Nuclear Safety
2024

A Review of Operating

Experience Involving Passive
Component Material Degradingin Periods of Extended andLong-Term Operation
CODAP Topical Report

NE A

Nuclear Safety

A Review of Operating Experience Involving
Passive Component Material Degrading in Periodsof Extended and Long-Term Operation
CODAP Topical Report© OECD 2024
NEA No. 7614
NUCLEAR ENERGY AGENCY
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FOREWORD

Foreword

The Nuclear Energy Agency (NEA) Component Operational Experience, Degradation and Ageing
Programme (CODAP) is the continuation of the 2002-2011 OECD/NEA Pipe Failure Data Exchange
Project (OPDE) and the 2006-2010 OECD/NEA Stress Corrosion Cracking and Cable Ageing Project(SCAP). In December 2020, 13 countries and economies (Canada, Czechia, Finland, France,
Germany, Japan, Korea, the Netherlands, the Slovak Republic, Spain, Switzerland, Chinese Taipei
and the United States) entered into an agreement for CODAP’s fourth term (2021-2023).
The project prepares topical reports to communicate insights and results from systematic
reviews of operating experience (OPEX) data on metallic passive component material degradationand failure. The topical reports provide summaries of trends and patterns in metallic material
performance in different operating environments and for different time periods.
This report provides technical insights regarding material degradation mechanisms that haveproduced failures during periods of extended operation and long-term operation. It was approvedby the NEA Committee on the Safety of Nuclear Installations (CSNI) on 8 December 2021 (seedocument NEA/SEN/SIN(2021)2/REV, not publicly accessible) and prepared for publication by the
NEA.

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ACKNOWLEDGEMENTS

Acknowledgements

This report is the product of work carried out over a three-year period since June 2019 and of six
meetings of the CODAP Management Board, chaired by Dr Jovica Riznic (Canadian Nuclear
Safety Commission). The list of members of the Management Board can be found in the tablebelow. This report could not have been produced without the valuable contributions of themany experts who helped to collect and assemble the information contained herein.

Jovica RIZNIC

CanadaJan DOTREL

Czechia

Marko MEHTONEN

FinlandFlorence CROMBEZFranceStephan FAUST

Germany

Kensaku ARAIJapan

Sung Sik KANG

Korea

Camilla MATTEOLINetherlands

Vladimír JURČO

Slovak Republic

Patricia FERNANDEZ-ANDUJAR

Spain

Michaela WEBER

Switzerland

Jec-Kong GONE

Chinese Taipei

Eric FOCHT

United States

Bengt LYDELL

United States

Diego ESCRIG FORANONEAA REVIEW OF OPERATING EXPERIENCE INVOLVING PASSIVE COMPONENT MATERIAL DEGRADING IN PERIODS OF EXTENDED AND LTO, NEA No. 7614, © OECD 2024

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TABLE OF CONTENTS

Table of contents

List of abbreviations and acronyms .................................................................................................. 11
Executive summary .............................................................................................................................. 17
Conclusions ........................................................................................................................................ 18
CODAP work for the fourth term (2021-2023) and beyond .......................................................... 18
CODAP data collection process........................................................................................................ 19
Chapter 1. Introduction ......................................................................................................................... 21
1.1

Objectives and scope ................................................................................................................ 21

1.2

Background and related work ................................................................................................. 21

1.3

Technical approach .................................................................................................................. 22

1.4

Nomenclature............................................................................................................................ 25

1.5

Report structure ........................................................................................................................ 25

Chapter 2. Temporal changes in material degradation .................................................................. 29
2.1

The CODAP-OPEX knowledge base......................................................................................... 29

2.2

Degradation mechanisms acting on process piping ............................................................ 38

2.3

Degradation mechanisms acting on cast austenitic stainless steel .................................. 46

2.4

Degradation mechanisms acting on reactor vessel internals ............................................ 46

2.5“Uncommon” and “unexpected” material degradation scenarios ..................................... 47

Chapter 3. Country/economy-specific overviews of material degradation
state-of-knowledge, research and operating experience ........................................... 49
3.1

Introduction............................................................................................................................... 49

3.2

Canada ....................................................................................................................................... 50

3.3

Czechia ....................................................................................................................................... 52

3.4

Finland ....................................................................................................................................... 55

3.5

France ......................................................................................................................................... 60

3.6

Germany..................................................................................................................................... 62

3.7

Japan ........................................................................................................................................... 63

3.8

Korea .......................................................................................................................................... 65

3.9

The Netherlands ....................................................................................................................... 68

3.10 Slovak Republic ......................................................................................................................... 69
3.11 Spain ........................................................................................................................................... 71
3.12 Switzerland ................................................................................................................................ 75
3.13 Chinese Taipei ........................................................................................................................... 82
3.14 United States ............................................................................................................................. 88

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Chapter 4. PEO/LTO operating experience data analysis ............................................................... 97
4.1

CODAP event database vs. WGIAGE conclusions ................................................................. 97

4.2

CODAP event database vs. PMDA/EMDA conclusions ......................................................... 97

4.3

Basic principles of PEO/LTO-OPEX data analysis ................................................................ 106

4.4

PEO/LTO-OPEX data analysis insights – safety-related piping ......................................... 110

4.5

PEO/LTO-OPEX data analysis insights – non-safety-related piping ................................. 115

4.6

PEO/LTO-OPEX data analysis insights – fire water system piping ................................... 117

4.7

PEO/LTO-OPEX data analysis insights – non-piping passive components...................... 118

Chapter 5. Conclusions and recommendations ............................................................................. 121
5.1

Main conclusions .................................................................................................................... 121

5.2

CODAP activities for the fourth term (2021-2023) and beyond ......................................... 123

References ............................................................................................................................................ 125
Annex A: Glossary of technical terms ............................................................................................ 133
Annex B: Examples of Reporting Systems/Routines for Passive Component Material
Degradation and Failure ................................................................................................ 143

List of figures
1-1:

The process for developing insights into long-term material degradation ................... 24

1-2:

Number of commercial reactor units in long-term operation in CODAP project
member countries/economies ............................................................................................. 28

2-1:

The piping operating experience by material type and time period .............................. 32

2-2:

The non-piping passive component OPEX by material type and time period............... 33

2-3:

Piping material degradation mechanisms observed during five decades ofcommercial nuclear power plant operation experience................................................... 34

2-4:

Piping OPEX by system group and age of plant at time of failure ................................... 35

2-5:

Piping OPEX by system group and for 0 ≤ T < 25 reactor operating years ...................... 35

2-6:

Piping OPEX by safety class and plant age ......................................................................... 36

2-7:

The PWSCC operating experience ....................................................................................... 36

2-8:

Reactor internals OPEX summary by plant type, component type and time period .... 37

2-9:

PWSCC in PWR reactor pressure vessel head penetrations (VHP) .................................. 37

2-10: The CODAP OPEX by NSSS design ....................................................................................... 39
2-11: The piping and non-piping passive component OPEX by CODAP member
country/economy ................................................................................................................... 39
2-12: Small-diameter pipe failure OPEX by time period, safety class and outside
diameter .................................................................................................................................. 41
2-13: Socket weld failure OPEX data by time period ................................................................... 41
2-14: Medium-diameter pipe operating experience by time period, safety class andoutside diameter .................................................................................................................... 42
2-15: Large-diameter (OD > 273.1 mm) pipe OPEX by time period and safety class ............... 42
2-16: Piping fatigue operating experience .................................................................................... 438

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2-17: Piping operating experience involving stress corrosion cracking ................................... 43
2-18: The VVER-440/213 passive component OPEX in CODAP organised by safety class ...... 44
2-19: Piping operating experience involving flow-assisted degradation mechanisms .......... 44
2-20: Piping operating experience involving corrosion failures ................................................ 45
2-21: Service water piping failure operating experience ............................................................ 45
2-22: Fire water piping operating experience .............................................................................. 46
3-1:

Steam generator collector DMW .......................................................................................... 54

3-2:

Dukovany-2 first-in-kind SEFW weld overlay .................................................................... 55

3-3:

The Fourth PSR schedule for the French 900 MWe reactor units .................................... 61

3-4:

The German piping OPEX after 25 years of operation....................................................... 63

3-5:

Piping OPEX as a function of plant age and NSSS-design ................................................ 75

3-6:

The code and non-code repair procedure........................................................................... 83

3-7:

Procedure for assessing leakage of safety-related piping ................................................ 84

3-8:

Geometry of corrosion-induced pipe flaw .......................................................................... 85

3-9:

NCSW leak locations adjacent to CCW heat exchanger ................................................... 87

3-10: Non-code repairs at Taiwanese nuclear power plants in 1993-2000 and 2007-2019 .... 87
3-11: Distribution of OPEX during the PEO for US plants vs. component age ......................... 92
3-12: Distribution of damage/degradation mechanisms for US plants during PEO ............... 92
3-13: Distribution of OPEX during the PEO across system groups ............................................ 93
3-14: Illustrations of (a) a typical Westinghouse baffle-former assembly and (b) crosssection of one octant of a typical Westinghouse baffle-former assembly showing
locations of bolt types ........................................................................................................... 95
4-1:

The BWR OPEX involving SCC of safety class 1 alloy 182/82 welds............................... 104

4-2:

Pipe failure rate as a function of the age of the piping at the time of failure ofsafety- and non-safety-related piping .............................................................................. 108

4-3:

The small-diameter (≤ DN50) pipe failure OPEX .............................................................. 108

4-4:

The BWR-specific IGSCC operating experience data....................................................... 109

4-5:

The IGSCC “incident rates” by time period and BWR design generation ..................... 109

4-6:

The safety related piping OPEX by plant type, safety class and pipe-diameter vs.
plant age at the time of pipe failure .................................................................................. 111

4-7:

The BWR and PWR primary piping system OPEX by event date ................................... 112

4-8:

RCPB piping OPEX by degradation mechanism and time period .................................. 112

4-9:

Normalised RCPB piping OPEX ........................................................................................... 113

4-10: Safety class 2 piping OPEX by system group and time period ....................................... 113
4-11: Safety class 3 piping OPEX by system group and time period ....................................... 114
4-12: Temporal trends in safety class 3 piping OPEX................................................................ 114
4-13: OPEX on HCF in safety-related piping ............................................................................... 115
4-14: Emergency diesel generator fuel oil and lubricating oil pipe failure OPEX .................. 115
4-15: Flow-assisted pipe wall thinning mechanisms ............................................................... 116
4-16: FAC operating experience ................................................................................................... 116
4-17: Fire protection water system pipe failure experience .................................................... 117
4-18: PWR vessel head penetration OPEX .................................................................................. 118

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TABLE OF CONTENTS

List of tables

10

1-1:

Results of WGIAGE questionnaire on PEO/LTO [32]........................................................... 23

1-2:

Commercial nuclear power plant operating licence bases in CODAP project
member countries/economies ............................................................................................. 26

2-1:

Effect of ageing management on material performance in SCC- andfatigue-susceptible environments ....................................................................................... 30

2-2:

Effect of ageing management on material performance in FAD-susceptible
and corrosive environments ................................................................................................ 31

2-3:

Degradation mechanisms by the time period in which a failure was discovered ........ 32

2-4:

Piping operating experience by time period, safety class and outside diameter .......... 40

3-1:

Candidate materials for first stage of R&D programme ................................................... 64

3-2:

Candidate materials for second stage of R&D programme .............................................. 64

3-3:

Examples of leak locations in the NCSW piping system .................................................. 86

4-1:

A synthesis of the results of the WGIAGE survey .............................................................. 98

4-2:

Comparison of PMDA/EMDA conclusions with CODAP event database insights.......... 99

4-3:

Comparison of PMDA/EMDA conclusions with CODAP event database insights(Part 2) .................................................................................................................................... 101

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LIST OF ABBREVIATIONS AND ACRONYMS

List of abbreviations and acronyms

AERM

Ageing effects requiring management

AHC

Access hole cover

AMP

Ageing management programme

ANVS

Autoriteit Nucleaire Veiligheid en Stralingsbescherming (Authority for Nuclear
Safety and Radiation Protection, Netherlands)

AOT

Allowed outage time

APAL

Advanced PTS analysis for LTO (Euratom)

AR

Action request

ART

Adjusted reference temperature

ASME

American Society of Mechanical Engineers

ASN

Autorité de Sûreté Nucléaire (Nuclear Safety Authority, France)

BAC

Boric acid corrosion

BFB

Baffle-former bolt

BKW

Bernische Kraftwerke AG (Switzerland)

BMI

Bottom mounted instrument nozzle

BNI

Basic nuclear installation

BRUTE

Barsebäck RPV material used for true evaluation of embrittlement

BWR

Boiling water reactor

BWRVIP

BWR vessel integrity programme

CAPS

CSNI activity proposal sheet

CASS

Cast austenitic stainless steel

CBBCore barrel bolt

CC

Component cooling

CFPR

Carbon fibre-reinforced polymer repair

CFR

Code of federal regulations

CHUG

CHECWORKS® users group

CNS

Centre for Nuclear Safety (Finland)

CNSC

Canadian Nuclear Safety Commission

CR

Condition report

CRDM

Control rod drive mechanism

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LIST OF ABBREVIATIONS AND ACRONYMS

12

CRIEPI

Central Research Institute of Electric Power Industry (Japan)

CTH

Chalmers University of Technology (Sweden)

CUF

Cumulative usage factor

CVC

Chemical and volume control

CW

Cold worked

DETEC

Federal Department of the Environment, Transport, Energy and Communications(Switzerland)

DHC

Delayed hydride cracking

DMW

Dissimilar metal weld

DPA

Displacement per atom

EAF

Environmentally assisted fatigue

EBO V2

Bohunice Nuclear Power Plant units 3 and 4 (Slovak Republic)

ECSCC

External chloride induced SCC

EDG

Emergency diesel generator

EFPY

Effective full-power year

EMDA

Expanded material degradation assessment

EMO

Mochovce Nuclear Power Plant units 1 and 2 (Slovak Republic)

ENS

Event notification system

ENSI

Federal Nuclear Safety Inspectorate (Switzerland)

ENVDEG

Environmental degradation conference

EOL

End of licence

EOLE

End of licence, extended

EPU

Extended power uprate

ESW

Essential service water

ETC

Embrittlement trend curve

FAC

Flow-accelerated corrosion

FAD

Flow-assisted degradation

FAMOS

Fatigue monitoring system

FAVOR

Fracture analysis of vessels – Oak Ridge

FEM

Finite element method

FFS

Fitness for service

FIV

Flow induced vibration

FSWOL

Full structural weld overlay

FW

Feedwater

GALL

Generic Aging Lessons Learned (NRC)GE

General Electric

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LIST OF ABBREVIATIONS AND ACRONYMS

GSKL

Gruppe der schweizerischen Kernkraftwerksleiter (Group of Swiss Nuclear Power
Plant Managers, Switzerland)

HCF

High-cycle fatigue

HDPE

High-density polyethylene

HPCI

High pressure coolant injection

HVAC

Heating ventilation, and air conditioning

HWC

Hydrogen water chemistry

HWR

Heavy water reactor

IA

Instrument air

IAD

Irradiation-assisted degradation

IAEA

International Atomic Energy Agency

ICES

INPO consolidated events database

ID

Inside diameter

IGALL

International Generic Ageing Lessons Learned (IAEA)

IGSCCIntergranular SCC

IHSI

Induction heat stress improvement

INCEFA

Increasing Safety in Nuclear Power Plants by Covering gaps in (EU)
Environmental Fatigue Assessment

INPO

Institute of Nuclear Power Operations (United States)

IPA

Integrated plant assessment

IR

Incident report

IRP

International review panel

IRS

Incident reporting system

IRSN

Institut de Radioprotection et de Sûreté Nucléaire (Institute of Radiologicalprotection and Nuclear Safety, France)

JIE

Jet impingement erosion

JNO

Japanese Nuclear Operators

KATAM

Katalogs der Alterungsmechanismen (Catalogue of Ageing Mechanisms,
Switzerland)

KEG

Kernenergiegesetz (Nuclear Energy Act, Switzerland)

KHNP

Korea Hydro and Nuclear Power Co.

KKB

Nuclear Power Plant Beznau (Switzerland)

KKG

Nuclear Power Plant Gösgen (Switzerland)

KKL

Nuclear Power Plant Leibstadt (Switzerland)

KTH

Royal Institute of Technology (Sweden)

KWU

Kraftwerk Union AG (Germany)

LBB

Leak-before-break

LCF

Low-cycle fatigue

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LIST OF ABBREVIATIONS AND ACRONYMS

LDIE

Liquid droplet impingement erosion

LLC

Low-leakage coreLR

Licence renewal

LTO

Long-term operation

LT-OAM

Long-term overall ageing management

LTOP

Low temperature overpressurisation

LWR

Light water reactor

LWRS

Light water reactor sustainability programmeMAMill-anneal

MBManagement BoardMC

Master curve

MEACTOS

Mitigating environmentally assisted cracking through optimisation of surface
condition.

MHI

Mitsubishi Heavy Industries

MIC

Microbiologically influenced corrosion

MRP

Materials reliability programme

MSR

Moisture separator reheater

NCD

Non-conformance disposition

NFPA

National Fire Protection Association

NMCA

Noble metal chemical addition

NRA

Nuclear Regulation Authority of Japan

NRC

Nuclear Regulatory Commission (United States)

NPS

Nominal pipe size

NSSC

Nuclear Safety and Security Commission (Korea)

NSSS

Nuclear steam supply system

NUREG

US Nuclear Regulatory Commission technical report designation

NWC

Normal water chemistry

OD

Outside diameterOL

Operating licence

OLNC

On-line noble chem

OPEX

Operating experience

PARTRIDGE Probabilistic analysis as a regulatory tool for risk-informed decision guidance

14

PAUT

Phased-array ultrasonic testing

PCSG

Pipe crack study group

PEO

Period of extended operation

PIP

Problem investigation process

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LIST OF ABBREVIATIONS AND ACRONYMS

PIRTPhenomena identification and ranking technique

PISA

Pressure vessel integrity and safety analysis

PLiM

Plant life management

PMDA

Proactive material degradation assessment

PSCR

Primary systems corrosion research

PSI

Paul Scherrer Institute (Switzerland)

PSR

Periodic safety review

PTS

Pressurised thermal shock

PWR

Pressurised water reactor

PWROG

PWR owners’ group

PWSCC

Primary water SCC

RHR

Residual heat removal

PZR

Pressuriser

RHWG

Reactor Harmonization Working Group (WENRA)

RVH

Reactor vessel head

RVI

Reactor vessel internals

RVID

Reactor vessel integrity database

RVLIS

Reactor vessel level indication system

RWCU

Reactor water cleanup

SAFIR

Safety of Nuclear Power Plants (Finnish National Research Programme)

SBP

Small bore piping

SCC

Stress corrosion cracking

SEFW

Super emergency feedwater system

SG

Steam generator

SICC

Strain-induced corrosion cracking

SIL

Service information letter

SME

Subject matter expert

SMILE

Studsvik material integrity life extension project

SRE

Selected representative event

SRP

Standard review planSS

Stainless steel

SVTI

Swiss Association for Technical Inspections

SW

Service water

SWR

Siedewasserreaktor (BWR)

T/C

Thermocouple

TAE

Thermal ageing embrittlement

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LIST OF ABBREVIATIONS AND ACRONYMS

16

TASC

Thermal stratification, cycling and striping

TF

Thermal fatigue

TGSCC

Transgranular SCC

TLAA

Time-limited ageing analysis

TLR

Technical letter report

TPC

Taiwan Power Company

TPR

Topical peer review

TSO

Technical support organisation

TT

Thermal treatment

UCL

Uniaxial constant loadÚJDÚrad jadrového dozoru (Nuclear Regulatory Authority of the Czech Republic)

UT

Ultrasonic testing

VHP

Vessel head penetration

VISA

Vessel integrity simulation analysis

WENRA

Western European Nuclear Regulators Association

WGIAGE

Working Group on Integrity and Ageing of Components and Structures (NEA)WGRISK

Working Group on Risk Assessment (NEA)

WPS

Warm pre-stress effect

WWER

Water-cooled, water-moderated energy reactor

XFEM

Extended finite element method

xLPR

Extremely low probability of rupture

ZIRP

Zorita Internals Research Project

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EXECUTIVE SUMMARY

Executive summary

The Nuclear Energy Agency (NEA) Component Operational Experience, Degradation and Ageing
Programme (CODAP) is the continuation of the OECD/NEA Pipe Failure Data Exchange Project(OPDE – 2002 to 2011) and the NEA Stress Corrosion Cracking and Cable Ageing Project (SCAP –2006 to 2010). In December 2017, 13 countries and economies (Canada, Czechia, Finland, France,
Germany, Japan, Korea, the Netherlands, the Slovak Republic, Spain, Switzerland, Chinese Taipei
and the United States) entered into an agreement for a third term for CODAP (2018-2020).
The CODAP project prepares topical reports to communicate insights and results fromsystematic reviews of operating experience (OPEX) data on metallic passive component material
degradation and failure. The data collection for this report ended in 2020, and only the validated
data were used. The topical reports provide summaries of trends and patterns in metallic
material performance in different operating environments and for different time periods.
During the April 2019 Working Group on Integrity and Ageing of Components and Structures(WGIAGE) meeting, it was decided to pursue an evaluation of material degradation OPEX during
periods of extended operation (PEO) and long-term operation (LTO) of nuclear power plants. The
objectives were to: a) document insights from evaluations of OPEX with passive metallic
components during PEO/LTO; and b) based on the results of these evaluations, develop
recommendations for possible enhancements to the CODAP data collection process. The results,
conclusions and recommendations that are documented in this report were obtained using thefollowing process:•

STEP 1. In 2015, the WGIAGE, operating under the Committee on the Safety of Nuclear
Installations (CSNI), published the results of a “Questionnaire on Long-term Operation ofCommercial Nuclear Power Plants” (NEA/CSNI/R(2015)13). In preparing this report, acomparison was made between the content of the CODAP event database and the resultsof the WGIAGE questionnaire. This step was taken in order to identify possible gaps inthe scope and content of CODAP relative to the WGIAGE qualitative ranking ofdegradation mechanisms considered to be important with respect to PEO/LTO.

•

STEP 2. The results of two expert panels on potential material degradation issues during
PEO/LTO were reviewed. The subject expert panels were convened by the Office of
Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). The
Proactive Material Degradation Assessment Expert Panel (PMDA, 2004 to 2006) identified
the material degradation scenarios that could affect plant systems for up to 40 years ofoperation, and the Expanded Material Degradation Assessment Expert Panel (EMDA, 2012
to 2014) extended the analysis time frame from 40 to 80 years. In preparing this report, acomparison was made between the CODAP and the results of the two expert panels. The
PMDA and EMDA expert panel members were drawn from the United States andinternational materials science community. The PMDA performed a comprehensive
evaluation of potential ageing related degradation modes for reactor internals, as well as
primary, secondary and some tertiary piping systems. The PMDA output has supported
the NRC staff evaluations of licensees’ ageing management programmes and theprioritisation of research needs. The EMDA broadened the analytical time frame to
80 years to encompass a potential second and third 20-year licence renewal. The EMDA
output has supported the NRC staff evaluations of licence renewal applications and theprioritisation of research needs.

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EXECUTIVE SUMMARY•STEP 3. In 2019, the NRC organised an international workshop on age-related degradationof reactor vessels and internals. During this workshop the participating organisations wereinvited to summarise the country/economy-specific state of knowledge, ongoing research
and associated operating experience. The technical information included in the workshop
presentations and the summary report was used to complement the country/economyspecific overviews of PEO/LTO material degradation issues in this report.

•

STEP 4. The material degradation issues as recorded in the CODAP event database werereviewed. This review included selected reactor vessel internals and safety-related andnon-safety-related piping system components.

•

STEP 5. A synthesis of the PEO/LTO material degradation insights based on the results ofSteps 1 through 4 was formulated. Because of the different plant licensing regimes
among the 13 CODAP project member countries and economies, this report considers
OPEX from plants that have been in operation for more than 25 years.

Conclusions
The technical and programmatic processes that have been implemented to address stresscorrosion cracking (SCC) problems appear to be effective. In comparison with the SCC OPEX
during the first 25 years of plant operation, a significantly lower SCC incident rate is noted forplants in PEO/LTO. A possible exception is the recent WWER-specific SCC OPEX on mediumdiameter safety Class 2 pressure boundary components.
Since the late 1980s, major piping pressure boundary failures attributed to flow-accelerated
corrosion (FAC) have been relatively rare events. This is attributed to steps taken to implement
effective secondary-side chemistry control (in PWRs), proactive piping replacements using FACresistant materials, and implementation of plant-specific FAC monitoring programmes. It isnoteworthy that the WWER-specific FAC experience appears to be different from that of PWR
plants. There is also a “gap” between the CODAP event database content with respect to FAC in
WWER plants and the ranking assigned to this material degradation mechanism in the WGIAGE
questionnaire.
Corrosion in safety-related raw water-cooling system piping and fire water system piping
poses unique ageing management challenges. Applications of non-metallic materials, advanced
metallic materials and composite repair technology are receiving increased attention in
mitigating corrosion effects on piping pressure boundary integrity.

CODAP work in the fourth term (2021-2023) and beyond
This report looks into observed age-dependent material degradation of piping systems and reactor
internals. While it is unclear whether the identified trends and patterns are representative acrossthe countries and economies participating in the CODAP project, validating these trends would betechnically challenging and time-consuming. Recommendations for potential future work include:
•

•

18

Selecting a single “system-material-environment” grouping and, for a limited and recent
time period (e.g. 2015-2020), to determine:–

the applicable country/economy-specific ageing management
requirements, regulations and operability determination procedures;–the country/economy-specific OPEX;–

the country/economy-specific regulatory inspection programmes and the processes
for documenting the findings from site inspections.

programme

Evaluating the WWER-specific FAC OPEX to determine if and how it differs from thecorresponding PWR-specific OPEX.

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The CODAP event database contains a significant population of fire protection water
system and service water system pipe failures in US plants. The project should consider
determining if this “bias” is attributed to unique US piping system designs, material
selections, operating environments and/or material ageing management programmes,
or whether it is more of a reflection of the data exchange process.

In this study, the evaluation of possible age-dependent material degradation was done in ahigh-level manner using simple “visual tests” of graphical plots. It is a first step in a more
rigorous analysis of the OPEX. The work done in producing this Topical Report would benefit
from a benchmarking exercise involving multiple TSOs under an “umbrella project”, e.g. bydeveloping a CSNI Activity Proposal Sheet (CAPS) entry with joint CODAP/WGIAGE/WGRISK
oversight and guidance.

CODAP data collection process
Since the 1960s the exchange and analysis of nuclear power plant operating experience has
been a point of focus for nuclear regulatory authorities, technical support organisations and thenuclear industry at large. The processes and systems for collecting and disseminating operating
experience data continue to evolve. For the year 2021 and beyond the CODAP Management
Board (MB) is considering the following activities to potentially enhance the data collection
process:•

Each project member to identify the current national routines for recording andsubmitting information on material degradation issues, including OPEX access.

•Each project member to report on the current national regulatory and industry routines/
practices and requirements for performing operability determinations or fitness forservice assessments. In particular, the question of how such evaluations provide
relevant OPEX data for submission to CODAP should be elaborated.

•

The CODAP Management Board has devised the term “selected representative event”(SRE) as a means for simplifying data entry. However, it has promoted an ad hoc
approach as opposed to systematic approach to the exchange of reactor internals OPEX.
It is therefore recommended that the list of BWR, CANDU, PWR and WWER reactor
internal parts of interest be more firmly defined and that a plan be developed for how to
conduct the future reactor internals OPEX exchange.

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INTRODUCTION

Chapter 1. Introduction

Since 2002, the NEA has operated an event database project that collects information on passive
metallic component degradation and failure. The scope of the database includes primarysystem piping components, reactor pressure vessel internals (“reactor components”), mainprocess and standby safety system piping, and support system piping (e.g. ASME III Class 1, 2
and 3, or equivalent) components, as well as non-safety piping components whose degradation
or failure can have significant operational impact. The CODAP project prepares topical reports
to communicate insights and results from analyses of operating experience data. This report
evaluates material degradation operating experience during periods of extended operation (PEO)
and long-term operation (LTO).

1.1 Objectives and scope
The objectives of this report are twofold. First, it aims to document insights from evaluations ofoperating experience (OPEX) with passive metallic components in commercial nuclear power
plants during PEO/LTO. Second, based on the results of these evaluations, it aims to develop
recommendations for possible enhancements to the CODAP data collection process.
The technical questions that are addressed in this report relate to the feasibility ofdetermining trends and patterns in possible age-dependent occurrence rates of passive
component failures. Also addressed are the national differences in materials ageing management
practices. The scope of the OPEX evaluations are as follows:•

safety-related and non-safety-related piping systems;•

fire water system piping 1; and•selected non-piping passive components, e.g. reactor internals.

1.2 Background and related work
Research on nuclear power plant materials’ ageing degradation has been carried out for morethan five decades, references [1]–[39]. CODAP prepares topical reports to communicate insights
and results from systematic reviews of operating experience with metallic passive component
material degradation. Since 2010 the following topical reports have been published:1.

•“Technical Basis for Commendable Practices on Ageing Management: Stress Corrosion
Cracking Mechanisms” (NEA/CSNI/R(2010)5) “ [21];•“Flow Accelerated Corrosion (FAC) of Carbon Steel and Low Alloy Steel Piping in
Commercial Nuclear Power Plants” (NEA/CSNI/R(2014)6) [30];This category was selected in order to develop an international perspective on the ageing managementof fire water systems. As an example, the US fire water system ageing management requirements aredefined in Section XI.M27 of NUREG-2191, Vol. 2, Generic Ageing Lessons Learned for Subsequent
License Renewal (GALL-SLR),” www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr2191.

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INTRODUCTION• “Operating Experience Insights into Pipe Failures in Electro-Hydraulic Control (EHC)and Instrument Air (IA) Systems” (NEA/CSNI/R(2015)6) [31];•“Operating Experience Insights into Pressure Boundary Component Reliability andIntegrity Management (RIM)” (NEA/CSNI/R(2017)3) [33];•“Operating Experience Insights into Below Ground Piping at Nuclear Power Plants”(NEA/CSNI/R(2018)2) [34];•“Basic Principles of Collecting and Evaluating Operating Experience Data on Metallic
Passive Components” (NEA/CSNI/R(2018)12) [35]; and•“A Review of the Post-1998 Experience with Thermal Fatigue in Heavy Water and Light
Water Reactor Piping Components” (NEA/CSNI/R(2019)13) [38].

During its April 2019 Working Group Meeting, the CODAP Management Board (MB) decided
to pursue an assessment of material degradation OPEX during extended or long-term operationof commercial nuclear power plants.

1.3 Technical approach
The results, conclusions and future activities that are documented in this report were obtainedusing a five-step approach (Figure 1-1):

22•

Step 1. The results from a questionnaire on LTO (see document NEA/CSNI/R(2015)13 [32])served as a basis for determining to what extent the CODAP event database can be used
to assess LTO trends in material performance. Prepared by the WGIAGE, this document
includes a qualitative ranking of material performance issues of concern, Table 1-1.
In preparing this report a comparison was made between the content of the CODAP
event database and the results of the WGIAGE questionnaire.

•

Step 2. A review of the results of two expert panels convened by the US Nuclear
Regulatory Commission on material degradation issues during LTO. The proactive
material degradation assessment (PMDA; 2004 to 2006) [17] identified the different
material degradation scenarios that could affect plant systems for up to 40 years ofoperation, and the expanded material degradation assessment (EMDA; 2012 to 2014)[27][28][29] extended the analysis time frame from 40 years to 80 years. In preparing thisreport a comparison was made between the CODAP event database and the results of
the PMDA and EMDA expert panels.

•

Step 3. In 2019, the US Nuclear Regulatory Commission organised an international
workshop on age-related degradation of reactor vessels and internals during which theparticipating organisations summarised their country/economy-specific state ofknowledge, ongoing research and operating experience [39]. The technical information
included in the workshop presentations and the workshop summary report was used toaugment the country/economy-specific overviews of PEO/LTO material degradation
issues in this report.

•

Step 4. Material degradation issues as recorded in the event database were reviewed.
This review included consideration of reactor vessel internals and safety-related(i.e. Safety Class 1 through 3) and non-safety-related piping system components. The
evaluation of possible age-dependent material degradation was done in a high-level
manner using simple “visual tests” of graphical plots.

•

Step 5. A synthesis of the PEO/LTO material degradation insights was formulated basedon the results of Steps 1 through 4.

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INTRODUCTION

Table 1-1: Results of WGIAGE questionnaire
on PEO/LTO [32]Country

Degradation mechanisms considered important to PEO/LTO

Ranking

CanadaCANDU Pressure Tube – fracture toughness reduction, delayed hydride cracking
and irradiation induced creep

High

CANDU Outlet Feeders in the Primary Coolant Circuit – FAC

High

SG Tubes and External Surfaces – Alloy 800 SCC, WastageHigh

Secondary Coolant Circuit – FAC

High

Valves, pumps and other components – various degradation mechanisms

High

CANDU Calandria vessel and its internals – various degradation mechanisms

MediumCzechia

Bolting – IASCC/SCC

Low

Ageing of reactor pressure vessels and reactor internals

High

FAC of high energy pipelines
Environmentally assisted fatigue of primary circuit metals
FinlandHigh
Medium

RPV radiation embrittlement, particularly in combination with hydrogen flaking
and high nickel content
Thermal ageing and SCC in RPV safe ends and other dissimilar welds
Thermals fatigue caused by stratification and thermal cycling in branchconnections (e.g. mixing tees)
Pitting and crevice corrosion near deposits, metallic contact points and stagnant
conditions

No ranking
provided

Boric acid corrosion in bolted closures of the primary circuit pressure boundary
SCC and IASCC in RPV internals
FranceIrradiation embrittlement of reactor pressure vessel (RPV)
Irradiation embrittlement of reactor vessel internals
Stress corrosion cracking of nickel alloys
Fatigue considering the environmental effects

High

Thermal ageing – e.g. RCS Hot Leg Steam Generator (S/G) inlet CASS elbows
External/internal corrosion of buried pipingJapan

Neutron irradiation embrittlement of reactor pressure vessel (RPV) and coreinternals

High

Irradiation assisted stress corrosion cracking (IASCC)High

Stress corrosion cracking mechanisms other than IASCC)High

Thermal ageing embrittlement of cast austenitic stainless steel (CASS)
components

MediumPipe wall thinning due to flow assisted degradation

MediumThe Netherlands

Thermal stratification leading to thermal fatigue in piping

MediumSlovak Republic

Internal/external corrosion of buried emergency service water piping

High

Switzerland

RPV embrittlement

High

United States

Ageing of reactor vessel internals

High

Thermal and neutron embrittlement of cast austenitic stainless steels (CASS)

Medium

Metal fatigue of components considering environmental effects

Medium

RPV embrittlement

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INTRODUCTION

Figure 1-1: The process for developing insights into long-term material degradation

At the time when the US NRC PMDA expert panel was convened, the average age of the UScommercial light water reactor (LWR) population was about 25 years (26.5 years for the BWR
plant population and 23.6 years for the PWR plant population). The objective of the PMDA expert
panel was to identify the potential key material degradation scenarios for the next 15 years ofoperation, up to the end of the original 40-year operating licence. At the time when the US NRC
EMDA expert panel was convened, the average age of the US LWR population was about 33 years.
The objective of the EMDA expert panel was to identify the potential key material degradation
scenarios for extended periods of operation, i.e. 40+20 years (the licence renewal domain) and60+20 years (the subsequent licence renewal domain).
Against this background, the review of the operating experience in the CODAP event
database was limited to events that occurred at plants 25 years old or older. The review
responded to the following questions:

1.

What was the passive component operating experience from 2004 to 2014 (the start dateof the PMDA expert panel and the conclusion of the EMDA expert panel)?

2.

Are the apparent and underlying causes of material degradation during PEO/LTO similar
to what was experienced during early plant life? If not, what are the main differences?

3.

Is it feasible to validate the WGIAGE questionnaire results [32] against the CODAP event
database? In other words, to what extent are any of the material degradation issues thatwere identified in the questionnaire also acknowledged and tracked in the CODAP
project?

4. Based on the results of the operating experience evaluations that are documented in thisreport, are there any gaps in the lists of degradation scenarios that were identified by thePMDA, EMDA or WGIAGE?

5. What, if any, procedural changes should be contemplated to ensure that the CODAP

project actively continues to track the long-term material performance for years to come?

24

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INTRODUCTION

There are well-known technical challenges in evaluating OPEX data. The quality of thedatabase insights that are obtained through systematic evaluations is strongly correlated withthe completeness of an OPEX database. From the point of view of PEO/LTO and possible trends
in the rate of material degradation, the “latency” in populating an OPEX database with failure
events presents the data analysts with another technical challenge. Since this report isconcerned with operating experience through the end of the calendar year 2019, and as further
explained in the next paragraph, there could potentially be latent gaps in the material
degradation knowledge base.
Reference [35] introduced the term “inherent latency” of structured data collection efforts.
Inherent latency refers to the amount of time required for OPEX to travel from the “originator”(i.e. nuclear power plant) to the data analyst, and this time could be considerable. The
referenced report made the contention that it takes about five years to ensure OPEX data
completeness. That is, five or more years could elapse before achieving high confidence in data
completeness for the previous five years of OPEX collection and evaluation. In other words,
around the years 2024 to 2025, and possibly later, the data mining for the previous five-year
interval (i.e. 2015-2019) would be expected to approach “saturation” (meaning high confidence
in the completeness of a database). Could the potential for “cliff-edge-effects” (i.e. small changes
in the OPEX data resulting in large results variation) influence the outcome of an analysis due
to database infrastructure factors? It depends on the maturity of inspection programmes andthe state of knowledge concerning certain degradation mechanisms, as well as on the“effectiveness” of the CODAP project in capturing possible new material degradation insights.
Considerations about the use of up-to-date failure data is intrinsically assumed to be factored
into any OPEX data analysis task.

1.4 Nomenclature
There are two regulatory approaches to extending the operating time beyond an original
licensing period. One approach is through a periodic safety review (PSR) and another approachis through licence renewal (LR). PSRs are performed to: (i) determine whether the plant complies
with its licensing basis; (ii) identify the extent to which the original licensing basis remains valid,in part, by determining the extent to which the plant meets current safety standards andpractices; (iii) provide a basis for implementing appropriate safety improvements, corrective
actions, or process improvements; and (iv) provide confidence that the plant can continue to be
operated safely.
The term “period of extended operation” (PEO) is used where an original operating licencemay be renewed for a predetermined period only. Approval of a PEO is preceded by a periodic
safety review and once the end of a PEO is reached the reactor is permanently shut down anddecommissioned. The term “long-term operation” (LTO) is used where an original operating
licence is extended through a formal licence renewal process. The United States has
implemented a LR process that proceeds along two tracks: one for review of safety issues andanother for environmental issues. The operating licence may be renewed multiple times inaccordance with specific regulatory requirements.
The PEO/LTO definitions of the CODAP project member countries and economies are
summarised in Table 1.2. As of mid-2021 the population of nuclear power plants in PEO/LTO(≥ 40 years of operation) in the CODAP project member countries and economies consisted of
81 reactor units (see Figure 1.2).

