveillance test has been performed to monitor changes in the fracture toughness properties of reactor vessel materials. Two issues in surveillance testing were introduced: (1) the low value of lead factor and (2) the 60-year design life. The surveillance capsule of the KSNP (Korean Standard Nuclear Plant) type reactor is located on the inside wall of the reactor vessel, and thus, the lead factor of a KSNP is less than that of a Westinghouse-type reactor. To overcome the problems caused by low lead factors at KSNP, the capsule withdrawal schedule was adjusted to obtain irradiation data at the end of licensed operation (i.e., 40 years). New surveillance capsules were fabricated using archive materials of KSNP reactors, then inserted into Westinghouse-type reactors, making use of their higher lead factor to obtain irradiation data at the extended life (60 years). The current surveillance test requirement is based on a 40-year design life as noted in Korean Nuclear Safety and Security Commission (NSSC) Notice 2017-20 and ASTM E185-82. Thus, it is difficult to apply the current requirement directly to the recently developed power plant with a 60-year design life. A draft bill of surveillance requirement to resolve this problem has been developed so that it can be applied to all reactors regardless of design life. This bill is in the process of rulemaking in South Korea.

2.16 Ongoing Researches in Age-Related Degradation of Reactor Materials in Korea (B-S Lee)

Bong-Sang Lee, KAERI, discussed ongoing research in age-related degradation of reactor materials in South Korea (ADAMS Accession No. ML19150A194). He noted that irradiation embrittlement of a high copper weld at higher neutron dose was characterized with surveillance test data from both Charpy and pre-cracked Charpy V-notch (PCVN) specimens. The transition shifts from PCVN fracture toughness data were almost the same as Charpy V-notch impact data for the high copper weld. High-flux experimental data obtained from research reactors were comparable to those from surveillance tests of Korean RPV steels. ETC models might have underpredicted the measured transition temperature shift (TTS) values for Korean RPV steels and welds, especially at higher fluence regions. Baffle-former bolts in the Kori-1 PWR have shown some defect indications from the final ISI signals. Those bolts will be extracted and investigated to identify a possible IASCC mechanism. Furthermore, the retired Kori-1 reactor components are going to be used for aging degradation studies under nuclear safety R&D projects. Advanced NDE technology and corrosion-related research are also important topics, with projects underway to improve the knowledge of NPP aging management.

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2.17 Operating Experience (OpE) on RV Internals, RV Head Penetrations, and RCS Small Bore Nozzles in Korea (J-S Yang)

Jun-Seog Yang, KHNP, presented OE of reactor head, reactor internals, and small-bore nozzles attached at the reactor coolant pipe (ADAMS Accession No. ML19150A195). He noted that in 2012, at Hanbit Unit 3, a Combustion Engineering (CE) two-loop plant, ultrasonic technology (UT) detected flaws in six control rod drive mechanism (CRDM) penetrations. All the reported flaws were oriented axially and located in the CRDM nozzles near the location of the J-groove weld toe. Reactor vessel (RV) heads of Hanbit Units 3 and 4 were replaced with Alloy 690/52/152 heads. Welds of RV head penetrations of Hanwul Units 3 and 4 and Hanbit Units 5 and 6 are scheduled to be overlay-welded with 52/152 weld metal. In 2015, Kori Unit 1, a Westinghouse Electric Corporation (WEC) two-loop plant, reported indications of cracking in baffle-former bolts for the first time in South Korea. In 2016, at Hanwul Unit 3, a CE two-loop plant, a leak was discovered on the sampling nozzle of reactor cooling system (RCS) hot-leg piping. The half-nozzle repair technique will be used on small-bore nozzles attached to RCS hot-leg piping and pressurizer heater sleeves. The Korean regulatory body required the utility to identify the root cause of bolt damage and to establish plans for the inspection and evaluation of RVIs to manage aging effects.

United Kingdom 2.18 UK Regulatory Experience in Materials Ageing (G. Hopkin)

Gareth Hopkin, Office for Nuclear Regulation (ONR), gave an overview of the United Kingdom (U.K.) regulatory experience in materials aging (ADAMS Accession No. ML19150A198).

Starting with a history of nuclear power in the United Kingdom, he progressed to ONR safety assessment principles, which include five different engineering principles for aging and degradation (EAD). For example, EAD.03 relates directly to materials aging, including irradiation embrittlement. He emphasized that the U.K. regulatory approach is goal setting and not prescriptive. The management of materials aging is to be addressed as part of the generic design assessment process. The ONR expectation for irradiation embrittlement of major vessels is that code compliance is not necessarily sufficient. For example, the United Kingdoms one civil LWR, Sizewell B, has a surveillance program containing Charpy and fracture toughness specimens, including pre-strained samples. For new builds, the expectations vary between technologies and reactor designs, depending on the nuclear safety significance and the importance of the degradation mechanism to overall integrity. New build reactors in the United Kingdom must demonstrate that they have suitable and sufficient surveillance programs. The requesting party must demonstrate an adequate understanding of the mechanism of irradiation embrittlement and how the specifics of the reactor design interact with this mechanism.

