Nuclear Safety
NEA/CSNI/R(2017)3
June 2017www.oecd-nea.org

Operating Experience

Insights into Pressure
Boundary Component Reliability
and Integrity Management
Topical Report by the Component
Operational Experience, Degradation
and Ageing Programme (CODAP)GroupUnclassified

NEA/CSNI/R(2017)3

Organisation de Coopération et de Développement Économiques
Organisation for Economic Co-operation and Development

15-Jun-2017________________________________________________________________________________________________________English text only
NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSNEA/CSNI/R(2017)3
Unclassified
Operating Experience Insights into Pressure Boundary Component Reliability
and Integrity Management
Topical Report by the Component Operational Experience,
Degradation and Ageing Programme (CODAP) Group

JT03416097

English text only

Complete document available on OLIS in its original format
This document, as well as any data and map included herein, are without prejudice to the status of or sovereignty over any territory, to thedelimitation of international frontiers and boundaries and to the name of any territory, city or area.

NEA/CSNI/R(2017)3

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTThe OECD is a unique forum where the governments of 35 democracies work together to address the economic, social andenvironmental challenges of globalisation. The OECD is also at the forefront of efforts to understand and to help
governments respond to new developments and concerns, such as corporate governance, the information economy and thechallenges of an ageing population. The Organisation provides a setting where governments can compare policy
experiences, seek answers to common problems, identify good practice and work to co-ordinate domestic and international
policies.
The OECD member countries are: Australia, Austria, Belgium, Canada, Chile, the Czech Republic, Denmark, Estonia,
Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Israel, Italy, Japan, Latvia, Luxembourg, Mexico, theNetherlands, New Zealand, Norway, Poland, Portugal, Korea, the Slovak Republic, Slovenia, Spain, Sweden, Switzerland,
Turkey, the United Kingdom and the United States. The European Commission takes part in the work of the OECD.
OECD Publishing disseminates widely the results of the Organisation’s statistics gathering and research on economic,
social and environmental issues, as well as the conventions, guidelines and standards agreed by its members.

NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membership consists of 31
countries: Australia, Austria, Belgium, Canada, the Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary,
Iceland, Ireland, Italy, Japan, Luxembourg, Mexico, the Netherlands, Norway, Poland, Portugal, Korea, Russia, the Slovak
Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The European
Commission and the International Atomic Energy Agency also take part in the work of the Agency.
The mission of the NEA is:– to assist its member countries in maintaining and further developing, through international co-operation, thescientific, technological and legal bases required for a safe, environmentally sound and economical use of nuclear
energy for peaceful purposes;– to provide authoritative assessments and to forge common understandings on key issues as input to government
decisions on nuclear energy policy and to broader OECD analyses in areas such as energy and the sustainable
development of low-carbon economies.
Specific areas of competence of the NEA include the safety and regulation of nuclear activities, radioactive waste
management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law
and liability, and public information. The NEA Data Bank provides nuclear data and computer program services forparticipating countries.
This document, as well as any statistical data and map included herein, are without prejudice to the status of or sovereignty over any
territory, to the delimitation of international frontiers and boundaries and to the name of any territory, city or area.
Corrigenda to OECD publications may be found online at: www.oecd.org/publishing/corrigenda.
© OECD 2017
You can copy, download or print OECD content for your own use, and you can include excerpts from OECD publications, databases and multimedia products in your own
documents, presentations, blogs, websites and teaching materials, provided that suitable acknowledgement of the OECD as source and copyright owner is given. All requests for
public or commercial use and translation rights should be submitted to neapub@oecd-nea.org. Requests for permission to photocopy portions of this material for public orcommercial use shall be addressed directly to the Copyright Clearance Center (CCC) at info@copyright.com or the Centre français d'exploitation du droit de copie (CFC)
contact@cfcopies.com.