1.5 Report structureThis report consists of six chapters and three appendices. A summary of the PEO/LTO operating
experience is given in Chapter 2. Chapter 3 gives examples of the country/economy-specificmaterial degradation OPEX during PEO/LTO. Chapter 4 documents the results of the PEO/LTO
operating experience data evaluations. A report summary with conclusions and future activities
is presented in Chapter 5. A list of references is given in Chapter 6.

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INTRODUCTION

Annex A is a glossary of technical terms. Annex B is a summary of the US reporting
requirements and routines for passive component degradation and failure.

Table 1-2: Commercial nuclear power plant operating licence bases
in CODAP project member countries/economies 2Country/
economyCanadaCanadian CANDU plants were originally designed assuming a nominal 30-year operating life. The plants arelicensed to operate for periods of a maximum of ten years followed by periodic safety reviews (PSRs).
Extended operation beyond 30 years is possible if accepted by the Canadian Nuclear Safety Commission(CNSC) based upon the results of a PSR. Long-term operation (for example for up to an additional 30 yearsof operation) typically requires major refurbishment activities which includes the replacement of theoriginal fuel channels and feeder piping and, in some cases, steam generators.

Czechia

Initial design life is 30 years for main components, such as steam generators, reactor coolant pumps andpressurisers. The design life of the reactor pressure vessel is 40 years. The operating licence is unlimited, but
the national regulator requires a safety review of the nuclear power plants every ten years and canauthorise continued operation of a nuclear power plant for another ten years. Improvements required as aresult of the PSR process, including updates to the safety standards, of operating experience feedback andof lessons learnt, must be implemented.

Finland

Periodic safety reviews are performed every ten years. The Loviisa Units 1 and 2 (WWER-type reactors) were
initially designed for a 30-year operating life. The units were subsequently relicensed for an additional
20 years. The current operating licence ends in 2027 and 2030, respectively. Olkiluoto-1/2 (ASEA-Atom
BWRs) were designed for a 40-year operating life. They have been relicensed to operate until 2038. The key
issues in the licence renewal are ageing management, organisational issues, deterministic and probabilistic
safety analyses, environmental fatigue, status of the safety improvements as well as matters relating to theenvironment, nuclear waste and nuclear fuel.

France

The initial design life is 40 years. As the plant ages, the safety authority conducts a periodic safety review(PSR) every ten years to ensure that a plant continues to be in conformity with the original safety standards
and with any additional requirements. Improvements required as a result of the PSR process, of updates tothe safety standards, of operating feedback and of lessons learnt, must be implemented.

Germany

As a legal basis for the operation of the German nuclear installations, the Atomic Energy Act (Atomgesetz,
AtG) was amended in 2002 with the aim to phase out the use of nuclear energy for the commercial
generation of electricity in a controlled and structured manner. The Act laid down the electricity
production rights for each nuclear installation. With the amendment of the AtG on 6 August 2011,
continued operation of eight nuclear installations for electricity generation (power operation) was
terminated, whereas additional dates for the latest possible termination of power operation were fixed forthe remaining nine nuclear installations. The last operating units to be permanently shut down at the end
2022 are Emsland, Isar-2 and Neckarwestheim-2. LTO is no longer an issue in Germany.

2.

26

Basic principles

For a comprehensive review, see NEA (2019), Legal Frameworks for Long-Term Operation of Nuclear Power
Plants, OECD Publishing, Paris, www.oecd-nea.org/jcms/pl_1515. Of particular relevance are Chapter 2(“Review of Approaches to Long-Term Operation”) and Chapter 3 (“Country Reports”).

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INTRODUCTION

Table 1.2: Commercial nuclear power plant operating licence bases
in CODAP project member countries/economies (cont’d)Country/
economyBasic principles

Japan

Under the new regulatory requirements of 8 July 2013 based on the revised Reactor Regulation Act, theoperating life of nuclear power reactors is set at 40 years, in principle. This may be extended one time only
for another 20 years if approval is obtained before the expiration of the initial 40 years. With theintroduction of this system, a 27 February 2013 NRA Commission Meeting decided that before any
extension could be agreed, the current status of the relevant plant would be assessed to determine
potential further anticipated deterioration during the proposed extension period. A maintenance policy forthis extension period must be included in an application, with particular attention paid to preventive
measures against ageing. Regarding commercial power reactors in operation for 30 years or more, amaterial degradation assessment and a long-term maintenance policy covering every ten years are alreadyrequired, based on the Reactor Regulation Act and related legislation.

Korea

The operator of nuclear installations is required to perform a periodic safety review every ten years fromthe date of operating licence issuance and submit the results to the Nuclear Safety and Security
Commission (NSSC). In case an operator wants to operate a nuclear installation beyond the design life, two
additional items such as lifetime evaluation for major components and radiological environmental impact
assessment are to be incorporated into the periodic safety review.

The
Netherlands

The operating licence for the Borssele Nuclear Power Plant was issued in 1973 and it did not contain apredetermined expiration date. This means that as long as the requirements (as stated in the regulations andthe operating licence) are fulfilled, the plant is allowed to operate. In 2006 the Dutch government signed anagreement (the “Covenant”) with the owners of the Borssele Nuclear Power Plant, which allows for operation
until the end of 2033, at the latest. In the meantime the Covenant conditions should be met, in addition to
the requirements of the licence. The aforementioned end-date of operation is also a requirement in article
15a of the Dutch Nuclear Energy Law.

SlovakRepublic

Based on the 2013 revision of the Atomic Act the operating licence is not limited in time, but the licenceholder must comply with the provisions of law, i.e. every ten years demonstrate by a periodic safety review(PSR) the readiness of the facility for further operation. The regulatory authority ÚJD SR can complement theoperating licence with specific conditions.

Spain

The Spanish nuclear power plants are subject to a system of operating permit renewal for a given period.
Likewise, every ten years the plants carry out a PSR, updating the situation of the continuous safetyassessment programmes that are performed systematically and analysing the applicability of the changes
occurring in the standards over the ten-year period (Condition Application Standards, NAC). Facilities
requesting long-term operation must include in their PSR an integrated ageing assessment andmanagement plan, among other additional requirements.

Switzerland

In the aftermath of the 2011 Fukushima Daiichi accident, the Swiss government decided to phase out
nuclear energy. Existing plants will continue to operate as long as they are considered safe by ENSI and fulfil
all legal and regulatory requirements in this respect. Assessments of LTO have been performed for two Swiss
nuclear power plants which have been in commercial operation for over 40 years. A detailed examination
demonstrated that the conditions for taking out of service a nuclear power plant are not yet and will not bereached by Beznau Nuclear Power Plant and Mühleberg Nuclear Power Plant within the next ten years.
Nevertheless, it is mandatory to continue with the scheduled ageing management, maintenance andbackfitting activities. Mühleberg Nuclear Power Plant was permanently shut down at the end of 2019 after
47 years of operation.

ChineseTaipeiThe nuclear power plants in Chinese Taipei have an initial design life of 40 years. According to theregulations, a life-extension application must be filed 5 to 15 years before the expiration of the operating
licence (OL). However, in response to the Fukushima Daiichi accident, the government’s national energy
policy was revised in November 2011 with a provision of no life extension beyond 40 years for all nuclear
units. As a result of this policy, Chinshan-1 was permanently shut down in December 2018 and Chinshan-2
was permanently shut down in July 2019.

United States

The Atomic Energy Act authorises the Nuclear Regulatory Commission to issue licences for commercial
power reactors to operate for up to 40 years. These licences can be renewed for an additional 20 years at atime. The period after the initial licensing term is known as the period of extended operation. Economic andantitrust considerations, not limitations of nuclear technology, determined the original 40-year term forreactor licences. As of the end of 2020, 49 reactor units were in an extended period of operation, operating
beyond 40 years.

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INTRODUCTION

Figure 1.2: Number of commercial reactor units in long-term operationin CODAP project member countries/economies

No. of nuclear power plants in commercial operation

70
BWR

60

60

CANDU

PWR incl. WWER

50

40
34
30

19

20

18

16

13

11
8

10
4

2

7

7
4

7

6
3

2

40-44

45-50

0
25-29

30-34

35-39

50-54

Age of reactor as of 31 December 2023

28

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Chapter 2. Temporal changes in material degradation

This chapter presents high-level summaries of the material degradation mechanisms that havebeen observed to act on metallic piping systems as well as certain non-piping passive
components. These degradation mechanisms are for the most part well understood and in somecases they have produced failures with significant operational impacts. Using the results fromroot cause evaluations and materials research, detailed ageing management processes havebeen implemented to prevent recurring failures or to mitigate the effects of active degradation
mechanisms. Systematic OPEX data evaluations provide insights into the effectiveness ofdegradation mechanism mitigation processes and monitoring programmes.

2.1 The CODAP-OPEX knowledge base
Extensive material performance information exists, and it covers field experience obtained from
more than five decades of commercial nuclear power plant operation. Organised by type ofmaterial, degradation mechanism and time period, the field experience is summarised inTables 2-1 through 2-3 and Figure 2-1. For non-piping passive components, corresponding toabout 10% of the total event population in CODAP, Figure 2-2 summarises the field experienceby component type and the time period in which a failure was observed. Only selected
representative events are included in CODAP. The four predominant types of metallic materials
in current use are:•Carbon steels are used extensively in primary- 1 and secondary-side piping systems. The
OPEX is extensive. For piping in a raw water operating environment there is an upwards
failure trend indicative of ageing management challenges.

•

Low-alloy steels are used extensively in balance-of-plant systems and as a replacement
for carbon steels in wet steam operating environments. Hundreds of failures have beenreported for low-alloy steels.

•

Nickel-base alloys are used extensively in primary systems of light water reactors inconnections between ferritic components and austenitic piping. The OPEX consists ofwell over 500 failures. Since the 1990s a downward trend is noted.

•

Stainless steels are used extensively in primary- and secondary-side piping systems. The
OPEX is extensive and with a clearly discernible and well understood downward failure
trend since the 1990s. There are multiple grades of stainless steels. Austenitic stainlesssteels are the most common.

The degradation mechanisms acting on the above listed material types are well understoodand managed through different types of reliability and integrity management (RIM)
programmes; reference [33]. For process piping systems the OPEX is organised by degradation
mechanisms; Figure 2-3. Summarised in Figures 2-4 through 2-9 are “selected OPEX excerpts”
from CODAP. These summaries do not account for the plant population that produced thefailure data in a given time period. A formalised assessment of ageing trends is beyond the scopeof this Topical Report.

1.

As an example, carbon steels are used for the CANDU primary heat transport system (HTS) feeder piping.

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Table 2-1: Effect of ageing management on material performancein SCC- and fatigue-susceptible environments
Degradation
mechanism (DM)ECSCCStress corrosion cracking

TGSCC

IGSCC

Steps taken to manage/
mitigate degradation(Post-1980s)

Ageing management
effect on OPEX (plant age> 25 years)

Multiple

Improved fabrication
practices, cleanliness, pipe
surface decontamination

Inconclusive (i.e. notpossible to make clear
distinction between the“early-life OPEX” vs.
“PEO/LTO OPEX”)

BWR primarywater (HWC,
NMCA)

Improved water chemistry
control, application of stress
improvement process(e.g. MSIP, peening), use of
SCC-resistant material,
application of FSWOL,
minimise number of welds)

Stabilised
austenitic SS
Alloy 82/182

HCF (smalldiameter lines)

LCF

Fatigue

Stainless steel(SS) – multiple
gradesOperatingenvironment(s)Unstabilised
austenitic SS

PWSCC

Thermalfatigue

Corrosion
fatigue/environmental
degradation

30

Material(s)Significant reduction noted
in SCC incident rates

Hightemperature
PWR primarywater

Use of SCC-resistant material(e.g. Alloy 690/52/152),
application of stress
improvement process,application of FSWOL

Multiple

Elimination of socket welds,
fatigue monitoring,
improved piping
configuration, installation ofvibration dampers

Inconclusive

Multiple

Multiple

Improved piping
configuration (e.g. added
supports), enhanced welding
technology

Inconclusive

Multiple

Hot/cold fluid
mixingFatigue monitoring,
improved NDE technology,
improved system operatingprocedures, piping systemre-configuration"Somewhat" inconclusive –to be further investigated,
for more details see forexample Reference [38]Multiple

Hightemperature/high-pressure
primary/
secondary sidesystemenvironment

Reconfiguration of piping(e.g. PWR feedwater nozzle
configuration). Fatigue
monitoring, improved plant
operating procedures

Inconclusive (only limited
OPEX exists, most of which
are from the 1970s to early
1980s). Relationshipbetween field experienceand experimental andtheoretical work yet to be
determined

Alloy
600/82/182

Multiple

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Table 2-2: Effect of ageing management on material performancein FAD-susceptible and corrosive environments

Flow assisted degradation (FAD)

Degradation
mechanism (DM)Material(s)Operatingenvironment(s)Steps taken to manage/
mitigate degradation

Ageing management
effect on OPEX
(plant age > 25 years)
Inconclusive – a function of
the effectiveness ofknowledge transfer

ErosionCavitation

Multiple

Multiple

Piping systemreconfiguration, improved
system operatingprocedures

ErosionCorrosion (E/C)

CS

MultipleUse of E/C-resistantmaterials

Inconclusive

Use of FAC-resistantmaterial, implementationof FAC monitoring
programme, improved
secondary-side waterchemistry

Significant reduction noted
for FAC-induced leaks andruptures

LAS, SS

Reconfiguration of piping,
improved operatingprocedures/operational
strategies (e.g. baseload vs.
load following)

Inconclusive; therelationships between
extended power uprate and
LDIE-susceptibility to be
determined

Al-Bronze,
CS, SS(e.g. 300Series)

Chemical treatment ofwater, use of corrosion
resistant material incl. nonmetallic material,reconfiguration of piping,
application of composite
repair technology1,
installation of cathodic
protection system, etc.

Inconclusive – however,high-alloy austenitic
stainless steels(e.g. AL-6XN®, 254-SMO®
and 654-SMO®) and HDPE
materials appear to perform
well in corrosive
environments

Flow-assisted
corrosion

CS, LAS
High-energy
piping, single or
two-phase flow
conditions

Liquid droplet
impingement
erosionCorrosion

Corrosion –GeneralCrevicecorrosion
De-alloying –Selective
leaching
Galvanic
corrosion

Raw water(e.g. brackish, lake,
pond, river, sea)

Microbial
corrosion
1. For more information, see for example https://adams.nrc.gov/wba (Accession No. ML20014E476; “30 to 96-inch Pipelines Upgrade
Case Study: Navigating Safety Related Pipeline Upgrades with CFRP” & ML20014E506, “Carbon Fiber Reinforced Plastic (CFRP) Repair –Future Applications and R&D Gaps”).

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Table 2-3: Degradation mechanisms by the time period in which a failure was discovered
OPEX summary by time period (No. of failures)

Degradation mechanism

1971-2005

2006-10

2011-152016-20

83

9

3

2

ECSCC – PWR – Piping

98

43

17

7

IGSCC – BWR – Piping

1 380

43

13

18

IGSCC – PWR – Piping

119

8

22

11

PWSCC – PWR – Piping
PWSCC – PWR – Vessel head penetration

92
119

47
10

6
29

14
13

TGSCC – BWR – Piping

130

14

0

0

TGSCC – PWR – Piping

34

4

7

2

HCF – BWR – Socket weld

144

9

12

9

HCF – PWR – Socket weld

457

40

25

25

HCF – All plant types and all piping systems

1 550

150

168

146

LCF – All plant types and all piping system

161

30

42

32

TF – BWR

73

5

5

3

TF – PWR

117

6

15

20

Corrosion – FP system

144

64

91

50

Corrosion – BWR – SW system

152

91

135

141

Corrosion – PWR – SW System

683

178

292

280

MIC – All plant types and all systems

426

169

371

415

Flow-assisted
degradation

Corrosion

FatigueStress corrosion cracking

ECSCC – BWR – Piping

FAC – BWR – All systems

422

69

36

20

FAC – PWR – All systems

1 305

159

67

29

Erosion-Cavitation – All plant types and all systems

124

41

33

39

Erosion-Corrosion - All plant types and all systems

314

86

60

60

Figure 2-1: The piping operating experience by material type and time period
10 000No. of material failures [Log10]

1 000

100

10

1
Al-bronze

Carbon steel
1970-7932

Copper1980-89

Low alloy steel Ni base alloys Stainless steel
1990-99

2000-09"Other"2010-19

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

RPV-VHP

Figure 2-2: The non-piping passive component OPEX by material type and time period
2010-19
2000-09
1990-99
1980-89

Ni-base alloy

Valve body

CASS

Cast ironAl-bronze

Pump casing

CASS

Ni-base alloy

Baffle former
BMI-nozzle
bolt

PWR CEDM
thermal Core shroud
sleeveweld
Cont. liner

PZR heater
sleeveCast ironNi-base alloy

Carbon steel

Stainless steel

Stainless steel

Stainless steel

Ni-base alloy

0

10

20

30

40

50

60

70

80

No. of failures

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-3: Piping material degradation mechanisms observed during five decadesof commercial nuclear power plant operation experience

SCC

TGSCC
SICC
PWSCC
IGSCC

Flow-assisted degradation

LDIE - Liquid droplet impingement erosion
Jet impingement erosion
Flow - Induced vibration - FIV
FAC - Entrance effect
Flow - Accelerated corrosion (FAC)
Erosion - Corrosion
Erosion - Cavitation
Erosion
Thermal fatigue - Stratification

Fatigue

Thermal fatigue - Cycling
Low-cycle fatigue
High-cycle fatigue
Corrosion - Fatigue
Microbiologically influenced corrosion
Galvanic corrosion

Corrosion

Dealloying - Selective leaching
Corrosion - Pitting
Corrosion - External - Corrosion under insulation
Corrosion - ExternalCorrosion - Crevice
Corrosion - General
0.1%

1.0%

10.0%

100.0%

Percent of total failure population [Log10]

34

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-4: Piping OPEX by system group and age of plant at time of failure
350

25-29 Years
30-34 Years

300

35-39 Years
No. of pipe failures250

40-44 Years
45-49 Years

200
150
100
50
0

Figure 2-5: Piping OPEX by system group and for 0 ≤ T < 25 reactor operating years

No. of pipe failures [Log10]

1 000

100

10

1

0-4 years

5-9 years

10-14 years

15-19 years

20-24 years

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-6: Piping OPEX by safety class and plant age
700
600

No. of pipe failures

500
400
300
200
100
0
25-29

30-34

35-39

40-44

45-49

Age of failed piping component at time of failure
ASME III Class 1

ASME III Class 2

ASME III Class 3

Non-safety

Fire water system

Figure 2-7: The PWSCC operating experience
7060

No. of PWSCC failures

50

40

30

20

10

0
1986-90

1991-95
Piping - DMW

36

1996-2000
RPV-VHP

2001-05
BMI-nozzle

2006-10

2011-152016-20PZR heater sleeve penetration

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-8: Reactor internals OPEX summary by plant type,component type and time period
ICI-tube
Fuel element alignment pin
PWR

CR cluster guide tube support pin
Core barrel support assemblyCore barrel
Bolting
Baffle bolt
Shroud support access hole cover weld
Shroud head hold-down bolt
RPV-internals pipe weld
Pipe
Jet pump support brace
Jet pump hold-down beam
BWR

In-core monitor housing
ICI-tube
CRDM connection
Core spray sparger
Core shroud weldCore shroud tie rod
Core shroud support
Control rod
0

5

10

15

20

25

30

35

40

No. of component failures2010-19

2000-09

1990-99

1980-89

1970-79

Figure 2-9: PWSCC in PWR reactor pressure vessel head penetrations (VHP)
80
70
60

No. of VHP failures

50
40
30
20
10
0
1990-94

1995-99

2000-04

2005-092010-142015-192020-24

Period in which a vessel head penetration failure was discovered
Crack - Part

Crack - Full

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

2.2 Degradation mechanisms acting on process piping
This section includes tabular and graphical summaries of the piping degradation and failure
OPEX. Represented in CODAP are multiple nuclear steam supply (NSSS) designs as well as
multiple design generations; Figure 2-10. Indicated in Figure 2-11 are the respective CODAP
member country/economy contributions to the OPEX database.
About 70% of the event population originates in the United States. This is not only due tothe relative number of reactors in the United States, but also fundamental differences in event
reporting requirements to the regulatory authorities among the member countries/economies.
The disproportionality is also attributed to the origin of the event database, the data collection
process as it was originally implemented, national regulations, and the ability to exchange
restricted data on material performance and failures.
The piping OPEX data are organised by the time-period in which a failure was observed, as
well as by the degradation mechanism that caused or contributed to a failure. The following
summaries are provided:

38•

Table 2-4. Summary of the piping operating experience by time period in which a failure
was observed, safety class and outside pipe diameter (OD) in mm. The fire water system
OPEX in this chart comes from Finnish [11][37] and US nuclear power plants.

•

Figure 2-12. Small-diameter piping OPEX. The predominant failure mechanisms arecrack initiation and propagation via fretting or high-cycle fatigue (HCF). In this summary
the term “small-diameter” refers to piping with outside diameter  ≤ 60.3 mm (≤ NPS2).
The chart includes butt welded and socket welded piping components.

•

Figure 2-13. This chart shows the small-diameter socket weld OPEX sub-population(987 failures).

•

Figure 2-14. Medium-diameter piping OPEX. In this summary the term “mediumdiameter” refers to piping with outside diameter in the range of 60.3 mm <  ≤ 273.1 mm(2 < NPS ≤ 10).

•Figure 2-15. Large-diameter piping OPEX  > 273.1 mm (> NPS10).

•

Figure 2-16. Summary of the OPEX involving the four different fatigue mechanisms:
1) corrosion fatigue, 2) high-cycle fatigue, 3) low-cycle fatigue, and 4) thermal fatigue.

•

Figure 2-17. Summary of the OPEX involving stress corrosion cracking (SCC).

•

Figure 2-18. Summary of the WWER-440/213 OPEX as recorded in CODAP. OPEX has beenrecorded for Bohunice-3/4, Dukovany-1/2/3/4, Loviisa-1/2, Mochovce-1/2 and Temelín-1/2.

•

Figure 2-19. Summary of the OPEX involving flow-assisted degradation; 1) erosioncavitation, 2) erosion-corrosion, 3) flow-accelerated corrosion (FAC), and 4) liquid droplet
impingement erosion (LDIE).

•

Figure 2-20. Summary of the OPEX involving corrosion failures, mainly in service water(SW) and fire protection (FP) water systems.

•

Figure 2-21. Summary of the OPEX involving service water system pipe failures; safetyrelated and non-safety-related piping.

•

Figure 2-22. Summary of the OPEX involving Fire Water system pipe failures. In this chart
the term “Corrosion – Various mechanisms” includes “general corrosion, pitting, crevice
corrosion, galvanic corrosion”.

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-10: The CODAP OPEX by NSSS design

CANDU/AECL - 2.8%
ASEA-Atom - 5.2%
Babcock & Wilcox/BBR - 4.5%
Combustion Engineering - 10.4%
Framatome - 3.2%
GE/GE-Hitachi/GE-Toshiba - 31.1%
KEPCO - 0.3%
Siemens/KWU - 6.6%
SKODA - 1%
WH/WH-MHI - 34.9%

Figure 2-11: The piping and non-piping passive component OPEX
by CODAP member country/economy

Netherlands - 0.1%
Slovak Republic - 0.3%
Chinese Taipei - 0.6%
Czechia - 0.7%
Finland - 1.3%
Spain - 1.5%
Korea - 1.8%
Switzerland - 2.4%
France - 3.7%
Canada - 4.9%
Japan - 6.2%
Germany - 8.1%
United States - 68.5%

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39

TEMPORAL CHANGES IN MATERIAL DEGRADATION

Table 2-4: Piping operating experience by time period,
safety class and outside diameter
Safetyclass

1

Outsidediameter [mm]Number of failures per time period
1970-74 1975-79 1980-84 1985-89 1990-94 1995-99 2000-04 2005-09 2010-14 2015-19∅ ≤ 21.3

2

14

27

14

9

25

18

22

13

8

21.3 < ∅ ≤ 60.3

18

44

54

77

60

78

67

64

47

60

60.3 < ∅ ≤ 114.3

18

20

31

13

24

17

22

35

11

11

114.3 < ∅ ≤ 273.1

2

13

91

40

101

47

9

7

19

348

223

55

151

123

65

17

10

∅ > 273.1

2

3∅ ≤ 21.3

2

10

24

13

21

20

15

29

6

4

21.3 < ∅ ≤ 60.3

29

131

131

130

161

125

117

76

67

40

60.3 < ∅ ≤ 114.3

6

33

31

143

27

23

16

17

27

10

114.3 < ∅ ≤ 273.1

7

43

113

53

55

44

32

24

36

9

∅ > 273.1

4

22

47

57

40

38

45

40

18

9

∅ ≤ 21.3

3

11

27

21

9

18

22

14

25

20

21.3 < ∅ ≤ 60.3

16

77

105

82

97

157

143

149

200

116

60.3 < ∅ ≤ 114.3

3

16

37

16

55

67

83

83

117

80

114.3 < ∅ ≤ 273.1

5

44

14

20

78

77

79

108

124

124

∅ > 273.1

2

9

19

83

89

75

127

124

173

158

9

14

17

11

11

9

10

6

2

∅ ≤ 21.3
4–
Balanceof plant

Fire
water

21.3 < ∅ ≤ 60.3

20

58

84

136

131

119

122

94

55

39

60.3 < ∅ ≤ 114.3

9

14

19

42

39

33

48

66

32

19

114.3 < ∅ ≤ 273.1

7

20

144

324

49

46

74

81

38

21

∅ > 273.1

7

14

123

331

46

73

77

86

41

13

∅ ≤ 21.3

1

21.3 < ∅ ≤ 60.3

4

4

7

6

2

11

18

9

60.3 < ∅ ≤ 114.3

5

1

4

10

7

8

23

7

1

114.3 < ∅ ≤ 273.1

1

5

8

9

20

17

31

42

27

∅ > 273.1

1

9

2

4

11

25

20

18

14

623

1 507

1 850

1 182

1 291

1 299

1 264

1 154

814

Total no. of pipe failures:

40

4

160

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TEMPORAL CHANGES IN MATERIAL DEGRADATIONASME III Class 1

ASME III Class 2

ASME III Class 3

Non-safety

Figure 2-12: Small-diameter pipe failure OPEX by time period,
safety class and outside diameter

21.3 < OD ≤ 60.3

2015-192010-14≤ 21.3

2005-09
2000-04

21.3 < OD ≤ 60.3

≤ 21.321.3 < OD ≤ 60.3

≤ 21.321.3 < OD ≤ 60.3

≤ 21.3
0

50

100

150

200

250

No. of pipe failures

Figure 2-13: Socket weld failure OPEX data by time period
200
BWR - 316
failures180

No. of socket weld failures

160
PWR - 957
failures

140
120
100
80
60
40
20
0
1970-74 1975-79 1980-84 1985-89 1990-94 1995-99 2000-04 2005-09 2010-14 2015-19
Time period in which socket weld failed

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41

TEMPORAL CHANGES IN MATERIAL DEGRADATIONASME III Class 1

ASME III Class 2

ASME III Class 3

Non-safety

Figure 2-14: Medium-diameter pipe operating experience by time period,
safety class and outside diameter

114.3 < OD ≤ 273.1

2015-192010-1460.3 < OD ≤ 114.3

2005-09
2000-04

114.3 < OD ≤ 273.1

60.3 < OD ≤ 114.3

114.3 < OD ≤ 273.1

60.3 < OD ≤ 114.3

114.3 < OD ≤ 273.1

60.3 < OD ≤ 114.3
0

20

40

60

80

100

120

140

No. of pipe failures by outside diameter (OD)

Figure 2-15: Large-diameter (OD > 273.1 mm) pipe OPEX by time period and safety class17
2015-19168

910Time period

41
2010-14

174

18
17

Safety Class 4
Safety Class 3

86
2005-09

40

124

65

Safety Class 1
77

2000-04

127

45
0

20

40

Safety Class 2

123
60

80

100

120

140

160

180

200

No. of pipe failures

42

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-16: Piping fatigue operating experience300

Corrosion
fatigue
High-cycle
fatigue

250

No. of fatigue failuresLow-cycle
fatigue
200

Thermalfatigue

150

100

50

0
1970-74 1975-79 1980-84 1985-89 1990-94 1995-99 2000-04 2005-09 2010-14 2015-19

Figure 2-17: Piping operating experience involving stress corrosion cracking
200
CISCC

180

ECSCC
IGSCC

160

PWSCC

No. of SCC failures

140

TGSCC

120
100
80
60
40
20
0
1990-94

1995-99

2000-04

2005-092010-142015-19

Time period in which failure occurred

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-18: The VVER-440/213 passive component OPEX in CODAP organised by safety class
45
40
35

No. of pipe failures

30
25
20
15
10
5
0
1980-89

1990-99

2000-092010-19

Period in which a pipe failure occurredClass 1

Class 2

Class 3

Non-safety

Fire water system

Figure 2-19: Piping operating experience involving flow-assisted degradation mechanisms

No. of failures attributed to flow-assisted degradation

300

250

200

150

100

50

0
1990-94

1995-99

2000-04

2005-092010-142015-19

Time periods in which failure occurred
Erosion-cavitation

44

Erosion-corrosion

FAC - Carbon steels

LDIE - Low-alloy or stainless steels

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-20: Piping operating experience involving corrosion failures
MIC
Graphitic corrosion

Corrosion Mechanisms

Galvanic corrosion
Dealloying - Selective leaching
Corrosion - Pitting
Corrosion - External - Below ground piping
Corrosion - Crevice
Corrosion - General
B/A-Corrosion
1

10

100

1 000

No. of corrosion failures2010-19

2000-09

1990-99

Figure 2-21: Service water piping failure operating experience

100

10

MIC

Graphitic corrosion

Galvanic corrosion

Erosion-corrosion

Erosion-cavitation

Dealloying - Selective leaching

Corrosion-pitting

Corrosion-externalCorrosion-crevice

1
Corrosion

No. of corrosion and FAD failures[Loh10]

1 000

Service water system
1970-79

1980-89

1990-99

2000-092010-19

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TEMPORAL CHANGES IN MATERIAL DEGRADATION

Figure 2-22: Fire water piping operating experience160

No. of fire water pipe failures

140
120
100
80
60
40
20
0
1970-79

1980-89

1990-99

2000-092010-19

Period in which a pipe failure occurred
Corrosion mechanisms other than MIC

MIC

FAD

FIV

Severe overloading / water hammer

2.3 Degradation mechanisms acting on cast austenitic stainless steel
Cast austenitic stainless steels (CASS) used in pump casings, valve bodies, pipe elbows, andother components in reactor coolant systems of HWRs and LWRs can suffer a loss in fracture
toughness due to thermal ageing embrittlement (TAE) after many years of service attemperatures in the range of 280-320°C. Thermal ageing of cast stainless steels at thesetemperatures causes an increase in hardness and tensile strength and a decrease in ductility,
impact strength and fracture toughness of the material. There are no failures (as in throughwall leaks) that are attributed to TAE, however.
The CODAP event database includes 41 failures involving CASS valve bodies. Included inthis event population is the discovery of non-surface penetrating intergranular stress corrosion
cracks in all eight DN600 CASS valves in the Reactor Recirculation System of Oskarshamn Unit 1(ASEA-Atom BWR) during the 1995 refuelling outage [10]. At the time of this discovery thereactor had been in commercial operation for 23 years 2. The CASS material was of Type EN
1.4552, which corresponds to ASTM A-351-CF8C. The longest measured IGSCC crack was
145 mm and the deepest crack was 14.5 mm. All flaws were removed using the boat samplingtechnique.

2.4 Degradation mechanisms acting on reactor vessel internals
Austenitic stainless steels (SS) are used extensively as structural alloys in the internal componentsof LWR pressure vessels because of their relatively high strength, ductility, corrosion resistanceand fracture toughness. Intergranular stress corrosion cracking (IGSCC) in austenitic piping was amajor issue for BWRs in the 1970s and 1980s. The susceptibility of reactor vessel internals (RVI) to
IGSCC was also recognised. Shroud cracking in 1993-1994 confirmed that IGSCC of RVI is asignificant issue for BWRs. The susceptibility of materials to IGSCC is enhanced by the effect ofradiation. Neutron irradiation increases the IGSCC susceptibility of austenitic stainless steels bychanging the material microstructure and microchemistry. This effect is referred to as irradiation-2.

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assisted stress corrosion cracking (IASCC). IASCC was not considered in the design of LWR coreinternal components, but it has been considered in addressing plant ageing and licence renewal
issues in the United States. Microstructural changes in austenitic SS due to neutron irradiationvary with the irradiation temperature, neutron fluence, flux and energy spectrum. Neutron
irradiation can decrease the fracture toughness of austenitic SS significantly. Fracture toughness
data on austenitic SSs irradiated in LWRs indicate:•

little or no loss of fracture toughness below ~0.5 dpa (displacement per atom); 3•

a substantial and rapid decrease at exposures of 1-5 dpa

In BWRs, in-service failures have been seen in the core shroud, jet pump assembly, top guideand core plate. Failures of baffle-former bolts have been observed in PWRs. In the United States,
BWR plant owners formed the BWRVIP (BWR Vessel and Internals Programme) in 1994, andsubsequently a similar initiative was established by the US PWR plant owners. The two
programmes issue annual inspection results summaries that are available on the US NRC website.

2.5 “Uncommon” and “unexpected” material degradation scenarios
In this section the terms “uncommon” and “unexpected” are used to characterise events thathave received attention by regulators, industry and/or material scientists because of theoperational impact, the modes of degradation and failure, and/or because of the potential effecton reliability and integrity management (RIM) practices. For illustrative purposes, seven eventsare described in further detail below. These events were selected for two reasons: (1) theyoccurred after more than 25 years of operation; and (2) they highlight possible ageing
management programme challenges.
1. December 2001 reactor pressure vessel head cooling pipe rupture at a German BWR plant;
CODAP Record #2747. At the time of the event the reactor had been in commercial
operation for 25 years. On 14 December 2001, control room staff received indications ofdrywell leakage. These indications were initially interpreted as a flange leakage. Further
evaluation of the control room indications resulted in the determination that the leakage
was from a non-safety related system. It was later determined that the affected system
was the RPV head cooling system (TC), a system only used during reactor shutdown. The
apparent leakage was located on an isolable section of DN100 stainless steel TC piping.
On 21 February 2002, an investigation team was chartered to evaluate the event. During
the initial drywell inspection on 18 February 2002, it was discovered that a large section(length of two to three metres) of DN100 stainless steel piping had ruptured. On the orderof 25 piping fragments could be located at the time. The initial hypothesis regarding thecause of rupture pointed to deflagration (hydrogen explosion) due to a local, highconcentration of hydrogen in the pipe. Possible issues relating to operating practices
were not considered in the root cause analysis.
2. In December 2014 multiple pinhole size leaks were discovered in stainless steel service
water supply piping to emergency diesel generator coolers at a Finnish PWR plant;
CODAP Record #5102. After about 35 years, the original service water piping had been
replaced with corrosion resistant material during the 2014 refuelling and maintenanceoutage. The replacement material was a high-alloy austenitic stainless steel especiallydeveloped for corrosive operating environments; stainless steel of Type EN 1.4547 (UNS
S31 254). The leak locations were at or adjacent to butt welds. Except for periods ofemergency diesel generator operation and surveillance testing, the raw water in theaffected piping is stagnant.
3. During the March 2018 refuelling outage at a US PWR, crack indications were observed on
the core shroud and the outside diameter of the core support barrel during MRP-227-A
inspections [24] that were conducted during the outage; CODAP Record #5079. At the time

3. 1 dpa (displacement per atom) is equated to a fast fluence of 6.7 x 1020 n/cm2.

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TEMPORAL CHANGES IN MATERIAL DEGRADATIONof the event, the reactor had been in commercial operation for 42 years. Cracks were foundon the OD surface of the core barrel in the beltline elevation using the enhanced visual
testing (EVT-1) method. The Owner informed industry and the NRC during the outage
about this first-of-a-kind EVT-1 inspection finding. Additional EVT-1 and UT inspections
were performed to further characterise the indications. One crack-like indication found inbase-metal adjacent to the middle-girth weld, 45 crack-like indications found in basemetal adjacent to the middle-axial weld. None of these indications were through-wall,
however.
4. In January 2019 through-wall leaks were discovered in the stainless steel large-diameterturbine cross-under piping at a Spanish PWR plant; CODAP Record #5061. At the time ofthe event, the reactor had been in commercial operation for 30 years. During a refuellingand maintenance outage in 1988 and in an effort to mitigate material degradation
through flow-accelerated corrosion the original carbon steel piping was replaced withnew piping made of stainless steel material (Type EN 1.4541). The turbine cross-underpiping is the large-diameter piping (e.g. 900-1 500 mm diameter) that carries wet steam
from the high-pressure turbine to the moisture separator reheater (MSR) and dry(-er)
steam from the MSR to the low-pressure turbine.

5. March 2019 brittle fracture of a DN25 pipe coupling in the reactor vessel level indication

system (RVLIS) at a US BWR. At the time of the event, the reactor had been in commercial
operation for 43 years. On 28 March 2019, while operating at 100% reactor power, thenarrow range reactor water level instrument (1-C32-LI-R606B) failed high; CODAP Record#5083. It is an instrument tap off the steam space of the reactor vessel. Drywell pressureand drywell floor drain leakage increased. Investigations inside the containment
determined that a DN25 coupling on line 1-B21-774 located on the steam side of a reactor
level condensing chamber experienced a 360° circumferential separation at theapproximate centre of the coupling [40]. This failure resulted in a primary system leakrate of approximately 0.5 kg/s. The post-event metallurgical report determined that the
coupling showed no evidence of localised plastic deformation. The coupling experiencedhydrogen embrittlement and did not exhibit a leak-before-break failure mechanism. The
coupling was made of a shape memory alloy material composed primarily of NickelTitanium-Iron (Tinel material). Examination of the failed coupling was conducted at a
metallurgical laboratory. Microhardness testing, visual microscopy, and scanningelectron microscopy were used to characterise the failed material. The examinations
confirmed that the failure was caused by hydrogen embrittlement. This was supportedby the transgranular cleavage on the fracture surface, high hardness values in the region
exposed to the process fluid, and a hydrogen rich environment, which are all consistent
with hydrogen embrittlement. The root cause of this event was that the selection of Tinel
was inappropriate for long-term application in a high temperature process that contains
elevated levels of hydrogen.

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Chapter 3. Country/economy-specific overviews of material degradation
state-of-knowledge, research and operating experience

This chapter summarises selected country/economy-specific material degradation issues withrespect to plant ageing management. For countries/economies with a fixed operating time limit(i.e. 40 years) the OPEX review addresses material degradation issues that have been observed atplants that have been in commercial operation for 25 years or longer. For countries/economies
without a predetermined operating licence time limit, the OPEX review addresses material
degradation issues that have been observed since a plant has been in commercial operation for
25 years or more, as well as degradation issues during the extended period of operation beyond
40 years.