2.19 State of Knowledge and Research Activities on RPV Materials in UK (G. Burke)

Grace Burke, Director of the Materials Performance Centre (MPC) at the University of Manchester, gave an in-depth tutorial on the evolution of current understanding of irradiation embrittlement of RPV materials (ADAMS Accession No. ML19150A199). She included studies on the properties and physical changes in neutron-irradiated steels and welds and emphasized model development (based on empirical and fundamental understanding) for predicting behavior. She also discussed research activities in the United Kingdom, focusing on irradiation-induced degradation of structural materials. She stated that there is no need to invoke late-blooming phases to explain irradiation behavior in RPV steels and welds; rather, 11

irradiation embrittlement can be explained by the evolution of solute-enriched clusters. She mentioned domestic and international collaborations that contribute to ongoing and proposed research. The U.K. National Nuclear Users Facilities, for example, provide government-supported nuclear R&D. In addition to gathering data from real (harvested) components, she proposed to pursue continued collaboration with the International Group on Radiation Damage Mechanisms, originally started with U.S. NRC guidance.

Hungary 2.20 AM and LTO-Related Activities of RPV and Its Internals and Other Primary Pressure Boundary Components at the Paks NPP (S. Ratkai)

Sandor Ratkai, of the MVM Paks NPP, discussed aging management and LTO-related activities for RPVs, internals, and other primary pressure boundary components (ADAMS Accession No. ML19150A191). AMPs for Hungarys VVER NPP are based on the NRC license renewal guidance documents, are updated periodically in accordance with IGALL, and take into account R&D results and U.S. and European OE. About 150 very detailed AMPs were developed for passive safety-related components, while active components are managed by a maintenance effectiveness monitoring process. Time-limited aging analyses (TLAAs) are also used in AMPs, and Mr. Ratkai described several examples. A new TLAA is being developed for void swelling and IASCC in RPV internals.

May 24, 2019Afternoon Session France 2.21 EDF Operating Experience RV Internals (R. Menand)

Roch Menand, Électricité de France (EDF), recalled EDF OE for RVIs, including wear of thermal sleeves (ADAMS Accession No. ML19150A189). Maintenance operations on components including guide tubes, split pins, CRDMs, and baffle bolts were used to illustrate the context for the EDF lifetime management policy and readiness for plant life extension. Next, he presented ongoing research on RVIs, such as flux thimble tubes, thermal sleeves, and baffle bolting.

Research on irradiation of RVIs includes an investigation to determine PWR conditions that can lead to void swelling.

2.22 Carbon Segregations in Heavy Forged Components (E. Viard)

Emmanuel Viard, of the Institut de Radioprotection et de Sûreté Nucléaire (IRSN), gave an overview of OE with carbon segregation in heavy forged components, as an assessment of nonconformance in some French PWR primary coolant systems (ADAMS Accession No. ML19150A190). Carbon contents exceeding maximum allowable values have been identified as the leading cause for low toughness values in low-alloy steel forgings. In response to the nonconformances, the French safety authority demanded a determination of the toughness of the segregated areas based on characterization of sacrificial components. For the future, Mr. Viard recommends that material quality demonstrations include drop weight tests, in addition to Charpy V-notch and compact tension (CT) fracture toughness tests. Additional conservatism on material properties is recommended to ensure that specimens are sufficiently representative, and that variability is addressed.

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Switzerland 2.23 Aging Management and LTO of NPPs in Switzerland: Status 2019 (R. Doering)

Ralph Doering of ENSI shared detailed OE related to RPV inclusions (Beznau NPP), core shroud cracking (Muhleberg NPP (KKM)), fatigue, and embrittlement at various Swiss NPPs (ADAMS Accession No. ML19150A196). Results of RPV fatigue monitoring in the form of current fatigue usage factors and the corresponding levels extrapolated to 60 years of operation indicate that the LTO of Swiss NPPs is not subject to any limitations as a result of RPV material fatigue. Similarly, results of surveillance programs in the form of fluence calculations and the corresponding levels of irradiation embrittlement, extrapolated to 60 years of operation, show that the LTO of Swiss NPPs is not subject to any limitations as a result of RPV irradiation embrittlement. Switzerland established a systematic aging management methodology in 1991.

The degradation mechanisms of concern include fatigue (low-cycle fatigue, high-cycle fatigue, thermal fatigue, thermomechanical fatigue, environmentally assisted fatigue (EAF)), irradiation embrittlement, stress-corrosion cracking (SCC), flow-accelerated corrosion (FAC), and others (corrosion, wear, erosion, thermal aging). Based on results of the extensive inspection programs, specific corrective actions have been taken for identified aging-related damage and degradation. With application of the AMP for RPVs, safe LTO for an operating period of 60 years (or, for KKM, until the final shutdown in December 2019) is ensured.

2.24 Operating Experience of a Swiss BWR (J. Heldt)

Jens Heldt, Kernkraftwerk Leibstadt NPP (KKL), shared detailed OE related to the Leibstadt NPP, especially associated with environmentally assisted cracking (EAC) observed in reactor water piping and SCC of dissimilar metal welds (ADAMS Accession No. ML19150A197). After weld overlay repairs of these cracks, fracture toughness tests and analysis of SCC susceptibility were conducted. A pilot application for a mechanical stress improvement process was prepared for the Swiss regulator. Mr. Heldt concluded by stating that the following are important for the assessment of SCC and EAC: understanding of the mechanisms and phenomena, disposition lines for crack growth, operational experience, fabrication history, and NDE capability.

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3 DISCUSSION OF KEY ISSUES During the workshop and in subsequent correspondence, the speakers were asked the following questions, with a request that they provide perspectives from their country:

(1) (a) What program or guidance forms the basis of your aging management approach (e.g., GALL, IGALL, other references)?