2

NEA/CSNI/R(2017)3

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSThe Committee on the Safety of Nuclear Installations (CSNI) shall be responsible for NEA programmes
and activities that support maintaining and advancing the scientific and technical knowledge base of thesafety of nuclear installations, with the aim of implementing the NEA Strategic Plan for 2017-2022 in its
field of competence.
The Committee shall constitute a forum for the exchange of technical information and for collaboration
between organisations, which can contribute, from their respective backgrounds in research, development
and engineering, to its activities. It shall have regard to the exchange of information between member
countries and safety R&D programmes of various sizes in order to keep all member countries involved inand abreast of developments in technical safety matters.
The Committee shall review the state of knowledge on important topics of nuclear safety science andtechniques and of safety assessments, and ensure that operating experience is appropriately accounted forin its activities. It shall initiate and conduct programmes identified by these reviews and assessments inorder to confirm safety, overcome discrepancies, develop improvements and reach consensus on technicalissues of common interest. It shall promote the co-ordination of work in different member countries thatserve to maintain and enhance competence in nuclear safety matters, including the establishment of joint
undertakings (e.g. joint research and data projects), and shall assist in the feedback of the results toparticipating organisations. The Committee shall ensure that valuable end-products of the technical reviews
and analyses are provided to members in a timely manner, and made publicly available when appropriate,
to support broader nuclear safety.
The Committee shall focus primarily on the safety aspects of existing power reactors, other nuclear
installations and new power reactors; it shall also consider the safety implications of scientific andtechnical developments of future reactor technologies and designs. Further, the scope for the Committee
shall include human and organisational research activities and technical developments that affect nuclear
safety.
The Committee shall organise its own activities. Furthermore, it shall examine any other mattersreferred to it by the Steering Committee. It may sponsor specialist meetings and technical working groups
to further its objectives. In implementing its programme, the Committee shall establish co-operative
mechanisms with the Committee on Nuclear Regulatory Activities in order to work with that Committee
on matters of common interest, avoiding unnecessary duplications.
The Committee shall also co-operate with the Committee on Radiological Protection and Public Health,
the Radioactive Waste Management Committee, the Committee for Technical and Economic Studies on
Nuclear Energy Development and the Fuel Cycle, the Nuclear Science Committee, and other NEA
committees and activities on matters of common interest.

3

NEA/CSNI/R(2017)3

FOREWORD

Structural integrity of piping systems is important for plant safety and operability. In recognition of this,information on degradation and failure of piping components and systems is collected and evaluated byregulatory agencies, international organisations (e.g. the Nuclear Energy Agency [NEA] and theInternational Atomic Energy Agency [IAEA]) and industry organisations worldwide. This information isoften used to provide systematic feedback to reactor regulation and research and development programmes
associated with non-destructive examination (NDE) technology, in-service inspection (ISI) Programmes,
leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications
involving passive component reliability.
Several Organisation for Economic Co-operation and Development (OECD) member countries haveagreed to establish the NEA Component Operational Experience, Degradation and Ageing Programme(CODAP) to encourage multilateral co-operation in the collection and analysis of data relating todegradation and failure of metallic piping and non-piping metallic passive components in commercial
nuclear power plants. The scope of the data collection includes service-induced wall thinning, partthrough-wall cracks, through-wall cracks with and without active leakage, and instances of significant
degradation of metallic passive components, including piping pressure boundary integrity. The NEA
Committee on the Safety of Nuclear Installations (CSNI) acts as an umbrella committee of the Project.
CODAP is the continuation of the 2002–2011 OECD/NEA Pipe Failure Data Exchange Project(OPDE) and 2006–2010 NEA Stress Corrosion Cracking and Cable Ageing Project (SCAP). OPDE wasformally launched in May 2002. Upon completion of the Third Term (May 2011), the OPDE project was
officially closed to be succeeded by CODAP. SCAP was enabled by a voluntary contribution from Japan.
It was formally launched in June 2006 and officially closed with an international workshop held in Tokyo
in May 2010. The majority of the member organisations of the two projects were the same, often being
represented by the same person. In May 2011, thirteen countries and economies signed the CODAP First
Term (2011-2014) agreement (Canada, Czech Republic, Finland, France, Germany, Korea (Republic of),Japan, Slovak Republic, Spain, Sweden, Switzerland, Chinese Taipei and United States). In December
2014, eleven countries and economies (Canada, Czech Republic, France, Germany, Korea (Republic of),Japan, Slovak Republic, Spain, Switzerland, Chinese Taipei and United States of America) agreed to enter
into an agreement for the CODAP Second Term (2015-2017). The CODAP Second Term work plan
includes tasks to prepare topical reports to foster technical co-operation and to deepen the understanding of
national differences in ageing management.
This third topical report is concerned with the effectiveness of pressure boundary component
reliability and integrity management (RIM). Specifically, this report addresses operating experience
insights related to less-than-adequate (LTA) RIM and its potential safety and operational impacts. The term
LTA-RIM is defined as events where degradation has progressed beyond acceptable limits in systems,
structures or components (SSCs) that have a RIM Programme. These LTA-RIM events have some safetysignificance. In this topical report the LTA-RIM definition is broadened to also include events where a
RIM Programme has resulted in a false positive; that is, it has identified degradation that either didn’t exist
or was not close to violating acceptance criteria. While such events needlessly expend resources and could
be considered LTA-RIM from an economic perspective, they do not have any safety significance.
Additionally, this report addresses selected international practices with respect to pressure testing,leak detection, ISI including NDE, and performance demonstration initiatives to improve the reliability of4