3.1 Introduction
The country/economy-specific information on material degradation issues during PEO/LTO
addresses three areas: 1) overview of the current state-of-knowledge, 2) ongoing research, and3) operating experience during LTO and solutions. The specific topics that are being addressed
include but are not limited to:•State-of-knowledge. Degradation issues and ageing management.
–

Description of the degradation mechanisms being managed during the period ofextended operations.

–

Description of the regulatory framework used to manage degradation/ageing during
the PEO/LTO. Specifically, the programme or guidance that forms the basis of therespective ageing management programme.

–

Description of the most significant technical issues related to the PEO/LTO.
–

1.

Plans, if any, to update the regulatory guidance to address any of these ageing
management issues.

–

Summaries of, or references to, any operating experience, research programmes orageing management programmes related to irradiation-induced void swelling or creep.

–

With respect to reactor internals and, if applicable, description of the use ofembrittlement trend curves (ETC) 1:–

Which trend curve is used and when was it implemented in the country/
economy’s regulations?

–

Are the ETCs used for predictive purposes or reliance placed on surveillance
results and with ETCs used for interpolating between surveillance results?

RPV & reactor internals material embrittlement trend curves (ETCs) are used to estimate the magnitudeof neutron irradiation embrittlement as a function of both exposure (fluence, flux, temperature) andmaterial composition variables.

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COUNTRY/ECONOMY-SPECIFIC OVERVIEWS OF MATERIAL DEGRADATION STATE-OF-KNOWLEDGE, RESEARCH AND OPERATING EXPERIENCE•

Ongoing research–•

Summarise the planned and ongoing research addressing degradation issues being
managed during the PEO/LTO.

Operating experience during the PEO/LTO and solutions–

Discussions on select operating experience from plant operation during the PEO/LTO.

3.2 Canada
In Canada, there are several material degradation issues that must be managed for safe LTO ofCANDU nuclear power plants. Many of these issues arose towards the end of the original 30-year
operating life of the CANDU reactors. Some examples are discussed below.

3.2.1 De-alloying of Monel steam generator tubes
The Pickering Nuclear Generating Station is comprised of six 515 MWe CANDU reactors, eachwith 12 steam generators using tubes made of Monel 400 SB-163/N04 400 material with nominal
outside diameter (OD) of 12.60 mm and 1.25 mm nominal wall thickness. The seamless tubes
were supplied in the annealed condition. These are recirculating steam generators with an
integral preheater above the cold-leg tube sheet, and the steam generators have eight carbonsteel (A245B) lattice-bar type of tube-to-bundle supports.
During the 2016 Pickering Unit 4 steam generator inspection outage a new outside surface,volumetric degradation mechanism (subsequently characterised as de-alloying degradation)
was detected by CTR-1 and X-Probe electromagnetic probes above the top preheater baffle plate
P19 in the cold-leg side of two steam generators. Based upon eddy current sludge mapping,these indications were located within the magnetite sludge pile on top of the upper preheaterbaffle plate; the height of the sludge pile extends for approximately 15 cm and it appears to beunchanged since 2001. The de-alloying features were localised and characterised as volumetricwall loss, but a few had depths of about 80% of the wall thickness, indicating they had initiatedand grown relatively rapidly since the degradation had not been detected in previous inspection
outages. The following activities were undertaken by the operator to characterise thedegradation and develop a management strategy [42]:•inspection scope expansion;•

examination of removed tubes;•

burst testing of removed tubes and fabricated specimens with simulated flaws;•

development of a flaw model and acceptance standards, that include:–

growth rate predictions;–tube plugging limits.

3.2.2 Steam generator tube support plate degradation
The Bruce Nuclear Generating Station is comprised of eight CANDU units providing
approximately 6 300 MWe in total. Each unit has eight steam generators (SG) and fourpreheaters (PH) configured into two circulating loops with four SGs and two PHs per loop. Unit 8
operates original recirculating steam generators containing 4 200 tubes each and external
preheaters containing 2 820 inverted U-tubes each. The tubing material for both steam
generators and preheaters is mill-annealed Inconel 600 alloy with design dimensions of
12.95 mm (0.510 inch) outside diameter (OD) and 1.12 mm (0.044 inch) wall thickness.
Three of the eight steam generators at Unit 8 have been experiencing significant degradationof the carbon steel tube support plates due to flow accelerated corrosion, which appears to be
attributable to the low chromium content of the material [43]. Replacement of the steam
generators is planned during the major component replacement outage planned to commence in

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2030. Until that time, the degradation of the supports is monitored through inspection activities
and there have been changes to feedwater chemistry control to reduce the rate of degradation.

3.2.3 Material degradation related to steam generator lay-up
In the mid-2000s, the alloy 600 steam generator tubes at the Bruce Nuclear Generating Station
were subject to outside diameter intergranular attack/stress corrosion cracking at the hot-leg
top of tube sheet region [44]. The degradation mechanism was attributed to aggressive reduced
sulphur species in the sludge pile arising from exposure to oxygen when the steam generators
were fully drained for maintenance activities. In 2007, the operator implemented controls to
reduce oxygen ingress during maintenance activities which has arrested the degradation. These
controls included only fully draining the steam generators for maintenance when necessary
and limiting the time which the steam generators are fully drained. These controls have been
adopted across the Canadian CANDU nuclear power plant fleet.

3.2.4 Buried piping
CANDU nuclear power plants have sections of buried piping on several safety significant
systems, including emergency core cooling systems, service water systems and fire protection
systems. Materials of construction included carbon steel, reinforced concrete pipe and cast iron.
The exterior of the piping was usually coated to protect against soil-side corrosion and often
sacrificial anode cathodic protection was often used. Many of these systems contained raw
service water. As the plants age the external coatings have been subject to deterioration leading
and anodes have been consumed leading to the potential for increased soil side corrosion.
Furthermore, the interior of the piping may have contained raw service water and been exposed
to microbiologically influenced corrosion (MIC). The Canadian utilities have implemented
buried piping programmes to assess the current condition of safety significant buried piping toconfirm the condition of piping, repair protective coatings, perform selective replacement of
piping (sometimes using non-metallic materials) and improve cathodic protection to support
LTO; see reference [34] for additional information.

3.2.5 Calandria and pressure tubes
In the CANDU®, RBMK and Fugen-ATR 2 reactors, the pressure tubes (PTs) serve the same
purpose as the pressure vessel in light water reactors. These tubes are long zirconium alloytubes that pass through the centre of the reactor core and therefore are subjected to neutron
irradiation, high temperatures and pressures, and corrosion from the heat transport fluid. For
these tubes to remain fit for service from an economic perspective, they have to maintain theirintegrity for at least 30 years at a 90% capacity factor.
The pressure tubes used in a CANDU reactor are made from zirconium alloyed with niobium;(Zr-2.5%Nb. During service the pressure tubes operate at temperatures between about 250°C and
310°C and with variable coolant pressures up to 11 MPa, corresponding to hoop stresses of up
to 130 MPa. The pressure tubes are subjected to a maximum flux of fast neutrons from the fuelof about 4E+17 n·m2 s−1. These operating conditions can lead to degradation of the pressure tube
material with respect to dimensional changes because of irradiation creep and growth,
deterioration in mechanical properties due to irradiation embitterment, initiation and growthof new flaws like fretting damage due to debris. Thermal creep, irradiation creep and irradiation
growth change the shape of pressure tubes during service. The tubes elongate axially, expand
in diameter and sag. The rate of deformation is linearly related to fast neutron flux.
The products of the chemical reaction between the pressure tube and the PHTS heavy water
are zirconium oxide and deuterium. The loss of metal from corrosion is structurally insignificant
but some of the deuterium is absorbed by the tubes. Deuterium also enters the tubes via the

2. Fugen-ATR was a Japanese 165 MWe Prototype Heavy Water Reactor. It began commercial operation in
March 1979 and it was permanently shut down on 29 March 2003.

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stainless steel end fittings at each end of the channels. It may also enter from the outside
surface of the tube if there is deuterium present in the annulus between the pressure tube andCalandria tube and the oxide on the pressure tube loses its effectiveness as an adequate barrier.
The increase in deuterium concentration increases the susceptibility to delayed hydride
cracking (DHC) [46] and may decrease the fracture toughness. Pressure tube cracking andfracture have been observed since the early 1970s. 3
An example of pressure tube fracture is an event at Pickering-A Unit 2. On 1 August 1983 apressure tube failed by developing a two-metre-long crack near the coolant outlet of the tube.
While at full power, a sudden loss of heavy water from the PHT system occurred. The loss rate
was about 17 kg/s. The failure was caused by the contact of the pressure tube and the Calandria
tube for a relatively long time (two to five years) due to installation error of the outlet garter spring.
The outside surface of pressure tube was cooled where it contacted the Calandria tube. The
deuterium built up in the operating pressure tube, then migrated to the cool areas and formed
zirconium hydride blisters. A crack was formed through four of the zirconium hydride blisters.
Canadian licensees periodically inspect pressure tubes in accordance with clause 12 of CSA
standard N285.4 (Periodic Inspection of CANDU Nuclear Power Plant Components). Among several
volumetric and dimensional inspection requirements, clause 12 of CSA N285.4 requires thedetermination of the pressure tube to Calandria tube gap and stipulates that the pressure tubes
are considered acceptable for operation only when “no pressure tube to Calandria tube contact”
is predicted to exist at the end of the next periodic inspection interval. When the screening criteriaof CSA N285.4 are not satisfied, N285.8 (“Technical Requirements for In-Service Evaluation of
Zirconium Alloy Pressure Tubes in CANDU Reactors”) provides additional requirements todisposition. Additionally, N285.8 requires a risk assessment for the entire core performed either
deterministically or probabilistically.

3.3 Czechia
In Czechia, there are four nuclear units of type WWER-440/V213 in operation at the Dukovanysite. The four Dukovany reactor units were commissioned in 1985-1987. Commissioned in 2000
and 2002, the two reactor units at Temelín are of type WWER-1 000/320. There are no fixed
operating time limits in Czechia.
The definitions and procedures concerning requirements for ageing management are foundin SÚJB Safety Guide BN-JB-2.1 “Ageing Management for Nuclear Installations.” Addressed inthis guide are the following activities:
1. Scoping and screening process of the SSCs that should be subject to ageing assessment;
2. Understanding dominant ageing effects/mechanisms of SSCs selected within thescreening process and finding or developing effective and usable methods for monitoringand mitigating their ageing effects; and3. Ageing degradation management of specific SSC by implementing effective measures in
the field of ISI, maintenance and management of facility operation.

3.3.1 The Czech AMP approach
The overall ageing management programme (AMP) for the Dukovany and Temelín nuclear power
plants is consistent with the relevant IAEA Safety Standards (e.g. Safety Series Guide SSG-48“Ageing Management and Development of a Programme for Long-Term Operation of Nuclear
Power Plants” [36]) and the WENRA Safety Reference Levels. The quality assurance process of theageing management programme is defined in document ČEZ_PG_0001 “Operational Ageing
Management Programme for NPPs.”

3.

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Limited to selected “representative events” the current version of the CODAP event database includes
three events that involve DHC-induced degradation of pressure tube integrity.

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In view of the fact that the Dukovany Nuclear Power Plant has already achieved its original
lifetime limit set by design and that ČEZ, a. s., declared a strategic goal for its nuclear power
plants to extend their life span by 20 to 30 years as a minimum, the LTO programme was
implemented in accordance with international best practice. Therefore, ČEZ, a. s. took an active
part in the IAEA programme called the safety aspects of long-term operation (SALTO) and has
been involved in the IAEA International Generic Ageing Lessons Learned (IGALL) Programme[26]. In accordance with the requirements of Section 49(1)s of the Atomic Act and Sections 11
and 12 of Decree No. 21/2017, an operational ageing management programme was implemented
by the nuclear power plant operator ČEZ, a.s.

3.3.2 Systems, structures and components subject to AM
The scope of the ageing management process is defined in Decree No. 21/2017 Coll. The
following should be included in the selection of systems, structures and components subject tothe ageing management process:•To fulfil the requirements of Decree No. 21/2017 Coll., the methodology ČEZ_ME_0987 –Selection and assessment of equipment for AM and LTO, the following criteria for theselection of systems subject to the ageing management process are set out:–

Equipment classified as safety class one, two, three according to Decree No. 329/2017
Coll.

–

Equipment with specific safety functions according to the methodology ČEZ_ME_0901;
classification of systems, structures and components of nuclear power plants interms of Nuclear Safety, i.e. equipment with criticality level 1 and level 2, determinedaccording to the methodology ČEZ_ME_0608 – SSC categorisation in the production
division.

–

Equipment classified as safety-relevant per PSA insights and as articulated in Decree
No. 162/2017 Coll.

–

Equipment recommended from global good practice, operating experience andoutcomes of common walkdowns.

3.3.3 Component-specific AMPs
To ensure the required service life of SSC, a graded approach is implemented to ageing
management according to the standard ČEZ_ST_0072, “Requirements for Nuclear Power Plant
Reliability Management.” The graded approach is selected on the basis of the strategy definedfor the care of SSC.
The list of equipment for which the ageing management is ensured by introducing thecomponent specific AMPs or by preparing TLAA analyses, results from the strategies for care of
SSCs and is given by a list of specific equipment (Category A), which is the outcome of theimplementation of activities under working procedure ČEZ_PP_0425. Selected componentspecific AMPs are itemised below:•

low-cycle fatigue – passive mechanical components;•

flow-accelerated corrosion – nuclear power plant secondary circuit piping;•radiation damage to reactor pressure vessels;•

service water piping;•welds and weld-HAZ susceptible to degradation;•

passive components of main circulation pumps (MCP);•

passive components of main isolation valves (MIV);•pressuriser (PZR);•

steam generators (SG);•safety class one piping;•high-energy pipelines.

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3.3.4 A Czech perspective on WWER age-dependent material degradation
A Czech perspective on “open material issues before extended operation of reactor vesselinternals” is outlined in Reference [49]. The first widely reported IASCC event in the internals of
WWER plants was the cracking of Loviisa-2 baffle bolts, which were discovered in September
2006 [50]. Several bolts were investigated as a result of NDE indications after 30 years ofoperation. The material of the bolts is cold worked ti-stabilised austenitic stainless steel (AISI
Type 321 steel).
The issue of high dose irradiation is considered more significant for WWERs than BWRs
because estimated doses can reach 120-150 dpa. It is known that material properties change inlong-term irradiation and saturation is expected after 5-15 dpa. Experimental data of material
irradiated at the thermal reactor conditions are available up to 80 dpa, and include data formechanical properties, the ultimate tensile strength and the fracture toughness. It is not fully
clear whether the material behaviour remains at the saturation level after more than 100 dpa.

3.3.5 PEO/LTO-OPEX and solutions
The WWER reactor of Type 440/213 has six primary loops. Each loop has a horizontal steam
generator (SG). This reactor type was designed in the 1970s and four units are in operation in
Czechia (Dukovany Units 1, 2, 3 and 4). The primary circuit is made of seamless titanium-stabilised
austenitic stainless steel piping of grade 08Ch18N12T steel. The steam generator (SG) shell is madeof carbon steel. There are dissimilar metal welds (DMWs) between the stainless steel piping andthe steam generators. There is one DMW on the steam generator hot collector (inlet side) and one
DMW on the steam generator cold collector (outlet side) for a total 12 DMWs per reactor unit.
After approximately 25 years of operation, NDE examinations performed at Dukovany Units 2
and 4 in 2012 and 2013, respectively, revealed circumferential stress corrosion cracks along thefusion line between the carbon steel and the first layer of the stainless steel weld butter; [51][52].
The degraded welds were completely removed followed by the installation of new welds withhigher chromium content stainless steel weld material. The DMWs are located at the lowest pointof respective SG in the area of closed pockets on the secondary side; Figure 3-1. Corrosion products
from the secondary side are deposited on the lowest point of the SG and it creates a corrosive
environment with higher concentrations of salts. This creates an environment that promotes
electrochemical corrosion due to a difference in the electrochemical potential of the carbon steel
and stainless steel. This in turn produces pitting which is a significant stress concentrator andpromotes stress corrosion cracking.

Figure 3-1: Steam generator collector DMW

Source: Reproduced from a presentation prepared for the
NEA WGIAGE Metal Subgroup Meeting, 11-13 April 2015;“Current Status of Weld Joints in CZ NPPs,” by Lubomír
Junek (IAM, Brno) and Zdeněk Čančura (ČEZ a.s.).

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Multiple instances of SCC of DMWs in the super emergency feedwater system (SEFW) piping
were identified at the Dukovany Nuclear Power Plant (and at the Bohunice Nuclear Power Plant
in the Slovak Republic) in 2015. The affected DMWs were completely removed and re-welded.
In response to continuing material degradation, a first-in-kind weld overlay (WOL) following thegeneral guidelines of ASME Code Case N-740-2 was installed in March 2018 on a DMW joining
the SEFW piping (OD = 89 mm) to the steam generators in Dukovany Unit 2 [53]; Figure 3-2. This
was the first WOL installed in Czechia and it was accepted by the Czech regulator and two
different inspection agencies following successful mock-up demonstrations, welding procedure
qualification, a non-destructive examination demonstration and weld residual stress analyses.

Figure 3-2: Dukovany-2 first-in-kind SEFW weld overlay

3.4 Finland
STUK Regulation on the Safety of a Nuclear Power Plant (STUK Y/1/2018) 4 stipulates that thedesign, construction, operation, condition monitoring and maintenance of a nuclear power
plant shall provide for the ageing of systems, structures and components (SSCs) important tosafety in order to ensure that they meet the design‐basis requirements with necessary safetymargins throughout the service life of the facility. Furthermore, systematic procedures shall bein place for preventing such ageing of SSCs which may deteriorate their operability, and for the
early detection of the need for their repair, modification and replacement. Safety requirements
and applicability of new technology shall be periodically assessed in order to ensure that thetechnology applied is up to date, and the availability of the spare parts and the system support
shall be monitored.

3.4.1 Regulatory AM framework
Issued in February 2019, Regulatory Guide YVL A.8 “Ageing Management of a Nuclear Facility” 5
imposes requirements on licensees related to the management of physical ageing as well as
obsolescence of SSCs and presents the regulatory oversight relevant to the licensees’ duties. The
Guide applies to all nuclear power plant life cycle phases and all SSCs important to nuclear andradiation safety. Regulatory requirements set in the Guide aim at ensuring both short and longterm operability and technological conformance of SSCs, whether in service or stand‐by.

4. STUK (2018), “Radiation and Nuclear Safety Authority Regulation on the Safety of a Nuclear Power Plant”,
www.stuklex.fi/en/maarays/stuk-y-1-2018.
5. STUK (2019), “Ageing Management of a Nuclear Facility”, 15.2.2019, www.stuklex.fi/en/ohje/YVLA-8.

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Annex A of YVL A.8 includes a list of “typical ageing mechanisms.” The key documents for the
regulator’s review are a conceptual plan for ageing management and an ageing management
programme along with construction and operating licence applications, respectively, andageing management follow‐up reports issued annually by the licensees.
In 2018 the two Finnish licensees TVO and Fortum issued their updated ageing management
programmes according to Guide YVL A.8. The role of comprehensive ageing management
programmes will be more emphasised as the operation of the nuclear power plant units is
extended beyond the original design lifetime. The design basis operability of SSCs has to be
maintained even though their degradation rate may be hard to anticipate. A generic lesson
learnt in Finland is that the closer nuclear power plants get to the end of their design lifetime,the more challenging it is for the licensees to start large and expensive investments tomodernise or modify the nuclear power plants.

3.4.2 Active research programmes
In accordance with Chapter 7a of the Finnish Nuclear Energy Act, the objective of the SAFIR2018
National Nuclear Power Plant Safety Research Programme is to ensure that, as new mattersrelated to the safe use of nuclear power plants arise, the regulatory authority possesses
sufficient technical expertise and other competence required for rapidly determining thesignificance of the matters. With respect to the structural safety and materials area the aim of
the research is to increase knowledge supporting the long-term and reliable use of nuclear
power plants, particularly with respect to the integrity of barriers and material issues that affectsafety. A summary of selected research results of the SAFIR2018 project [54] is given below.

3.4.2.1 Analysis of fatigue and other cumulative ageing effects to extend lifetime (FOUND)
The project FOUND (2015-2016) 6 was a cross-disciplinary assessment of ageing mechanisms forsafe management and extension of operational plant lifetime. It developed deterministic,
probabilistic and risk-informed approaches in computational and experimental analyses. The
focus areas were:•remaining lifetime and long-term operation of components having defects;•

susceptibility of BWR RPV internals to degradation mechanisms;•

fatigue usage of primary circuit, with emphasis on environmental effects andtransferability;•

fatigue and crack growth caused by thermal loads, with emphasis on modelling;•elaborations on certain aspects of RI-ISI methodology 7;•

nuclear power plant piping analysis methods;•

residual stresses in BWR nuclear power plants.

3.4.2.2 Long-term operation aspects of structural integrity (LOST)
The main objective of LOST is to develop methods and tools to assess the structural integrity ofprimary circuit components, considering particularly the long-term operation aspects. LOST is
divided into two work packages, where the first work package is carried out in SAFIR2018 andthe second work package in SAFIR2022, in the AMOS project. The first package focuses on RPV

6.
7.

56

VTT (2019), “FOUND Analysis of Fatigue and Other Cumulative Ageing to Extend Lifetime”,
http://safir2018.vtt.fi/finalseminar/22032019/3_2_FOUND%20final%20seminar.pdf.
Examples of work products include VTT-R-04644-17 (Computation of Conditional Large Early Release
Probabilities of Piping Components) and VTT-R-03666-16 (Computation of Consequences of Piping
Component Failures in PRA Software); these reports are available from https://cris.vtt.fi/en/searchAll.

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safety; 1) fast fracture in the upper shelf region and 2) pre-investigations for BREDA (Barsebäck
Research and Development Arena). The second package focuses on dissimilar metal welds;
1) residual stresses 2) materials characterisation and 3) numerical simulations.
Fast fracture testing was performed in a temperature above the range covered by the 5%
master curve, known as the upper shelf region. Specimens were cooled from 300°C to the room
temperature at a cooling rate of 2°C/s (rate determined from centre location of the specimen).
The effect of quick cooling on tearing resistances of the material, the J-R curves, were examinedin reference [55]. It was concluded that for the first time, a testing method was used for theassessment of cooling rate effect on fracture toughness in upper shelf region – extreme fast
fracture was not observed during testing of the fast cooling transient. However, conclusions
were based on the obtained load-displacement data and further analysis is required. SAFIR2022
will continue this work package in the AMOS project. Remarks were also made that thedeveloped method should be applied for an actual RPV-material in the future.
Pre-investigations for BREDA retained testing of fracture toughness with 4 mm thickminiature C(T) specimens for characterisation of the T0-temperature of respective materials(miniature specimens are essential in LTO-perspective since there is a limited amount of
nuclear power plant materials available in surveillance programmes for monitoring of ageing of
the RPV). The use of miniature C(T) specimens is validated by comparing the initiation locationsof the brittle fracture to standard sized specimens’ initiation locations. The initiation location
with respect to the mid-plane of miniature C(T) specimens has the same distribution as
conventional, standard sized specimens, and side-grooves did not show a significant effect on
the obtained T0; these results validate the use of miniature C(T) specimens. Large data sets,
regarding obtained T0 with miniature and thicker C(T) specimens for various materials, showthat the values are in the same range; see reference [56].
However, based on the previous work reported in open literature, the following was
concluded: In almost all the previous investigations, the materials characterisation included
chemical, hardness, macrostructure and fracture toughness analysis in the through-thickness
direction. These analyses are necessary to explain the variation of the material properties inthickness direction. Taking this into account, analyses were not comprehensive enough to draw
definitive conclusions.

3.4.2.3 Mitigation of cracking through advanced water chemistry (MOCCA)
Corrosion problems in the PWR secondary circuit are mostly related to deposition of magnetite
particles in the steam generator (SG) inside surfaces and the enrichment of impurities into
crevices under these deposits. About 70% of tube failures are caused by stress corrosion cracking(SCC), which is a localised phenomenon enabled by certain conditions, e.g. susceptible material,
sufficiently high stress levels and localised aggressive environments. MOCCA studies the effectof alternative water chemistry on:•

the feed water line corrosion rate;•

the tendency of magnetite to deposit into the SG;•

the lead (Pb) concentrations, which may interact adversely within the deposits.

A previously used method to minimise the SG feed water line corrosion is to use an oxygen
scavenger, hydrazine (N2H4). However, there is a distinct possibility that, because of the healthand environmental risks related to the use of hydrazine, the EU will in the future pass a directive
forbidding its use. Therefore, different measures to mitigate the above-mentioned detrimental
effects of deposits, e.g. variation of secondary side circulation pH, searching good oxygen
scavengers (as replacement to hydrazine) and variation of surface (electrical) charge of particles,
were also studied in MOCCA; see references [57][58].
It was concluded that the two possible candidates to replace hydrazine as an oxygen
scavenger were found to be carbohydrazine and iso-ascorbic acid. However, their efficiency as
oxygen scavengers was about 50% of that of hydrazine. The effect of different pH of water

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chemistry regimes was found to mitigate the deposits into SG’s when the pH was more alkaline.
The effect of surface charge of particles was suggested to play an insignificant role in magnetite
deposition process.
It has been a previous concern that lead may promote SCC in carbon steel made SG’s. In
this work, MOCCA, it was shown that the addition of 100 ppm of lead was in fact beneficial tocarbon steel under both acidic and alkaline crevice conditions; see reference [59]. Without theaddition of lead, it was found that the carbon steel was more susceptible to SCC under acidic
conditions than under alkaline crevice conditions.

3.4.2.4 Thermal ageing and EAC research for plant life management (THELMA)
The project THELMA, thermal ageing and EAC (environmental assisted cracking) research for plantlife management, deals with nuclear materials behaviour in light water reactor environments. The
development of thermal ageing of stainless steels, both welds in wrought material and castmaterial, have been investigated in THELMA – in other words, the focus of the study was theageing of austenitic primary circuit materials. The application of an electrochemical measurement
method, double loop electrochemical potentiokinetic reactivation (DL-EPR), was applied todetermine material ageing in the study as well. One of the objectives of THELMA was to increase
the knowledge on material behaviour in nuclear power plant environments, which is needed toassist STUK, licensees and other entities in questions concerning plant life management, material
degradation assessment and failure analysis. The following topics were studied:•thermal ageing of cast stainless steel;•thermal ageing and short-range ordering (SRO) of alloy 690;•

RPV weld brittle fracture initiation;•

characterisation of irradiated stainless steel materials.

Measurements were conducted on hot leg (HL) and cold leg (CL) materials of the SG, when
the stainless cast materials were examined with DL-EPR; see reference [60]. Results showed thecomplexity of ageing and that the DL-EPR can be used for the evaluation of the degree of thermal
ageing in materials that have aged at higher than operational temperatures, so it may not be so
useful for plant-aged materials. It was concluded that there is a need for a method capable ofdetermining the degree of thermal ageing from plant components without the need todismantle the component. It has been noted that a method called transient granting (TG)
spectroscopy has been developed at MIT, United States, that has given promising preliminary
results that are in line with the DL-EPR’s observation of a stronger “ageing response” in the CL
material compared to the HL material.
In the case of alloy 690, there is a clear influence of SRO on the hardness. This was verified inseven different material conditions, i.e. ageing temperatures and post-treatment states (e.g. coldworked, solution annealed). The combination of intergranular (IG) carbide precipitation andgrowth in combination with SRO is deemed detrimental to PWSCC resistance, and it would
deserve further investigations on the matter with more commercial alloy 690 materials; seereference [61].

3.4.2.5 NDE of primary circuit components (WANDA)
The ageing of the passive components of nuclear power plants is monitored through in-service
inspections (ISI), which are heavily related to condition monitoring by NDE methods. The
WANDA project focuses on three important aspects of proactive ageing management: early
detection of deterioration, monitoring of deterioration and application of prognostics for theestimation of remaining component life. NDE is one of the recommended tools for the early
detection of deterioration of nuclear power plant materials.

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COUNTRY/ECONOMY-SPECIFIC OVERVIEWS OF MATERIAL DEGRADATION STATE-OF-KNOWLEDGE, RESEARCH AND OPERATING EXPERIENCEWANDA was divided into two work packages. The first package focused on early detection
and inspection reliability. This included research on plausible ways to detect material
deterioration and efficient ways to evaluate the probability of detection (POD) curves for ISI. In
work package two, the focus was on concrete infrastructure. The main objective of work package
two was to bring concrete NDE research on a par with metal NDE.
The new approach to determining the POD-curves was introduced in WANDA with promising
results; physical flaws were developed with eFlaw technology using only limited amounts of real
physical flaws. With eFlaw, scanned UT-data from a single sample was digitally altered andextended to an unlimited number of examples of data for POD determination. With the ability tocreate flawed data files on demand, it was possible to gather hit/miss data from various inspectors
and to generate POD-curves on these data – even with the very limited number of actual physicalflaws, inspectors were able to find flaws and generate a plausible POD-curve; see reference [62].
Concrete NDE research was done by building a mock-up wall, loosely representing the crosssection design of the containment wall of Olkiluoto-2 containing steel reinforcements, steel liner
and tendon ducts. During construction of the mock-up wall, numerous common types ofsimulated defects were embedded within the wall and the slab to determine how the current
state-of-the-art NDE techniques are able to determine various forms of degradation in nuclear
power plant concrete structures. It was also concluded that the mock-up wall enables future
research on concrete NDE and opens the possibility of training future experts; see reference [63].

3.4.2.6 RPV integrity research
As part of a Nordic nuclear safety research programme [64], the Technical Research Centre of
Finland (VTT), Chalmers University of Technology, and the Royal Institute of Technology (KTH)
have continued earlier efforts regarding extraction, mechanical and microstructural testing, andanalysis of materials from a retired reactor pressure vessel (RPV). The objective of the study is
to increase the current knowledge base on the correctness of the existing surveillance
programmes, as well as the influence of long-term thermal ageing of the materials. During 2016,a baseline study was performed to prepare the basis for a test programme to analyse the asaged material properties of the RPV from the decommissioned Barsebäck Unit 2. 8
Included in the SAFIR2022 (2019-2022) research plan [65] is the BRUTE Project, “Barsebäck
RPV Material Used for True Evaluation of Embrittlement”, which is intended to pioneer the new
infrastructure of the VTT Centre for Nuclear Safety (CNS) hot cells, and to determine theproperties of RPV material after thermal ageing and neutron irradiation. All methods used areverified and commissioned in the new CNS hot cell environment.
Mechanical tests (e.g. impact, fracture toughness and tensile tests) will be performed todetermine the thermal and neutron embrittlement in high-Ni reactor pressure vessel weld metal,
typical for Nordic RPVs. The microstructure is to be characterised to obtain an improved
understanding of factors affecting brittle fracture and embrittlement. The results will be used tocompare the results from surveillance programmes, used for assessment of the embrittlement
using specimens manufactured at the time of the RPV manufacturing with results from material
from the component itself. The data will be used to update the existing prediction curve(s) for
embrittlement and improve the understanding of both neutron and thermal embrittlement.

8. The Swedish Barsebäck 2 BWR began commercial operation in 1977 and was permanently shut down in
2005. Following this, the BREDA project (Barsebäck Research & Development Arena) was initiated in 2016
with the purpose of assessing the ageing of RPV materials. In the BREDA project, cylindrical trepans witha diameter of 200 mm were extracted from beltline welds subjected to a neutron irradiation from thebeltline as well as from welds from the RPV head. This extraction was completed during 2018. The BRUTE
project was launched in 2018 as part of the SAFIR2018 programme, and an application for its
continuation in the SAFIR2022 programme has been submitted.

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3.4.3 PEO/LTO-OPEX and solutions
TVO has two power units in Finland: The OL1 and OL2 BWR units are of type AA-IV BWR-2500.
They were designed by Asea-Atom (AA) of Sweden. OL1 reached the 40-year design age in 2018
and OL2 will do so in 2020. Both units were granted a licence to operate until 2038. During licence
renewal of OL1/OL2, the owner/operator re-calculated thermal transients of the processes.
It was concluded that no changes are required to be implemented in the processes. However,there are known locations where high cycle thermal fatigue has induced cracks. These are:•

FW/RHR (312/321-system) mixing tees where cracks up to a few mm depth occurred after
25 years of operation; see Topical Report 6 [38] for more detailed information. The cyclic
thermal load occurs during hot standby when cold water is injected to keep the reactor
water level in the allowable range. These tees were recently replaced. Process changes areunder investigation to mitigate these thermal cycles by changing the circulation of cold
feedwater from FW/RHR (312/321) mixing tees to go through systems RWCU/RHR (331/321).

•

FW-distributors (213.1-system), which distributes the feedwater evenly to the reactor: On
the short-term thermal shields were added to protect thermally loaded piping. On thelong-term changes have been made to the spargers and these cycles no longer occur.

•

Core-spray pipelines (213.3-system). These pipelines were thermally loaded due to theabove-mentioned FW-distributor problem. The cracks have arrested, but recently someof the existing cracks opened due to a test where cold water was injected into these pipes.

Thermal fatigue in the above-mentioned locations is monitored with ISI. Assessment of therisk for thermal fatigue can also be managed by analysis. In addition to deterministic fatigue
evaluation, TVO has developed a probabilistic risk assessment method (PAMPRI) to evaluate therisk for piping or component damage as well as the risk for core damage. With this approach all
significant damage mechanisms are considered. Although the programme itself is ready,validation and implementation of PAMPRI is still ongoing.

3.5 France
The nuclear power plants currently in operation in France comprise a total of 56 pressurised
water reactors (PWRs). The two oldest units, Fessenheim Unit 1 and Unit 2, were permanently
closed in 2020. One unique feature of the French nuclear power plant fleet is its standardisation,with many technically similar reactors spread over 19 nuclear sites. Each site includes two to
six PWRs. The nuclear reactor fleet is divided into four NSSS designs: CP0 Series (900 MWe), CPY
Series (900 MWe), P4 and P’4 Series (1 300 MWe) and N4 Series (1 450 MWe). There are four
reactors in the N4 Series; two units at Chooz and two units at Civaux.

3.5.1 Regulatory AM framework
According to the French regulatory framework, ageing management must be demonstrated,
relying on OPEX feedback, the maintenance provisions, and the possibility of either repairing or
replacing the components. Other factors must also be considered, in particular physicalphenomena which can modify the characteristics of the equipment, depending on theirfunction or their conditions of use. Consideration must therefore be given to the deteriorationof replaceable items and the lifetime of non-replaceable items. The periodic safety review is an
opportunity for an in-depth examination of the effects of ageing on the equipment. The
regulatory requirements concerning ageing management are summarised as follows:•60

The periodic review process contained in articles L. 593-18 and L. 593-19 of theEnvironment Code. The Environment Code defines the conditions for a periodic safety
review of the nuclear facilities and stipulates the role of each of the stakeholders:
including a) the licensee, which is responsible for the safety of its facilities, b) ASN, whichis responsible for ensuring that the licensee implements the means needed to ensure ahigh level of safety, and c) IRSN, which is ASN’s technical support organisation.

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Article R. 557-14-2 of the Environment Code for nuclear pressure equipment (reactor
vessels in particular).

•

The provisions of the order of 30 December 2015 relative to nuclear pressure equipment.

•

Article 2.5.1 of the order of 7 February 2012 setting the general rules for basic nuclear
installations (BNIs), requiring that “design, construction, testing, inspection andmaintenance provisions ensure that this qualification is maintained for as long asnecessary”.

•

Provisions of the order of 10 November 1999 relative to the monitoring of operation of
the main primary system and the main secondary systems of nuclear pressurised water
reactors.

•

ASN positions.

•

The ageing management provisions of the “PWR Design Guide” produced jointly by the
ASN and IRSN and published on 18 July 2017.

3.5.2 Fourth periodic safety review of 900 MWe reactors
In 2018, assessments were initiated for the periodic safety review of 900 MWe series reactors
associated with the fourth ten-yearly inspections (VD4 900); see Figure 3-3. These reactors werecommissioned between 1979 and 1988. The IRSN assessed the measures implemented by EDF
to successfully manage the ageing of these reactors and presented its report to the advisorycommittee for reactors (GPR) and the advisory committee for nuclear pressure equipment(GPESPN). It also continued to assess design-basis accident studies, Level 1 and 2 PSA andstudies of internal and external hazards as well as severe accident studies, with a view topresenting its conclusions to the GPR in 2019 at the four dedicated meetings.

Figure 3-3: The fourth PSR schedule for the French 900 MWe reactor units

Dampierre-2
Gravelines-1
Dampierre-1
Blayais-1
Bugey-5

Tricastin 1
2020

St. Laurent-B2
Gravelines-4
Gravelines-2
Dampierre-3
Chinon-B1
Blayais-2
Bugey-3

2021

Bugey-2
Bugey-4
Tricastin-2

2022

2023

Gravelines-3
Tricastin-3

2024

Cruas-1
Blayais-4
Blayais-3
2025

2026

Cruas-3
Dampierre-4
St. Laurent-B1
Tricastin-4

Gravelines-6
Cruas-2Chinon-B3
2027

2030

Chinon-B2
Cruas-4
Gravelines-5

2031

Chinon-B4

Note: This timeline indicates the year in which a PSR is expected to be completed.

In addition, the in-service performance of 900 MWe reactor vessels between the two tenyearly inspections was subject to a first review, which was presented to the GPESPN and wasconcluded in 2019 by a second assessment. This generally entails assessing the relevance anddemonstrative nature of the studies submitted by EDF, evaluating the adequacy of the changes
proposed by EDF with regard to the objectives of the VD4 900 safety review (in particular working
towards the safety objectives defined for third-generation reactors), and ensuring that thereactors can continue to be operated safely over the coming decade.

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3.5.3 Insights from the PSR 2018-2020
The approach to controlling ageing implemented by EDF is based on a process of examining
structures, systems or components and the way in which their integrity or functionality may beaffected by an ageing mechanism. This process accounts for the operating and maintenanceprovisions in force, as well as repair or replacement considerations.
The IRSN considers that this process is satisfactory with a view to extending the operating
life of the 900 MWe reactors beyond 40 years, even if the processing of experience feedback andthe anticipation of decisions to be taken will have to be improved (the generalisation to all
reactors of lessons from a particular reactor on corrosion, fouling of piping, etc.).
EDF's maintenance strategy aims to decide on the actions to be taken with a view tocontrolling the ageing or obsolescence of systems, structures and components. These actions
may relate to preventive maintenance (routine) and exceptional maintenance or advancestudies of repair or replacement solutions, integrating research and development actions whereappropriate. The IRSN considers that EDF must improve its preventive maintenanceprogrammes to better guarantee the compliance of its installations. This requires better
anticipation, a strengthening of the integration of safety issues into risk analyses, animprovement in the quality of execution, controls and requalification, as well as the speed ofdetection of deviations. The commitments made by EDF concerning exceptional maintenanceare satisfactory, with, for example, the replacement of the cylinder heads of the back-up
generator sets by the end of 2025. Finally, concerning obsolescence, the EDF process is likely toguarantee a satisfactory and lasting treatment.
As part of the extension of the operating life of the reactors, a specific approach concerns
the replaceable equipment that would remain in place beyond the life selected during theirinitial qualification. EDF has defined a progressive qualification programme which concerns
some electrical and mechanical equipment, which is satisfactory given the commitments made.
The complete demonstration of this progressive qualification will be acquired once the resultsof the expertise and tests of materials to be taken on-site have been obtained.
Some structures and equipment, if they are not replaceable, are the subject of a suitability
check for continued operation. In this regard, based on the files examined, the IRSN considers
that the serviceability of the reactor pressure vessel (RPV) of reactors No. 1 at Tricastin and No. 2
at Bugey, the first reactors concerned by the RP4-900 review, has been demonstrated for ten
years of operation beyond the fourth ten-yearly outages. For the other reactors, additional
information is still necessary to reach conclusions. Likewise, additional controls and studies areplanned for the primary and secondary circuits. The elements presented by EDF for the civilengineering structures, the internal structures of the RPV, the primary pump units (hydraulicpart), the general electrical installation cables, the instrumentation and control system and theelectrical feedthroughs of the enclosure wall are generally satisfactory, with some supplements
provided by EDF. For the reactors at the Bugey Nuclear Power Plant, the IRSN considers it
necessary for EDF to implement a programme to check the condition of the shell screws of thevessel internals.