(b) What do you believe are the most significant technical issues related to long-term operation (LTO)? (Provide a list below and, if appropriate, applicable references.)

(c) Does your country have any plans to update its regulatory guidance to address any of these aging management issues?

(2) Radiation-induced void swelling and creep are degradation mechanisms of potential concern during LTO. Could you briefly summarize or reference any operating experience, research programs, or aging management programs related to void swelling or creep?

(3) (a) What is your current embrittlement trend curve (ETC) for reactor pressure vessel steels, and when was it implemented into your regulations?

(b) Does your country use ETCs for predictive purposes, or do you rely on surveillance results and then use ETCs for interpolating between surveillance results?

(4) Does your country have any plans to use additive manufacturing (or other advanced manufacturing techniques) for repair/replacement of components? (Provide brief summary below and if appropriate applicable references.)

The NRC staff compiled the answers that were received into Tables 4-7 below.

Table 3-1 captures verbatim the answers, received by e-mail from each participating country, for Questions 1 (a, b, and c) regarding aging management .

Table 3-1. Aging Management Approach Country Responses to Questions 1 (a, b, and c)

(a) Aging Management (AM) is based upon methodology elaborated by IAEA. The Czech general approach of IAEA to AM represented by Safety Standard, Safety Republic Fundamentals and Safety Requirements and consequently safety Guides was transformed into CEZ control documentation (documentation concerned with ageing management). The generic approach to AM is defined in Specific Safety Requirements No. SSR 2/2 Safety of Nuclear Power Plants: Commissioning and Operation. The guidance of Specific Safety Guide No. SSG-48: Ageing Management and Development of Programme for Long Term operation of Nuclear Power Plant (and its predecessors) was used for implementation of EZ AMPs.

The compliance of AMPs with general world practice is assured by participation in international project as IGALL (International Generic Aging Lessons LearnedIAEA), benchmarking performed by European Nuclear Safety Regulators Group ENSREG, regular information exchange between operators of VVER NPPs and experts from supporting technical organisations (Czech Republic, Slovakia, Hungary) during regularly performed workshops. The 14

Country Responses to Questions 1 (a, b, and c) information from ongoing international research projects with possible connection to aging of reactor pressure vessel are evaluated on an annual basis. The AM process was reviewed several times during SALTO (Safety Aspects of Long-Term Operation) missions by IAEA experts.

Evaluation of reactor pressure vessel integrity (with respect to specific degradation mechanisms is performed according to the NTD ASI (Normative Technical Documentation of the Czech Association of Mechanical Engineers),

version 2017. The first edition of NTD ASI was issued in1996. It was based mainly upon Russian standard PNAE-G-7-002-86 and it represented first step of harmonisation with ASME approach.

The next important modernisation of NTD ASI took place in 2008 edition where the VERLIFE 2003Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation was developed within the 5th Framework Program of the European Union and incorporated into Czech technical standards. By that edition the new development in fracture mechanics and approaches of PWR codes were included together with IAEA PTS guidelines. In the 2013 the NTD ASI was upgraded according to outputs of 6th Framework Program COVERSSafety of WWER NPPs of the European Union in 2008updated version of VERLIFE 2003.

At this moment the further NTD ASI development is ongoing based on results from IAEA NULIFEPlant Life Management of NPPs project, where the VERLIFE procedure was further modernised and extended with collaboration of experts of PWR operating countries as well as of VVER operating countries including Russian experts. The next edition of NTD ASI is prepared for issue in 2019.

(b) For LTO it is necessary to know response of the material to the long term operation loads (fatigue, environment, irradiation). To obtain proper data for evaluation and prediction of the passive structure and components state during whole operation it is necessary to perform experimental program using material aged during real operation of the NPP. The most significant technical issue is change of the RVI material properties during long term operation (IASCC, RI swelling and creep, LEA formation).

The other issue could be long term temperature aging for specific kind of steels.

Degradation of concrete.

(c) In Czech Republic (with collaboration of the other countries) the project aiming for harvesting material (steel) from decommissioned unit Bohunice (Slovak Republic) has been started. Based upon material acquired the experimental program to provide information about in-situ aged components material shall be performed based on supporting funding.

(a) Nuclear Regulation Authority (NRA) guidelines [see Table 8, below, for Japan document references]:

• The Guide for Extension of Operational Period on Commercial Power Reactors [1]

• The Standard Review Plan for Extension of Operational Period on Commercial Power Reactors [2]

• The Guideline on Implementing Measures for Aging Management at Commercial Nuclear Reactors [3]

• The Standard Review Guideline for Aging Management at Commercial Nuclear Reactors [4]

• Utility guidelines

• Code on Implementation and Review of Nuclear Power Plant Ageing Management Program 2015 [5]

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Country Responses to Questions 1 (a, b, and c)

(b) NRA considers neutron irradiation embrittlement of reactor pressure vessels (RPVs), and aging of electrical penetration of primary containment vessels as long-term operation related research issues [6].

(c) NRA does not currently have plans to update the regulatory guidelines listed in 1(a).

(a) (The Korean NSSC Notice #2017-29 describes the requirements for scoping, South screening, and assessment of aging management of Korean nuclear power Korea plants, which refers to NUREG-1801 (GALL)).

National R&D programs are also underway to improve the capabilities of aging management in addition to gathering the operating experiences.