NEA/CSNI/R(2017)3
NDE techniques. The purpose of integrity management is to prevent the occurrence of piping through-wallleaks as well as to monitor passive metallic component degradation. RIM Programme utilises risk insights
to augment or enhance existing deterministic integrity management programme. Through a systematic
examination of the operating experience as recorded in the CODAP event database, the field experience
with the different RIM strategies is evaluated in order to draw qualitative and quantitative insights about
integrity management reliability. This topical report also includes an introduction of an approach toevaluate the effect of different integrity management strategies on pressure boundary component
reliability.
The CODAP event database includes numerous records where degraded or failed conditions are
attributed to LTA implementation of RIM, including NDE. Especially noteworthy are some recent 2010 –2016 occurrences of passive component degradation or failure that are attributed in part to LTA
implementation of RIM. When RIM failures occur, one or more of the following factors are often present:

Accepting a rejectable flaw indication for continued operation. This could be due tomisinterpretation of NDE results.



Rationalising away detected defects.

Using improperly qualified NDE technique.



Poorly implementing qualified procedures.

Poorly implementing owner-defined inspection programme.

Missing a flaw with a qualified procedure. A procedure may not sufficiently document the basis
for the examination details used to inspect for a specific, previously observed degradation
mechanism. There may also be inadequate administrative controls for augmented inspections
and disposition of inspection results.

Based on the results of an evaluation of the CODAP database content, the number of through-wallleakages could be decreased by the following actions:

Utilising properly qualified UT techniques.

Periodically reviewing and independently validating UT-scanning grids used in inspection of
piping components.

Optimising the ISI programme on the basis of probabilistic RIM methods.

The report has been prepared by the CODAP Project Review Group (PRG), with technical support
from the CODAP Operating Agent.

ACKNOWLEDGEMENTS
The CODAP-PRG gratefully acknowledges the beneficial independent review of this report by Mr Carl
Latiolais (EPRI), and Ms Carol Nove and Dr Rob Tregoning (both with the US NRC).

5NEA/CSNI/R(2017)3

TABLE OF CONTENTSACKNOWLEDGEMENTS .......................................................................................................................5
ABBREVIATIONS AND ACRONYMS ..................................................................................................81.

INTRODUCTION ............................................................................................................................12

1.1
1.2
1.3
1.4
1.5

Background and related work ........................................................................................................12
Objectives and scope ......................................................................................................................14
Nomenclature .................................................................................................................................15
Disclaimer-+...................................................................................................................................17
Report structure and reading guide ................................................................................................17

2.

RELIABILITY AND INTEGRITY MANAGEMENT PRACTICES ...........................................19

2.1
2.2
2.3
2.4
2.5
2.6
2.7
2.8
2.9
2.10
2.11

ASME Section XI, Appendix VIII .................................................................................................19
Canadian Standards Association periodic inspection programme .................................................20
French RSE-M code .......................................................................................................................22
Nuclear Safety Standards Commission (KTA) ..............................................................................22
Korea NSSC and KEPIC MI ..........................................................................................................23
European Network for Inspection and Qualification .....................................................................24
NDE qualification methodologies ..................................................................................................25
Owner Defined Integrity Management Programme .......................................................................28
Code Case N-729-1 ........................................................................................................................29
Code Case N-770-1 ........................................................................................................................31
NDE of small-diameter piping .......................................................................................................35

3. RIM INFLUENCES IN CODAP ........................................................................................................37
3.1
3.2
3.3

Database structure and coding guideline........................................................................................37
Method of detecting degradation ...................................................................................................38
High-level CODAP database summary .........................................................................................41

4.