3.6 Germany
In Germany, a fixed shutdown date for each plant is mandated by law, with the end of 2022
being the final date for the plants to be shut down. Considering the final shutdown dates, nooperating plant in Germany will exceed the design lifetime of 40 years. Indeed, none of theplants which were already shut down in the past have exceeded the 40-year design lifetime.
The longest operating time in the past and the longest estimated operating time for respectiveplant currently in operation is about 37 years. The German piping OPEX for plants that havebeen operated for ≥ 25 years is summarised in Figure 3-4. Included in this OPEX are failures
discovered during periods beyond final shutdown and during decommissioning.

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Figure 3-4: The German piping OPEX after 25 years of operation
20
18

No. of pipe failures

16
14
12
10
8
6
4
2
0
25

26

27

28

29

30

31

32

33

34

35

36

37

38

39

40

41

42

43

Age of plant at the time of failure
RCPB

Safety Class 2

Safety Class 3

Non-safety

One of the most challenging issues for LTO is the neutron embrittlement of the RPV beltline,
especially in PWRs. In the design of the RPV of German 1 300 MW PWRs, the water gap between
the core barrel and RPV wall was increased compared to other Western PWR RPVs to reduce theneutron fluence at the RPV wall. Together with the early implementation of the low-leakage
core strategy, the end-of-life fluence in German PWR RPV beltlines is calculated to be around3∙1 018 n/cm2 (E > 1 MeV) and hence around one order of magnitude lower than for other
Western PWR RPVs. The ductile-to-brittle transition temperature of German PWR RPVs is stillaround 0°C or below. From a technical point of view, neutron embrittlement would not be anissue, not even for LTO.

3.7 Japan
Nuclear power plant operating periods are limited to 60 years in Japan. Most of the Japanese
nuclear power plants experienced, or continue to experience, prolonged shutdown periods after
the Great Japan Earthquake of 2011. These prolonged shutdown periods have significantly reduced
the remaining licensed operating times for the nuclear power plants. As a result, the 11 Japanese
nuclear operators (JNOs) have completed technical assessments of ageing-related degradation for
nuclear power plant infrastructure originating from the long-term shutdown period. In 2018, theElectric Power Research Institute (EPRI) published the results of an independent review of the JNO
technical assessments; see reference [66]. The independent review focused on: 1) low-cycle fatigueof Class 1 components, 2) neutron irradiation embrittlement, 3) irradiation assisted stresscorrosion cracking, and 4) thermal ageing of cast austenitic stainless steels (CASS). The following
conclusions were reached:1.

Low-cycle fatigue (LCF). Since no transient events are expected to occur during longterm shutdown, long-term plant shutdowns will not cause significant effects on LCF.

2.

Neutron irradiation embrittlement. No embrittlement effects from neutron fluence are
expected because of the absence of a fission reaction in the reactor core during longterm shutdown. Furthermore, during the long-term shutdown, no pressurised thermaltransients, at either high pressure, low temperature or both, are possible. Thus, nopressurised thermal shock (PTS) events that challenge the integrity of the RPV under
brittle conditions, nor any events that could induce a ductile failure via an RPV rupture,are predicted to occur. It is therefore concluded that there are no consequential ageing
effects or damage that occurs to the RPV during the long-term shutdown.

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3.

Irradiation assisted stress corrosion cracking. Neither SCC nor IASCC are relevant forreactor components exposed to low temperatures below 60°C (140°F) during a long-term
shutdown. Also, during shutdowns, while the reactor is not critical, there is no neutron
irradiation acceleration of SCC (i.e. IASCC) due to the absence of neutron fluence.

4.

Thermal ageing of CASS. Based on the existing body of research, for CASS plantcomponents where temperature exposure is below 250°C (482°F), no reduction in
fracture toughness is expected. The ageing effect of thermal embrittlement at lower
temperatures is not considered relevant and long-term ageing effects below 250°C(482°F) do not need to be considered for CASS components. Therefore, there is no ageing
effect associated with thermal ageing of CASS materials during a long-term shutdown.

3.7.1 Active research programmes
The Division of Research for Reactor System Safety, in the Secretariat of the Nuclear Regulation
Authority (NRA) will conduct a safety research project using materials from decommissioned
plants to confirm the adequacy (conservatism) of past results of safety research programmesapplied in the confirmation of the adequacy of the ageing management technical evaluationand the review of the extension of operational period [67]. As of March 2020, the S/NRA/R has
been conferring with utilities on the following items to start the research project since theirco-operation is essential [68]:•candidate units, candidate materials and available time frame, among other things;•

the possibility of disclosing the research information that utilities have already acquired
to avoid duplication of resources;•

legal issues such as transportation and the status of material like waste or researchmaterial.

Tables 3-1 and 3-2 show the candidate metal materials for this project. From 2020 to 2024
the plan of this research project for metal ageing has included testing or test preparation toevaluate RPV (reactor pressure vessel) embrittlement, fracture toughness of BWR core internals,
the effectiveness of preventive maintenance of BWR core internals and the integrity ofthermally aged CASS (cast austenitic stainless steel) [69].

Table 3-1: Candidate materials for first stage of R&D programmeMaterials

Structures, systems and componentsst

1 stage of R&D Programme (2020 – TBD)
Low alloy steel (BWR)RPVCASS (BWR)

Primary loop recirculation (PLR) pump casing

Stainless steels

PLR piping where induction heat stress improvement (IHSI) was carried out
Core shroud (BWR)

Table 3-2: Candidate materials for second stage of R&D programmeMaterials

Structures, systems and componentsnd

2 stage of R&D Programme (TBD – 2024)
Low alloy steel (PWR)

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RPV

CASS (PWR)

Main coolant piping

Ni-based alloy

Instrumentation tube where peening was carried out

Stainless steels

Baffle-Former bolt, core barrel

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3.7.2 PEO/LTO-OPEX and solutions
As one example of recent OPEX, on 16 February 2017, during the 17th inspection period of
Shimane Unit 2 (BWR/5), a crack was found in the weld joint (Alloy 82) of one of the access holecovers (AHCs) while performing a visual inspection of the RPV internals using an underwater
camera; CODAP Record #5168; see reference [70]. The inspection was part of the ageing
management technical evaluation for the 30th year of operation of the reactor unit.
Both access hole covers (including a part of the shroud support plate) were cut and removed
for detailed investigation. As a result of the investigation, cracks were found in the weld jointof both access hole covers. Moreover, the cause of the cracks was determined to be stresscorrosion cracking due to the overlap of hardening in the vicinity of the weld joint of the accesshole cove; water quality deterioration in the gap between the shroud support plate and theaccess hole cover and tensile residual stress in the vicinity of the weld joint of the access holecover Upon investigation of the shroud support plate side in the reactor, no cracks remained,
and the extent of crack propagation was confirmed. Therefore, an integrity evaluation of thecore support structure (shroud support) was performed, and it was confirmed that the core
support function was not affected. The corrective action involved replacements with new accesshole covers that do not have a welded structure (bolt fastening structure).

3.8 Korea
Korea has 24 nuclear power plants in commercial operation: 21 PWR units and 3 CANDU-6 units.
An additional four units are under construction. The oldest reactor unit, Kori-2 has been in
operation since 1983. A total of nine (9) reactor units have been in operation for more than 25 years.
Periodic safety reviews (PSRs) are conducted to address the cumulative effects of plant ageing,
operating experience and the evolution of science and technology. In particular, the assessmentand management of plant ageing is one of the major areas of work. It includes identification of
the systems, structures and components (SSCs) for ageing management, assessment of ageing
effects and planning of the ageing management implementation programme.

3.8.1 Ageing management technology reports
Prepared by Korea Hydro and Nuclear Power Co. (KHNP), two AM technology reports were
accepted by the Nuclear Safety and Security Commission (NSSC) in 2019. The specifictechnology report under Article 100 paragraph 1 of the Enforcement Decree of the Nuclear Safety
Act shall refer to reports containing the following information: (1) methodologies andcomputing codes related to technical information concerning the selection of a site for a nuclear
reactor facility, design, manufacture, construction, pre-service test, trial operation, operationand disassembly; (2) information concerning safety that can be applied repeatedly for an
identical purpose; and (3) information that provides the basis for preparing documents to be
attached to the application for a permit related to a nuclear reactor facility. Examples of ageing
management guidelines include:•

Integrated Guidelines for Management of Alloy 600 Locations (Review Period: 31 August
2015 until 30 June 2019). Cracks were continuously observed in the component with Alloy
600 materials; an example is the boric acid leakage in the reactor upper head of the Hanbit
Unit 3 in November 2012 (CODAP Event Record #4584). The NSSC required an integrated
ageing management plan for alloy 600 material of KHNP. Accordingly, KHNP establishedthe comprehensive management plan for the integrity of alloy 600 material. KHNP applied
for specific technology report approval in August 2015 and received the permission in 2019.
This report includes a non-destructive testing programme for each device using alloy 600
material, defect repair maintenance, preventive maintenance and water quality
management to prevent PWSCC.

•

Methodology for Fatigue Monitoring of ASME Class 1 Components in LWR Environments(Review Period: 8 September 2017 until 31 October 2019). Although the conservative
fatigue life design is applied to safety class 1 equipment, it is difficult to predict the actual

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fatigue life of equipment of a plant in LTO, because transients are not considered. In Korea,
a fatigue monitoring system is applied for the quantitative fatigue life management of
Safety Class 1 equipment as part of ageing management of the PSR. In July 2013 KHNP
applied for a specific technical subject report to establish and operate the Nuclear Power
Plant Fatigue Monitoring System (NuFMS) and was approved; see reference [71]. However,recent studies have shown that the fatigue life of equipment tends to decrease in thereactor coolant environment of LWRs compared to an air environment. Thus, the needs
for fatigue evaluation of Safety Class 1 equipment have been raised considering theenvironmental impact of the reactor coolant. In this regard, US NRC describes the use ofRegulatory Guide 1.207 Rev.01 9 and NUREG/CR-6909 [72], to reflect the Environmental
Fatigue Correction Factor in the air environment fatigue assessment. In Korea, to reflect
the environmental fatigue monitoring methods in the existing fatigue monitoring system(NuFMS), this specific technology report was issued and approved.

3.8.2 Statutes and requirements for periodic safety review of long-term operation
Periodic safety reviews (PSRs) are performed by the power plant operator (Korea Hydro andNuclear Power Co., KHNP) to assess the integrity and safety of operating nuclear power plants
in accordance with the Nuclear Safety Act. The PSRs are submitted to the regulatory body forevaluation. The Korean nuclear regulatory body consists of the NSSC and its technical support
organisation, the Korea Institute of Nuclear Safety (KINS). The highest law related to PSR is the
Nuclear Safety Act 10:•

Nuclear Safety Act – Article 23 (Periodic Safety Review)–

The operator of the nuclear power plant reactor shall periodically review the safetyof the nuclear power reactor and related facilities, and report the results thereof to
the Commission under the conditions as prescribed by the Presidential Decree.
(Intermediate omission)–

The Commission may, when the results of the periodic safety review referred to in
Paragraph (1) and the safety measures taken based on such results are deemed
insufficient, order the operator of the relevant nuclear power reactor to correct or
supplement such insufficiencies.

–

Matters concerning the method and contents of review, etc. in Paragraph (1) shall beprescribed by the Presidential Decree.

The details (timing, contents, etc.) of the PSR are described in the Enforcement Decree of the
Nuclear Safety Act (Presidential Decree), Enforcement Regulation of the Nuclear Safety Act(Ordinance of the Prime Minister). Each operator of a nuclear power reactor shall comprehensively
review the safety of the reactor facilities every ten years from the date they have obtained an
operating licence. The requirements related to ageing and degradation management are as below:
Article 37 11 of Presidential Decree, Paragraphs 1, 2 and 7.
•

Actual status of safety-related structures, systems, and devices;•

Ageing (referring to physical or chemical processes that will result in the degradation ofsystems, structures and equipment in a nuclear power plant over time and with use)management plans.

9.

NRC (2018), “Guidelines for Evaluating the Effects of Light Water Reactor Water Environments in Fatigue
Analyses of Metal Components”, RG 1.207, www.nrc.gov/docs/ML1631/ML16315A130.pdf.
10. Nuclear Safety Act, 대한민국 영문법령 (https://klri.re.kr).
11. Enforcement Decree of the Nuclear Safety Act (Presidential Decree), Article 37 “Details of Periodic Safety
Review”, 대한민국 영문법령 (https://klri.re.kr).

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The details above are described in Article 20 of the Ordinance of the Prime Minister,
Paragraphs (1) 2 and 7. Based on the law of PSR, the KINS provided the Safety Review Guidelines
for Periodic Safety Review.
The contents of ageing management are located in Chapter 7 of this guideline. The targets ofageing management are the concrete containment structure; the nuclear safety related concrete
structures; and nuclear facility components. To review the ageing mechanism and phenomenonof each facility, these documents are referenced: KEPIC MNZ App. W (corresponding to or ASMEBPVC Section III, Appendix W), IAEA SRS-82 (IGALL), NUREG-1801(GALL).
The evaluation target components of a PWR are the:–reactor vessel;–

reactor internals;–

control element drive mechanism(CEDM);–

steam generators;–pressurisers;–

reactor coolant piping;–reactor coolant pumps;–

supports and pipe whip restraints;–

valve system;–pump system;–

pressure vessels;–

heat exchangers;–

HVAC components;–

diesel generator system;–

fire protection system;–

secondary side water system; and–turbine system.

3.8.3 Ongoing R&D related to ageing management at KINS
The project called Regulatory Technologies for Ageing Management of Crack Defect in Operating
Nuclear Power Plants, which was started in 2018, is in development. The objective of theresearch is to develop regulatory technologies for ageing detection, integrity assessment andrepair methods of major nuclear components in long-term operation. The scope of this projectis as follows:•••

Development of regulatory techniques for crack evaluation and ageing management:–

Residual weld stress to dissimilar metal weld parts considering the phasetransformation;–

Ultrasonic crack signal analysis techniques based on artificial intelligence;–

Non-linear ultrasonic technique for cast austenitic stainless steel (CASS) thermal
ageing.

Development of regulatory technologies for component integrity assessment considering
environmental factors:–

Reactor vessel safety assessment considering long-term irradiation;–

Damage evaluation of reactor vessel internals;–

Development of regulatory tools for probabilistic analysis.

Development of regulatory technology for weld integrity and repair weld safety in reactor
components.

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3.9 The Netherlands
The Netherlands has a single nuclear power plant in operation, the two loop Siemens PWR of
Borssele. This power plant went into operation in 1973 and was granted a LTO licence in 2013
for a lifetime extension of 20 calendar years, bringing the lifetime from 40 to 60 years. Currently,
there is no plan to extend the lifetime of the power plant further than 2033. The Dutch Nuclear
Energy Act states in article 15a that the power plant of Borssele will operate until 2033 with nopossibility of licence renewal.
In the LTO licence, the subject of ageing is broadly addressed. The licence contains somespecific requirements for the licensee, the most relevant of which are listed below:•

To submit to the regulatory authority an implementation plan regarding the extension
and improvement of the existing ageing management programme with extra ageing
management measures specifically thought for the period between 40 and 60 years of
the lifetime of the reactor. The implementation plan should be based on thecomprehensive Ageing Management Review for passive SSC important to safety, the PSR
assessment of Safety Factors 10 (Organisation) and 12 (Human Factor), the outcome of
the SALTO PR in 2012 and the regulatory review.

•

To enhance the existing ISI-programme with specific items to be performed in the firstyears of LTO.

•

To assure that active components important for safety are adequately addressed in thepreventive maintenance or monitoring programme of the plant.

•

To confirm the safety margin of the RPV against brittle fracture with additional
surveillance capsules.

•

To confirm the safety margins against fatigue failure for specific components andperform yearly monitoring of fatigue loading with a dedicated monitoring system.

•

To carry out yearly monitoring and determination of the residual lifetime of electric
equipment for design base accidents to preserve the qualification of this equipment.

3.9.1 Operating experience during LTO
RPV Irradiation Surveillance Programme (SOP)
According to the conditions in the LTO licence, the licensee must estimate the safety margin tobrittle fracture of the RPV by means of an additional irradiation surveillance programme, withthe goal of evaluating the transition temperature to brittle behaviour for 60 years of operation.
To estimate the changes in material properties of the RPV, several samples of the ferritic steelof the vessel material were irradiated in the reactor and retrieved after the desired equivalent
lifetime for destructive testing. Specimens from the base materials of the RPV beltline as well
as from the weld material of the circumferential core weld of the RPV were tested. The
harvesting and testing of the specimens was carried out following the guidelines of the German
standard KTA 3203, issue 11/17. The second specimen set, irradiated until an equivalent lifetimeof at least 55 EFPYs, was retrieved in 2018 and the results of the destructive testing are currently
under review by the regulatory authority. So far, results are encouraging: Based on the neutron
flux average over cycles with full low leakage core loading and low leakage with MOX fuelassemblies, the maximum fluence value at the most irradiated position of the RPV after
55 EFPYs is determined to be 3.20 x 1019 cm-2. An evaluation of uncertainties shows that even in
the most conservative case the fluence will still be smaller than the RPV design fluence of
3.5 x 1019 cm-2. To determine the brittle to ductile reference temperatures both Charpy V- andthree-point bending specimens (Master Curve) were used. With both concepts (RTNDT and RTT0)high safety margins have been proven for 60 years of operation.

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Degradation of baffle-to-former bolts
Baffle-to-former bolts (BFBs) are part of the reactor vessel internals of a pressurised water
reactor. These bolts connect the vertical baffle plates that frame the core region to the horizontal
formers, thus forming a connection between the baffles and the core barrel. While all these bolts
are mostly addressed as baffle bolts in the international community, a distinction is generally
made between baffle-to-former bolts (BFBs) and core barrel-to-former bolts (CBBs) in Siemens/
KWU plants. The region between the barrel and the baffle plates is cooled by the primary water,
flowing through holes in the formers. In the Dutch power plant of Borssele (Siemens/KWU
design) the formers, and thus the bolts, are cooled with an upward flow. The plant is equippedwith seven formers and two sizes of baffle bolts: M16 bolts in the first and the seventh former
and M12 bolts in all the others.
The baffle-to-former bolts of the Borssele Nuclear Power Plant have a history of degradation
due to SCC. The bolts where originally made of nickel alloy X-750. In response to the operational
experience of other S/KWU reactors in the 80s, the bolts were replaced in 1988 with new bolts
made of austenitic stainless steel 1.4571, which is comparable to 316Ti steel and should haveless sensitivity to stress corrosion cracking. The lowest row of bolts (in the seventh former) was
not replaced. In 2006 degradation of the replaced bolts was observed via visual inspection. Thesefindings led to the replacement of 39 bolts in the same year. The design and material of the newbolts has not been changed.
It has been confirmed that Inter-Granular SCC was the leading mechanism in the
degradation of the S/KWU bolts. Indeed, sensitisation at the grain boundary, even though
reduced by using titanium, is possible. In general, this corrosion mechanism is enhanced by the
local presence of impurities, such as chloride and sulphate, and of dissolved oxygen. As such,dissolved oxygen is hardly present in the primary water. However, residual amounts of oxygen
could be introduced in the primary circuit, during refuelling for instance. Local conditions (lowflow rate, accumulation of dissolved oxygen) will favour radiolysis and then create an oxidising
environment in the shaft, promoting IGSCC.
At KCB, Oxygen is kept at a low level during refuelling and gassed out when restarted, avoiding
then the use of hydrazine. Currently the baffle–to-former bolts are visually inspected every three
years. There are several damaged bolts and they are mostly concentrated on the highest positions(above the core). The next visual inspection is planned during the refuelling in 2020.
The two-loop PWR of Borssele is one of the earlier reactors built by Siemens/KWU. The
design as built is unique, limiting thus the comparison with other nuclear power plants. The
design of the core structure is, however, based on the Westinghouse design and is also used in
the other nuclear power plants of the Siemens/KWU reactor fleet. With the “Atomausstieg” inGermany, the number of reactors with identical baffle-to former bolts is reduced.
The Dutch nuclear regulatory authority is setting up a supervision strategy on this specifictopic to challenge the licensee regarding the chosen inspection methodology and its frequency.

3.10 Slovak Republic
In the Slovak Republic, there are currently four nuclear units of type WWER 440/V213 inoperation, two units at the Bohunice site (EBO V2) and two units at the Mochovce site (EMO).
Two additional WWER 440/V213 units at the Mochovce site are under construction (EMO V2).
The EBO V2 units were commissioned in 1984 and 1985, respectively. The two EMO units werecommissioned in 1998 and 1999, respectively.

3.10.1 Regulatory AM framework
A legal framework for ageing management is established in regulatory Decree No. 33/2012 on
Periodic Safety Review. Referring to AM, the regulatory authority issued safety guide BNS
I.9.2/2019 “Ageing Management of Nuclear Power Plants, Requirements”, which specifies andsupplements the requirements of Decree No. 33/2012. This guide is based on IAEA and WENRAA REVIEW OF OPERATING EXPERIENCE INVOLVING PASSIVE COMPONENT MATERIAL DEGRADING IN PERIODS OF EXTENDED AND LTO, NEA No. 7614, © OECD 2024

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recommendations and provides guideline on development, establishment and implementationof ageing management programmes for safety-related SSCs.
Based on the regulator Decision No. 1012/2013, the licensee is obliged to submit on a yearlybasis the following information referring to the lifetime of main components:•

Remaining service life of the reactor pressure vessel and main components, including
critical temperature of the brittle fracture of the RPV.

•

Assessment of the critical brittle temperature of the RPV based on surveillance
specimens programme.

•

Results of in-service inspection programmes.

The effectiveness of the ageing management process is evaluated every ten years within
the Periodic Safety Review (PSR). The licensee is required to implement an action plan forcorrective actions resulting from the PSR.

3.10.2 AM approach
For implementation of ageing management, an operating organisation issued themethodological guide “Ageing Management of SSC in NPPs”. This guide – an overall ageing
management programme – is valid for all nuclear units in the Slovak Republic (both in operation
and under construction). The guide is consistent with the relevant IAEA Safety Standards(SSG-48 “Ageing Management and Development of a Programme for Long-term Operation of
Nuclear Power Plants”) and WENRA SRL and defines roles and responsibilities which arenecessary for proper implementation of ageing management.
For specific SSCs, the individual ageing management programmes are developed. Theseageing management programmes are also valid for all nuclear power plant units in the Slovak
Republic. Individual ageing management programmes are developed with consideration of a
60-year operation time frame.

3.10.3 Systems, structures and components subject to AM
According to national requirements, all safety important SSCs in safety class one to four are
subject to ageing management. In addition, the operator can enlarge the scope of SSC subject
to AM, e.g. taking into account the economical viewpoint. For each nuclear power plant, lists ofsystems, structures and components for ageing management are developed.

3.10.4 Component-specific AMPs
For specific SSCS, the operating organisation developed and implemented individual ageing
management programmes. A list of these component-specific amps is provided below:

70•

reactor pressure vessel;•

steam generators;•

main circulation pumps;•

main isolating valves;•primary piping;•pressurisers;•

secondary piping;•essential service water piping;•main condensers;

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reactor internals;•

electrical cables;•

reactor building;•diesel generator station;•

forced draft cooling towers;•

building of chemical treatment water of nuclear power plant EBO V2;•

auxiliary building;•turbine hall and basement of turbo-generator.

The specific type of AMP which is not component-oriented is the amp for monitoringcorrosion of SSCS.

3.10.5 PEO/LTO-OPEX and solutions
A current technical issue being addressed for the WWER 440/V213 is the modification ofdissimilar metal welds (DMWs) of steam generators in the Bohunice Nuclear Power Plant. DMWs
are specific joints between the stainless-steel piping and the carbon steel steam generator body.
A general problem associated with DMWs after 30 years of operation is the difficulty of making
repairs and their relatively short service life. Modification is focused on diameters from 18 to
90 mm, and it includes removal of the original DMWs and replacement with new weldmentswith dissimilar weld metal. The required quality of weldments with DMW will be achieved byusing a unique welding procedure (production in workshop by machining devices). In addition,
protection of the weld root from the secondary medium will be achieved by application of aprotective nickel layer. The above-mentioned factors have ensured the required reliability of
new DMWs during long-term operation. The modifications are planned from 2021 to 2025.
Also, a modification of DMW of feed water pipeline with diameters of 270 mm inside thesteam generator is being prepared. In this case, the original DMW will be removed by the flange
joint according to the solution applied and verified on steam generators of WWER 440/V213
units in other European countries, i.e. Czechia and Hungary.

3.11 Spain
Since the 1990s, considerations of ageing management have led to a series of recurrent activities
at the Spanish plants that have arisen from the operation of the plant itself and to respond to
regulatory requirements. These requirements were initially established based on the specificconditions set out in the operating licence and later, from 2009, in CSN Safety Instruction IS-22 12,
which includes the requirements for ageing management and long-term operation.
The main objective of the first ageing management activities applied by the Spanish plants
was to preserve the option of extending operation beyond the foreseen design lifetime, withsimilar approaches to those implemented at US pilot plants for the application of 10 CFR 54(“Requirements for Renewal of Operating Licensees for Nuclear Power Plants”). 13 The
requirements established in IS-22 apply to all plant operating conditions, from initial start-up
to definitive shutdown, for which reason common and specific requirements are defined foreach of the two plant lifetime phases, i.e. design lifetime of the facility (40 years) and long-term
operation (LTO).

12. CSN Instruction IS-22, Revision 1, of July 1st 2009, on safety requirements for the management of ageing
and long-term operation of nuclear power plants, https://piramidenormativa.sne.es/Repositorio/CSN/is-22i.pdf.
13. Available on the NRC website at www.nrc.gov/reading-rm/doc-collections/cfr/part054.

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COUNTRY/ECONOMY-SPECIFIC OVERVIEWS OF MATERIAL DEGRADATION STATE-OF-KNOWLEDGE, RESEARCH AND OPERATING EXPERIENCEAs shown below, the Spanish fleet of operating nuclear power plants consists of seven units
at five sites. None of the operating units has reached LTO, but all of them have been operating
more than 30 years and some are close to LTO:•

Trillo-I (PWR-KWU 3-LP), first criticality in May 1988;•

Vandellós II (PWH 3-LP), first criticality in November 1987;•

Cofrentes (GE BWR/6), first criticality in August 1984;•

Ascó (WH 3-LP), Unit I first criticality in June 1983 and Unit II in September 1985;•

Almaraz (WH 3-LP), Unit I first criticality in April 1981 and Unit II in September 1983.

Two Spanish nuclear power plants have reached the definitive shutdown phase: Santa Maríade Garoña Nuclear Power Plant (GE BWR/3 Mark I, was permanently shut down on 2 August 2017)
and José Cabrera-1 (Zorita Nuclear Power Plant, PWR 1-Loop-Westinghouse was permanently shut
down on 30 April 2006).

3.11.1 Zorita reactor internals research project (ZIRP)The ZIRP project was an R&D project in which the CSN participated together with entities fromother countries (EPRI Materials Reliability Program (MRP), US NRC, SSM (Swedish Radiation
Safety Authority), Tractebel, and AXPO [operator of Beznau Nuclear Power Plant]). The project
has assessed the state of stainless steel Type 304 corresponding to certain reactor internals of
the Zorita Nuclear Power Plant (1968-2006, 26 EFPY) to characterise the effects of neutron
irradiation on the mechanical and microscopic properties of stainless steel, irradiated underservice conditions, in order to increase awareness of the effects of irradiation towards the endof the life of the nuclear power plants (40 and 60 EFPY, and beyond). Therefore, material samples
were extracted from the baffle plates inside the Zorita RPV, from which a series of specimens
were manufactured and have been subjected to the different characterisation tests. The
different specimens that have been obtained from the extracted material had irradiation doses
ranging from about 10 dpa to almost 50 dpa.
After an initial engineering study carried out in Spain (Tecnatom developed the vesseltemperature analysis and Gas Natural Fenosa performed neutron creep analysis), the material
that was extracted from the Zorita vessel was transported to the laboratories of Studsvik(Sweden) where mechanical tests were carried out, while microstructural tests were carried out
at MHI laboratories in Japan (samples were sent from Studsvik to MHI). Itemised below are thematerial tests performed to date:•

Mechanical testing:
–

–

–

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Tensile testing:–Tensile testing of 10, 30 and 50 dpa material at room temperature and 320°C;–

Significant radiation hardening observed at all doses.

Fracture toughness:–

Fracture toughness testing of 9, 24 and 35 dpa material at 320°C in air and 9 dpa
material also in PWR environment at 320°C;–

Low fracture toughness values observed;–

PWR environment showed slight reduction in fracture toughness.

Crack initiation:–

Crack initiation testing of UCL and O-ring specimens of ~40-45 dpa material to
9 000 hours;–

No failures of O-ring specimens;

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Failure of UCL specimens ranging from 60% to 100% YSirr with failure timeincreasing with decreasing stress.

Crack growth rate testing:–

Crack growth rate testing of 9, 24, and 41 dpa material at three temperaturesand 2 K levels in a PWR environment;–

Low crack growth rates with no apparent K dependency;–

Rapid crack growth rates after decrease in test temperature or operational upset(chemistry or load).

Microstructural testing:
–

–

–

–

Void swelling:–

Transmission electron microscopy (TEM) examination of nine areas including
highest temperature and highest dose regions;–

Maximum average swelling value was 0.08% at the location of the highest
temperature.

Microstructure:–

Grain size typical of Type 304 stainless steel;–Gamma prime phase and Frank loop sizes and densities quantified;–

Precipitates analysed.

Hardness:–

Vickers hardness testing (1.0 kg and 0.1 kg) of three areas;–

Results typical of irradiation-hardened Type 304 material.

Gas content:–

Hydrogen (thermal conductivity) and helium (mass spec) gas analyses of
specimens from three locations;–

Microstructural bubbles.

The ZIRP project was scheduled to be completed at the end of 2016, but there were somedelays in the Studsvik testing and some of them were finished during 2018. EPRI issued the final
report at the end of 2019, with which the project is fully concluded (Materials Reliability
Programme: Zorita Internals Research Project [MRP-440] Testing of Highly-Irradiated Baffle Plate
Material 3002016015 Final Report, October 2019) 14.
The project has been a pioneer in its field since, at the time of its initiation, the propertiesof this steel in real irradiated state (not in experimental reactor) were not available, or almost
not available worldwide, at doses up to approximately 50 dpa. The project lasted almost 15 years,
within which the characterisation tests represented only a relatively small part. The large
number of years required to perform the project was due to, among other factors, thecomplexities associated with the participation of different nations, the need to plan all the steps
properly in advance due to irradiated material handling, and, especially, the difficulties
associated with the experiments themselves.
In relation to the results obtained, the tests constitute an important milestone in theknowledge of the properties of the irradiated material (of type 304 stainless steel) and are ofgreat interest, especially for those countries that plan to operate their nuclear power plants in
LTO (beyond the original design life).

14. The report is available at no cost to funding project members only.

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3.11.2 ESW inspection plan
All Spanish nuclear power plants have implemented dedicated inspection plans (IPs) to monitor
and control the degradation phenomena affecting the essential service water (ESW) piping, bymeans of an ageing management programme (AMP) based on the NUREG-1801 AMP.XI-M20“Open-cycle Cooling Water System.” Each Spanish nuclear power plant has ad-hoc IP based on
the defects found due to each degradation mechanism, which guarantees the functionality,
operability and structural integrity of the ESW system.
As a result of material degradation due to corrosion, MIC, galvanic corrosion, or corrosion
under deposits, all the Spanish nuclear power plants have currently established IPs using UT or
PAUT (Phased Array UT) to monitor pipe thickness, both in areas already inspected and in new
areas, to detect the growth of the indications and/or the appearance of new indications. To
control and mitigate the loss of material, the IPs consider the previous OPEX and identifies themost susceptible locations, including both preventive and corrective actions, such as:•

Selection and prioritisation of lines, following the EPRI Technical Report 1010059 “Service
Water Piping Guideline” [73].

•Improvement of the monitoring and control of degradation mechanisms. This includes
inspection equipment, inspection criteria and inspection processes. Also included arecontingency measures in case of active leakage, flaw evaluation strategies, repair
strategies and validation of new ASME Code Cases.

•

Monitoring of the evolution of pipe wall loss and consideration of using less susceptible
material system operational changes to reduce the progression of pipe wall loss, using
longer circulation times through piping, and changes to valving operations.

Based on the minimum thickness established, the IP defines ranges of remaining thickness
and establishes actions to be taken in relation to the thickness measured, as well as sample
extensions, evaluations (ASME Code Case N-513-3/4, “Evaluation Criteria for Temporary
Acceptance of Flaws in Moderate Energy Class 3 Piping”) or substitutions, temporary repairs andinspections in replaced, recharged or repaired sections. Laboratory analyses are carried out todetermine the cause of the defect when applicable.
There is also continuous monitoring of the corrosion rate by means of corrosion probes,
spools, or corrosion coupons, and where it is possible, chemical and biological treatment of thesystem and operational practices have been optimised. In some cases, the inner lining of thesections containing stagnant water will be carried out as well as inspections after thesemitigation measures are considered.

3.11.3 Vibrational fatigue AMP
In response to a socket weld failure that occurred in 2018 (CODAP Record #5026), some licensees
have developed an augmented non-destructive examination programme for small-diameter
piping susceptible to high-cycle fatigue. The programme is to be implemented prior to the LTO.
The new inspections will be included in the AMP based on GALL AMP-XI.M35 15 “One-time
Inspection of ASME Code Class 1 Small-Bore Piping.” All Spanish nuclear power plants aredeveloping a validated and qualified procedure to inspect socket welds. Additionally, a specific
AMP for vibrational fatigue is being developed in some Spanish plants, which will include thefollowing inputs:•

The most susceptible locations within the reactor coolant system (RCS), chemical andvolume control system (CVCS) and emergency core cooling system (ECCS).

•

Inspection of socket welds in ASME III Code Class 1, 2 and 3 pipelines of diameter < DN100.

15. NRC (2010), “Generic Aging Lessons Learned (GALL) Report”, NUREG-1801, Rev. 2, US Nuclear Regulatory
Commission.

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Proximity assessment of rotating/pulsating equipment to susceptible lines wheretransmission of vibrations can favour high-cycle fatigue (HCF) cracking.

•Prioritisation in the selection of non-insulated lines in case of fatigue cracking in orderto facilitate repair without affecting the plant operability.

3.12 Switzerland
Three different reactor types are in commercial operation in Switzerland: the GE-BWR/6, theKWU/Siemens 3-loop PWR, and the Westinghouse 2-loop PWR. A fourth reactor type, a BWR/4
at KKW Mühleberg, was permanently shut down at the end of 2019 and is undergoing
de-commissioning. Summarised in Figure 3-5 is the international piping OPEX applicable to thefour NSSS-types. In 1991 the Swiss regulatory body implemented requirements for plant ageing
management. Subsequently the GSKL (Group of Swiss Nuclear Power Plant Operators) initiatedthe development of plant-specific ageing management programmes.

Figure 3-5: Piping OPEX as a function of plant age and NSSS-design180
BWR - BWR/4
BWR - BWR/6
PWR - Siemens/KWU
PWR - WH-2LP

160

No. of pipe failures

140
120
100
80
60
40
20
0
25-29

30-34

35-39

40-44

45-49

Age of plant [years] at the time of pipe failure

3.12.1 State of knowledge regarding degradation mechanisms during LTO
The degradation mechanisms considered to be of most relevance to LTO include fatigue
mechanisms, irradiation embrittlement, stress corrosion cracking (SCC) and flow-accelerated
corrosion (FAC). Other mechanisms like corrosion, wear, erosion and thermal ageing
embrittlement are also taken into account.
For the monitoring and assessment of fatigue, three of the four Swiss nuclear power plants
are using automated fatigue monitoring systems: FatiguePro 16 at Leibstadt (KKL) since 2007,

16. IAEA (n.d.), Advanced Approach For Fatigue Monitoring, IAEA-CN-194-040, https://inis.iaea.org/
collection/NCLCollectionStore/_Public/43/070/43070840.pdf?r=1&r=1

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COUNTRY/ECONOMY-SPECIFIC OVERVIEWS OF MATERIAL DEGRADATION STATE-OF-KNOWLEDGE, RESEARCH AND OPERATING EXPERIENCEWESTEMSTM 17 at Beznau-1/2 (KKB) since 2002 and FAMOSi 18 at Gösgen (KKG) since 2014.
At Mühleberg (KKM), fatigue relevant data were recorded followed by a manual data evaluation.
The fatigue evaluations combine stress-based and cycle-based approaches. The stress-based
approach is done using measurements of temperature-time histories to identify transients.
Prior to the installation of fatigue monitoring systems, a simpler and more conservative cyclebased approach was used. This is referred to as baseline fatigue and is estimated from thenumber of transients based on design specifications.
Environmental effects on fatigue must be considered according to ENSI-B01 [74]. The resultsof the fatigue monitoring in the form of current fatigue usage factors and the corresponding
levels extrapolated to 60 years of operation are submitted to ENSI in an annual report. Currently,
the results of the fatigue monitoring indicate that the long-term operation of the Swiss nuclear
power plant is not subject to any limitations because of fatigue. Potential future challenges arepreventive actions to reduce the probability of small-diameter piping fatigue damage caused byvibrations and, possibly, operating plants in a load-following mode 19.
Another important issue related to LTO is irradiation embrittlement of reactor pressure
vessel materials. The ageing management for embrittlement includes the assessment of
neutron fluence, which is reduced by use of low leakage core (LLC) configuration, and the testingof surveillance specimens. The determination of Adjusted Reference Temperature (ART) isbased on the requirements of ENSI-B01. This guideline references three ART determination
methods. Method I is based on US NRC Regulatory Guide 1.99 Rev. 2 20, and Methods IIA and IIB
are based on the master curve (MC) approach according to ASTM E 1921 21. Method IIA is theapplication of MC approach for irradiated material, while method IIB combines the MC approachfor unirradiated material with the “Charpy-shift model.”The embrittlement has to meet the so called “DETEC 22 Criteria” (ART < 93°C in ¼ wall depth,
Charpy energy ≥ 68 J). For the PWR units pressurised thermal shock (PTS) analyses are performed
by deterministic fracture mechanics studies. The current PTS-Analyses are guided by the KTA
3203 methodology. 23 For the normal operation, the licensees must specify pressure-temperature
limiting curves. The P-T-limiting curves are reviewed and, if necessary, updated periodically.
Stress corrosion cracking (SCC) remains an important issue with respect to LTO.
Components affected by SCC in the past included the core shroud in Mühleberg, the baffle bolts
in Gösgen as well as in Beznau Units 1 and 2, the RPV head penetration welds in different plants,
the dissimilar metal welds at the pressuriser in Gösgen and the N5 RPV core spray nozzle inLeibstadt. Some of these components were substituted by less susceptible materials. One
mitigation action, especially for BWRs, is the application of hydrogen water chemistry combinedwith noble metal injection. All relevant areas that are susceptible to SCC are part of the periodic
inspection programme.