(b) 1) Cracking phenomena: PWSCC, IASCC, ODSCC

2) Environmental assisted fatigue analysis Design criteria at the time of construction may be different from the present design criteria. In that case, how to apply the current license basis to the existing plants for the long-term operation is important. Environmental assisted fatigue, for example, was not considered in the design status because it was not required at that time, however, it might be considered if it is re-evaluated for long-term operation

3) Integrity assessment of the embrittled componentsRPV & RVI Effects of large seismic loadings on the embrittled RVI structure has been focusing on the integrity assurance of fuel elements.

(c) South Korea does not have a specific plan to update the regulatory guidance of aging management. However, several research projects are underway to improve technology of the aging management including the above issues.

(a) The basis is the GALL document Sweden (b) 1) Fracture toughness of the RPV beltline weld. Fracture toughness properties of the Stainless steels in the internals. Wear of internals, such as Lower Radial Support and Flux Thimble Tubes.

(c) The Swedish regulator SSM does not have any detailed regulatory guidance. Its up to the licensee to address these issues and justify adequate aging management and report it to the regulator.

(a) Basis for AMP are the national laws (nuclear energy act, nuclear energy Switzerland ordinance and DETEC ordinance Preliminary shut down of NPP) and the IAEA NS-G-2.12/SSG-48.

Other references are:

• IAEA Technical Reports Series No. 448, Plant Life Management for Long Term Operation of Light Water Reactors, 2006.

• IAEA-EBP-SALTO, Safety Aspects of Long Term Operation of Water Moderated Reactors, 2007.

• IAEA Safety Reports Series No. 15, Implementation and Review of a Nuclear Power Plant Ageing Management Programme, 1999.

• IAEA TecDocs No. 1361, 1470, 1471, 1556, 1557.

EUR 22763, Development of a European Procedure for Assessment of High Cycle Thermal Fatigue in Light Water Reactors: Final Report of the NESC-Thermal Fatigue Project, 2007.

Results by other programmes like iGALL and Operating Experience are taken into account.

(b) Most relevant, we expect, are embrittlement for the older reactors especially the PWR, fatigue including vibrational fatigue, SCC and FAC.

Challenges are implementation of updated regulatory requirements (like increased hazards or safety margins).

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Country Responses to Questions 1 (a, b, and c)

(c) All currently relevant issues are addressed. AMP will [be] periodically reviewed according to the state-of-the-art and OpEx.

The AMP-Guideline ENSI-B01 will be checked for the need of updates according to the new IAEA SSG-48 (2018) and the results of the ENSREG Topical Peer Review 2017.

(a) The guidance for aging management for license renewal (operation from 40 to U.S.A.

60 years) is provided in (as supplemented by license renewal-interim staff guidance documents (LR-ISG)):

• Generic Aging Lessons Learned (GALL) Report (NUREG-1801, Rev. 2),

December 2010

• Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR, NUREG-1800, Rev. 2), December 2010

• The guidance for aging management for subsequent license renewal (operation from 60 to 80 years) is provided in (and will be supplemented by SLR-ISG documents)

• Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report (NUREG-2191), July 2017

• Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR) (NUREG-2192), July 2017 (b) The most significant issues for SLR are:

• Reactor pressure vessel neutron embrittlement at high fluence

• Irradiation-assisted stress corrosion cracking of reactor internals and primary system components

• Concrete and containment degradation

• Electrical cable qualification, condition monitoring and assessment Material degradation issues are also periodically identified through operating experience. Some recent examples include selective leaching, control rod drive mechanism thermal sleeve degradation, and baffle former bolt failures.

(c) The NRC will update the SLR guidance for these topics and as needed for any other new information, using the SLR-ISG document process. Several SLR-ISGs are in process as of autumn 2020, related to mechanical, electrical and structures portions of the SLR guidance, and PWR RVI components (see https://www.nrc.gov/reading-rm/doc-collections/isg/license-renewal.html for more information on SLR-ISGs and LR-ISGs (the latter of which update the GALL report and the SRP-LR)).

Table 3-2, below, captures the responses from each country to Question 2, regarding radiation-induced void swelling and creep, degradation mechanisms of potential concern during LTO.

Table 3-2. Radiation-Induced Void Swelling and Creep Country Responses to Question 2 Regular evaluation of the neutron irradiation load (fluxes, fluences) of the RVI material including assessment of the gamma heating (for each campaign)included in the RPV and RVI AMP.

Czech Computational assessment of the RVI swelling and creep development (1 x 5 year)

Republic based upon real operational historyconfiguration of the fuel elements in the core, campaignincluded in the RPV and RVI AMP.

Measurement of the macroscopic geometry changes of the core shroudunder preparation, planned to the 2021-2022.

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Country Responses to Question 2 Failure events caused by radiation induced void swelling and creep have not been reported to NRA.

Japan A previous Japanese safety research program developed a creep evaluation formula based on creep tests of baffle former bolts under irradiation [7].

We have yet no experiences or concerns of void swelling or creep in PWR components.

South Those might be a significant degradation mechanism in CANDU fuel channels.

Korea The baffle plates has been VT tested for void swelling and these test will be followed up every 4 year, no indications.

Some material studies of Ringhals material has been done related to these issues, for example:

Sweden Edwards et. al. J. of Nuclear Materials, vol 384, (2009), pp. 249-255.

A. Jensen, B. Forssgren, P. Efsing, B. Bengtsson and M. Molin, Examination of highly irradiated stainless steels from BWR and PWR reactor pressure vessel internals, Proceedings frn Fontevraud 7, Avignon, Frankrike, SFEN, 2010.