FAILURE OF RIM TO DETECT FLAWS ...................................................................................51

4.1
4.2
4.3

An overview of RIM failures in CODAP ......................................................................................51
Selected (1999-2015) RIM failures................................................................................................54
Data analysis insights regarding LTA-RIM ...................................................................................60

5.

ASSESSMENT OF INFLUENCE OF RIM ON PASSIVE COMPONENT RELIABILITY .......61

5.1
5.2
5.3

Problem statement ...........................................................................................................................61
Models of RIM reliability ...............................................................................................................62
Probability of detection ...................................................................................................................64

6. SUMMARY AND CONCLUSIONS ...................................................................................................66
6.1
6.2

Summary .........................................................................................................................................66
Conclusions .....................................................................................................................................66

6

NEA/CSNI/R(2017)3
6.3

Recommendations ...........................................................................................................................67

7. REFERENCES .........................................................................................................................................69
APPENDIX A ...............................................................................................................................................74
APPENDIX B................................................................................................................................................79
APPENDIX C................................................................................................................................................82
APPENDIX D ...............................................................................................................................................86

7

NEA/CSNI/R(2017)3

ABBREVIATIONS AND ACRONYMS

AFCEN

Association française pour les règles de conception, de construction et de surveillance en
exploitation des matériels des chaudières électronucléaires

AFWS

Auxiliary feed water system

AMP

Ageing management programme

ASME

American Society of Mechanical Engineers

ASNT

American Society for Non-destructive Testing

AUT

Automated ultrasonic testing

BMV

Bare metal visual inspection

BPVC

Boiler and pressure vessel code

BWR

Boiling water reactor

BWRVIP

BWR vessel and internals project

CANDU

Canadian deuterium uranium

CCWS

Component cooling water system

CDF

Core damage frequency

CI

Completeness index

CINDE

Canadian Institute for Non-Destructive ExaminationCIQB

CANDU Inspection Qualification Bureau

CNRA

Committee on Nuclear Regulatory Activities

CNSC

Canadian Nuclear Safety Commission

CODAP

Component Operational Experience, Degradation and Ageing Programme

COG

CANDU Owners Group

CORDEL WG

Co-operation in Reactor Design Evaluation and Licensing Working Group

CRDM

Control rod drive mechanism

CRP

Conditional rupture probability

CSA

Canadian Standards Association

CSD

Cold shutdown

CSNI

Committee on the Safety of Nuclear Installations

CSTF

Codes and Standards Task ForceCSS

Containment spray system
8

NEA/CSNI/R(2017)3
CSWG

Codes and Standards Working GroupCVCS

Chemical and volume control System

DEGB

Double-ended guillotine pipe break

DIN

Deutsches Institut für Normung

DMW

Dissimilar metal weld

DN

Diamètre nominal

ECCS

Emergency core cooling system

ECT

Eddy-current testing

EDF

Électricité de France SA.

EMDA

Expanded material degradation assessment

EMQ

ENIQ Methodology for Qualification

ENIQ

European Network for Inspection and Qualification

ENSI

Eidgenössisches Nuklearsicherheits-inspektorat

EPRI

Electric Power Research Institute

EQMD

European qualification methodology document

ESWS

Essential service water system

FAC

Flow-accelerated corrosion

FR

Federal register

FSAR

Final safety analysis report

FSWOL

Full structural weld overlay

FW

Feedwater

GALL

Generic ageing lessons learnt

HSB

Hot standby

HSD

Hot shutdown

IAEA

International Atomic Energy Agency

IGSCC

Intergranular stress corrosion cracking

IAGE

CSNI Working Group on Integrity and Ageing of Components and Structures

IGALL

International GALL

IHSI

Induction heat stress improvement

IPEC

Indian Point Energy Centre

ISI

In-service inspection

ISO

International Organisation for Standardisation

JRC

Joint Research Centre

IVVI

In-vessel visual inspection

KEPIC

Korea Electric Power Industry Code9

NEA/CSNI/R(2017)3
KTA

Kerntechnische Ausschuss

LOCA

Loss-of-coolant accident

LPT

Liquid penetrant test

LTA

Less-than-adequate

MDEP

Multinational Design Evaluation Programme

MRP

Materials Reliability Programme

MSIP®

Mechanical Stress Improvement Process®

MSR

Moisture separator reheater

NDE

Non-destructive examination

NDT

Non-destructive testing

NEA

Nuclear Energy Agency

NFPA

National Fire Protection Association

NIFG

NDE Improvement Focus Group

NPP

Nuclear power plant

NRC

Nuclear Regulatory Commission

NSSC

Nuclear Safety and Security Commission (of Korea)