17. NRC (2012), “WCAP-17577”, Revision 0 (Topical Report on ASME Section III Piping Fatigue Analysis
Utilising the WESTEMSTM Computer Code), US Nuclear Regulatory Commission, www.nrc.gov/docs/
ML1206/ML12061A377.pdf.
18. NDT net (2012), “Structural health monitoring of power plant components based on a local temperature
measurement concept”, www.ndt.net/article/ndtnet/2013/65_Rudolph_Rev1.pdf.
19. IAEA (2012), Non-baseload Operation in Nuclear Power Plants: Load Following and Frequency Control Modes of
Flexible Operation, Nuclear Energy Series No. NP-T-3.23, www.iaea.org/publications/11104/non-baseloadoperation-in-nuclear-power-plants-load-following-and-frequency-control-modes-of-flexibleoperation (Section 5.2 of the IAEA report addresses “Phenomenological Aspects of Flexible Operation.”)
20. NRC (1988), “Radiation Embrittlement of Reactor Vessel Materials”, RG 1.99, www.nrc.gov/docs/ML0037/ML003740284.pdf.
21. Standard Test Method for Determination of Reference Temperature for Ferritic Steels in the Transition
Range.
22. Federal Department of the Environment, Transport, Energy and Communications.
23. KTA (2017), “Surveillance of the Irradiation Behaviour of Reactor Pressure Vessel Materials of LWR
Facilities”, Nuclear Safety Standards Commission (KTA), KTA-3203 (2017/11); Salzgitter, Germany,
www.kta-gs.de/e/standards/3200/3203_engl_2017_11.pdf.

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Consideration of flow-assisted corrosion (FAC) is integrated into the periodic inspection
programmes. The wall thickness of potential affected areas of these piping is measured byultrasonic testing.

3.12.2 Regulatory AM framework
The Swiss ageing management framework is based on the Nuclear Energy Act 24, Nuclear Energy
Ordinance 25, DETEC Ordinance “Suspension of a NPP Operating Licence”, and the IAEA Specific
Safety Guide SSG-48 [26]. Results of other programmes such as IGALL [36] and national andinternational OPEX are also considered. The operating licence for the Swiss nuclear power
plants is not limited to a fixed period, but there are conditions for suspending plant operation
if certain requirements are not met. The periodic safety review, when reaching 40 years ofoperation, must include an LTO safety evaluation based on licensees’ safety cases. The
regulation requires continued backfitting measures to improve nuclear safety or to reduce
nuclear risks. The required TLAAs include the following issues:•

RPV Integrity (irradiation surveillance programme, neutron fluence assessment, PTS
analyses) based mainly on ENSI-B01, US NRC Regulatory Guide 1.99 Rev. 2 and KTA 3203 26.

•

Fatigue analyses for pressure vessels and piping, mechanical and thermo-mechanicalbased on ENSI-B01, 10 CFR 50.54 27, NUREG/CR-6909 [72] and NUREG-1800 [75].

•Leak-before-break analyses for pressure vessels and piping based on NUREG 0800
SRP 3.6.3 28, NUREG-1061 [76], Reg. Guide 1.45 29 and KTA 3206 30.

Other TLAAs are safety analyses for the nuclear steam supply system, corrosion monitoring
and corrosion assessment of the steel containment, and structural analyses of the concreteparts of the containment.
With the ratification of the Nuclear Energy Act (KEG) and the Nuclear Energy Ordinance(NEO) in 2005, more detailed conditions for AMP were established at the legislative level. The
KEG and the NEO are accompanied by the ordinance for safety-classified vessels and piping in
nuclear power plants, and the DETEC Ordinance on “Suspension of a NPP Operating Licence” 31,
which became effective in 2008.
The latter requires that a nuclear power plant be shut down if the integrity of the physicalbarriers is compromised by ageing mechanisms. The definition of non-acceptance criteria is,for the RPV embrittlement, an adjusted reference temperature of ≥ 93°C or an upper shell Charpy
energy < 68 J, for degradation of primary piping or of the primary containment a wall thickness
below the original design thickness, and for primary piping the violation of allowable fracture
mechanics-based limits by cracking. For LTO the licensee has to demonstrate compliance withall these criteria for ageing over the expected operation time by TLAA. The Swiss Regulator ENSI
has issued different regulatory guidelines related to LTO. In particular, these are the regulatory
guideline on Ageing Management ENSI-B01 [74] in 2011, the regulatory guideline on Periodic
Reporting ENSI-B02 [77] in 2008, and the regulatory guideline on Periodic Safety Review (PSR)
ENSI-A03 in 2014 [78].

24. Kernenergiegesetz (KEG) 732.1, 2003, www.fedlex.admin.ch/filestore/fedlex.data.admin.ch/eli/cc/2004/723/20230901/de/pdf-a/fedlex-data-admin-ch-eli-cc-2004-723-20230901-de-pdf-a-1.pdf.
25. Swiss Nuclear Energy Ordinance, 2004, www.fedlex.admin.ch/eli/cc/2005/68/en.
26. KTA (2017), “Surveillance of the Irradiation Behaviour of Reactor Pressure Vessel Materials of LWR
Facilities”, KTA 3203 (2017/11), Nuclear Safety Standards Commission (KTA), Salzgitter, Germany.
27. US NRC Regulations, Title 10, Code of Federal Regulations, Section 50.54, Conditions of Licenses.
28. NRC (2007), “Leak-Before-Break Evaluation Procedures”, www.nrc.gov/docs/ML0636/ML063600396.pdf.
29. NRC (2008), “R.G. 1.45: Guidance on Monitoring and Responding to Reactor Coolant System Leakage”,
Rev. 1.
30. KTA (2014), “Break Preclusion Verifications for Pressure-Retaining Components in Nuclear Power
Plants”, KTA 3206 (2014/11), Nuclear Safety Standards Commission (KTA), Salzgitter, Germany.
31. See www.bfe.admin.ch/bfe/en/home/supply/nuclear-energy.html.

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COUNTRY/ECONOMY-SPECIFIC OVERVIEWS OF MATERIAL DEGRADATION STATE-OF-KNOWLEDGE, RESEARCH AND OPERATING EXPERIENCEENSI-B01 defines the specific requirements for AMP. The scope of the SSCs to be covered in
the AMP includes all safety classified structures and electrical components and all safety class 1
to 3 vessels, piping and related mechanical components (e.g. pumps, valves, supports andattachments). It defines the methods used for the verification of the embrittlement resistanceof the RPV (mandatory time-limited ageing analyses, or TLAA, Annex 5) and the scope of thecomponents to be covered in the fatigue monitoring and assessment programme (mandatory
TLAA, Annex 6). Further it requires basic AMP documents to be prepared by the licensees. These
are generic industry guidelines (by GSKL, the Swiss Nuclear Power Plant Owners Group), the
KATAM (catalogue of ageing mechanisms) and the so-called fact sheets (“Steckbriefe”), whichsummarise the basic data, tables of ageing mechanisms relevant for the individual components,
the component history, and the periodical updates to be done by the licensee. The KATAM is a
proprietary document prepared by the GSKL. The KATAM is a comprehensive compendium ofpotential materials ageing mechanisms for mechanical equipment. Originally issued in 1991,
the KATAM, which is in German only, is updated periodically.
ENSI-B02 defines the requirements for annual reporting. This includes an overview of changes
performed in the AM-documents, the results of the fatigue surveillance programme, the results of
the review of the operating experience and the follow-up of the current state of science andtechnology, a description of the consequences of these results on the AMP and the evaluation of
the efficiency of the AMP based on failure statistics and maintenance indications. The guideline
ENSI-A03 defines additional requirements on reporting within the PSR. This includes a review ofthe design and qualification of the in-scope SSCs, an assessment of technological obsolescenceand the review of existing time limited ageing analyses with respect to LTO.
Based on the results of the Swiss participation in the ENSREG topical peer review process 32 33[79] and the IAEA Specific Safety Guide SSG-48 [16] for AMP, ENSI will review the regulatory
guidelines ENSI-B01 and ENSI-B02.

3.12.3 Significant technical issues related to PEO/LTO
From a Swiss regulatory perspective, the most relevant technical issues concern embrittlement
for the older reactors, especially the PWRs, fatigue including vibrational fatigue, SCC and FAC.
A key challenge is the implementation of updated regulatory requirements (like increased
hazards or safety margins). The AMP addresses all currently relevant issues and will be
periodically reviewed based on the state-of-the-art of technology and on OPEX. The guidelines
ENSI-B01 and ENSI-B02 will be reviewed to determine the need for updates consistent with IAEA
SSG-48 and the results of the ENSREG topical peer review 2017.
Currently, irradiation induced void swelling or creep is not an issue and there is no ongoing
research programme. The phenomenon is mentioned in the KATAM, but not addressed separately.
ENSI intends to monitor the state-of-the-art knowledge on the issue and the international OPEX.
If necessary, it will be included into the AMP.
With respect to reactor internals, the applied embrittlement trend curve (ETC) is the one
defined by US NRC Regulatory Guide 1.99 Revision 2. It was implemented in the Swiss regulatory
framework in 2008 through the DETEC Ordinance “Suspension of a NPP Operating Licence” andin 2011 through ENSI-B01. The ETC of RG 1.99 was used prior to 2008, but it was not a mandatory
requirement. If surveillance data are available (which is the case for all Swiss nuclear power
plants), ETC is used only for interpolating/extrapolating these results.

32. EU Topical Peer Review 2017 on “Ageing Management of Nuclear Power Plants”, www.ensreg.eu/news/
eu-topical-peer-review-2017-ageing-management-nuclear-power-plants.
33. The plant-specific Topical Peer Review reports are available at www.ensi.ch/en/documents/documentcategory/gutachten-ensi.

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3.12.4 Active research programmes
The ongoing research projects funded by the Swiss regulator are the projects LEAD and
PROACTIV, performed by the Paul Scherrer Institut (PSI). The NORA III project (Noble Metal
Deposition Behavior in Boiling Water Reactors) was completed in 2018. Furthermore, ENSI aswell as other Swiss organisations participate in the PIONIC and the IAEA IGALL projects.
The LEAD Project (LTO concerns environmental assisted degradation), scheduled to run
from 2018 to 2020, is a continuation of the SAFE and SAFE II projects 34. The research topics are
SCC initiation in austenitic ni-alloys and stainless steel, environmental effects on fracture andon fatigue and investigations of synergies and superposition of ageing mechanisms. The project
has connections to other research activities like the EU Horizon 2020 projects MEACTOS 35 and
INCEFA+ 36 and the ICG-EAC Round Robin on SCC initiation 37 . Results from the forerunnerprojects significantly affected the technical content of Revision 2 of BWRVIP-233 [80].
The PROACTIV project, scheduled to run from 2019 to 2021, is a follow-up of the PROBAB
project and its predecessors PISA-I and II 38. It investigates “Probabilistic Integrity Assessments
for Critical RPV Locations and Piping” under consideration of active degradation mechanisms,
experimental validation of XFEM with respect to application to crack growth simulations andfracture toughness evaluation of RPV-steels by use of small specimen. The project includes aparticipation in PARTRIDGE [81]. A Swiss participation in three OECD NEA Benchmark Projectson
leak-before-break, extended finite element method (XFEM) and leak rate calculations areincluded in the PROACTIV project as well. 39
The recently completed NORA project (2010 to 2019) 40 investigated the platinum deposition
behaviour in BWRs and its possible impact on structural materials and fuel element performance[82][83]. It led to a deeper understanding of the mechanisms of SCC mitigation by noble metal
application. One important result of direct practical use is a further optimisation of EPRI’s OLNC
procedure.
Project ZINC, a continuation of the NORA project, was initiated in 2019. It will investigate
water chemistry related issues with respect to the interactions between zinc and noble metal
injection.
ENSI has expressed interest in supporting the international SMILE project 41. Furthermore, PSI
and ENSI have expressed interest in participating in two proposed EURATOM/EU projects; APAL
(Advanced PTS Analysis for LTO) and FRACTESUS (Fracture mechanics testing of irradiated RPVsteels by means of sub‐sized specimens). APAL is planned to develop advanced probabilistic PTS
assessment methods, improved quantification of safety margins for LTO and the development ofbest-practice guidance for advanced PTS analysis with respect to LTO. FRACTESUS has the goal tojoin European and international efforts to incorporate the foundation of small specimen fracture
toughness validation into codes and standards to allow the application and acceptance by various
national regulatory bodies.

34. PSI (n.d.), “Safe LTO in the context of environmental effects on fracture, fatigue & EAC”,
www.psi.ch/sites/default/files/import/lnm/SafeEN/SAFE.pdf.
35. CORDIS (n.d.), “Mitigating Environmentally Assisted Cracking Through Optimisation of Surface
Condition”, https://cordis.europa.eu/project/id/755151.
36. CORDIS (n.d.), “Increasing Safety in NPPs by Covering gaps in Environmental Fatigue Assessment”,
https://cordis.europa.eu/project/id/662320.
37. LWRS (2016), “SCC Initiation in Alloy 600 and Alloy 690”, https://lwrs.inl.gov/Materials%20Aging%20and%20Degradation/SSC_Initiation_in_Alloy_600_and_Alloy_690.pdf.
38 ENSI (2012), PISA, Pressure Vessel Integrity and Safety Analysis, www.ensi.ch/en/2012/03/09/pisa.
39. Additional details can be found in ENSI “Erfahrungs- und Forschungsbericht 2018” pp 30-32 (German
language only).
40. PSI (n.d.), NORA (Noble Metal Deposition Behavior in Boiling Water Reactors): NanoTech for an Efficient
Mitigation of Stress Corrosion Cracking in BWRs – A Decade of Research on NobleChem, www.psi.ch/en/lnm/nora.
41. Refer to Table 3-1 of this report.

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3.12.5 PEO/LTO-OPEX and solutions
Three reactors (KKB1 & 2, KKM) entered into LTO during the period from 2009 to 2012. A fourth
reactor (KKG) entered LTO at the end of 2019. For the fifth reactor (KKL), LTO will begin in 2024.
For KKM (Mühleberg), LTO was approved by the regulator in 2012 for continued operation until
the end of calendar year 2019. For KKB-1&2 (Beznau), a first term LTO was approved for theperiod 2008 until 2018 (Unit 1) and 2010 until 2020 (Unit 2). The documents for the second 10-year
LTO period were submitted to the regulator in 2018. The KKG project for the first LTO period was
submitted to the ENSI by the end of 2018. Summarised below is a Swiss perspective on material
degradation during LTO.
Irradiation embrittlement: Four reactors have concluded the surveillance programmes forradiation embrittlement. It is continued only in KKL. KKL has tested two sets of surveillance
specimen. The third capsule was to be extracted in 2020 and is expected to be tested in 2021.
The two tested sets cover more than 24 years of operation. The third capsule will cover morethan 36 years of operation. The embrittlement in KKL is very low. The adjusted reference
temperature (ART) established using the RG 1.99 method is 9°C and the minimum upper shelf
Charpy energy is at least 119 J. Additional observations of relevance include:•

KKM has tested three sets of specimens. The embrittlement for the base material
and the core weld is low. Leading in KKM is the weld V2 with an ART of 62°C outside
the core region because of the elevated copper content. Due to the permanent shutdown
in December 2019, one remaining surveillance capsule will not be tested. The untested
surveillance specimen will be used for future research activities.

•

In KKG, all irradiation specimens have been removed and tested. Because of the highlead factor of around 3 in the PWR reactor manufactured by KWU, a theoretical operation
time of more than 100 years would be covered. The predicted ARTs for 60 years for theleading base material II are lower than 33°C established by the conservative RG 1.99
method.

•In KKB-1 all six and in KKB-2 five of six surveillance capsules have been removed andtested. The coverage for both units exceed the expected operation time of 60 years. The
highest values of embrittlement are found in the base material of ring C in Unit 1.
Applying the method IIA according to regulatory guideline ENSI-B01, an extrapolated
RTRef of 74°C for 60 years of operation was calculated in a depth of 1/4 of the wall
thickness. With the more conservative method IIB, the value of RTRef is 83°C. The UT
indications in the PRV base material of KKB Unit 1 discovered in 2015, identified as Al2O3
inclusions from manufacturing, do not influence the reference temperature. 42 With
many investigations within a safety case it was shown that these inclusions do not havea negative effect on fracture toughness.

It can be concluded that all Swiss reactors fulfil the requirements of the DETEC Ordinance(the 93°C shut down criteria). The existing results show that the long-term operation of the
Swiss nuclear power plants is not subject to any limitations because of RPV irradiation
embrittlement. PTS analyses showed for all three Swiss PWR that the allowable RTPTS (withconsideration of warm pre-stress effect, WPS) is higher than the 93°C DETEC shut down
criteria.
Baffle-former bolt degradation at Beznau-1/2 and Gösgen: Based on negative international
operating experiences and problems during performing ultrasonic testing due to the geometryof the bolt head, the licensee decided to replace preventive numerous baffle-former bolts at
Beznau units 1&2 in 2009 and 2010. The replacements were done based on a calculation of an
optimised pattern. In all, 192 bolts in Unit 2 and 195 bolts in Unit 1 were replaced. The

42. ENSI (2018), ENSI Review of the Axpo Power AG Safety Case for the Reactor Pressure Vessel of the Beznau
NPP Unit 1, www.ensi.ch/wp-content/uploads/sites/5/2018/03/14H2573-Rev.-1-1.pdf.

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replacement bolts have an optimised notch contour to reduce stresses and are made of steel
AISI 316 CW. Additional observations of relevance include:•

At KKM four actions were taken to address the issue: 1) periodic UT and visual testing;2) numerical simulations; 3) optimisation of water chemistry conditions; and 4) amechanical modification.

•

In 1996, as a precautionary safety action against unexpected fast crack growth, pull rods
and radial stabilisers were installed at four positions. After 2014, based on new
knowledge from numerical analyses, it was planned to replace the four pull rods by a
new design with six pull rods. After the decision for the final shutdown in 2019 thismodification was not realised.

•

In response to a regulatory request, an intensive periodical inspection programme forthe cracks was initiated. In 1993 qualified systematic (1 or 2-year inspection interval) VT
and UT (using a NSSS vendor inspection system, inspection from inner diameter) werestarted. In 1996 the visual inspection was expanded to also include the newly installedpull rods. In 2011 a new UT-system was qualified and introduced. This system enhanced
the coverage, and it allowed the measurement of crack depth. The inspection was done
from outer diameter surface. 43•

In 2014 for the first time transversal cracks or so called “off-axis flaws” were detected.
The last inspection of the core shroud, including off-axis cracks, was done during theoutage in 2018. It confirmed that for the circumferential flaws, no further crack
propagation had occurred since previous inspections in 2015 and 2016. For the off-axis
cracks no further crack propagation was detected. The actual time period of crack
initiation is not fully clear.

•

A third action to guarantee the structural integrity of the core shroud was to carry out
numerical simulations. In 1994 this action was started with a simplified fracture
mechanics-based model to provide screening criteria. With respect to the LTO thesimplified model was replaced in 2011 by a three-dimensional finite element model for
fracture mechanics analyses. The cracks were conservatively modelled as through-wallcracks with their measured crack length. The leading load cases were combinations ofseismic loads from the most adverse direction in combination with one failed tie rod andoperational loads.

•

In 2013, after the UT was able to size the crack depth, the cracks were realised as
circumferential cracks with a depth of 60% wall thickness. This kind of modelling has
the advantage that all crack lengths are covered if there is no growth in depth direction
above 60%. The model was upgraded in 2014 to include the off-axis flaws. As additionalload, case acoustic loads were added. In connection with new seismic loads, an improved
model of the reactor building was created. This model integrates the major componentsof the nuclear steam generation system into the building model.

•

In 1992/93 the conductivity of the reactor water was optimised to around 0.1 mS/cm. In
2000 the hydrogen water chemistry was complemented by platinum injection and in2005 improved based on the Online Noble Chem method (ONLCTM) developed by General
Electric. 44 While the simple hydrogen water chemistry didn’t show significant effects on
the crack propagation, the usage of ONLCTM correlates with significant lower crack
growth rates. As a result of continued research efforts (projects NORA I to III) a further
optimisation of the platinum-feed-in strategy was achieved [82]. No further crack
propagation was detected since 2009.

43. ENSI (n.d.), Fissures in the KKM Core Shroud: ENSI’s Previous Assessment is Confirmed,
www.ensi.ch/en/fissures-in-the-kkm-core-shroud-ensis-previous-assessment-is-confirmed.
44. Hettiarachchi (n.d.), NobleChemTM for Commercial Power Plant Application, https://inis.iaea.org/
collection/NCLCollectionStore/_Public/34/062/34062664.pdf.

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Overview of SCC of a BWR RPV feedwater nozzle (CODAP Event ID 4 500): SCC of a dissimilar
metal weld of an RPV feedwater nozzle was identified during an in-service inspection performed
in 2012 per procedure SVTI-NE-14 45. A full structural weld overlay was applied using the corrosionresistant Inconel 52M. The qualification and implementation took place according to therequirements of design code ASME III, Subsection NB as well as ASME Code Case N-740-2 46 andguideline ENSI-G11 [84]. As part of the evaluation of the indications on the feedwater nozzle, arevised maintenance and ageing management concept for the pipe connection welds of all RPV
nozzles was prepared and submitted to ENSI. This essentially comprises shortened inspection
intervals and special remedial actions (MSIP, or Mechanical Stress Improvement Process).
Currently, an application for approval of the technical specification for the MSIP is being preparedbased on ASME III, Subsection NB, ASME Code Case N-770-1 47 and guideline ENSI-G11.

3.13 Chinese Taipei
The Taiwanese experience with non-code repairs of moderate-energy piping is used as an
example of how a frequently occurring ageing degradation mechanism is managed. Based on
the US NRC Generic Letters 90-05 48 and 91-18 49, the implementation of non-code repairs to stop
leakage without a plant shut down for repairs has been established. Reference [47] summarises
the regulatory experience from the time of programme implementation in 1993 up to the year
2000. This section summarises the post-2000 regulatory experience.

3.13.1 Overview
Regulations and plant technical speciﬁcations (TS) require that the structural integrity of safetyrelated piping be maintained according to Section XI of the ASME boiler and pressure vesselcode. The information concerning pipe ﬂaw behaviours must be recorded in non-conformance
disposition (NCD) reports and submitted to the regulatory authority for the evaluation of thestructural integrity and the corrective actions taken. To implement the code repairs, usually theplant is required to shut down. However, when leakage is discovered in Code Class 3 moderateenergy piping during operation, the guidelines for non-code repairs provided by generic letter
90-05, “Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3
Piping,” and US NRC Inspection Manual Chapter (IMC) 0326, “Operability Determinations andFunctionality Assessments for Conditions Adverse to Quality or Safety” can be applied without
the plant being shut down. Under the request of the regulatory authority, the licensee (Taiwan
Power Company, TPC) has established the procedures of the code and non-code repairs fornuclear safety-related piping as shown in Figure 3-6.

45. SVTI (2005), Festlegung NE-14 Wiederholungsprüfungen von nuklear abnahmepflichtigen mechanischen
Komponenten der Sicherheitsklassen 1 bis 4, Revision 6, www.svti.ch/fileadmin/SVTI/NUK/NE-14.pdf.
46. ASME Code Case N-740-2, “Dissimilar Metal Weld Overlay for Repair or Mitigation of Class 1, 2, and 3
Items.”
47. ASME Code Case N-770-1, “Alternative Examination Requirements and Acceptance Standards for Class
1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler
Material with or Without Application of Listed Mitigation Activities”, 2009.
48. NRC (1990), Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping
www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1990/gl90005.html.
49. NRC (1991), Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of
Degraded and Nonconforming Conditions and on Operability, www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1991/gl91018.html.

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Figure 3-6: The code and non-code repair procedure

3.13.2 Non-code repair of code class 3 moderate-energy piping
If a leak is discovered in a code class one, two or three component while conducting in-service
inspections, maintenance activities, or during plant operation, the licensee should declare the“component” inoperable (this not necessarily results in the relevant “system” being inoperable)
and the component should be repaired or replaced according to the requirements of corrective
measures in ASME Section XI IWA-5 250. The repair/replacement work should be completed
within the speciﬁc allowed outage times (AOTs) in the TS. If not, the limiting condition for
operation (LCO) must be entered upon the discovery of pressure boundary leakage. A non-code
repair of code class three piping is applicable until the next scheduled outage exceeding 30 days,
but no later than the next scheduled refuelling outage.
Temporary non-code repairs of code class three piping in high energy systems, that is, themaximum operating temperature exceeds 200°F or the maximum operating pressure exceeds
275 psig, must have load-bearing capability like that provided by engineered weld overlays orengineered mechanical clamps. Generic letter 90-05 does not allow requests for high energy codeclass three piping based on repairs such as encapsulation of leaking pipes in cans using liquidsealants, clamps with rubber gasketing, or non-engineered weld overlays (patches). For temporary
non-code repairs of code class three piping in moderate energy systems, that is, other than highenergy systems, TPC may consider non-welded repairs. Furthermore, the structural integrity ofthe temporary non-code repair of code class three piping should be assessed periodically.

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For code class three moderate-energy piping, a ﬂaw evaluation must be provided todemonstrate the structural integrity of the through-wall cracked piping and then implement a
non-code repair without plant shut down and prior written relief from the regulatory authority.
If the result of the evaluation meets the acceptance criteria in generic letter 90-05, thecomponent can be considered as operable until the next outage schedule. The licensee should
then submit the evaluation report to the regulatory authority for review and approval in one
month to obtain the written relief of this non-code repair. However, a code repair must be
performed during the next up-coming refuelling outage.
At the request of the Atomic Energy Council, in 1993 the TPC established the process of the
non-code repair for the moderate-energy class three piping. This process is illustrated in
Figure 3-7. Most of the piping that required non-code repairs for leakage was in the seawater
piping system with internal epoxy or rubber liner. The fracture of the inner liners due toageing/degradation resulted in direct attack of the inner surface of piping by the seawater.
A crack due to corrosion could penetrate the wall thickness and cause leakage.

Figure 3-7: Procedure for assessing leakage of safety-related piping

3.13.3 Flaw evaluation
When a leakage in code class three moderate-energy pipe components is identiﬁed, the ﬂaw
geometry must be characterised by volumetric inspection. The ﬂaw geometry has to beconsidered to account for measurement uncertainties and limitations. The structural integrityof the ﬂawed piping is assessed by the fracture mechanics approach provided by GL 90-05. The“LBB.NRC Analysis Method for Circumferentially Through-Wall Cracked Pipes Under Axial Plus
Bending Loads” [48] is used to evaluate the through-wall ﬂaw independent of the orientation
with respect to the piping, as shown in Figure 3-8.

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Figure 3-8: Geometry of corrosion-induced pipe flaw

Based on linear elastic fracture mechanics and assuming a pipe thickness of “𝑡𝑡𝑚𝑚𝑚𝑚𝑚𝑚 ”, thestress intensity factor “K” resulting from the flaw under the applied load is:
K = 1.4 × S × F√𝜋𝜋𝜋𝜋(3.2)

c = a/(πR)(3.3)

F = 1 + A𝑐𝑐1.5 + 𝐵𝐵𝑐𝑐 2.5 + 𝐶𝐶𝑐𝑐 3.5(3.1)

R = mean pipe radius
A = −3.26543 + 1.52784r − 0.072698𝑟𝑟 2 + 0.0016011𝑟𝑟 3(3.4)

C = −3.18609 + 3.84763r − 0.18304𝑟𝑟 2 + 0.00403𝑟𝑟 3(3.5)

B = 11.36322 − 3.91412r + 0.18619𝑟𝑟 2 − 0.004099𝑟𝑟 3
𝑟𝑟 =

𝑅𝑅

𝑡𝑡𝑚𝑚𝑚𝑚𝑚𝑚

(3.4)

(3.6)

Where
t 𝑚𝑚𝑚𝑚𝑚𝑚 = the code-required minimum wall thickness
2a = the length of the through-wall ﬂaw

If the length 2a exceeds either 75 mm or 15% of the length of the pipe circumference, theﬂaw is not acceptable by this approach. The stress S at the ﬂawed location should be determinedfrom the combination of deadweight, pressure, thermal expansion and safe-shutdown
earthquake. A safety factor of 1.4 should be applied to the stress. The value of “r” should bewithin 4 and 16. If it is outside this range, “LBB.NRC” assumes either 4 or 16 as appropriate. For
the stability of the crack, the value of the calculated stress intensity factor K should be less than
35 or 135 ksi √in corresponding to ferritic steel or austenitic stainless steel, respectively.

3.13.4 Types of non-code repairs

The considerations for temporary repairs should be to avoid further degradation and impractical
periodic inspections. The typical acceptable temporary repairs suggested by GL 90-05 are clamps
with rubber gaskets and encapsulation of leaking pipes in cans using liquid sealants. The patch
welding to the cracked area of the pipe is not acceptable to prevent crack propagation near theoriginal ﬂaw regions due to thermal stresses during welding process.

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3.13.5 Monitoring and augmented inspection
If the ﬂaw still exists in the cracked piping component after the temporary repair, the crack
growth must be monitored during the operation and the procedures given below must be
followed continuously:

1. Increase the frequencies of the walkdown inspection for the qualitative assessment of
the leakage through the temporary repair.

2. Apply volumetric examination of the defect area at the speciﬁc period. At least ﬁve of

the most susceptible locations shall be examined to evaluate the global degradation of
the affected region.

If either of the actions mentioned above does not ensure the effectiveness of the temporary
repair, a code repair is required to maintain the structural integrity of the flawed pipe component.

3.13.6 Implementation of non-code repairs
The pipe components in the nuclear service coding water (NSCW) system leaked several times
as listed in Table 3-2. The pinholes induced by the throttle ﬂow near the valve caused the innerrubber fracture and the corrosion of the pipe wall in CCW heat exchanger as shown in Figure 3-9.
To prevent these drawbacks, the stainless liners were installed, and 316 stainless pipe
components were selected to resist the corrosion. The oriﬁce plates were also embedded in thepipe components to reduce the throat ﬂow. For the period 2007 to 2019, Figure 3-10 shows thenumber of non-Code repairs at Chinshan-1/2 (GE BWR-4), Kuosheng-1/2 (GE BWR-6) andMaanshan-1/2 (WH-3LP).

Table 3-3: Examples of leak locations in the NCSW piping system

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Figure 3-9: NCSW leak locations adjacent to CCW heat exchanger

Figure 3-10: Non-code repairs at Taiwanese nuclear power plants in 1993-2000 and 2007-201930

No. of non-code repairs

25
20
15
10
5
0
ESW/NSCW CCW/CSCW ESW/NSCW CCW/CSCW ESW/NSCW CCW/CSCW
ChinsanKuosheng
1995-2000

Essential
chillerFW

Maanshan2007-2019

3.13.7 Conclusions
Surveys of the non-conformance documents indicated that piping liner failure causing
corrosion of the piping is the most common ageing degradation mechanism. Due to thesuccessful applications of fracture mechanics analysis and temporary non-code repairs, thestructural integrity of the safety-related class three moderate pipe components with cracks is
conﬁrmed to be able to prevent the leakage without the shutdown of the reactors. The non-code
repaired piping components are to be replaced in the next scheduled outage. Complete
procedures including the inspection programme, ﬂaw evaluation, repair and quality control
have been established in Chinese Taipei to enhance the safety of long-term nuclear operation.

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3.14 United States
There are 93 (as of the end of 2021) nuclear power reactors licensed to operate in the United States(62 PWRs and 31 BWRs). The average age of these reactors is about 39 years old. As of the end of2021, 94 reactors have received renewed licences (9 units have subsequently shut down), and at
the time of writing this report, the NRC has approved the subsequent licence renewal applications
for six units. Ageing management is a major focus area of licence renewal and subsequent licence
renewal (to authorise plant operation for 80 years). The US NRC and the US nuclear industry are
actively managing and monitoring the effects of ageing on nuclear plant systems, structures andcomponents through the US regulatory framework and via voluntary industry initiatives andprogrammes. The NRC issued NUREG-1800 and NUREG-1801 as companion documents forreviewing licence renewal applications, and NUREG-2192 and NUREG-2191 as companion
documents for reviewing subsequent licence renewal applications. These documents are
discussed below. A significant document issued by industry, NEI 03-08, “Guideline for theManagement of Materials Issues,”50 outlines the policies and practices the industry commits tofollow for managing materials ageing issues through voluntary initiatives. Several industry issue
programmes are responsible for implementing the NEI 03-08 guidance. The issue programmes
include, for example, the BWR vessel and internals project (BWRVIP), the BWR owners’ group, thematerials reliability programme (MRP), the steam generator management programme (SGMP) and
the PWR owners group.

3.14.1 Building the state of knowledge regarding the management of material
degradation during LTO
Long-term operation of the US nuclear reactor fleet is being addressed by stakeholders that
include both industry and the NRC through various initiatives and programmes in preparation
for licensee renewal and subsequent licensee renewal. In 2006, the NRC published
NUREG/CR-6923, “Expert Panel Report on Proactive Materials Degradation Assessment,” or thePMDA, to address operations up to 40 years [17]. To address 80 years of operation, the NRC
published NUREG/CR-7153, “Expanded Materials Degradation Assessment (EMDA),” in 2014 [27].
The PMDA and the EMDA identified materials and components where future degradation may
occur in specific light water reactor systems, and gaps in knowledge. In 2019, the NRC organised
an International Workshop on Age-Related Degradation of Reactor Vessels and Internals. The
workshop participants addressed the state of knowledge, operating experience and research
activities related to reactor pressure vessel embrittlement at high fluences, degradation ofreactor vessel internals for operating periods up to 80 years and degradation related to LTO ofother safety-significant pressure boundary components [39].
To assist with regulatory decision-making regarding ageing-related materials degradation,
the NRC conducts targeted confirmatory research on age-related degradation to ensure theeffectiveness of existing ageing management programmes and inform the development ofrevised guidance. The NRC research programme is summarised in NUREG-1925 and is updated
every two years [88].
The US Department of Energy (DOE) Office of Nuclear Energy’s (NE) light water reactor
sustainability (LWRS) programme conducts research to, in part, extend the operating life of US
nuclear power plants. The LWRS programme is focused on (1) developing knowledge tounderstand the ageing of materials, structures and components; (2) applying that knowledge tosupport long-term operations; and (3) researching new technologies that enhance performance,
economics and safety. The LWRS programme addresses the following technical areas: plantmodernisation, flexible plant operation and generation, risk-informed systems analysis, materials
research and physical security. The DOE publishes an annual report providing an overview of the
LWRS programme and its accomplishments [89]. More information on the LWRS programme,
including reports published on each of the technical areas, can be found on the DOE NE website 51.

50. Nuclear Energy Institute (2010), “Guidelines for the Management of Materials Issues”, NEI 03-08 [Rev 2],
www.nrc.gov/docs/ML1010/ML101050337.pdf.
51. DOE NE website: www.energy.gov/ne/office-nuclear-energy.

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Industry, through EPRI, developed the materials degradation matrix (MDM) [90] to identify
potential degradation mechanisms, knowledge of degradation mechanisms (how well they are
understood), and strategic long-term operation issues. EPRI also developed and maintains Issue
Management Tables (IMTs) [91][92] to assess the consequences of failure, identify gaps in
guidance, and prioritise and direct R&D efforts related to materials degradation. The MDM and
IMTs are used by industry to manage materials degradation/ageing issues through the various
issues programmes mentioned above.
The NRC, DOE and EPRI developed and maintain a research roadmap that includes research
they are conducting related to reactor systems materials. The purpose of the roadmap is to help
co-ordinate research activities among the various US stakeholders to provide focus and reduce
unnecessary duplication. The roadmap is based on information from various sources which
include EPRI’s MDM and IMTs.
Based on the activities of the US stakeholders described above, several ageing mechanisms
are being addressed during the licence renewal and subsequent licensee renewal periods. The
following are examples: reactor pressure vessel material radiation embrittlement, stress corrosion
cracking (including primary water stress corrosion cracking and irradiation assisted stresscorrosion cracking), thermal ageing, fatigue (high-cycle, low-cycle and thermal), selective leaching,flow accelerated corrosion and microbiologically influenced corrosion. The NRC has developed aregulatory framework to manage these, and other, ageing mechanisms, that is described in thenext section.