Currently irradiation induced void swelling or creep is not an issue of the AMP and there is no running research programme.

The phenomenon is mentioned in KATAM (catalogue of potential ageing mechanisms),

Switzerland but not addressed separately.

ENSI will follow the state-of-the-art and the international OpEx. If necessary it will be included into the AMP.

NRC and the U.S. industry have been interested in research activities related to both void swelling and creep. Void swelling of high-fluence baffle plates were assessed as part of the Zorita Internals Research Program (ZIRP). There has been no operating experience where void swelling has been indicated as a causal factor in the material/component degradation. While the NRC has not been actively involved in creep/stress relaxation research, the US industry, through EPRI, has been engaged in several recent projects. Stress relaxation does play a fundamental role in U.S.A. baffle-former-bolt failures. An extensive test matrix was developed to develop fundamental material stress relaxation data under representative temperature and irradiation conditions to support the improvement of baffle-former-bolt degradation models to inform both inspection and replacement activities. Aging management of void swelling and creep are managed by GALL Report (NUREG-1801, Rev. 2)

AMP XI.M16A, PWR Vessel Internals, and GALL-SLR Report (NUREG-2191)

AMP XI.M16A, PWR Vessel Internals.

Table 3-3 compiles responses from participating countries to Question 3 about radiation embrittlement and their (a) establishment and (b) application of embrittlement trend curves (ETCs).

Table 3-3. Embrittlement Trend Curves Country Responses to Question 3 (a, b)

(a) At this moment the current NTD ASI development is ongoing based on results from IAEA NULIFEPlant Life Management of NPPs project, where the VERLIFE procedure was further modernised and extended with collaboration of experts of PWR operating countries as well as of VVER operating countries including Czech Russian experts. The next edition of NTD ASI is prepared for issue in 2019.

Republic New prediction formulae of trend curves are based on results from the analysis of the database of surveillance specimen test results (after re-analysis of neutron fluence and reconstitution of specimens realized within European projects TACIS, TAREG and Russian projects).

(b) We use ETC for predictive purposes.

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Country Responses to Question 3 (a, b)

(a) Current Japanese embrittlement trend curve is specified in the Japan Electric Association Code Method of Surveillance Tests for Structural Materials of Nuclear Reactors (JEAC4201-2007) [2013 Supplement] [8]. It was implemented into Japan Japanese regulation in 2015.

(b) In Japan, ETC is used for interpolating between surveillance results when the fluence of the inner surface of RPV is greater than 2.4x1019 n/cm2 (E > 1 MeV) [9].

(a) Embrittlement trend curve (ETC) in RG-1.99 (rev. 2) has been used as a Korean trend curve since RG-1.99 (rev. 2) was issued. ETC from 10 CFR 50.61a is only used for a comparative purpose.

South (b) South Korea has used the embrittlement trend curve in RG-1.99 (rev. 2) to predict Korea RTNDT and decrease in upper shelf energy at the design phase. While the plant operates, the trend curve of the specific reactor vessel can be adjusted by using the surveillance test results. In South Korea, all PWRs have their own surveillance specimens (capsules) as well as archive materials.

(a) Its up to the licensee to justify the use of trend curve and implemented margin to the regulator. Because of RPV welds with high Ni- and Mn-content, Ringhals use Sweden an in-house ETC fitted to surveillance data.

(b) For weld material we rely on surveillance data and interpolating between results.

(a) The applied ETC is the one defined by RG 1.99 Rev. 2.

It was implemented in Swiss Regulation since 2008 (DETEC Ordinance Preliminary shut down of NPP) and 2011 (RegGuide ENSI-B01).

Switzerland The RG 1.99 ETC was used before 2008 too, but it was nonmandatory.

(b) If surveillance data are available (which is the case for all Swiss NPP), ETC is used only for interpolating/extrapolating these results.

(a) The methodology provided in Regulatory Guide 1.99, Rev. 2 (May 1988) is used to evaluate neutron embrittlement of RPV steels. Because use of a regulatory guide is not required, implementation is on a plant-specific basis, generally in plant-specific responses to Generic Letter 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations, issued on July 12, 1988. This guidance is often used to determine the allowable pressure-temperature envelopes for normal operations and BWR leak-testing that are required to satisfy Appendix G to 10 CFR Part 50Fracture Toughness Requirements. Additionally, the procedures documented in RG 1.99, Rev. 2 are directly incorporated into 10 CFR 50.61Fracture Toughness Requirements for U.S.A. Protection Against Pressurized Thermal Shock Events (1984), which requires licensees to demonstrate that the likelihood of reactor pressure vessel rupture during a pressurized thermal shock event is insignificant.

(b) The methodology of RG 1.99, Rev. 2, uses the results of surveillance testing (Regulatory Position C.2 of the RG), if the data are credible as defined in the RG, to best fit the ETC that is provided in Regulatory Position C.1 of the RG to the surveillance data. If the surveillance testing is not deemed credible, then the ETC is determined based solely on Regulatory Position C.1 using either known or generic information about the material. In either case, the ETC can be used to extrapolate the predicted shift in RTNDT beyond the fluence values of the surveillance results.

Table 3-4 contains responses to Question 4 regarding plans to use additive manufacturing (or other advanced manufacturing techniques) for repair/replacement of components.

19

Table 3-4. Advanced Manufacturing Technologies Country Responses to Question 4 Czech Yes, validation of welding methodology for RPV wall repair was performed.

Republic Answer from Central Research Institute of Electric Power Industry (CRIEPI):

Japan The Japanese utilities and CRIEPI have not developed repair and replacement technology that utilizes advanced technology, and have no plan for it.

We plan to apply the additive manufacturing (AM) methods to the production of various reactor components, which include a nuclear fuel support grid, a lower flow plate, South safety-grade valves and some parts of CRDM. National R&D programs are underway to Korea evaluate the potential use of AM components in the reactor environments. [The South]

Korean regulatory body is interested in advanced manufacturing techniques and thus, observing industry research results.

A Vattenfall R&D project has been started together with the Swedish industry to Sweden manufacture some test specimens and investigate the material properties. No results have yet been published.

The ENSI as regulator does not plan or initiate the usage of such technologies. The initiative for the use of advanced manufacturing techniques has to come from the licensees and/or manufacturers. We currently have no knowledge of such plans. ENSI Switzerland assumes these technologies offers attractive opportunities and that there will be interest in their future application. Therefore, ENSI will make an effort to keep up to date on advances in these technologies.

The future use of advanced manufacturing techniques (AMT) in component repair and replacement in operating plants in the USA is currently uncertain. At least one component, a non-safety related thimble plug device (TPD), manufactured using laser powder bed fusion additive manufacturing, was installed at a plant in spring 2020. It is U.S.A. apparent that the U.S. nuclear industry is considering using AMTs for localized and surface applications such as localized repair, component hard-facing, and corrosion protection. Bulk AMT applications are also being considered for producing components that are no longer available from the original equipment manufacturer or that have excessive lead-times when manufactured using conventional methods.

Table 3-5 provides additional supporting references supplied by the participants from Japan for the responses summarized in Tables 4-7.

Table 3-5. Additional Japanese References References Cited in Response to Questions 1, 2, 3, and 4 Reference (only in Japanese):

[1] Nuclear Regulation Authority, The Guide for Extension of Operational Period on Commercial Power Reactors, September 20, 2017.

https://www.nsr.go.jp/data/000069250.pdf

[2] Nuclear Regulation Authority, The Standard Review Plan for Extension of Operational Period on Commercial Power Reactors, April 2016.

https://www.nsr.go.jp/data/000147250.pdf

[3] Nuclear Regulation Authority, The Guideline on Implementing Measures for Aging Management at Commercial Nuclear Reactors, September 20, 2017.

https://www.nsr.go.jp/data/000069249.pdf

[4] Nuclear Regulation Authority, The Standard Review Guideline for Aging Management at Commercial Nuclear Reactors, September 2, 2016.

https://www.nsr.go.jp/data/000168877.pdf

[5] Atomic Energy Society of Japan, Code on Implementation and Review of Nuclear Power Plant 20

References Cited in Response to Questions 1, 2, 3, and 4 Ageing Management Program 2015, AESJ-SC-P005:2015, March 2016.

[6] Nuclear Regulation Authority, Draft of field of safety research to be promoted in the future and its implementation policy, September 3, 2019.

https://www.nsr.go.jp/data/000275651.pdf

[7] Japan Nuclear Safety organization, Research report related to evaluation technique of irradiation assisted stress corrosion cracking (IASCC) JFY 2008, September 2009.

http://warp.da.ndl.go.jp/info:ndljp/pid/10207746/www.nsr.go.jp/archive/jnes/atom-pdf/seika/000014676.pdf

[8] The Japan Electric Association, Method of Surveillance Tests for Structural Materials of Nuclear Reactors (JEAC4201-2007) [2013 Supplement], May 14, 2014.

[9] Nuclear Regulation Authority, Technical evaluation report related to the Japan Electric Associations Method of Surveillance Tests for Structural Materials of Nuclear Reactors (JEAC4201-2007) [2013 Supplement], October 2015.

http://www.nsr.go.jp/data/000125554.pdf 21

4 WORKSHOP

SUMMARY

The International Workshop on Age-Related Degradation of Reactor Vessels and Internals held May 23-24, 2019, at NRC Headquarters, focused on materials degradation of safety-related components during LTO, including RPV embrittlement and the degradation of RPV internals and piping due to irradiation. Participants included representatives from 31 regulatory, research, and industry organizations in 11 countries (Table 1). Presentations addressed the state of knowledge, research activities, and OE related to RPV embrittlement at high fluence levels, as well as degradation of RPV internals and other safety-significant primary pressure boundary components during operation for up to 80 years (Table 2). RES staff also worked with EPRI to access research insights and industry OE from its members, including both domestic and international NPP operators. Following the prepared presentations, the workshop participants were invited to participate in panel discussions addressing the prognosis for effective aging management and any additional research needs and to submit additional information in correspondence. Section 3 of this report includes these responses. The individual presentations from the workshop are summarized in Section 2, and slides are condensed in Appendix B. The participants presentation files are publicly available in ADAMS at Accession No. ML19150A174.

While the workshop presenters varied greatly among government, academic, and industry organizations, their experiences with materials aging management were remarkably similar.

Each organization recognized several materials degradation challenges that would need to be addressed to enable LTO of power reactors, and they discussed technical and regulatory approaches to understanding and mitigating these challenges. The key technical issues identified as concerns for LTO were consistent across organizations and consistent with the high-priority topics being pursued by the NRC. Specifically, the main structures and components recognized as having materials challenges in extended operating lifetimes are the RPV, RVIs and piping, concrete, and cables.