3.14.2 Regulatory framework
The Atomic Energy Act and Nuclear Regulatory Commission (NRC) regulations limit commercial
power reactor licences to 40 years but permit the operating licences to be renewed. The original
40-year term was selected based on economic and antitrust considerations rather than technicallimitations; however, many of the technical safety evaluations were subsequently based on a
40-year operating period. The decision to seek licence renewal rests entirely with the nuclear
power plant owners and typically is based on the plant’s economic situation and whether it cancontinue to meet NRC requirements.
The provisions of Title 10, Code of Federal Regulation (CFR) Part 5452 allow an operating licenceto be renewed for 20 years and with no limit on the number of times a licence can be subsequently
renewed, provided it is justified, and that the effects of age-related degradation will be adequately
managed during the period of extended operation. The earliest that a licensee can submit a licence
renewal application is 20 years before the expiration of its current licence. Therefore, a licensee iseligible to apply for an initial licence renewal once it has operated for 20 years. A licensee is eligible
to apply for a subsequent licence renewal (SLR) once it enters the initial period of extended
operation (i.e. the 20-year period beyond its initial 40-year licence period). In 2000, the NRC issued
the first renewed licences for the Calvert Cliffs Nuclear Power Plant and the Oconee Nuclear
Station. As of the end of 2021, 94 reactors had received renewed licences (nine units havesubsequently shut down). In 2019, 2020 and 2021, the NRC approved the SLR applications for
Turkey Point Units 3 and 4, Peach Bottom Units 2 and 3 and Surry Units 1 and 2, respectively. The
NRC had received four other SLR applications (North Anna Units 1 and 2, Oconee Units 1, 2 and 3,
Point Beach Units 1 and 2, and St. Lucie Units 1 and 2) at the time this report was drafted.
Licence renewal (40 to 60 years): “Licence renewal” (LR) is the process used in the UnitedStates for a nuclear power plant to request renewal of the plant’s operating licence for an
additional 20 years of operation via a licence renewal application (LRA). The NRC issued
NUREG-1800, “Standard Review Plan for Review of Licence Renewal Applications for Nuclear
Power Plants,” (i.e. SRP-LR) to provide NRC staff with guidance on how to review licensee
applications to extend operating licences from 40 years to 60 years. The SRP-LR covers scoping
and screening methods for identifying systems, structures, and components subject to ageing
management review and time-limited ageing analyses. NUREG-1800 [95] references NUREG-1801,

52. Available on the NRC website at: www.nrc.gov/reading-rm/doc-collections/cfr/part054.

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COUNTRY/ECONOMY-SPECIFIC OVERVIEWS OF MATERIAL DEGRADATION STATE-OF-KNOWLEDGE, RESEARCH AND OPERATING EXPERIENCE“Generic Aging Lessons Learned (GALL) Report” [96], which evaluates existing programmes todetermine if they are adequate for managing identified ageing effects during the period ofextended operations or require changes. The GALL Report describes acceptable ageing
management programmes (AMPs) for the structures and components within the scope of the SRPLR. Applicants can reference the GALL Report in their applications and treat it as an approved
Topical Report. As stated in the GALL Report, “however, if an applicant takes credit for a
programme in the GALL, it is incumbent on the applicant to ensure that the conditions andoperating experience at the plant are bounded by the conditions and operating experience forwhich the GALL Report programme was evaluated.” The initial GALL report and SRP-LR were
issued in 2001 and were revised in 2005 and 2010.
The guidance in the GALL report and the SRP-LR have been modified using the Interim Staff
Guidance process (ISG-LR), as described at www.nrc.gov/reading-rm/doc-collections/isg/licenserenewal.html.
Subsequent licence renewal (60 to 80 years): The term “subsequent licence renewal” (SLR)
refers to the second (or subsequent) renewal of a licence that was previously renewed. For
example, SLR may approve continued operation for the period from 60 to 80 years. SLR also refers
to the process for applicants to apply for the staff to review and evaluate. The term “period ofextended operation” or PEO is used in 10 CFR 54.3 in the definition of “integrated plant assessment”(IPA). This term describes plant operation beyond the initial 40-year licence term; for example,
plant operation from 40 to 60 years under a renewed licence. The period from 60 to 80 years isreferred to as the “subsequent period of extended operation.” A framework like the licence
renewal process was established for reviewing applications for subsequent licence renewal.
NUREG-2192, “Standard Review Plan for Review of Subsequent Licence Renewal Applications for
Nuclear Power Plants” [97], was issued in 2017, and the companion reference NUREG-2191,“Generic Aging Lessons Learned for Subsequent Licence Renewal (GALL-SLR) Report” [98], waspublished previously in 2015. Like the GALL Report, the GALL-SLR Report contains AMPs that werereviewed specifically for SLR and modified as necessary.
The GALL-SLR report and the SRP-SLR have been modified using the ISG-SLR process, as
described at www.nrc.gov/reading-rm/doc-collections/isg/license-renewal.html.

3.14.3 Materials research in support of LTO
The development of the GALL report (NUREG-1801), the technical bases for the SRP-LR(NUREG-1800), involved periodic operating experience reviews (domestic and international), andthe consideration of comments from stakeholders in the process (e.g. licensees and the Nuclear
Energy Institute). As a result, revisions subsequent to NUREG-1801, Rev. 0 incorporated changes
that made it more efficient and responsive to current operating experience. The GALL Report is a“living” document meant to capture lessons learnt from licence renewal reviews and operating
experience, as is the GALL-SLR Report. The NRC seeks to continuously improve its understandingof materials ageing mechanisms and their effects on plant systems, structures and components
and to incorporate the information into relevant regulatory documents via the ISG-LR and ISGSLR process referenced in the previous paragraph.
The Office of Nuclear Regulatory Research (RES) provides data, techniques and methods to
the US Nuclear Regulatory Commission’s (NRC’s) regulatory offices to support their reviews ofmaterial performance-related licensing submittals and safety issues. The confirmatory researchon materials performance focuses on both the development of methodologies needed tosupport regulatory actions and the work supporting the technical bases for codes and standards
development. A common theme in this work is a proactive approach to the management ofageing degradation. As industry’s interest in licence renewal for operation beyond 60 years
increases, the staff has begun to assemble technical information on ageing phenomena that canaffect materials in nuclear power plants and to develop technical guidance for the staff’s review
of subsequent licence renewal (SLR) applications. Outlined below are the key research areas.
•

90

Steam generator tube integrity: Research is underway to develop a technical basis forsteam generator tube integrity to support regulatory decisions and code applications.

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The goal is to ensure appropriate inspection intervals while still maintaining public
safety. To provide this basis, research is focused on two areas: (1) inspection reliability
and in-service inspection technology; and (2) the evaluation and experimental validationof tube integrity prediction modelling of known degradation modes.
•

Reactor pressure vessel (RPV) integrity and internals: Safe plant operation relies onmaintaining the structural integrity of the RPV during routine operations and postulated
accident scenarios. Two key capabilities underpin the assessment of RPV structural
integrity: (1) the ability to predict the behaviour of cracked components under loading; and(2) the ability to predict the effects of irradiation embrittlement on the fracture toughnessof RPV steels. Ongoing NRC efforts are aimed at understanding the adequacy of existing
approaches and predictive procedures, as well as surveillance requirements to address theintegrity of the RPV under long-term operation. For RPV internals, ongoing research is
focused on (1) irradiation-assisted degradation (IAD) of RPV internals to assess the
structural and functional integrity of pressurised-water reactor internal components; and(2) performing confirmatory analysis of the performance of RPV internal materials duringextended operation up to 80 years. Research is being conducted on harvested ex-plantmaterials as well as on representative materials irradiated in test reactors.

•

Piping degradation: To better understand how the reactor is influenced by thephenomenon of primary water stress corrosion cracking (PWSCC), the NRC is conducting
confirmatory testing on both crack initiation and crack growth rate for susceptible
materials. Additional research is focused on the development of analysis methods andcomputational tools, and the performance of experimental testing to assess the impact of
PWSCC on the overall safety of piping systems of the reactor coolant pressure boundary.
The NRC also is assessing impacts of PWSCC on the leak-before-break behaviour of piping
systems.

•

Non-destructive evaluation (NDE): Current NDE research is focused on the areas ofeffectiveness and reliability of NDE methods, assessment of NDE modelling tools, andevaluation of human factors on NDE performance; [100][101][102][103]. The results ofongoing research are used to support regulatory decisions associated with NDE and inservice inspection (ISI) of safety-related systems. The research efforts also support
review of relief requests and proposed changes to ASME Code requirements.

•

Probabilistic codes: Through a co-operative agreement, the NRC’s Office of Nuclear
Regulatory Research and the Electric Power Research Institute (EPRI) developed the xLPR
(Extremely Low Probability of Rupture) probabilistic code to evaluate leak-before-breakanalysis requirements for primary pressure piping systems per the US Nuclear Regulatory
Commission (NRC) Standard Review Plan 3.6.3.

3.14.5 Operating experience during the PEO
A search in the CODAP database for US plants with OPEX during the PEO was conducted basedon components with 40 or more years of service at the time of the event (e.g. leakage, flaw
detection). The search resulted in 69 records. Figure 3-11 is the distribution of the events in
CODAP versus component age and shows that most of the events occurred between 40 and
45 years of component service. The data in Figure 3-11 is not normalised by the population ofplants based on age, which would be required to determine event frequencies.
Figure 3-12 shows the distribution of damage/degradation mechanisms for US plants during
PEO. Figure 3-13 shows the specific systems involved in US plant OPEX during the PEO. Systems
with four or more events recorded in CODAP include: chemical and volume control, condenser
system, reactor coolant system (PWR), reactor pressure vessel and service water system. The data
in Figures 3-12 and 3-13 are not normalised. Hence, using the data to develop event frequencies
requires an understanding of how events are characterised in CODAP. In certain cases, forexample, a single inspection of a component piece/part may result in multiple cracking events
being entered in the database.

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COUNTRY/ECONOMY-SPECIFIC OVERVIEWS OF MATERIAL DEGRADATION STATE-OF-KNOWLEDGE, RESEARCH AND OPERATING EXPERIENCE

Figure 3-11: Distribution of OPEX during the PEO for US plants vs. component age
18
16
14
12

Events in CODAP

10
8
6
4
2
0
40

41

42

43

44

45

46

47

48

49

50

Component age bins

Figure 3-12: Distribution of damage/degradation mechanisms for US plants during PEO
Thermal fatigue - cycling
Overstressed
TGSCC
PWSCC
IGSCC
IASCC
Hydrogen embrittlement
Low-cycle fatigue
High-cycle fatigue
Flow-assisted degradation
Fretting
Corrosion - All mechanisms
0

92

2

4

6

8

10

12

14

16

18

20

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Figure 3-13: Distribution of OPEX during the PEO across system groups
RCPB

System groups

Reactor auxiliary systems
Service water
Feedwater and condensate
Safety injection and recirculation
Steam systems
Misc. support systems
S/G-system
0

5

10

15

20

25

30

35

40

No. of failures in CODAP

Baffle-former bolt cracking OPEX: An example of RPV materials degradation being managed
during the PEO is the SCC of baffle-former bolts in PWR plants. Through the EPRI materials
reliability programme (MRP) and the BWRVIP, the NRC receives biennial summaries of the BWR
and PWR reactor internals inspection results, respectively. These public domain summaries areavailable from the NRC website. 53 Below is an overview of the PWR baffle-former bolt OPEX.
These reports identify the reactor internals inspected and generally include the date andfrequency of inspection, the inspection method used, a summary of results including repair or
replacement activities and flaw dimensions.
The core baffle is a portion of the reactor vessel internals and is located within the core
barrel and functions to direct the coolant flow through the core and provide some lateralsupport to the fuel assemblies. Figure 3-13 shows a typical Westinghouse baffle-former
assembly and the locations of various bolt types. Vertical baffle plates are bolted to the edges ofhorizontal former plates, which are attached to the inside surface of the core barrel. There are
typically eight levels of former plates located at various elevations within the core barrel. The
bolts that secure the baffle plates to the former plates are referred to as baffle-former bolts. To
cool the baffle structure, some water flowing through the reactor vessel is directed between thecore barrel and the baffle plates (through holes in the former plates) in either a downward
direction (down-flow configuration), or an upward direction (up-flow configuration).
Degradation of baffle-former bolts was first noted in 1988 in the French Bugey Unit 2 reactor.
The NRC communicated operating experience on baffle-former bolt degradation to US licensees
in Information Notice (IN) 1998-11, “Cracking of Reactor Vessel Internal Baffle Former Bolts in
Foreign Plants,” dated 25 March 1998 [105]. Subsequent inspections in the US plants identified
limited degradation at most. As a preventive measure, some of these plants also replaced someof their bolts, including those without indications, to an evaluated pattern to ensure structural
integrity of the baffle-former assembly, using a different stainless-steel alloy and an enhanced
geometry to reduce stress risers in the bolt geometry.

53. See https://adams.nrc.gov/wba; Accession No. ML19232A181 (2018 report for US BWRs) and
ML18204A161 (2018 report for US PWRs).

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The degradation of the baffle-former bolts is attributed to irradiation assisted stresscorrosion cracking. Baffle-former bolts are subjected to significant stresses and irradiation overyears of plant operation. PWRs with a down-flow configuration place additional stress on thebaffle-former bolts because of the pressure differential across the vertical baffle plates. At thistime, the most significant degradation of the baffle-former bolts has been observed inWestinghouse four-loop PWR reactors with the down-flow configuration and bolts made ofType 347 stainless steel. Seven reactors in the United States match this description.
In 2011, the EPRI MRP issued MRP-227-A, “Pressurised Water Reactor Internals Inspection
and Evaluation Guidelines” [24], which includes an inspection of the baffle-former bolts during
the time frame when bolt degradation is most likely to appear, as shown by operating
experience. The NRC endorsed MRP-227-A in 2012. The guidelines of MRP-227-A provide for thedevelopment of an ageing management programme for PWR reactor vessel internals that meets
the NRC requirements for issuance of a renewed operating licence.
From spring 2016 through spring 2017, four US nuclear power plants identified many Type 347
stainless steel baffle-former bolts with indications of degradation during ultrasonic inspections
following MRP-227-A guidelines. From spring 2017 through spring 2019, several other US nuclear
power plants also performed ultrasonic inspections following MRP-227-A guidelines but identified
relatively few stainless steel baffle-former bolts with indications of degradation. In general, theseplants replaced potentially degraded bolts and proactively replaced bolts that did not exhibit
indications with Type 316 stainless steel bolts to improve the design.
The NRC performed a risk-informed evaluation of the safety impact that the degradation of
baffle-former bolts could present to operating reactors. The NRC concluded at that time that
this issue did not require the immediate shutdown of any facilities. Between 2016 and 2017, EPRI
issued interim guidance on baffle-former bolt inspections in Westinghouse-design PWRs. The
NRC’s assessment of this interim guidance is discussed in “Staff Assessment of EPRI MRPInterim Guidance on Baffle Former Bolts,” dated 20 November 2017. 54 Analyses were performed
to determine the material condition of the baffle-former bolts that were removed from theseplants to provide additional insights on the degradation mechanism. The US nuclear industry
formed a working group to consider potential changes to the MRP-227-A inspection regime andplans to incorporate the interim guidance in future revisions to MRP-227.
On 21-22 May 2019, the NRC met with representatives from the nuclear industry for an
exchange of technical information on the industry’s materials programmes. 55 During this public
meeting, EPRI provided an update on baffle-former bolting inspection results. In all instances
except one, the inspection results have been as, or better than, expected. The exception occurredin the spring of 2019 at a 4-loop down-flow unit. Visual inspections identified significant
degradation of baffle-former bolts. Based on these results, the unit performed ultrasonic testing
inspections of all original baffle-former bolts. The ultrasonic testing inspections identified asignificant number of baffle-former bolts with indications of degradation and a significant
clustering of degraded bolts. At the time of the public meeting, the utility was planning to replaceapproximately 272 baffle-former bolts and was considering other potential future actions in fall2020.

54. NRC (2017), “Staff Assessment of EPRI MRP Interim Guidance on Baffle Former Bolts,” www.nrc.gov/
docs/ML1731/ML17310A861.pdf.
55. Amberge, K.J. (2019), “Update Baffle-Former-Bolting (BFB) Inspection Results in the United States”, apresentation made at the EPRI-NRC technical information exchange meeting on 21-22 May 2019;https://adams.nrc.gov/wba/view (Accession No. ML19134A161).

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Figure 3-14: Illustrations of (a) a typical Westinghouse baffle-former assemblyand (b) cross section of one octant of a typical Westinghouse baffle-former
assembly showing locations of bolt types

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Chapter 4. PEO/LTO operating experience data analysis

This chapter gives the results of the PEO/LTO operating experience data evaluations. The data
evaluations were performed in three phases. First, through comparisons of the content of the
CODAP event database against the observations and results of the WGIAGE survey [32], in
Section 4.1. Next, a comparison was made of the CODAP event database against the findings ofthe two expert panels on material degradation [17][27], in Section 4.2. Mainly, these comparisonswere done to identify possible gaps in the CODAP event database as well as to identify ways inwhich the project can improve its communications to the material science community. Finally,a simplified OPEX assessment was performed to identify the age-dependent trends and patternsof different combinations of systems, materials and degradation mechanisms; see Sections 4.3
through 4.7.

4.1 CODAP event database vs. WGIAGE conclusions
Specific to LWR operating environments, a synthesis of the WGIAGE conclusions [32] is foundin Table 4-1. According to the survey, embrittlement of RPV material and RPV internals wasrecognised by four organisations as being of “high importance” with respect to regulation of
PEO/LTO. With respect to RPV material, irradiation embrittlement is extensively monitored
throughout the nuclear industry through surveillance programmes to measure changes in theproperties of actual vessel materials due to the irradiation environment. The CODAP event
database includes OPEX data on RPV head and bottom head penetrations. According to thecurrent Terms of Reference for the project, CODAP does not address RPV minor fabrication
defects such as the 2012 Doel-3 and 2015 Beznau-1 RPV base material flaw indications.
Three respondents to the WGIAGE questionnaire assigned a “high importance” to material
degradation scenarios related to corrosion mechanisms such as “pitting and crevice corrosion
near deposits, metallic contact points and stagnant conditions.” These corrosion mechanisms
are extensively addressed in the CODAP event database.

4.2 CODAP event database vs. PMDA/EMDA conclusions
In comparing the PMDA/EMDA conclusions and the CODAP event database content nofundamental gaps have been identified. CODAP is concerned with material degradation
mechanisms that have produced failures. The key observations from this comparison are as
follows:•

The long-term effectiveness of the various degradation mitigation strategies does
warrant a continued and systematic OPEX data evaluation effort.

•

The environmental effects on fatigue are a research area that continues to receive
significant attention. The related OPEX is sparse, however. The continued OPEX data
mining efforts would be of significant value in supporting the future research activities.

•

Considerable R&D has been directed to furthering the TAE knowledge base, including
extensive mechanical testing of CASS components that have been removed from service.
The conjoint requirements of TAE are well understood. No TAE failures have yet beenreported, however.

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Table 4-1: A synthesis of the results of the WGIAGE survey
Degradation mechanisms considered
important to PEO/LTO of light water
reactors (BWRs, PWRs and WWERs)1

Ranking

No. ofrespondents(of 8 total)2

How is this degradation mechanism
addressed in CODAP?

RPV irradiation embrittlement, particularly
in combination with hydrogen flaking andhigh nickel content

High

5

Medium1

Irradiation embrittlement of RPV internals

High

5

CODAP includes “selected
representative events”

High

3

The related OPEX is extensively coveredin the database. NEA/CSNI/R(2018)2 [34]summarises the below groundcorrosion OPEX

High

2

Medium2

Irradiation assisted SCC

High

1

CODAP includes “selected
representative events”

SCC of Ni-base alloys

High

1

The related OPEX is extensively coveredin the database. For additional details
refer to NEA/CSNI/R(2010)5 [21].

Thermal fatigue caused by stratificationand thermal cycling in branch connections(e.g. mixing tees)

High

1

Medium1High

1 (WWER)

Medium1 (WWER)High

1

Medium2

High

1

Pitting and crevice corrosion near deposits,
metallic contact points and stagnant
conditions.
Internal/external corrosion of below
ground piping
Thermal and neutron embrittlement ofcast austenitic stainless steels (CASS)

FAC of high energy pipelines

Environmentally assisted fatigue ofprimary circuit metals
Boric acid corrosion in bolted closures ofthe primary circuit pressure boundary

No service induced OPEX is available

Only limited OPEX in the current versionof the event database

The related OPEX is extensively coveredin the database. For additional details
refer to NEA/CSNI/R(2019)13 [38].
For WWERs the database includes some
“representative events” and potentiallythis OPEX does not reflect the “high” or“medium” priority in the WGIAGE
Survey3
Only a few records in the database andlimited to corrosion fatigue/SICC failures
Yet, bolted connections are not part ofthe scope of the CODAP event database

1. CANDU-specific degradation mechanisms are not addressed in this table.
2. Eight CODAP project member countries/economies responded to the survey.
3. The CODAP event database includes 13 CZ records covering the period 1985 to 2013 and 4 SK records covering the period 2000 to
2010.

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Table 4-2: Comparison of PMDA/EMDA conclusions
with CODAP event database insights
BWR primary system components
PMDA: 25 < T ≤ 40 Years

EMDA: 40 < LTO ≤ 80 years

CODAP OPEX data insights

SCC of Alloy 82/182 weldments used injoints between ferritic and austenitic
components (e.g. thermal sleeves,attachment pads) and particularly
alloy 182 under NWC conditions.

SCC was identified as a high knowledge,
moderate Susceptibility mode of
degradation for alloy 182/82 weldments
in BWR HWC environments. Reduced
fracture resistance in alloy 182/82 welds
at lower temperatures has been noted inlaboratory testing although additional
mechanistic understanding is needed.

Extensive OPEX coverage in the database.
As an example, refer to Figure 4-1 for the
BWR-specific OPEX.

SCC of Type 304/316 stainless steel heataffected zones (even under HWC
conditions) at moderate fluences.

The following knowledge gaps wereidentified: 1) impact of irradiation on
fracture toughness and SCC in both NWC
and HWC environments, 2) SCC
susceptibility at very long operation
periods, particularly in NWC
environments, and 3) Cumulative impactof fatigue on corrosion and component
integrity, particularly weldolets, sockolets,and components in the upper coreinternals.

The CODAP database is structured so thatmaterial degradation of reactor internals
can be tracked. So far, no systematic
process has been implemented for thecollection of failure data on reactor
internals, however.

Fatigue of type 304/316 stainless steel insteam dryers under NWC and HWC
conditions. This is probably designspecific, but as of now, is not quantified
across the whole fleet.

Not explicitly addressed beyond notingthat the steam dryer holds down brackets
are alloy 182.

CODAP includes selected representative
event reports on BWR steam dryers. So
far, no systematic process has been
implemented for the collection of failure
data on reactor internals, however.

SCC cracking of alloy 600 components
and heat affected zones throughout thereactor coolant system, particularly under
NWC conditions.

SCC was identified as a high knowledge,high susceptibility mode of degradation
for alloy 600 in BWR NWC environments
by the expert panel. This is a known issue
for these alloys in high potential
environments.

Most, if not all, BWR plants have
implemented HWC to mitigate SCC
susceptibility; refer to the first line item ofthis table.

SCC of Alloy X-750 components
throughout the reactor coolant system
under both NWC and HWC conditions.

Extended service will result in increased
time exposed to the high-temperaturewater environment and under stress.
Further, for core internals, an increased
fluence will be experienced due to longer
service and power uprates. PIRT scoringfor X-750 in the high Knowledge, highSusceptibility category due to problems
identified in service.

The first Alloy X-750 failures (in-vessel
bolting) were observed in Germany in thelate 1970s. The most recent and a first-ofkind failure of X-750 were reported in the
United States in February 2006. It
involved a core shroud tie rod uppersupport.

Cast stainless steel components(e.g. CF8M brackets and guide rods andCF3 control rod guides), may experience a
reduction in fracture resistance as a resultof thermal ageing. For the most part,however, the BWR RCS components
operate at lower temperatures than PWR
RCS components and, therefore, possible
problems would be expected to become
evident in PWRs earlier than in BWRs.
Also, the possibility of SCC occurring incast stainless steel components should
not be dismissed without further
investigation of the possible synergistic
effect of thermal ageing on thisdegradation mode.

The effects of long-term thermal ageingfor extended operating periods may drivechanges in mechanical or corrosion
performance that are relatively unknown.

No thermal ageing failure data has been
identified for inclusion in CODAP. The
definition of TAE failure modes is an issue
for the CODAP-MB to consider.

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Table 4-2: Comparison of PMDA/EMDA conclusions
with CODAP event database insights (cont’d)
BWR primary system components
PMDA: 25 < T ≤ 40 Years

EMDA: 40 < LTO ≤ 80 years

CODAP OPEX data insights

The long-term effects of HWC on SCC are
not yet clear and it is also not clear thatthe low potentials needed for mitigation
are in fact achieved at all the highsusceptibility locations. The role ofhydrogen in the stress corrosion cracking
process of austenitic stainless steels is
critical but is expected to be less than inhigh strength nickel-base alloys. The
effect of hydrogen on fatigue and its
interaction with ripple loading anddynamic strain ageing was noted as apotential contributor to degradation
under HWC conditions.

Continued "potential knowledge gap"
regarding the SCC susceptibility at very
long lifetimes.

This “degradation scenario” is being
monitored through the review of OPEX.
Relevant information has been reported
at conferences such as EDF FAC
conferences, Fontevraud1 through
Fontevraud8 and ENVDEG1 through
ENVDEG19 (Environmental Degradationof Materials in Nuclear Power Systems). To
date, CODAP does not contain any
explicit information, however.
Noteworthy is the observation of theSwedish National Ageing Management
Action Plan (SSM 2019:17) regarding
HWC. The Swedish BWR plants withinternal recirculation pumps (Forsmark1/2/3 and Oskarshamn-3 abandoned
HWC mitigation in the 1990s. According
to the Owners “… it is almost impossible
to create the mitigating environmental
conditions in the bottom of the pressure
vessel.”www.stralsakerhetsmyndigheten.se/en/p
ublications/reports/safety-at-nuclearpower-plants/2019/201917.

Accelerated environmental effects on
fatigue of austenitic stainless steels andnickel alloys at low electrochemicalcorrosion potentials under HWC
conditions. These effects have been
demonstrated in the laboratory andtherefore are anticipated to apply in thefield.

Continued "potential knowledge gap"
regarding the cumulative impact of
fatigue on corrosion and component
integrity at very long lifetimes.

No OPEX available as of yet.

Stress corrosion cracking of severely cold
worked or mechanically strained stainlesssteels in BWR water, e.g., weld shrinkage
strains in heat affected zones, cold bent
elbows.

The EMDA Report includes extensive
details on this subject.

Extensive OPEX included in CODAP. The
role of cold work on the susceptibility ofstress corrosion cracking (SCC) in metallic
nuclear materials, especially austenitic
stainless steels, and nickel-base alloys, hasbeen recognised for the past several
decades and the first operational eventsoccurred during the early days ofcommercial nuclear energy production.

“Abusive” grinding or machining during
weld preparation or surface finishing andtheir adverse effects on stress corrosion
and fatigue crack initiation.

Not explicitly addressed. However, theEMDA Expert Panel refers to the adverse
effects of improper weld preparation or
surface finishing.

Extensive OPEX on this topic can beextracted from CODAP. See for example
Reference [21].

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Table 4-3: Comparison of PMDA/EMDA conclusions
with CODAP event database insights (Part 2)
PWR primary system components
PMDA: 25 < T ≤ 40 years

EMDA: 40 < LTO ≤ 80 years

CODAP OPEX data insights

SCC of alloy 82/182 weldments
throughout the primary system andespecially in the highest temperaturecomponents such as the pressuriser.

The expert panel identified only one
potential knowledge need for alloy 182/82 in
PWR environments. Specifically, fracture
resistance in alloy 182/82 welds at lower
temperatures has been noted in laboratory
testing although limited mechanistic
understanding has been established.

Addressed extensively in CODAP event
database and the Topical Report on SCC[21].

SCC of alloy 600MA and 600TT steam
generator tubes and alloy 600 forgedcomponents and, especially, cold
worked or mechanically strainedmaterial such as steam generator tube
expansion transitions, small radius
U-bends, and the high-temperaturecomponents of the pressuriser.

SCC is a known issue for alloy 600 in all alloy
forms. SCC has been observed in bothprimary and secondary-side applications for
all forms of alloy 600 in service. The expert
panel scored all 12 categories of SCC in thehigh knowledge, high susceptibility
grouping. alloy 600 in the MA form wasgiven the highest possible susceptibility
score, consistent with operational
experience. Most plants have or will be
performing steam generator replacements
and in the process are upgrading from Alloy
600 to Alloy 690.

The CODAP work scope considers alloy
800 and alloy 690 SG tube degradation.
The recent (2018 and 2019) German OPEX
involving degradation of alloy 800 tubes
is included in CODAP. As of end of
calendar year 2019, the CODAP event
database includes two (2) data records on
SG tube degradation.

Corrosion of low-alloy steel
components in boric acid concentrateoriginating from primary water leaks,
specifically in the annuli between thealloy 600 penetrations and the lowalloy steel reactor vessel or pressuriser.

Several trends for low alloy steels in PWR
environments were identified via the PIRT
process: 1) BAC of low alloy steel was
identified as a high Knowledge, highSusceptibility mode of degradation, but onlyin the event of a leak of primary coolant. Thisis a well-known form of degradation, and 2)
Crevice corrosion, pitting, and FAC of low
alloy steel were identified as highKnowledge, moderate Susceptibility modesof degradation, but only in the event of aloss of water chemistry control. These are
well-known forms of degradation. Stresscorrosion cracking and fatigue are possible
for these alloys, but unlikely in service.
Changes in loading or increases in chemical
conditions may drive increased susceptibility
over a long operating period.

This is covered in the CODAP event
database.

Irradiation-induced creep and SCC ofaustenitic stainless steels at more than
0.5 dpa, (e.g. baffle bolts, and otherhigh strength fasteners) and swellingin internals components that reachhigher temperatures and much higher
doses.

Several knowledge gaps for 316 SS in PWR
environments were identified via the PIRT
process: 1) Effect of irradiation on fracture
toughness, irradiation creep, swelling, andSCC, and 2) SCC susceptibility at very long
lifetimes. Additionally, while scored in thehigh Knowledge category, the panellists
noted that the cumulative effect of fatigue
on corrosion and component integrity
deserved additional examination due topossible changes and uncertainties in cyclic
and flow-induced loading over extended
service condition.

Extensive baffle former OPEX exists.
Selected representative events areincluded in the CODAP event database.

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Table 4-3: Comparison of PMDA/EMDA conclusions
with CODAP event database insights (Part 2) (cont’d)
PWR primary system components
PMDA: 25 < T ≤ 40 years

EMDA: 40 < LTO ≤ 80 years

CODAP OPEX data insights

Fatigue due to unanticipated vibrationand thermal fluctuations, e.g. in socket
welds (which occur throughout theRCS), primary circuit deadlegs(e.g. abandoned pipe sections), andbaffle bolts after irradiation-induced
relaxation of pre-stress.

Fatigue and corrosion fatigue are known
issues in reactor service. As service life
increases, so does the total number ofloading cycles experienced by a component.
Further, power uprates may also increase
flow-induced cyclic loading and impact
component lifetime. The CUF must be
evaluated for extended service. Allenvironments for 316 SS were scored in thehigh knowledge, moderate susceptibility
grouping.

Although not yet done, a correlation (ifany) between power uprates and thefatigue incident rate can in principle bedetermined using the CODAP event
database. This could be a topic for a
future Topical Report.

Although alloy 690TT has exhibited
good resistance to SCC compared withalloy 600 MA during more than 16
years of use in steam generator tubing,some panel members questioned its
long-term resistance in other PWR
primary water applications(e.g. pressure vessel penetrations). The
potential for this degradation increases
when considering effects of cold workand the use of the alternate welding
alloys (alloys 52 and 152) which
present welding challenges associatedwith, for example, lack of fusion andductility dip cracking.

No knowledge gaps were identified for alloy
690 under subsequent operating periods in
PWR environments following the PIRT
scoring activity. The panellists did note that
SCC, fatigue cracking, and pitting should be
minimal for alloy 690 in service, althoughgood water chemistry must be maintained.
SCC is a known issue for alloy 690 in all its
forms. SCC has not been observed to date in
actual service, although it has been
observed in some laboratory experiments.

Terms of reference of the 3rd project term
clearly states that any “service-induced
material failures involving alloy 690 oralloy 52/152 shall be considered forinclusion in the event database.”
According to NUREG-1841, “As of
31 December 2004, it [Alloy 690] wasbeing used in about 43% of the operating
PWRs in the US. Of the 577 070 alloy 690
TT tubes placed in service, only 333 tubes(0.06%) have been plugged after
approximately 173 calendar years ofoperation. Most of these tubes (65%)
were plugged prior to placing the steam
generators in service. The dominant inservice degradation mode, responsible forabout 24% of the plugged tubes, hasbeen wear caused by a support structure
or loose part. No corrosion or cracking
had been detected as of the time thisreport was prepared.”

SCC of severely mechanically strained
stainless steels in PWR primary water,e.g., weld shrinkage strains in heataffected zones, pressuriser heater
cladding and cold-bent piping elbows.

Not explicitly addressed in the PIRT scoring
activity.

No OPEX has yet been identified.

Cyclic thermal loading in the primarycircuit hot leg caused by the differencein temperature between primarycoolant streams exiting the centre andperiphery of low leakage cores which,despite turbulence, can persist up to
the steam generator channel head.
While studies have shown that thiscyclic loading does not produce
significant thermal fatigue issues, thepossible effect of ripple loading onother degradation mechanisms (suchas stress corrosion cracking) could be aconcern.

Not explicitly addressed in the PIRT scoring
activity.

No OPEX has yet been identified.

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Table 4-3: Comparison of PMDA/EMDA conclusions
with CODAP event database insights (Part 2) (cont’d)
PWR primary system components
PMDA: 25 < T ≤ 40 years

EMDA: 40 < LTO ≤ 80 years

CODAP OPEX data insights

Secondary side degradation (e.g. SCC) of steam
generator tubes by potentially corrosive
species not previously appreciated to be
widespread in the secondary system and which
may also affect “more resistant” tubing
materials such as alloy 690.

Ferritic stainless steel grade 405 and 409 are also
used in the secondary side of steam generators in
PWRs. The expert panel rated these materials inthe secondary environment for SCC and crevice
corrosion. In both cases, the panellists noted theimportance of specific water conditions, which
will drive susceptibility for these alloys.

Refer to Section 3.2.3 of thisreport.

Potential degradation related to end of fuelcycle primary water chemistry, especiallyduring cycle stretch-out, when theconcentration of boron may be reduced below
the minimum recommended value, thereby
leading to possible localised corrosion due toexcess LiOH, in, e.g. the boiling crevices in thepressuriser.

Not explicitly addressed in the PIRT scoring
activity.

No OPEX has yet been
identified.

Accelerated environmental effects on fatigue of
austenitic stainless steels and nickel alloys at
low electrochemical corrosion potentials, as
demonstrated in the laboratory and therefore
anticipated to apply in the field.

Not explicitly addressed in the PIRT scoring
activity.

No OPEX has yet been
identified.

Less-than-anticipated fatigue resistance ofsome components due to the current
uncertainty in the characterisation of the cyclicloading conditions, particularly thermalloading, and the material response under thosespecific conditions.

Not explicitly addressed in the PIRT scoring
activity.

No OPEX has yet been
identified.

Reduced fracture resistance and increased
susceptibility to stress corrosion cracking ofcast stainless steels (and perhaps ferritecontaining stainless-steel cladding and welds)
as a result of thermal ageing.

While CASS components have an excellent
performance record, extended service will resultin increased exposure time to the primary water
environment and stress. The expert panel
considered eight different categories of SCC forcast austenitic alloys in PWR environments. The
panel scored the CASS and HAZ in primary waterin the low Knowledge, moderate Susceptibility
category. As above, this was driven by uncertainty
in the effects of thermal ageing over an extended
operating period on microstructure andmechanical performance. Other categories wereranked at lower susceptibility, although thermal
ageing effects were considered by the panel as
the key factor for this mode of degradation in allenvironments. One potential knowledge gap for
CASSs was identified for both PWR and BWR
environments using the PIRT data. Specifically,
the effects of long-term thermal ageing forextended operating periods may drive changes inmechanical or corrosion performance that are
relatively unknown.

The current version of thedatabase includes
12 records involving CASS
components, none of which
has TAE as cause of failure.

“Abusive” grinding or machining during weld
preparation or surface finishing and theiradverse effects on stress corrosion and fatigue
crack initiation.

Not explicitly addressed in the PIRT scoring
activity.

This is covered in CODAP;
extensive OPEX exists.
Details can be found in the
Topical Report on SCC [21].

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Figure 4-1: The BWR OPEX involving SCC of safety class 1 alloy 182/82 welds
100

94

No. of alloy 182/82 SCC failures

90
80
70
60
50
40
30
20

25
20
12

11

10
0
1970-79

1980-89

1990-99

2000-092010-19

Note: In interpreting this OPEX summary it should be noted that by approximately 1990 most BWR operators had implemented (or werein the process of implementing) hydrogen water chemistry (HWC). Also, by this time, other types of IGSCC mitigation plans had been
implemented, or were in the process of being implemented.

4.2.1 Note on SCC-susceptibility of alloy X-750
According to Reference [86]:“Alloy X-750 with high temperature solution annealing (1 093°C) and single stage (704°C)
HTH ageing treatment is very resistant to SCC in high temperature primary water if any
surface oxide layer due to heat treatment is removed by machining. However, this SCC
resistance can be reduced by the presence of sulphide inclusion clusters in the matrix.
The presence of certain surface layers built up during the last step of the manufacturesuch as the ageing treatment. If the final step is the ageing treatment, the nature of theatmosphere of the furnace has a direct influence on the behaviour of the material whenstressed in high temperature primary water. Treatment under vacuum can induce SCC.
Heat treatment under nitrogen or machining after heat treatment leads to the best
resistance to SCC. The exposure of these surfaces to primary water at high temperature
modifies the surface conditions. Stress corrosion cracking may be then triggered whentensile stresses are applied. This phenomenon can be accounted for by the formation ofgrain boundary damage in the sub-surface matrix during the ageing heat treatment or
during the exposure to primary water depending on the composition of the surface layer.
The composition of the surface film may also have a direct influence on the initiation
phase for SCC. This effect of the surface layer on the SCC initiation explains thebeneficial effect of polishing the control rod tube guide pins in the French PWRs andsuggests that a better resistance may be obtained by oxidising the surface in a welldefined environment to create a surface layer enriched in Cr that protects the subsurface matrix against internal oxidation.”

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSISIn Germany, Alloy X-750 material degradation was first discovered in 1978 at KWB-A(Biblis A). 1 An ultrasonic examination revealed that several of the Inconel X-750 baffle-former
bolts were defective. When the bolts were removed, some of them were found to be broken andothers had cracks of different sizes. During 1980 through 1989, flawed baffle-former bolts werediscovered in another four reactor units: KWB-B (Biblis Unit B), GKN-1 (Neckarwestheim-1), KKU(Unterweser) and KKS (Stade). These PWR units were commissioned from 1972 to 1978. Studiesof the cracks showed that the cause of damage was IGSCC. The crack locations in the damaged
baffle-former bolts were in the transition between the bolt head and the necked-down shaft(extending into the hexagonal socket), within the necked-down shaft and in the transition
between the shaft and the thread and in the thread itself. All the baffle-former bolts were
examined by ultrasonic testing. Damage to Inconel X-750 baffle-former bolts was subsequently
found in other German plants as well. No defects were found in bolts made from austenitic
stainless steel.
In the German PWR plants the baffle-former bolts made of Inconel X-750 (a total of over
5 100 bolts) were replaced with bolts made of material 1.4571 (corresponding to Type 316Ti, cold
worked material). In the newer plants constructed after the early detection of this kind ofcracking, bolts made from the alternative material were used. However, in 2005 someindications in the austenitic baffle-former bolts were detected at several Siemens/KWU-type
PWR plants [105].
Reference [86] documents the French operating experience with Inconel X-750 bolts. After
11.6 EFPYs, 5-of-80 core barrel bolts were reported broken in 1984 during in-vessel visualinspections at Chooz-A. These bolts were replaced with the same material, but heat treated forbetter SCC resistance. Metallurgical examinations attributed the failures to intergranular SCC.

4.2.2 Note on PWR CRDM thermal sleeve degradation
Wear on CRDM thermal sleeves was initially noted in 2007 at a Westinghouse designed plant,
while performing vessel head penetration J-groove weld examinations. Since this discovery,
similar wear has been noted at other Framatome and Westinghouse designed PWR units; mostrecently in August 2019. The wear indications on the thermal sleeves were in the area where
the thermal sleeve exits the CRDM head adapter tube. The wear is attributed to the thermal
sleeve contacting the inside diameter of the CRDM head adapter tube due to flow-induced
motion of the thermal sleeve.