Approaches to the development and use of ETCs for RPV embrittlement did vary somewhat.

For example, while the United States requires licensees to demonstrate that the likelihood of RPV rupture during a pressurized thermal shock event is insignificant, technical guidance is in nonmandatory documents, and implementation is on a plant-specific basis. Some licensees use ETCs to interpolate between surveillance data points, while others also use ETCs for extrapolation beyond available data.

The workshop provided an opportunity to share information among international counterparts on approaches to aging management, including AMPs and related research being pursued by international organizations. These were found to be consistent with the NRCs approach. The staff learned that international counterparts are evaluating similar technical issues, as their reactors enter periods of extended operation.

22

APPENDIX A WORKSHOP ATTENDEES Table A-1 lists the workshop attendees, along with their organization and e-mail address.

Table A-1. Workshop Attendees First (Given) Last Organization E-Mail Name (Family)

Name David Alley NRC david.alley@nrc.gov Kyle Amberge EPRI kamberge@epri.com Xavier Amoros IDOM xavier.amoros@idom.com Taku Arai CRIEPI arait@criepi.denken.or.jp Kensaku Arai NRA kensaku_arai@nsr.go.jp Johan Blomstroem Ringhals AB johan.blomstroem@vattenfall.com Mike Burke EPRI mburke@epri.com Grace Burke University of m.g.burke@manchester.ac.uk Manchester Xiang (Frank) Chen ORNL chenx2@ornl.gov Ganesh Cheruvenki NRC ganesh.cheruvenki@nrc.gov Michel De Smet Tractebel michel.desmet@tractebel.engie.com Matt DeVan Framatome Inc. matt.devan@framatome.com Hana Dlouha SUJB hana.dlouha@sujb.cz Ralph Doering ENSI ralph.doering@ensi.ch Charles Dupuit Tractebel charles.dupuit@tractebel.engie.com Jakub Ertl CEZ jakub.ertl@cez.cz Edison Fernandez NRC edison.fernandez@nrc.gov Eric Focht NRC eric.focht@nrc.gov Istvan Frankl NRC istvan.frankl@nrc.gov Victor Garcia IDOM vmgarcia@idom.com Serguei Gavrilov SCK CEN serguei.gavrilov@sckcen.be Brian Harris NRC brian.harris2@nrc.gov Jens Heldt KKL jens.heldt@kkl.ch Allen Hiser NRC allen.hiser@nrc.gov Keith Hoffman NRC keith.hoffman@nrc.gov Gareth Hopkin ONR gareth.hopkin@onr.gov.uk Amy Hull NRC amy.hull@nrc.gov Dong-Yuk Kim KINS eastation@kins.re.kr Yong-Beum Kim KINS ybkim@kins.re.kr Dong Yuk Kim KINS ky32kdy@kins.re.kr Marlies Lambrecht SCK CEN mlambrec@sckcen.be Olivier Lareynie NRC/ASN olivier.lareynie@asn.fr Yiu Law NRC yiu.law@nrc.gov Bong-Sang Lee KAERI bongsl@kaeri.re.kr Heather Malikowski Exelon heather.malikowski@exeloncorp.com James Medoff NRC james.medoff@nrc.gov Roch Menand EDF roch.menand@edf.fr A-1

First (Given) Last Organization E-Mail Name (Family)

Name Carol Moyer NRC carol.moyer@nrc.gov Masaki Nagai CRIEPI nagai@criepi.denken.or.jp Drew Odell Exelon andrew.odell@exeloncorp.com Gabor Petofi IAEA g.petofi@iaea.org Jeff Poehler NRC jeffrey.poehler@nrc.gov Iouri Prokofiev NRC iouri.prokofiev@nrc.gov (phone)

Matthew Ralstin Southern Nuclear mfralsti@southernco.com Company Sandor Ratkai MVM Paks NPP ratkai@npp.hu Ali Rezai NRC ali.rezai@nrc.gov Kerstin Richnau Ringhals AB kerstinrichnau@vattenfall.com Dave Rudland NRC david.rudland@nrc.gov Jolana Rydlova SUJB jolana.rydlova@sujb.cz Simon Sheng NRC simon.sheng@nrc.gov Todd Sherman Entergy tsherm1@entergy.com Tae-Kwang Song KINS tksong@kins.re.kr Casper Sun NRC cxs14@nrc.gov Brian Thomas NRC brian.thomas@nrc.gov Rob Tregoning NRC robert.tregoning@nrc.gov John Tsao NRC john.tsao@nrc.gov Diederik Van Nuffel Bel V diederik.vannuffel@belv.be Emmanuel Viard IRSN emmanuel.viard@irsn.fr Leah Walters Entergy lwalter@entergy.com Chris Wax Arizona Public christopher.wax@aps.com Service Emma Wong EPRI ewong@epri.com Jun-Seog Yang KHNP jsyang88@khnp.co.kr Mark Yoo NRC mark.yoo@nrc.gov Sho Yoshinaga JEPIC-USA yoshinaga-sho@jepic-usa.org Miroslav Zamboch UJV Rez miroslav.zamboch@ujv.cz A-2

APPENDIX B PRESENTATION SLIDES B.1 Aging Management and Subsequent License Renewal in the United States (A. Hiser)

B-1

B-2 B-3 B-4 B-5 B-6 B-7 B-8 B-9 B-10 B-11 B-12 B-13 B-14 B-15 B.2 NRCs Aging and Materials Research Activities (R. Tregoning)

B-16

B-17 B-18 B-19 B-20 B-21 B-22 B.3 U.S. Nuclear Electric Power Generation Industry Management of Age-Related Degradation (M. Burke)

B-23

B-24 B-25 B-26 B-27 B-28 B-29 B-30 B-31 B-32 B-33 B-34 B-35 B-36 B-37 B-38 B-39 B-40 B.4 Overview of Metals Research in LWRS Program MRP (F. Chen)

B-41

B-42 B-43 B-44 B-45 B-46 B-47 B-48 B.5 Overview of Safety Research on Metal Aging Due to Neutron Irradiation in S/NRA/R (K. Arai)

B-49

B-50 B-51 B-52 B-53 B-54 B-55 B-56 B-57 B-58 B.6 CRIEPI Research Activities on Neutron Irradiation Embrittlement of RPV and Core Internals (T. Arai)

B-59

B-60 B-61 B-62 B-63 B-64 B-65 B-66 B-67 B-68 B-69 B-70 B-71 B-72 B.7 Czech Approach to Ageing Management of Reactor Pressure Vessel and Reactor Vessel InternalsState of Knowledge (J. Ertl)

B-73

B-74 B-75 B-76 B-77 B-78 B-79 B-80 B-81 B-82 B-83 B-84 B-85 B-86 B-87 B.8 UJV Activities in International Research Projects in the Field of RPV and RVI (M. Zamboch)

B-88

B-89 B-90 B-91 B-92 B-93 B-94 B-95 B-96 B.9 Measurement of Core Shroud at NPP Temelin (M. Zamboch)

B-97

B-98 B-99 B-100 B-101 B-102 B-103 B-104 B-105 B-106 B-107 B-108 B-109 B-110 B-111 B-112 B.10 Belgian R&D on Environmental Effects on Materials Degradation in LWRs (S. Gavrilov)

B-113

B-114 B-115 B-116 B-117 B.11 Repair of Doel 1 NPP Reactor Vessel Head Penetrations (C. Dupuit)

B-118

B-119 B-120 B-121 B-122 B.12 Belgian R&D Using the Enhanced Surveillance Strategy for RPV Embrittlement Assessment (M. Lambrecht)

B-123

B-124 B-125 B-126 B-127 B-128 B-129 B-130 B-131 B.13 Doel 1 & 2 Upper Plenum Injection Line Issue (M. De Smet)

B-132

B-133 B-134 B-135 B-136 B-137 B-138 B.14 Inspection of Control Rod Guide Assemblies in Belgian NPPs (C. Dupuit)

B-139

B-140 B-141 B-142 B-143 B-144 B-145 B.15 Current Status of Aging Management on Reactor Vessels in Korea (focusing on surveillance test) (T-K Song)

B-146

B-147 B-148 B-149 B-150 B-151 B-152 B-153 B-154 B-155 B-156 B-157 B.16 Ongoing Researches in Age-Related Degradation of Reactor Materials in Korea (B-S Lee)

B-158

B-159 B-160 B-161 B-162 B-163 B-164 B-165 B-166 B.17 Operating Experience (OpE) on RV Internals, RV Head Penetrations, and RCS Small Bore Nozzles in Korea (J-S Yang)

B-167

B-168 B-169 B-170 B-171 B-172 B-173 B-174 B-175 B-176 B-177 B-178 B-179 B-180 B.18 UK Regulatory Experience in Materials Ageing (G. Hopkin)

B-181

B-182 B-183 B-184 B-185 B-186 B-187 B-188 B.19 State of Knowledge and Research Activities on RPV Materials in UK (G. Burke)

B-189

B-190 B-191 B-192 B-193 B-194 B-195 B-196 B-197 B-198 B-199 B-200 B-201 B-202 B-203 B-204 B-205 B-206 B-207 B-208 B.20 AM and LTO-Related Activities of RPV and Its Internals and Other Primary Pressure Boundary Components at the Paks NPP (S. Ratkai)

B-209

B-210 B-211 B-212 B-213 B-214 B-215 B-216 B-217 B-218 B-219 B-220 B-221 B-222 B-223 B-224 B-225 B-226 B-227 B-228 B-229 B-230 B-231 B-232 B-233 B-234 B-235 B-236 B-237 B.21 EDF Operating Experience RV Internals (R. Menand)

B-238

B-239 B-240 B-241 B-242 B-243 B-244 B-245 B-246 B-247 B-248 B-249 B-250 B.22 Carbon Segregations in Heavy Forged Components (E. Viard)

B-251

B-252 B-253 B-254 B-255 B-256 B-257 B-258 B-259 B-260 B.23 Aging Management and LTO of NPPs in Switzerland: Status 2019 (R. Doering)

B-261

B-262 B-263 B-264 B-265 B-266 B-267 B-268 B-269 B-270 B-271 B-272 B-273 B-274 B-275 B-276 B-277 B-278 B-279 B-280 B-281 B-282 B-283 B-284 B-285 B-286 B-287 B-288 B-289 B-290 B-291 B-292 B-293 B-294 B-295 B-296 B.24 Operating Experience of a Swiss BWR (J. Heldt)

B-297

B-298 B-299 B-300 B-301 B-302 B-303 B-304 B-305 B-306 B-307 B-308 B-309 B-310 B-311 B-312 B-313