4.2.3 Note on raw water-cooling system piping OPEX
The safety-related raw water-cooling piping operates in a corrosive environment. The piping
system designs vary extensively. The system takes cooling water from a freshwater lake,reservoir, river or ocean. Depending on the specific system design, multiple important safetyfunctions may be cooled directly by raw water (“open-loop” system), or indirectly by a closedloop component cooling water system.
The safety significance of raw water-cooling piping integrity is evaluated within the scopeof the internal flooding (IF) probabilistic safety analysis (PSA) element of plant-specific PSA.
Insights from the review of IF-PSA studies indicate that for some plants a break of a raw watercooling pipe is a major contributor to core damage frequency (CDF). Therefore, the quantitative
assessment of raw water-cooling pipe break frequency in some cases is an important aspect of
IF-PSA.

1.

Rubel, H., J. Tautz and G. Micheel (1987), “Inconel~X-750 als Werkstoff für Kerneinbauten und Kernbauteile– Bisherige Erfahrungen und Austauschaktionen,” Proc. 13. MPA-Seminar, Stuttgart, Germany, 8-9 October
1987, https://inis.iaea.org/search/search.aspx?orig_q=RN:20059527.

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4.3 Basic principles of PEO/LTO-OPEX data analysis
The opportunity to identify and evaluate the potential effects of ageing on passive metallic
component integrity was made possible by two important factors. The first is the commitmentof the NEA and its members countries to support technical activities in the areas of structural
integrity and ageing management (e.g. WGIAGE 2 ) and operating experience data exchange(e.g. CODAP). The second is the availability of insights from the application of analysis methods
and techniques to explore operating experience data in a structured manner. Thus, identifying
trends in operating experience is made possible by the continuous and systematic applicationof relevant data analysis methods combined with a well maintained OPEX database.
There are multiple technical approaches to the analysis of age-dependent component
failure rates [106]. In this report only simple “visual tests” of graphical plots are performed toobtain high-level insights into age-dependent material performance. It is a first step in a more
rigorous analysis of the OPEX data. The objectives of ageing factor analysis account for:•

Ageing plant fleet. Numerous nuclear plants have entered into PEO/LTO. Ageing factor
assessments measure the effectiveness of an existing ageing management programmewith respect to long-term material performance. Specifically, a detailed identification
and assessment of temporal trends would generate insights into the effectiveness ofmaterial ageing management and to determine whether event recurrence patterns exist.

•

Renewal processes. Piping systems are routinely replaced-in-kind or upgraded, forexample, by replacing original material with material known or assumed to be resistantto degradation. Also, there is plant-to-plant variability in piping system design, which
impacts the material degradation propensity.

•

New operating experience data. The operating experience with metallic passive
components is continuously being updated. It needs to be determined if the datacollection process is sufficient to support quantitative ageing factor assessments.

•

Enhancements in reliability and integrity management (RIM). Plant life extension
initiatives together with applications of non-destructive examination (NDE) techniques
and inspection qualification processes continue to evolve. Embedded in the operating
experience data are effects of NDE and changes in the reporting of pipe failures. It needs
to be determined if the OPEX data reflects improvements in RIM programme
implementation or unanticipated changes in material performance.

As seen in Figures 4-2 and 4-3, the occurrence rate of pipe failure shows distinct negativeand positive trends over the nuclear power plant lifetime. In Figure 4-2 the pipe failure rate wascalculated as the total number of observed pipe failures in an arbitrarily selected 3-year interval
over the total number of plants in that interval that could produce a failure. The OPEX that isrepresented in Figure 4-2 covers all systems and all safety classes. Listed below are someobservations:

2.

106

•

An inverted “bathtub curve” may be discerned from this data processing step. However,further analysis of the OPEX is required to reach conclusions about ageing-dependency.
Included in Figure 4-4 are BWR- and PWR-specific insights about degradation
mechanisms and their implications on plant ageing management. For example, the BWR
plants experienced an increasing pipe failure rate during the first ten years of operation.
This increase was attributed to IGSCC. Subsequent regulatory requirements and industry
initiatives to address IGSCC have had significant mitigating effects. The IGSCC causal
factors are well understood and controlled.

•

An uptick and increase in the calculated pipe failure rate is observed for plants that haveoperated for >35 and >41 years, respectively. This increase is mainly attributed tocorrosion mechanisms acting on raw water piping systems, e.g. circulating water system,
service water system and fire water system piping. Examples of corrosion mechanisms
include crevice corrosion, microbiologically influence corrosion and pitting corrosion.

See the dedicated WGIAGE page on the NEA website: www.oecd-nea.org/wgiage.

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Age-dependent pipe failure rate estimation can be performed according to different
analysis strategies to obtain piping reliability parameters as a function of the age of an affectedpiping component at the time of failure observation, or as a function of the temporal changes
in the piping operating experience. The term “temporal change” means that descriptive failure
statistics (mean, upper/lower bound, etc.) change over time. The temporal trends may be due toageing, such as change in the physical properties of piping material (e.g. thinning or cracking),
but may also be due to changing reporting routines and the data collection processes.
In the analysis of piping reliability the term “cohort effect” is sometimes used to describe
variations in observed structural integrity factors (e.g. onset of crack initiation and subsequent
crack growth) as a function of operating time, plant age, plant design generation anddegradation mitigation implementation strategies (e.g. full structural weld overlay, peening,
induction heating stress improvement). The commercial nuclear power plant designs have
evolved, and those reactor units designed in the 1960s and commissioned in the early 1970s
exhibited quite different service experience histories than reactors designed and commissionedin later time periods.
As one example, cohort effects relating to intergranular stress corrosion cracking (IGSCC) of
boiling water reactor (BWR) primary system unirradiated stainless steel piping are particularly
strong. Those BWR units commissioned in the 1960s and 1970s exhibited a very high IGSCC
incidence rate, which was attributed to a lack of recognition of the relationships between theoperating environments (e.g. water chemistry), material characteristics (e.g. carbon content)
and stress conditions. The worldwide service experience with IGSCC in Code Class 1 piping issummarised in Figures 4-4 and 4-5 (a total of 1 126 IGSCC failure records). The service experienceis organised by “IGSCC Class” where each class corresponds to a uniquely defined event
population:•

IGSCC-1 (Gen I) corresponds to the first BWR design generation (essentiallydemonstration plants). These units were brought online in the 1960s and except for one
unit, they have all been decommissioned.

•

IGSCC-2 (Gen II “Early”) corresponds to the second BWR design generation; plantscommercialised in early 1970s. Several units have entered into an extended period of
operation, beyond 40 years, and a few of units are undergoing decommissioning.

•

IGSCC-3 (Gen II “Midi”) corresponds to the third BWR design generation. Multiple IGSCCmitigation projects were implemented in the latter part of the 1980s.

•

IGSCC-4 corresponds to two specific BWR NSSS design generations (SWR69 and SWR72,
respectively) that utilised stabilised austenitic stainless steel materials. This class is limited
to German plants for which IGSCC-mitigation consisted of a re-design of the primarypiping systems. The steps taken have significantly reduced the IGSCC susceptibility.

•

IGSCC-5 (Gen II “Late”) corresponds to BWR designs for which IGSCC-mitigationconsideration was an integral part of the original piping design (e.g. use of nuclear grade
austenitic stainless steel) and plant operating practices (e.g. enhanced primary waterchemistry control).

Not only do Figures 4-4 and 4-5 include five different sets of OPEX data; the charts show
how the IGSCC incidence rates have changed over time. Correlating the five event populations
with corresponding exposure terms (plant population that generated the failures, operatingtime and weld populations) provides a basis for establishing “a priori” failure rate distributions
for each of the IGSCC categories.

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Figure 4-2: Pipe failure rate as a function of the age of the piping
at the time of failure of safety- and non-safety-related piping

No. of failuers per plant in age interval "A"

6

5

4

3

2

1

0
24-26

27-29

30-32

33-35

36-38

39-41

42-44

45-47

Plant age interval "A" at the time of failure
BWR - 1 270 events

PWR - 2 275 events

No. of small-diameter (≤ DN50) pipe failures in time interval "T"

Figure 4-3: The small-diameter (≤ DN50) pipe failure OPEX
400
350
300
250
200
150
100
50
0
1970-79

1980-89
RCPB piping

108

1990-99
Safety Class 2

2000-092010-19

Safety Class 3

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSIS

Figure 4-4: The BWR-specific IGSCC operating experience data
1 000

No. of IGSCC events

100

10

1
1970-74
IGSCC-1 (Gen I)

1975-79

1980-84

1985-89

IGSCC-2 (Gen I early)

1990-94

1995-99

IGSCC-3 (Gen II 'midi')

2000-04

2005-092010-14 2015-2019

IGSCC-4 (German data)

IGSCC-5 (Gen II late)Figure 4-5: The IGSCC “incident rates” by time period and BWR design generation

Fraction of total IGSCC event population

1.0E+01.0E-1

1.0E-2

1.0E-3

1.0E-41970-74
IGSCC-1 (Gen I)

1975-79

1980-84

IGSCC-2 (Gen I early)

1985-89

1990-94

1995-99

IGSCC-3 (Gen II 'midi')

2000-04

2005-092010-14 2015-2019

IGSCC-4 (German data)

IGSCC-5 (Gen II late)

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4.4 PEO/LTO-OPEX data analysis insights – safety-related piping
An ever-present data analysis issue is concerned with the completeness of the OPEX data, thestatistical significance of an event population, and the applicability of the service experience
across different plant design generations. Figures 4-6 through 4-14 summarise the safety-related
piping OPEX. The following observations are made with respect to the ability of the CODAP event
database to support material degradation issues:

110•

Figure 4-6: The OPEX is a summary of the piping OPEX for plants that have been in
operation for 25 years or longer. The data is organised by plant type, safety class andnominal pipe size. For BWRs and PWRs the bulk of the OPEX is associated with safetyclass 3 piping systems, which mainly comprise reactor auxiliary systems (e.g. component
cooling and spent fuel pool cooling) systems and support systems (e.g. instrument air andessential service water).

•Figure 4-7: BWR and PWR primary pressure boundary piping (safety class 1) OPEX. Two
key observations are: First, there is a decline in medium- and large-diameter pipedegradation and failure of BWR-piping since the early-1990s, and this is well understood;
Reference [21]. Second, small-diameter fatigue-induced pipe failures continue to cause
operational impacts (e.g. plant shutdowns).

•

Figure 4-8: Same event population as in Figure 4-7 but with the pre-1995 OPEX screened
out. The OPEX is organised by time period in which a failure was observed and bymaterial degradation mechanism. Noteworthy are the seemingly increasing trends inhigh-cycle fatigue (HCF) and thermal fatigue (TF) failures.

•

Figure 4-9: Same event population (safety class 1) as in Figure 4-7. The event population
has been normalised against the plant population that produced the failures. The uptick
noted from plant age > 39 years is almost exclusively due to small-diameter (≤ DN50)
pipe failures.

•

Figure 4-10: Summary of the OPEX on safety class 2 piping located inside the containment.

•

Figure 4-11: Summary of the safety-related support system piping OPEX. Support
systems at each plant are unique. Extensive operating experience exists for the essential
service water (ESW) system. In some plants important safety functions (e.g. safeguard
equipment room coolers, emergency diesel generator coolers, component cooing water
heat exchangers) are cooled directly by service water (i.e. raw water). Other plants haveimportant safety functions served by a closed loop component cooling water system.

•

Figure 4-12: Same data as in Figure 4-10, but normalised against the 1995-99 time period
to show the temporal trends in the pipe failure data.

•

Figure 4-13: Summarised in this chart is the high-cycle fatigue OPEX, which in this caseencompasses all safety-related systems; safety classes 1 through 3. The event population
has been normalised against the plant population that produced the failures.

•

Figure 4-14: Summarised in this chart is pipe failure OPEX associated with emergency
diesel generators (EDGs). The OPEX includes two classes of piping system boundaries:
1) diesel fuel oil supply piping; and 2) lubricating oil piping.

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSIS

Figure 4-6: The safety related piping OPEX by plant type, safety classand pipe-diameter vs. plant age at the time of pipe failure

3

Large-Diameter
Medium-Diameter
Small-Diameter

2

PWR

Large-Diameter
Medium-Diameter
Small-Diameter

1

Large-Diameter
Medium-Diameter
Small-Diameter3

Large-Diameter
Medium-Diameter
45-49
40-44
Medium-Diameter

35-39

2

CANDU

Small-Diameter

30-34

Small-Diameter

25-29

1

Medium-Diameter
Small-Diameter3

Large-Diameter
Medium-Diameter
Small-Diameter

2

BWR

Large-Diameter
Medium-Diameter
Small-Diameter

1

Large-Diameter
Medium-Diameter
Small-Diameter
0

20

40

60

80

100

120

140

160

Note: Small-diameter: ∅ ≤ 50 mm; medium-diameter: 50 < ∅≤ 250 mm; large-diameter: ∅ > 250 mm.

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Figure 4-7: The BWR and PWR primary piping system OPEX by event date
400

No. of RCPB pipe failures

350
300
250
200
150
100
50
0
1970-74

1975-79

1980-84

1985-89

1990-94

1995-99

2000-04

2005-092010-14

BWR Small-Diameter

BWR Medium-Diameter

BWR Large-DiameterPWR Small-Diameter

PWR Medium-Diameter

PWR Large-Diameter

2015-19

Figure 4-8: RCPB piping OPEX by degradation mechanism and time period
70

No. of RCPB pipe failures

60
50
40

HCF
LCF

30

TF
SCC

20
10
0
1995-99

2000-04

2005-092010-142015-19

Event date112

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSIS

Figure 4-9: Normalised RCPB piping OPEX
No. of pipe failures per reactor unit and time interval "T"

5
4.5
4
3.5
3
2.5
2
1.5
1
0.5
0
0-2

3-5

6-8

9-11 12-14 15-17 18-20 11-23 24-26 27-29 30-32 33-35 36-38 39-41 42-44 45-47
Age of reactor unit at the time of failure
BWR - 1 447 failures

PWR - 671 failures

Figure 4-10: Safety class 2 piping OPEX by system group and time period
STEAM

PWR

SIR
RAS
FWC
STEAM

BWR

SIR
RAS
FWC
0

20

40

60

80

100

120

140

160

No. of pipe failures
1995-99

2000-04

2005-092010-142015-19

Note: FWC = feedwater system; RAS = reactor auxiliary systems (e.g. chemical and volume control, steam generator blowdown, reactor
water cleanup); SIR = safety injection and recirculation; STEAM = main steam.

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSIS

Figure 4-11: Safety class 3 piping OPEX by system group and time period

2015-20192010-2014

2005-2009

2000-2004

1995-1999
1

10

100

1 000

SW - Service water

1995-1999
279

2000-2004
294

2005-2009331

2010-2014473

2015-2019506

IA - Instrument air

14

17

14

19

17

CC - Component cooling

32

27

22

22

17

No. of pipe failures in indicated period

Figure 4-12: Temporal trends in safety class 3 piping OPEX
No. of pipe failures in indicated period normalised against
1995-1999

2
1.8
1.6
1.4
1.2
1
0.8
0.6
0.4
0.2
0
2000-2004

2005-2009CC

114

2010-2014IA

2015-2019SW

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSIS

Figure 4-13: OPEX on HCF in safety-related piping
200
180

No. of HCF failures in period "T"

160
140
120
100
80
60
40
20
0
1970-74

1975-79

1980-84

1985-89

1990-94
BWR

1995-99

2000-04

2005-092010-142015-19

PWR

Figure 4-14: Emergency diesel generator fuel oil and lubricating oil pipe failure OPEX
30

No. of EDG pipe failures in period "T"

25

20

15

10

5

0
1970-79

1980-89
Diesel fuel oil pipe

1990-88

2000-092010-19

Lubricating oil pipe

4.5 PEO/LTO-OPEX data analysis insights – non-safety-related piping
This section addresses OPEX associated with the “balance-of-plant” (BOP) piping, which in thisreport includes the steam cycle piping systems within the turbine building. The second Topical
Report [30] summarised the operating experience with flow-accelerated corrosion (FAC;
Figure 4-15) in the CODAP member countries/economies. The Topical Report covered the OPEX
for the period 1970 through 2012. Figure 4-16 summarises the FAC failure experience throughthe end of calendar year 2020.

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSIS

Figure 4-15: Flow-assisted pipe wall thinning mechanisms

All plant operators have implemented a long-term FAC monitoring programme to prevent
piping failures in high energy (two-phase and single-phase) carbon steel piping systems. The
programmes are developed by each utility using plant-specific conditions, industry-wide
operating experience, engineering judgement, NDE techniques and computer analysis of highenergy carbon steel systems.
The FAC operating experience data is recorded in different types of information systems
such as condition reports, work orders, and databases of FAC monitoring programmes.
Reportable occurrence reports or licensee event reports capture the major structural failures. In
addition, information notices (or equivalent documents) are issued by regulators to inform
licensees about generic issues. 3 Members of the CHECWORKS® Users Group (CHUG) exchange
operating experience data biannually. CHUG does not operate a formal FAC OPEX database,however.

Figure 4-16: FAC operating experience
900

No. of replaced FAC components

800
700
600
500
400
300
200
100
0
1971-1980

1981-1990
BWR

1991-2000
CANDU

PWR

2001-20102011-2020

WWER

3. In 2019 the US Nuclear Regulatory Commission issued Information Notice 2019-08, Flow-Accelerated
Corrosion Events, www.nrc.gov/docs/ML1906/ML19065A123.pdf.

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PEO/LTO OPERATING EXPERIENCE DATA ANALYSIS

4.6 PEO/LTO-OPEX data analysis insights – fire water system piping
The fire protection (FP) water system is important to plant safety. The assessment of thestructural reliability of the piping is an integral aspect of internal flooding probabilistic safetyassessments. Major leaks and breaks can cause internal flooding and spraying of, for example,
vital electrical equipment. The source of fire water can be the ultimate heat sink (e.g. untreated
raw water from a lake or river) or domestic water stored in a tank. 4 The fire water system piping
OPEX is summarised in Figure 4-17 and it is limited to through-wall defects that have resulted
in minor to significant pipe through-wall flow rates. CODAP does not include OPEX on blockages
of FP piping due to build-up of corrosion products.

Figure 4-17: Fire protection water system pipe failure experience
200
180
160

No. of pipe failures

140
120
100
80
60
40
20
0
1970-79

1980-89

1990-99

2000-092010-19

Period in which a pipe failure was discovered
In-plant

Buried / below ground

Most plants have implemented “MIC monitoring programmes” which include the use of
NDE techniques, chemical treatment of fire water, and periodic high-velocity flushing of certain
pipe sections. In terms of length of piping and in-piping water volume, the typical fire water
piping system is very extensive. A typical piping system consists of a large-diameter ring header(pipe which encircles the plant buildings), large-diameter in-plant distribution header, mediumdiameter riser piping and small-diameter sprinkler head piping. Except for piping directly
associated with the FP pump (including keep-fill pump) suction, discharge and recirculationlines, the predominant degradation mechanism acting on the FP piping is microbiologically
influence corrosion (MIC), pitting and crevice corrosion and uniform corrosion. Pinhole leaks are
relatively frequent occurrences. Unless there is a plant operational impact, there is no process
for external reporting of such occurrences.

4. Domestic water (or potable water) is used for indoor and outdoor household purposes (cleaning,
washing). It refers to water that is self-supplied (e.g. from a well) or water withdrawn directly from amunicipal water distribution system.

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Depending on the FP piping system design, water hammer may be a concern for the piping
integrity. Some plants systems have experienced severe water hammer pressure pulses [107].
These severe water hammer events have occurred in so-called pre-activation systems wherecertain piping sections are normally kept empty and only filled upon a demand. Many plants
have implemented design modifications to prevent water hammer in these systems. Examples
of such design modifications include the installation of vacuum breakers at the top of criticalpiping system risers, and replacement of electric FP pump bypass valves with relief valves inaccordance with the US National Fire Protection Association (NFPA) standard NFPA-25 [108].

4.7 PEO/LTO-OPEX data analysis insights – non-piping passive components
Except for PWR vessel head penetration (VHP) ageing degradation (Figure 4-18), in CODAP, thepopulation of non-piping passive component failure events is limited to “selected representative
events” (SREs) 5. Based on the scarce data in CODAP it is therefore not possible to develop any
generalised OPEX-based conclusions regarding the age-dependency of the various material
degradation scenarios for reactor internals, or other types of passive components.

Figure 4-18: PWR vessel head penetration OPEX
80

No. of cracked vessel head penetrations

70
60
50
40
30
20
10
0
1991-1995

1996-2000

2001-2005

2006-20102011-20152016-2020

Period in which a cracked vessel head penetration was discovered

5.

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In the CODAP project the term “SRE” has a specific meaning. The term is used to indicate that adegraded condition of specific and well characterised attributes also applies to multiple degradedlocations (i.e. component boundaries) within a specific system. A single record for which all relevant
database fields have been filled in also is representative of all the other degraded/flawed locations
within that system. As an example, during in-service inspection 24 reactor vessel head penetrations
were found to have recordable flaw indications. Following a complete flaw size characterisation using
a qualified NDE technology and structural evaluation, 18 of the 24 vessel head penetrations were
determined not to be fit for continued operation. Instead of creating 18 records, only a single SRE iscreated leaving it up to a future database user to expand the database content as needed.

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There are two groups of SREs: 1) Extensive legacy OPEX available and dating from the 1970s
but not generally accessible for detailed analysis, and 2) recent first-of-a-kind OPEX for which it
is known that degradation progresses slowly for a specific combination of component, material
and environment. An example of the former is baffle-former bolt (BFB) degradation. The
population of damaged and replaced BFBs is in the many thousands, and it would therefore
require a formidable effort to establish a comprehensive and (in a statistical sense) complete
BFB OPEX database. In the current version of the database there are only a handful of records
on BFB failures, including:•a single SRE on alloy X-750 BFB failures discovered in the 1970s;•

a single SRE on Ti-stabilised austenitic stainless steel BFB failures in the early 2000s;•

a few SREs on unstabilised austenitic stainless steel BFB failures post-2005.

An example of the latter SRE-type would be material flaws discovered in the outsidediameter PWR core barrel (refer to Section 2.5 of this report). The current (mid-2021 version of
the CODAP event database) includes a single first-of-a-kind event. This material ageing
degradation discovery was made after 33.7 EFPYs of operation. To ensure a more rigorous“treatment” of reactor internals in CODAP, this type of event could potentially be a benchmark
for the expanded consideration of reactor internals in the CODAP project.

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119CONCLUSIONS AND RECOMMENDATIONS

Chapter 5. Conclusions and recommendations

The objective of this report was to document observations from evaluations of the operating
experience with passive metallic components in commercial nuclear power plants during periodsof extended operation (PEO) and long-term operation (LTO). Specifically, this report addresses
trends and patterns in material degradation as a function of plant age The data collection period
for this report ended in 2020, and only the validated data was used for this report. It is a topic thatfor a very long time has received considerable attention from the point of view of reactor
regulations with respect to ageing management programmes, codes, and standards with respect
to reliability and integrity management, and the many supporting research programmes. In
preparing this report, simple visual examinations of graphical plots were made to reachpreliminary insights about the different material degradation mechanisms as a function of plantage. Summarised below are the conclusions and recommendations of this study.

5.1 Main conclusions
A five-step approach was used to reach conclusions about age-dependent material degradation
mechanisms during PEO/LTO. The CODAP event database content was compared and contrasted
against the results of two expert panels on material degradation [17][27], a questionnaire on longterm operation by the Working Group on Integrity and Ageing of Components and Structures(WGIAGE) [32], and the 2019 “International Workshop on Age-Related Degradation of Reactor
Vessels and Internals” [39]. Below are the conclusions on how the technical scope and content of
the CODAP event database addresses age-dependent passive component material degradation.

WGIAGE vs. CODAP
Several respondents to the NEA WGIAGE questionnaire assigned a “high priority” to the following“technical areas of mutual interest related to ageing-dependent degradation of materials duringlong-term operation of nuclear power plants”:•

irradiation embrittlement of reactor pressure vessel materials;•

irradiation embrittlement of reactor internals.

The CODAP project collects information on material degradation of reactor internals. So far,the data exchange has focused on selected representative events. The project does recognise
the importance of irradiation assisted, age-dependent degradation of reactor internals. The
complexity in accessing the relevant OPEX is also recognised. Invariably the reactor internals’
non-conforming conditions are discovered during in-service inspections and the resulting
findings and material flaw evaluations are generally not readily accessible for third party review;see recommendations below.
Several of the respondents to the WGIAGE questionnaire also assigned a “high priority” to“pitting and crevice corrosion near deposits, metallic contact points and stagnant conditions,
including internal/external corrosion of below ground piping.” This class of material degradation
impacts the pressure boundary integrity of raw water-cooling systems, e.g. ESW and FP water
systems. Both system groups are important for safe plant operation. Although CODAP includes a
sizeable population of OPEX for these system groups, the technical content is skewed in that it
mainly consists of US OPEX, thus reflecting the associated national routines and requirements forreporting and evaluating corrosion failures. This could potentially impede the ability to reach

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121CONCLUSIONS AND RECOMMENDATIONS

generic conclusions about the importance of ESW and FP age-dependent degradation that are
applicable across all CODAP member countries/economies and plant types. It could be that the US
OPEX is an “outlier”.
Two of the respondents assigned high-to-medium priority to flow-accelerated corrosion(FAC) in high-energy piping. From a CODAP-perspective it is unclear why this is so; seerecommendations below.

PMDA/EMDA/WGIAGE vs. CODAPThe US NRC Office of Nuclear Regulatory Research has convened two expert panels on agedependent material degradation. The proactive material degradation assessment (PMDA; 2004
to 2006) [17] identified the different material degradation scenarios that could affect plantsystems for up to 40 years of operation. The expanded material degradation assessment (EMDA;
2012 to 2014) [27] extended the analysis time frame from 40 years to 80 years. In comparing thePMDA/EMDA and WGIAGE questionnaire results with CODAP, four questions were raised andthe associated observations are listed below:
1. What was the passive component operating experience during the 2004-to-2014 time
period; the start date of the PMDA expert panel and the conclusion of the EMDA expert
panel?
–A small sample of the OPEX for the given period is summarised in Figure 5-1. It showsa small but noticeable uptick in fatigue failures and an increase in corrosion failures.

2. Are the apparent and underlying causes of material degradation during PEO/LTO similar
to those experienced during early plant life? If not, what are the main differences?
–

Low-cycle fatigue, high-cycle mechanical fatigue and thermal fatigue continue tocause operational challenges. Hence, there is no fundamental difference between
early life and PEO/LTO. An increasing failure trend is observed for safety class 3 raw
water-cooling water piping 1 . In contrast, flow-accelerated corrosion (FAC)
monitoring programmes and stress corrosion cracking (SCC) mitigation processes
appear to have a positive impact on the longer-term material performance. CODAP
has published two reports on these topics [30][21]. New OPEX should determine theneed for producing Topical Report updates.

–

WWER reactors are in operation in three of the current CODAP member countries/
economies (Czechia, Finland and the Slovak Republic). In the WGIAGE questionnaire,
FAC acting on high-energy piping was assigned a “high” priority by one country.
CODAP includes only very limited OPEX on FAC in WWER plants, however. 2

3. Is it feasible to validate the WGIAGE questionnaire results [32] against the CODAP event
database? In other words, to what extent are the material degradation issues identified
in the Questionnaire Results also acknowledged and tracked by the CODAP project?
–

It is unclear whether the respondents to the WGIAGE questionnaire were aware of or
had access to the CODAP event database. Nevertheless, the purpose of CODAP is tocollect field experience data on degradation mechanisms that have produced failures.
CODAP does not collect data on minor RPV fabrication flaws.

4. Based on the operating experience evaluation results that are documented in this Topical
Report, are there any gaps in the lists of degradation scenarios that were identified by
the PMDA and EMDA expert panels or the WGIAGE Questionnaire?
–

1.
2.

122

No gaps identified.

Also referred to as essential service water (ESW) or nuclear service water (NSW).
A possible explanation relates to the extensive use of field segmentation and welding of pipe elbows
for installation in WWER secondary-side piping systems. Segmented pipe elbows are more prone to
FAC-induced wall thinning than is the case for the more common prefabricated cold- or hot-formed
elbows with smooth interior pipe wall surface.

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CONCLUSIONS AND RECOMMENDATIONS

Figure 5-1: CODAP OPEX for the PMDA/EMDA period (2004-2014) 3
120
RCPB fatigue
RCPB SCC

100

Class 3 COR
Class 3 FAD

No.of failures

80

Class 3 fatigue
Class 3 SCC

60
40
20
0
2004

2005

2006

2007

2008

2009

2010

2011

2012

2013

2014

The “Workshop” vs. CODAP
CODAP is concerned with degradation mechanisms that have produced failures. The intent of the“Workshop” was to exchange information on ongoing research on age-dependent degradation ofreactor vessels and internals. The CODAP project exchanges reactor internals OPEX on selected
representative events.

5.2 CODAP activities for the fourth term (2021-2023) and beyond
CODAP activities for the fourth term fall into three categories: 1) improvements to dataexchange protocols; 2) validation of the data analysis observations; and 3) benchmarking andresearch topics.

Data exchange protocol
Since the 1960s the exchange and analysis of nuclear power plant operating experience has
been a point of focus for nuclear regulatory authorities, technical support organisations and thenuclear industry at large. The processes and systems for collecting and disseminating operating
experience data continue to evolve. To paraphrase Y. Benoist 4, OPEX analysis and feedback is
an ongoing activity that is technically challenging and workload intensive. For the year 2021
and beyond, the CODAP Management Board (MB) is considering the following activities topotentially enhance the data collection process:
•

3.
4.

Identifying the current national routines for recording and submitting material
degradation information, including OPEX access, especially for those “issues” that do notcause a direct impact (e.g. reactor trip/turbine trip) on routine plant operation.
Information on events that result in reactor trip is recorded in readily available and

The PMDA Expert Panel was convened during 2004 to 2006 and the EMPDA Expert Panel was convened
during 2012 to 2014. COR = corrosion mechanisms (various), FAD = flow-assisted degradation mechanisms(various), SCC = stress corrosion cracking.
From a presentation made at the 6th EUROSAFE Forum, 8-9 November 2004. The theme of the conference
was “Learning from Experience – A Cornerstone of Nuclear Safety.” More information is available atwww.eurosafe-forum.org/eurosafe2004.

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123CONCLUSIONS AND RECOMMENDATIONS

accessible licensee event reports. Annex B summarises examples of how material
degradation OPEX is documented in the United States.
•

Reporting on the current national regulatory and industry routines/practices andrequirements for performing operability determinations or fitness for service assessments.
In particular, the question of how these evaluations provide relevant OPEX data forpotential submission to CODAP should be elaborated.

•

Developing an enhanced list of BWR, PWR and WWER reactor internal piece parts andconsidering the development of a plan to conduct the future reactor internals OPEX
exchange.

The national reporting routines and requirements for documenting passive component
degradation are extensive and multifaceted (see for example Annex B). Subject matter expertise(SME) is needed to extract and evaluate the extensive national material degradation and failure
information. While data access limitations may exist, the time needed to perform data mining
and evaluations and the language barriers (i.e. translating non-English highly technicaldocuments into English) could be the critical path inhibitors from the point of view of populating
the CODAP event database with new information.

Validating the observations of this Topical Report
The report investigated age-dependent material degradation of piping systems and reactor
internals. It is not clear if the identified trends and patterns are representative across all CODAP
member countries/economies. One way of validating the conclusions of this study is to:•

Select a single “system-material-environment” grouping for a limited and recent time
period (e.g. 2015-2019 ) to determine:–

The applicable country/economy-specific ageing management
requirements, regulations and operability determination procedures;–

The country/economy-specific OPEX;–

The country/economy-specific regulatory inspection programmes and findings fromsite inspections.

programme•

Evaluate the CANDU- and WWER-specific FAC OPEX to determine if and how it differsfrom the corresponding PWR-specific OPEX.

•The CODAP event database contains a significant population of FP and SW pipe failures
in US plants. Possible technical issues to address include:–

Determining if this “bias” is attributed to unique US piping system designs, material
selections, operating environments, regulations, reporting, and/or material ageing
management programmes, or whether it is more a reflection of the data exchange
process.

–

For plants that remain in extended periods of layup, addressing how the integrity of
piping systems is verified prior to a possible plant restart.

Benchmarking and research topics
The evaluation of possible age-dependent material degradation was done in a high-level
manner using simple “visual tests” of graphical plots. It is a first step in a more rigorous analysisof the OPEX. The CODAP Management Board will consider continuing the work done inproducing this Topical Report as a formalised benchmark involving multiple TSOs under an“umbrella project”, e.g., by developing a CSNI Activity Proposal Sheet (CAPS) entry that involves
CODAP/WGIAGE/WGRISK oversight and guidance. This could set the stage for widened CODAP
database access and usage and could encourage and engage early career material scientists,
fracture mechanics specialists and PSA engineers in the study of passive component integrity.

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Issue Management Tables, BWRVIP-167, Revision 4, 3002018319, Palo Alto, CA.
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Assessment of NEI 03-08, Revision 2, “Needed” Interim Guidance Regarding Baffle-Former Bolt
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REFERENCES[100] D’Agostino, A., S. Morrow, C. Franklin and N. Hughes (2017), Review of Human Factors Researchin Non-Destructive Examination, Office of Nuclear Regulatory Research, US Nuclear Regulatory
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Metal Welds, PNNL-28090, Pacific Northwest National Laboratory, Richland, WA.
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Stress Corrosion Cracks, PNNL-28427, Pacific Northwest National Laboratories, Richland, WA.
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Technologies, PNNL-28643, Pacific Northwest National Laboratories, Richland, WA, US.
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Power Plants,” Paper #485, Proc. 19th Int. Conference on Environmental Degradation of Materials in
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GLOSSARY OF TECHNICAL TERMS

Annex A
Glossary of technical terms

Annealing (or recovery annealing). Annealing is a heat treatment process used mostly to increase
the ductility and reduce the hardness of a material. This change in hardness and ductility is a
result of the reduction of dislocations in the crystal structure of the material being annealed.
Annealing is often performed after a material has undergone a hardening or cold working
process to prevent it from brittle failure or to make it more formable for subsequent operations.
ASME Section XI, Division 2. Issued in July 2019, this Division provides the requirements for thecreation of the reliability and integrity management (RIM) programme for advanced nuclear
reactor designs. The RIM programme addresses the entire life cycle for all types of nuclear power
plants. It requires a combination of monitoring, examination, tests, operation and maintenancerequirements that ensures each structure, system and component (SSC) meets plant risk andreliability goals that are selected for the RIM programme.
Backfitting. According to NUREG-1409 Revision 1 (2010), backfitting occurs when the NRC imposes
new or changed regulatory requirements or staff interpretations of the regulations or
requirements on nuclear power reactor licensees, certain nuclear power reactor applicants, or
select nuclear materials licensees. Backfitting is an integral part of the regulatory process and may
be needed when the staff addresses safety or security issues. The NRC may take a backfitting
action only after conducting a formal, systematic review to ensure that the action is defined andjustified. This process ensures discipline, predictability and optimal use of NRC and licensee
resources. The backfitting requirements are found in 10 CFR 50.109, “Backfitting,” 70.76,“Backfitting,” 72.62, “Backfitting,” and 76.76, “Backfitting.” Provisions analogous to the backfitting
requirements, referred to as issue finality provisions, are set forth in 10 CFR Part 52, “Licences,
Certifications, and Approvals for Nuclear Power Plants.”
Below ground piping. Buried piping in contact with soil or concrete and underground piping thatis below grade but is contained within a tunnel or vault such that it is in contact with air and islocated where access for inspection is restricted.
Boat sampling technique (BST). A technique developed for obtaining samples from the surface of
any operating component. The technique is a non-destructive surface sampling technique as it
does not cause any plastic deformation or thermal degradation of the operating component.
BST can be used, remotely and in water-submerged condition, with the help of a handlingmechanism. The samples are boat-shaped, having three mm maximum thickness and require
about three hours to get scooped from a location. The samples are used for metallurgicalanalysis to confirm the integrity of the operating component. BST incorporates mainly sampling
module, handling mechanism and electric and pneumatic sub-systems.
BONNA® pipe. A thin steel pipe embedded in reinforced concrete. It has rebar or a heavy wire
mesh embedded in the OD concrete.
Buried piping. Piping that is below grade and in direct contact with soil. Buried piping is provided
with corrosion protection such as coating and cathodic protection.
BWR core shroud access hole cover (AHC). Jet pump BWRs are designed with access holes in the
shroud support plate, which is located at the bottom of the annulus between the core shroud
and the reactor vessel wall. Each reactor vessel has two such holes that are located 180 degrees
apart. These holes are used for access during construction and are subsequently closed bywelding a plate over the hole. The covers and shroud support ledge are Inconel Alloy 600
material. The connecting weld material is also Inconel 600 (Alloy 182 or 82). In the worst case,

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GLOSSARY OF TECHNICAL TERMS

access hole cover cracking could progress through wall and cause the cover to detach partiallyor completely. A substantial flow path from the bottom head into the annulus region would becreated, impacting core flow distribution during normal operation. The distribution would be
detectable at significant levels. Such cracking would impact the boundary which assures 2/3
core coverage following a LOCA event. The consequence of cracking is high.
BWRVIP reactor internals AMP. A set of ageing management guidance provided by the EPRIBWRVIP issue programme for managing age-related degradation in BWR reactor internals. The
set of applicable BWRVIP guidance includes all NEI 03-08 “Mandatory” and “Needed” guidance
elements applicable to BWR reactor internals except for repair-based guidance. It also includes
administrative controls and procedures ensuring that guidance is revised as needed to address
new OE and other relevant data. Annex A of BWRVIP-315NP (2019) provides a list of theapplicable BWRVIP guidance documents associated with the BWRVIP reactor internals AMP.
Calandria. The “calandria” is the cylindrical reactor vessel that contains the heavy water moderator.
It is penetrated from end-to-end by hundreds of calandria tubes, which accommodate thepressure tubes containing the fuel and coolant.
Carbon fibre reinforced polymer (CFRP) repair. A large-diameter carbon steel pipe repair/rehabilitation technique for degraded (e.g. thinned) large-diameter carbon steel pipe. Speciallydesigned layers of carbon fibres and glass fibres are bonded to the internal surface of the piping.
Cathodic protection (CP). A corrosion protection technique in which the potential difference isapplied to buried piping from an external power source or a more anodic material (sacrificial
anode) for the purpose of making the piping behave in a cathodic manner. Using CP, thecorrosion rate is normally reduced to an acceptable level.
Cavitation. Cavitation damage may occur when there is a flowing liquid stream that experiencesa drop in pressure followed by a pressure recovery. Such a pressure drops (i.e. the difference
between the upstream pressure and the downstream pressure) can occur in valve internals
where the flow has to accelerate through a small area. As the fluid moves through the restricted
area, the fluid velocity increases, and the pressure decreases as shown by the momentum
equation (i.e. Bernoulli’s theorem). If the local pressure passes below the vapour pressure at theliquid temperature, small bubbles are formed. When the downstream pressure rises above thevapour pressure, these bubbles collapse. The collapse of the bubbles causes high local pressuresand very high local water jet velocities. If the collapsing bubbles are close enough to a solid
surface, damage to that surface will occur. The collapse of the numerous bubbles generates
noise and vibration. Most often, cavitation causes most of its damage by vibration (e.g. cracked
welds, broken instrument lines, loosened flanges). The erosion caused by cavitation also
generates particles that contaminate the process fluid.
CFRP system: A buried piping rehabilitation and repair technique. It consists of high strengthcarbon fibre fabrics and/or glass fibre fabrics, fully saturated in a two-part 100% solids epoxy
matrix. These laminates are bonded both longitudinally and circumferentially to the interior
surface of the pipe forming a structural lining within the pipe. This lining can be designed toreplace the degraded portions of the existing system without reliance on the degraded piping
for the life of the repair, except at the terminal ends of the repair.
Charpy impact test. Also known as the Charpy V-Notch test, it is a standardised high strain-rate
test which determines the amount of energy absorbed by a material during fracture.
Concrete encased piping (CEP). Below ground piping that is embedded in concrete. The piping is
not easily extracted nor is the interior pipe surface readily accessible for inspection. The CEP
category also includes piping recessed in plant building floors.
Crevice corrosion. Crevice corrosion occurs in a wetted or buried environment when a crevice orarea of stagnant or low flow exists that allows a corrosive environment to develop in a component.
It occurs most frequently in joints and connections, or points of contact between metals andnonmetals, such as gasket surfaces, lap joints and under bolt heads. Carbon steel, cast iron, lowalloy steels, stainless steel, copper, and nickel base alloys are all susceptible to crevice corrosion.
Steel can be subject to crevice corrosion in some cases after lining/cladding degradation.

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Cumulative usage factor (CUF). The design basis for Class 1 nuclear power plant components
subjected to cyclic service is in ASME Section III. Transient events that the plant might credibly
experience are evaluated to establish a design basis for plant equipment. The fatigue analyses
rely on the definition of design basis transients that envelope the expected cyclic service andthe calculation of a cumulative usage factor (CUF). The CUF is the ratio of the number of cycles
experienced by a structure or component divided by the number of allowable cycles of thatstructure or component. In accordance with ASME Section 111, Subsection NB, the cumulative
usage factor must be less than 1.0.
Cured-in-place-pipe (CIPP). A BP temporary repair method. A resin-saturated felt tube made ofpolyester, fibreglass cloth or a number of other materials suitable for resin impregnation, isinverted or pulled into a damaged pipe. It is usually done from the upstream access point(manhole or excavation).
De-alloying (selective leaching). As defined by NACE, “de-alloying” or “selective leaching” refers to
the selective removal of one element from an alloy by corrosion processes. A common example
is the dezincification of unstabilised brass, whereby a weakened, porous copper structure is
produced. The selective removal of zinc can proceed in a uniform manner or on a localised(plug-type) scale. It is difficult to rationalise dezincification in terms of preferential Zn
dissolution out of the brass lattice structure. Rather, it is believed that brass dissolves with Zn
remaining in solution and Cu replating out of the solution. Graphitic corrosion of grey cast iron,
whereby a brittle graphite skeleton remains following preferential iron dissolution, is a further
example of selective leaching. During cast iron graphitic corrosion, the porous graphite networkthat makes up 4-5% of the total mass of the alloy is impregnated with insoluble corrosion
products. As a result, the cast iron retains its appearance and shape but is weaker structurally.
Testing and identification of graphitic corrosion is accomplished by scraping through thesurface with a knife to reveal the crumbling of the iron beneath.
Delayed hydride cracking (DHC). Delayed hydride cracking is a subcritical crack growth mechanism
occurring in zirconium alloys as well as other hydride-forming materials that requires theformation of brittle hydride phases at the tip of a crack and subsequent failure of that hydride
resulting in crack extension. Hydrogen in solution in the zirconium alloy is transported to thecrack tip by diffusion processes where it precipitates as a hydride phase. When the precipitate
attains a critical condition, related to its size and the applied stress intensity factor, KI, fractureensues, and the crack extends through the brittle hydride and arrests in the matrix. Each stepof crack propagation results in crack extension by a distance approximately the length of thehydride.
Double-walled pipe. A double-walled pipe is a secondary contained piping system. It is a pipewithin-a-pipe, or encased in an outer covering, with an annulus (interstitial space) between thetwo diameters. The inner pipe is the primary or carrier pipe and the outer pipe is called thesecondary or containment pipe.
Enhanced visual examination (EVT-1). The EVT-1 method is intended for the visual examination of
surface breaking flaws. Any visual inspection for cracking requires a reasonable expectation thatthe flaw length and crack mouth opening displacement meet the resolution requirements of theobservation technique. The EVT-1 specification augments the VT-i requirements to provide more
rigorous inspection standards for stress corrosion cracking. EVT-1 is also conducted in accordancewith the requirements described for visual examination (i.e. VT-1) with additional requirements(such as camera scanning speed). Any recommendation for EVT-1 inspection will require
additional analysis to establish flaw-tolerance criteria, which must consider potential
embrittlement due to thermal ageing or neutron irradiation. Acceptance criteria methodologies tosupport plant-specific augmented examinations are documented in WCAP-17096-NP 1.

1.

Westinghouse Electric Company LLC (2009), Reactor Internals Acceptance Criteria Methodology and Data
Requirements, WCAP-17096-NP, Cranberry Township, PA.

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Erosion cavitation (E-C). This phenomenon occurs downstream of a directional change or in thepresence of an eddy. Evidence can be seen by round pits in the base metal and is often wrongly
diagnosed as FAC (see below). Like erosion, E-C involves fluids accelerating over the surface ofa material; however, unlike erosion, the actual fluid is not doing the damage. Rather, cavitation
results from small bubbles in a liquid striking a surface. Such bubbles form when the pressureof a fluid drops below the vapour pressure, the pressure at which a liquid becomes a gas. Whenthese bubbles strike the surface, they collapse or implode. Although a single bubble imploding
does not carry much force, over time, the small damage caused by each bubble accumulates.
The repeated impact of these implosions results in the formation of pits. Also, like erosion, thepresence of chemical corrosion enhances the damage and rate of material removal. E-C has
been observed in PWR stainless steel decay heat removal and charging system piping.
Erosion/corrosion (E/C). “Erosion” is the destruction of metals by the abrasive action of moving
fluids, usually accelerated by the presence of solid particles or matter in suspension. Whencorrosion occurs simultaneously, the term “erosion-corrosion” is used. In the CODAP event
database, the term “erosion/corrosion” applies only to moderate energy carbon steel piping(e.g. raw water piping).
Equivalent break size (EBS). The calculated size of a hole in a pipe given a certain through-wallflow rate and for a given pressure.
Fatigue usage factor. The ratio of the number of cycles anticipated during the lifetime of acomponent to the allowable cycles.
Flashing. Flashing occurs when a high-pressure liquid flows through a valve or an orifice to aregion of greatly reduced pressure. If the pressure drops below the vapour pressure, some of theliquid will be spontaneously converted to steam. The downstream velocity will be greatly
increased due to a much lower average density of the two-phase mixture. The impact of thehigh velocity liquid on piping or components creates flashing damage.
Flow accelerated (or assisted) corrosion (FAC). FAC is “a process whereby the normally protectiveoxide layer on carbon or low-alloy steel dissolves into a stream of flowing water or water-steam
mixture.” 2 It can occur in both single phase and two-phase regions. The cause of FAC is a specific
set of water chemistry conditions (e.g. pH, level of dissolved oxygen), and there is no mechanicalcontribution to the dissolution of the normally protective iron oxide (magnetite) layer on theinside pipe wall.
Full structural weld overlay (FSWOL). A structural reinforcement and stress corrosion cracking(SCC) mitigation technique through application of a SCC-resistant material layer around theentire circumference of the treated weldment. The minimum acceptable FSWOL thickness is
1/3 the original pipe wall thickness. The minimum length is 0.75√(R×t) on either side of thedissimilar metal weld to be treated, where R is the outer radius of the item and t is the nominal
thickness of the item.
Fusion (or heat fusion). In the context of HDPE piping design and installation, fusion techniques
are used to join pipes together. It is a welding process used to join two different pieces of athermoplastic. This process involves heating both pieces simultaneously and pressing them
together. The two pieces then cool together and form a permanent bond.
Grayloc region. An inspection area at or near the mechanical connection between a PHTS feeder
tube and a fuel channel of the CANDU primary system. Grayloc® is a trade name of a type of
mechanical connector that provides metal-to-metal seals in piping systems.
Harvesting (of irradiated materials). Accessing and segmentation of irradiated metallic materials
from commercial nuclear reactors that have ceased operations. The harvesting is done
according to detailed research plans. Detailed information on the research planning and its

2.

136

Dooley, R.B and V.K Chexal (2000), “Flow-accelerated corrosion of pressure vessels in fossil plants”,
International Journal of Pressure Vessels and Piping, Volume 77, Issues 2-3, https://doi.org/10.1016/S03080161(99)00087-3.

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GLOSSARY OF TECHNICAL TERMS

practical implementation is documented in PNNL-27120 Revision 1 (“Criteria and Planning
Guidance for Ex-Plant Harvesting to Support Subsequent Licence Renewal,” Pacific Northwest
National Laboratory, 2019) and ORNL/TM-2016/240 (“Light Water Reactor Sustainability Program:
Report on the Harvesting and Acquisition of Zion Unit 1 Reactor Pressure Vessel Segments,” Oak
Ridge National Laboratory, 2016).
Holiday in pipe coating. A holiday is a hole or void in the coating film which exposes buried piping
to corrosion.
Hydrostatic pressure test. A pressure test conducted during a plant or system shutdown at a pressure
above nominal operating pressure or system pressure for which overpressure protection isprovided.
INCEFA+ (INcreasing Safety in Nuclear Power Plants by Covering Gaps in Environmental Fatigue
Assessment). The INCEFA+ project is intended to deliver new experimental data and new
guidelines for assessment of environmental fatigue damage to ensure safe operation of European
nuclear power plants. Austenitic stainless steels will be tested for the effects of mean strain, holdtime and material roughness on fatigue endurance. Testing will be in nuclear light water reactor
environments. The three experimental parameters were selected in the framework of an in-kindproject during which the current state of the art for this technical area was developed. The dataobtained will be collected and standardised in an online fatigue database with the objective oforganising a Comité Européen de Normalisation (CEN) workshop on this aspect. The gaps inavailable fatigue data lead to uncertainty in current assessments. The gaps will be targeted so thatfatigue assessment procedures can address behaviour under conditions closer to normal plant
operation than is currently possible. The project term is 1 July 2015 until 30 June 2020. 3
Inclusion. An “inclusion” is a non-metallic impurity such as slag, oxide and sulphide that is present
in the original ingot. During rolling of billets into bar stock, impurities are rolled in a lengthwise
direction. These direction-oriented inclusions in the finished product are generally referred to as
non-metallic inclusions or “stringers”. These stringers may be surface or subsurface and areusually short in length and parallel to the grain flow.
Indication. The definition of the term “indication” as it applies to NDE is: “A response or evidenceof a response disclosed through NDE that requires further evaluation to determine its true
significance.” 4
In-service pressure test. A system pressure test conducted to perform visual examination VT-2
while the system is in service under operating pressure.
Latent failure. A degraded material condition that may lie dormant for a long period before
leading to a visible flaw (e.g. through-wall crack, active leakage).
Leak detection system. Instrumentation and controls that use various temperature, pressure, leveland flow sensors to detect water and steam leakages in selected reactor systems and to initiate
annunciation and provide isolation signal (in certain cases) to limit leakage from the reactor
coolant pressure boundary when limiting leakage conditions exists.
Leakage pressure test. A system pressure test conducted during operation at nominal operatingpressure, or when pressurised to nominal operating pressure and temperature.
Lead factor. The ratio of the fast neutron fluence at the centre of the surveillance capsule to thepeak fast neutron fluence at the inside surface of the reactor pressure wall.
Less-than-adequate (LTA). In the context of a root cause analysis of RIM programme deficiencies,
the term “less-than-adequate” (LTA) is used to characterise a procedure that lacks somethingessential to successfully to perform an activity.

3.
4.

https://cordis.europa.eu/project/id/662320.
NRC (2012), 0444-E306 Nondestructive Examination (NDE) Technology and Codes, www.nrc.gov/docs/
ML1214/ML12146A160.html.

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Limiting condition for operation (LCO). According to Technical Specifications 5, a LCO is the lowest
functional capability or performance level of a piece of equipment required for safe operation of a
nuclear plant. When a LCO cannot be met, the reactor must be shut down or the licensee must
follow any remedial action permitted by the Technical Specifications until the condition can bemet.
Liquid droplet impingement (LDI). Liquid droplet impingement is caused by the impact of highvelocity droplets or liquid jets. LDI usually occurs when a two-phase stream experiences a highpressure drop (e.g. across an orifice on a line to the condenser). When this occurs, there is anacceleration of both phases with the liquid velocity increasing to the point that, if the liquid
strikes a metallic surface, damage to the surface will occur. The main distinction between
flashing and LDI is that in flashing the fluid is of lower quality (mostly liquid with some steam),
and with LDI, the fluid is of higher quality (mostly steam with some liquid).
Liquid penetrant examination. Liquid penetrant examination (LPT) uses liquids to detect cracks inmaterials. In the mid and late 1930s, Robert and Joseph Switzer worked with processes
incorporating visible coloured dyes in the penetrant to give better contrast. In 1941 they
introduced processes using fluorescent dyes which, when viewed under a black light, produced
contrasts superior to those obtainable with the visible dyes. The fluorescent method was quickly
accepted by the military for aircraft part examination. Since then, the use of both colourcontrast and fluorescent penetrants has spread to practically all fields of manufacturing, andnew and improved PT products are constantly being developed.
Low-frequency electromagnetic testing (LFET). This technique measures the changes inelectromagnetic fields while the scanner passes over the metal. Defects and corrosion maps arecalculated, and video displayed in real-time, high resolution, 3-D colour graphics that can besaved for further data analysis or permanent record archiving. Very low frequency magnetic
signals are not affected by iron oxide or any non-magnetic surface deposits, which allows for
accurate testing on base metals in piping.
Low-leakage core (LLC). A reactor core loading pattern in which the more burned fuel is placed on
the periphery of the core to achieve lower radial neutron leakage and to reduce fuel pin peakburnup.
Marshall Study Group. A group of experts chaired by Dr W. Marshall (U.K. Atomic Energy
Authority) and established to review the factors determining the integrity of LWR pressure
vessels. A first report, “An Assessment of the Integrity of PWR Pressure Vessels,” was publishedin 1976. A second, updated report was published in 1982.
Master curve (MC). The MC approach is used to assess irradiated reactor pressure vessel material
fracture toughness parameters; for details see for example IAEA Technical Report Series No. 429,
Guidelines for Application of the Master Curve Approach to Reactor Pressure Vessel Integrity in
Nuclear Power Plants (2005).
MEACTOS (Mitigating environmentally assisted cracking through optimisation of surface
condition). Environmentally assisted cracking (EAC) is one of the major failure modes occurringin light water reactors (LWRs). The condition of surfaces exposed to the primary coolant plays
a main role in the susceptibility of components to EAC. However, many national andinternational guidelines and standards do not address surface condition of critical components
in nuclear power plants. The goal of the MEACTOS project is to improve the safety and reliabilityof Generation II and III nuclear power plants by improving the resistance of critical locations,
including welds, to EAC through the application of optimised surface machining and improved
surface treatments. The project end date is 31 August 2021. 6
Mechanical Stress Improvement Process (MSIP®). A patented process that was invented, developed,
and first used in 1986 by NuVision Engineering Inc. for mitigating stress corrosion cracking in
nuclear plant weldments. MSIP® works by using a hydraulically operated clamp which contracts

5.
6.

138

“Betriebshandbuch” in German.
See https://cordis.europa.eu/project/id/755151.

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the pipe on one side of the weldment. A typical tool design consists of a specially designed
hydraulic box press for bringing the clamp halves together. By contracting the pipe on one sideof the weldment, the residual tensile stresses are replaced with compressive stresses.
NDE qualification. In the context of NDE, qualification includes technical justification, which
involves assembling all the supporting evidence for inspection capability (results of capability
evaluation exercises, feedback from site experience, applicable and validated theoretical models,
physical reasoning), and may include practical trials using deliberately defective test pieces.
NUREG-1061 (Report of the US Nuclear Regulatory Commission Piping Review Committee). This reportdeals with six topical areas: event combinations, response combinations, stress limits anddynamic allowables, water hamer loadings, relief valve opening and closing loads and piping
vibration loads. Recommendations prepared by the staff were based on consultant position
papers and industry comments and treat revisions to present NRC requirements and directions
for future research. Foreign information was obtained from sources in Belgium, Canada, France,
Italy, Japan, Sweden and the Federal Republic of Germany. In addition, the report contains
qualitative value impacts for the proposed recommendations. This report was developed over a
period of approximately one year, and partially fulfils and complies with the requirements of
the 13 July 1983 memorandum from the Directors of the Offices of Nuclear Reactor Regulation
and Nuclear Regulatory Research to NRC's Executive Director for Operations. 7
Optimised weld overlay (OWOL). A subset of the full structural weld overlay (FSWOL) process. It
has been developed for larger geometries (e.g. RCS hot and cold leg nozzles) where FSWOL
application becomes too time consuming for a typical refuelling outage. The optimised weld
overlay thickness is less than that of a full structural weld overlay in order allow completion in
the time available in a typical refuelling outage for the larger geometries.
Phenomena identification and ranking technique (PIRT). PIRT is a systematic way of gathering
information from experts on a specific subject, and ranking the importance of the information,
to meet some decision-making objective. It has been applied to many nuclear technology issues
including material degradation assessment to help guide research or develop regulatory
requirements.
Probability of detection (POD). It is the probability that a flaw of a certain size will be detected and isconditional on factors such as wall thickness, NDE personnel qualifications and flaw orientation.
Radiographic examination. A non-destructive testing (NDE) method of inspecting materials for
hidden flaws by using the ability of short wavelength electromagnetic radiation (high energy
photons) to penetrate various materials.
Reactor vessel embrittlement. Embrittlement caused by the reactor vessel material exposure tohigh energy neutron flux from the reactor core. The area most likely to be affected by neutron
embrittlement is the beltline region of the reactor vessel due to its proximity to the core.
Embrittlement is characterised by a gradual reduction in the pressure vessel’s fracture
toughness. If this reduction in toughness were to continue to progress, and a crack existed, it ispossible that a brittle fracture of the vessel could occur under postulated events such aspressurised thermal shock (PTS) and low temperature overpressurisation (LTOP).
Red brass. An alloy of copper, tin and zinc. Proportions vary but 88% copper, 8-10% tin, and 2-4%
zinc is an approximation.
Reliability and integrity management (RIM). Those aspects of the plant design and operational
phase that are applied to provide an appropriate level of reliability of SSCs and a continuing
assurance over the life of the plant that such reliability is maintained. The most recent editionof the ASME Boiler and Pressure Vessel Code Section XI was issued in 2019. Division 2 of the2019 Edition provides the requirements for the creation of the reliability and integrity
management (RIM) programme for advanced nuclear reactor designs. The RIM programme
addresses the entire life cycle for all types of nuclear power plants. It requires a combination of

7.

See www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1061/#abstract.

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monitoring, examination, tests, operation, and maintenance requirements that ensures eachstructure, system, and component (SSC) meets plant risk and reliability goals that are selected
for the RIM programme 8.
Selective leaching. Also referred to as de-alloying, demetalification, parting and selective
corrosion, selective leaching is a corrosion type in some solid solution alloys, when in suitableconditions a component of the alloys is preferentially leached from the material. The less noble
metal is removed from the alloy by a microscopic-scale galvanic corrosion mechanism. The
most susceptible alloys are the ones containing metals with high distance between each otherin the galvanic series, e.g. copper and zinc in brass.
Sockolet. The weldolet (see below) and the sockolet both have a “run end” and a “branch end”.
The run end connects to the larger diameter pipe, and the branch end connects to the smaller
diameter pipe. The connection to the run pipe is a fillet weld for both the weldolet and thesockolet. They differ, however, at the connection to the branch pipe. The weldolet has a bevelled
end to butt weld to the branch pipe (which will also have a bevelled end), whereas the sockolet
has a socket into which a plain end branch pipe is inserted, then fillet welded together.
Solid particle erosion (SPE). SPE is damage caused by particles transported by the fluid stream rather
than by liquid water or collapsing bubbles. If hard, large particles are present at sufficiently highvelocities, damage will occur. In contrast to LDI, the necessary velocities for SPE are quite low.
Surfaces damaged by SPE have a very variable morphology. Manifestations of SPE in service
usually include thinning of components, a macroscopic scooping appearance following thegas/particle flow field, surface roughening (ranging from polishing to severe roughening,depending on particle size and velocity), lack of the directional grooving characteristics of abrasion,and in some but not all cases, the formation of ripple patterns on metals.
Stainless steel grade 405. A ferritic high-alloy stainless steel with chromium content of about 12%.
Stainless steel grade 409. A ferritic steel that offers good mechanical properties and hightemperature corrosion resistance. It is commonly considered as a chromium stainless steel,with applications that demand weldability. Grade 409 steels are also available in highly
stabilised forms, such as grades S40930, S40920 and S40910. The stability of these grades isprovided by the presence of niobium, titanium, or both, in the composition of steels.
Subsequent licence renewal (SLR). A licensing process by the US NRC that would authorise nuclear
power plants to operate beyond the 60 years of the initial licence and the first licence renewal.
Subsequent licence renewals would also be for 20 years. The NRC has developed guidance forstaff and licensees specifically for the subsequent renewal period.
Super emergency feedwater system (SEFW). The SEFW system is specific to WWER-type reactors. It
is designed to respond to total loss of heat sink accident scenarios (i.e. total loss of normal andauxiliary feedwater). The system injects feedwater directly into the steam generator secondaryside by separate stainless-steel feedlines from demineralised water tanks.
Thermal ageing embrittlement (TAE). A time-and temperature-dependent change in the
microstructure of the material which leads to a reduced ductility and deterioration of the fracture
toughness and the impact properties. The material will show an increased embrittlement overtime.
Thermal striping. Incomplete mixing of high temperature and low temperature fluids near thesurface of structures with subsequent fluid temperature fluctuations giving rise to thermal
fatigue damage to wall structures.
Thermal treatment (TT) process. Steam generator tubing is thermally treated at about 705°C for
15 hours to relieve fabrication stresses and to further improve the microstructure. The thermal
treatment process promotes carbide precipitation at the grain boundaries and diffusion ofchromium to the grain boundaries.

8.

140

See www.ndt.net/article/NDE-Nuclear2019/papers/2.A.01_ASME_Section_XI,_Division_2_Rewrite.pdf.

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GLOSSARY OF TECHNICAL TERMS

Time-limited ageing analysis (TLAA). In the context of nuclear power plant licence renewal, a TLAA
is formal engineering analysis of the effect of ageing on the structural integrity of systems,
structures and components.
Tinel coupling. A pipe coupling made of Tinel material (half titanium and half nickel). The
coupling is machined to predetermined dimensions with the I.D. smaller than the tube O.D.
Passing a mandrel through the coupling, I.D. while submerged in liquid nitrogen at cryogenic
temperature will expand the coupling. It will remain as expanded while submerged in liquid
nitrogen. When the coupling is installed to couple two pieces of tubing together, it willautomatically return to its predetermined dimensions when exposed to room temperature. The
shape memory is the result of a change in the crystal structure of the alloy known as reversible
austenite to martensite phase transformation.
Trepan. Trepanning removes a solid cylindrical core of material by cutting around it.
Tritium. Tritium is a naturally occurring radioactive form of hydrogen that is produced in theatmosphere when cosmic rays collide with air molecules. As a result, tritium is found in verysmall or trace amounts in groundwater throughout the world. It is also a byproduct of the
production of electricity by nuclear power plants. Tritium emits a weak form of radiation, a lowenergy beta particle like an electron. The tritium radiation does not travel very far in air andcannot penetrate the skin.
Tritium in nuclear power plants. Most of the tritium produced in nuclear power plants stems froma chemical, known as boron, absorbing neutrons from the plant's chain reaction. Nuclear
reactors use boron, a good neutron absorber, to help control the chain reaction. Toward that
end, boron either is added directly to the coolant water or is used in the control rods to control
the chain reaction. Much smaller amounts of tritium can also be produced from the splitting of
Uranium-235 in the reactor core, or when other chemicals (e.g. lithium or heavy water) in the
coolant water absorb neutrons. Like normal hydrogen, tritium can bond with oxygen to form
water. When this happens, the resulting "tritiated" water is radioactive. Tritiated water (not tobe confused with heavy water) is chemically identical to normal water and the tritium cannot
be filtered out of the water.
Uniaxial constant load (UCL) Test. The UCL tests use tensile-type specimens to assure precise
applied stress values by direct measurement to determine Young’s modulus, yield stress,
ultimate tensile strength and elastic strain energy density.
Unified numbering system (UNS). An alloy designation system in use in North America. It consistsof a prefix letter and five digits designating a material composition. For example, a prefix of S
indicates stainless steel, C indicates copper, brass or bronze alloys.
Visual examination. The oldest and most used NDE method is visual testing (VT), which may be
defined as “an examination of an object using the naked eye, alone or in conjunction withvarious magnifying devices, without changing, altering, or destroying the object being
examined.” 9 Per ASME XI, there are three different VT methods; VT-1, VT-2 and VT-3.
Warm pre-stress (WPS) effect. The WPS effects were identified in the 1960s. According to researchand tests, the apparent fracture toughness of a ferritic steel can be elevated in the fracture modetransition: regime if the specimen is first “pre-stressed” at an elevated temperature, e.g. througha pressurised thermal shock. Once a specimen is subjected to a certain Kapplied and has not failed,
the temperature can be reduced, and the specimen will remain intact despite the fact that theprocess of reducing the temperature has also reduced the initiation fracture toughness (KIc or
KJc) to values smaller than Kapplied. In the past five decades, the physical mechanisms responsible
for the WPS effect have been identified, studied extensively and validated. 109.

NRC (2012), “0444-E306 Nondestructive Examination (NDE) Technology and Codes”, www.nrc.gov/docs/
ML1214/ML12146A160.html.
10. Section 7.7.1.1 in NUREG-1806, Vol. 1; Technical Basis for Revision of the Pressurized Thermal Shock(PTS) Screening Limit in the PTS Rule (10 CFR 50.61), 2007.

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Weld inlay. A mitigation technique defined as application of PWSCC-resistant material(Alloy 52/52M) to the inside diameter of a dissimilar metal weld that isolates the PWSCCsusceptible material (Alloy 82/182) from the primary reactor coolant.
Weldolet®. The most common of all branch connections and is welded onto a larger-diameter
pipe. The ends are bevelled to facilitate this process, and therefore the “weldolet” is considered
a butt-weld fitting. Weldolets are designed to minimise stress concentrations and provide
integral reinforcement.
XFEM. The extended finite element method (XFEM) is a finite element analysis (FEA) tool thatallows for mesh-independent analysis of discontinuities and singularities and can be used tosimulate crack growth in complex geometries.
Zorita Internals Research Project (ZIRP). The nuclear power plant known as José Cabrera (morecommonly referred to as “Zorita”) was Spain’s first nuclear power plant and began operation in
July 1965. The 160 MWe Westinghouse-designed pressurised water reactor was permanently
disconnected from the national power grid in April 2006 after 38 years of operation. It was asingle-loop PWR. In 2013, baffle plate material and core barrel weld material were removed fromthe nuclear power plant during the decommissioning activities. These materials provide thefoundation for several EPRI materials reliability programme (MRP) research efforts, including
the Zorita Internals Research Project (ZIRP). The objective of ZIRP is to study stainless steel
ageing degradation to generate correlations between the radiation and temperature history of
the materials and the changes in mechanical properties.

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EXAMPLES OF REPORTING SYSTEMS/ROUTINES FOR PASSIVE COMPONENT MATERIAL DEGRADATION AND FAILURE

Annex B
Examples of reporting systems/routines for passive
component material degradation and failure

Table B-1: Examples of US reporting routines and requirements forcommunicating/documenting degraded or failed passive components
Reportinglevel

Report/databasetype

Level of technical content from the pointof view of material science, PSA
applications, etc.

3rd party accessibility for purposes of
R&D and/or operational support

US Nuclear Regulatory Commission
Licensee event
reportHigh-level – requires interpretation andclassification by a SME - high reporting
thresholds. Since the 1990s only a few LERs of
relevance to CODAP are issued on an annual
basis; < 10 per year

Public domain; covers 1980 to date;https://lersearch.inl.gov/LERSearchCriteria.
aspx. A large number of pre-1980 LERs areavailable from https://adams.nrc.gov/wba/

Event notification(EN) system

High-level event tracking system forincoming notifications of significant nuclear
events with an actual or potential effect onplant safety. An EN may result in a LER. On anannual basis, only a few ENs of relevance to
CODAP are issued

Public domain; all ENs from 1 January 1999
onwards are available at:www.nrc.gov/reading-rm/doccollections/event-status/event/en.html

Identifies potential generic issues that aresafety significant and require technicalresolution

www.nrc.gov/reading-rm/doccollections/gen-comm/Generic
communications –Generic letters,
information notices
Regulatory
requirements
and guidance

Generic letter 90-05
Request for
temporary repair

Applicable to moderate-energy piping, theoperability evaluations that are part of the90-05 relief request are comprehensive.
Excellent source of pipe failure information

Public domain; https://adams.nrc.gov/wba/

10 CFR 50.55a(z)(2)– Application of
ASME Code Cases

The operability evaluations that are part of
the ASME Code Case relief request arecomprehensive. Excellent source of pipe
failure information

See below under "ASME Code Cases". Any
Code Case Relief request that is submitted
to the NRC for approval is also available forreview by 3rd party.

Biannual Problem
Identification and
Resolution
Inspection Report

Regulatory review of a licensee’s corrective
action program and the licensee’s
implementation of the program to evaluate
its effectiveness in identifying, prioritising,evaluating, and correcting problems, and to
confirm that the licensee is complying with
NRC regulations and licensee standards forcorrective action programmes

Public domain; https://adams.nrc.gov/wba/

Quarterly
integrated
inspection reports

Significant repository for information on
passive components failures. Requires
interpretation and classification by a SME

Public domain;
https://lersearch.inl.gov/IRSearchCriteria.as
px

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Table B-1: Examples of US reporting routines and requirements forcommunicating/documenting degraded or failed passive components (cont’d)Reporting level

Report/database typeLevel of technical content from thepoint of view of material science,
PSA applications, etc.

3rd party accessibility for purposes of
R&D and/or operational support

US nuclear industry level

Institute of
Nuclear PowerOperations(INPO)INPO Consolidated
Events Database (ICES).
Participating plants
submit ARs, CRs, WORs,
etc. – mainly freeformat narrative
information – to asearchable database

Varying - data extraction,interpretation and classification canbe time-consuming

Access restricted to ICES participatingorganisations. With respect to theOECD/NEA ICDE Project, an NRC-INPO MoU
has enabled a time-limited database accessin support of the project

Nuclear Energy
Institute (NEI)

NEI 07-07: Industry
Ground Water
Protection Initiative –Final GuidanceDocument, 2007

Reporting of below ground/buried
pipe failures causing > 0.38 m3
unintended release

Annual radioactive effluent release reportsubmitted by licensees to the US NRC

EPRI-CHUG:
Presentation material
from the biannual
CHUG meetings.
Informal exchange of
FAC information(summary of FAC
failures, information on
the use of NDE
technology)

Varying - data extraction,interpretation and classification canbe time-consuming

EPRI-MRP, multiple
programmes – reactor
internals, thermal
fatigue, ageing
management

NRC receives biennial summaries of
the BWR and PWR reactor internals
inspection results. These summaries
are available on the NRC website

Plant engineering/longterm operations/riskand safety management
programmes

Extensive source of material
degradation and failure information
organised by system, degradation
mechanism, etc. These programmes
build on active technical input fromthe sponsoring US and international
member organisations. Extensive
repository of relevant PEO/LTO
material degradation information

Access restricted to funding organisation

NRC/EPRI-MRP
Annual Technical
Information
Exchange
Meetings

Presentation materialsummarising R&D
status, results of reactor
internals inspections,
thermal fatigue
inspections, etc.

High-level

All presentation material available throughthe NRC Public Document Room

GE service
information letter(SIL), andequivalentbulletins from the
PWROG

Information on material
degradation issues in
response to recent
OPEX

Chronological lists of events withsome information on method of
discovery, extent of degradation,
flaw size data, etc. Plant identities
not revealed

The SILs can be downloaded from the NRC
Public Document Room; for example see
www.nrc.gov/docs/ML0501/ML050120032.
pdf

Electric Power
Research Institute(EPRI)

144

Access restricted to funding organisation.
Non-proprietary versions of EPRI-MRP
reports can be downloaded from the NRC
public website

A REVIEW OF OPERATING EXPERIENCE INVOLVING PASSIVE COMPONENT MATERIAL DEGRADING IN PERIODS OF EXTENDED AND LTO, NEA No. 7614, © OECD 2024

EXAMPLES OF REPORTING SYSTEMS/ROUTINES FOR PASSIVE COMPONENT MATERIAL DEGRADATION AND FAILURE

Table B-1: Examples of US reporting routines and requirements forcommunicating/documenting degraded or failed passive components (cont’d)Reporting level

Report/database typeLevel of technical content from thepoint of view of material science,
PSA applications, etc.

3rd party accessibility for purposesof R&D and/or operational support

Plant-level
Action request (AR), condition
report (CR), corrective action
request (CAR), Incident report(IR), Non-conformance report(NCR), problem investigation
process (PIP), repair/
replacement plant (RRP), workorder request (WOR),
operability determination, andstructural evaluation inaccordance with applicable
ASME Code Case

Details on location of flaw/leak withreference to P&ID, isometric drawing– oftentimes accompanied byphotographic records, root cause
analysis results, NDE results,
metallographic data, flaw size data,
leak/flow rates, spatial effects, risk
significance, safety significance.
Piping fabrication data, operationalparameters, piping design
information (material, dimensional
data)

Restricted/proprietary. US NRC
Resident Inspectors have access to
information on request

Shiftly (once every 8/12-hour
shift) RCS inventory massbalance calculation, Chem.
Lab trending of inflows to Aux.
Building / Reactor Building
sumps and holding tanks

Logs, databases (various types)

Restricted/proprietary. US NRC
Resident Inspectors have access to
information on request

Containment/drywell leakdetection alarm(≥ 0.1 gpm/6.3E-3 kg/s)

Control room logs/computer
printouts, Excel spreadsheets –holding tank inflow, details ofthrough-wall leak rates as a functionof time and location, etc.

Restricted/proprietary. US NRC
Resident Inspectors have access to
information on request

In-service
inspection (ISI)
per ASME XI –mandatory
inspections

List of “recordable
indications”, operability
determinations/fitness-forservice (FFS) evaluations,
ASME XI relief requests (formore details see “Regulatory”
above. Within 90 days of theconclusion of a refuellingoutage, the plant owner isrequired to submit an ISI
summary report (“Owner's
Activity Report”) to the US
NRC. This report itemises all
inspections performed (CodeClass 1, 2 and 3), any
operability determination is
documented

The ASME relief requests include all
relevant information to meet CODAP
event data input requirements.
Included are “Abstract ofRepair/Replacement Activities
Required for Continued Service” and“Items with Flaws or Relevant
Conditions That Required Evaluation
for Continued Service”

The ISI summary reports are publicly
available via the NRC PublicDocument Room;https://adams.nrc.gov/wba/

Owner defined ISI
programmes

FAC, Erosion-Corrosion, MIC
and thermal fatigue
programmes

NDE reports, “system health reports”

Restricted/proprietary. US NRC
Resident Inspectors have access tothe information upon request

During routine
operation –Shiftly walkdown
inspection

Leak detection –containment/dry
well

A REVIEW OF OPERATING EXPERIENCE INVOLVING PASSIVE COMPONENT MATERIAL DEGRADING IN PERIODS OF EXTENDED AND LTO, NEA No. 7614, © OECD 2024

145

EXAMPLES OF REPORTING SYSTEMS/ROUTINES FOR PASSIVE COMPONENT MATERIAL DEGRADATION AND FAILURE

Table B-1: Examples of US reporting routines and requirements forcommunicating/documenting degraded or failed passive components (cont’d)Reporting level

Report/database typeLevel of technical content fromthe point of view of materialscience, PSA applications, etc.

3rd party accessibility for purposes of
R&D and/or operational support

American Societyof MechanicalEngineers (ASME)

Numerous Code Case havebeen developed for theoperability determination of
degraded piping; e.g. N-513(Evaluation Criteria for
Temporary Acceptance ofFlaws in Moderate Energy
Class 3 Piping), N-561(Alternative Requirements
for Wall Thickness
Restoration of Class 2 andHigh Energy Class 3 Carbon
Steel Piping), N-562(Alternative Requirements
for Wall Thickness
Restoration of Class 3
Moderate Energy Carbon
Steel Piping), N-666 (Weld
Overlay of Class 1, 2, and 3
Socket Welds

Details on location of flaw/leakwith reference to P&ID, isometricdrawing – oftentimes
accompanied by photographic
records, root cause analysis results,
NDE results, metallographic data,
flaw size data, leak/flow rates,
spatial effects, risk significance,
safety significance. Piping
fabrication data, operationalparameters, piping design
information (material, dimensionaldata). Structural integrity
evaluation (inputs/results/
interpretation) is part of therequired documentation. The
ASME relief requests include all
relevant information to meet
CODAP event data input
requirements

The Code Case relief requests that are
submitted to the NRC for approval arealso available for 3rd party review. Wherea Code Case has been approved forplant-specific use, an associatedoperability determination is kept on fileon site and available for review by NRC
Resident Inspector. Hence, a 3rd part
review is not possible unless theunderlying technical information(e.g. fitness-for-service evaluations) has
been placed in the NRC Public Document
Room

System healthreports (or systemengineers’
notebooks)

Access or Excel-based
databases that are part ofthe overall ageing
management programme.
The system health reports
provide summaries of all
through-wall leaks

High-level, intended for trendingof material degradation issues,
effectiveness of degradation
mitigation programmes

Restricted/proprietary. US NRC Resident
Inspectors may have access to theinformation upon request

146

A REVIEW OF OPERATING EXPERIENCE INVOLVING PASSIVE COMPONENT MATERIAL DEGRADING IN PERIODS OF EXTENDED AND LTO, NEA No. 7614, © OECD 2024

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Review of Operating Experience Involving

Passive Component Material Degrading in
Periods of Extended and Long-Term Operation

This publication addresses trends and patterns in material degradation as a function of nuclear power
plant ageing, which is a key topic in the regulation of ageing management programmes, codes andstandards, reliability and integrity management, and the many supporting research programmes. It
documents insights reached from operating experience involving passive metallic components incommercial nuclear power plants during periods of extended operation and long-term operation.
These evaluations are enabled by an NEA joint project entitled the Component Operational Experience,
Degradation and Ageing Programme (CODAP). Established in 2002, the project has developed acomprehensive database on the operating experience with material degradation mechanisms that haveproduced failures of piping and selected non-piping passive components. Examinations of graphicalplots were performed to reach insights about the different material degradation mechanisms as afunction of plant age. The report also includes recommendations for future work.

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NEA No. 7614