OPDE

OECD Pipe Failure Data Exchange Project

OWOL

Optimised Weld Overlay

PDA

Performance Demonstration Administrator

PDI

Performance Demonstration Initiative

PMDA

Proactive Materials Degradation Assessment

POD

Probability of Detection

PRG

Project Review Group

PSA

Probabilistic Safety Assessment

PT

Pressure Test

PWR

Pressurised Water Reactor

PWSCC

Primary Water Stress Corrosion Cracking

RCC-M

Règles de Conception et de Construction des Matériels Mécaniques des Ilots Nucléaires
PWR

RCP

Reactor Coolant Pump

RCPB

Reactor Coolant Pressure Boundary

RCS

Reactor Coolant System

RHRS

Residual Heat Removal System

RI-ISI

Risk-informed in-service inspection

RIM

Reliability and Integrity Management

10NEA/CSNI/R(2017)3
RISMET

Risk-Informed In-Service Inspection Methodologies

RMS

Root Mean Square

RPV

Reactor Pressure Vessel

RT

Radiographic Testing

SAFT

Synthetic Aperture Focus Technique

SCAP

Stress Corrosion and Cable Ageing Project

SEAS

Slovenské Elektrárne AS

S/GSteam Generator

SLR

Subsequent Licence Renewal

SQL

Structured Query Language

TGR

Task Group on Risk

TSO

Technical Support Organisation

UNS

Unified Numbering System

UT

Ultrasonic TestingVT

Visual Examination

WNA

World Nuclear Association

WOR

Weld Overlay

XML

Extensible Mark-up Language

11

NEA/CSNI/R(2017)3

1. INTRODUCTION

Since 2002, the NEA has operated an event database project that collects information on passive metallic
component degradation and failures. The scope of the database includes primary system piping
components, reactor pressure vessel internals, main process and standby safety systems, and support
systems (i.e. American Society of Mechanical Engineers (ASME ) Code Class 1, 2 and 3, or equivalent)components, as well as non-safety-related (non-Code) components whose degradation or failure can havesignificant operational impact. With an initial focus on piping systems components (the OPDE Project), thescope of the project in 2011 was expanded to also address the reactor pressure vessel and internals as wellas certain other metallic passive components that are susceptible to damage or degradation. In recognitionof the expanded scope, the Project Review Group (PRG) approved the transition of OPDE to a new,expanded Component Operational Experience, Degradation and Ageing Programme (CODAP). The
CODAP 2011-2014 and 2015-2017 work programme include the preparation of topical reports to foster
technical co-operation and to deepen the understanding of national differences in-plant ageing
management.
This report represents the third CODAP topical report and focuses on the effectiveness of reliability
and integrity management (RIM) programme. Through an examination of the operating experience as
recorded in the CODAP event database, the field experience with the different RIM strategies is evaluatedin order to draw qualitative and quantitative insights about the effectiveness of RIM to ensure pressure
boundary component reliability. Throughout this report the term “RIM” is used broadly to describe allprocesses that are in place to monitor the structural integrity of reactor components.
1.1 Background and related work
Structural integrity of piping components and systems and non-piping passive components such as thereactor pressure vessel and internals is important for plant safety and operability. In recognition of this,information on degradation and failure of metallic piping and non-piping passive components is collected
and evaluated by regulatory agencies, international organisations (e.g. NEA and IAEA) and industry
organisations worldwide. This information is used to provide systematic feedback, for example, to reactor
regulation and research and development programme associated with ageing phenomena, non-destructive
examination (NDE) technology, in-service inspection (ISI) programme, leak-before-break evaluations,risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component
reliability.
Assessment of passive component service experience data has been an integral element of regulatory
and industry programme to address long-term operation and nuclear plant licence renewal. Examples ofprogramme addressing long-term operations include the Proactive Materials Degradation Assessment(PMDA), Expanded Materials Degradation Assessment (EMDA), Generic Ageing Lessons Learned(GALL), GALL for Subsequent Licence Renewal (SLR) and International Generic Ageing Lessons
Learned (IGALL)1. A common feature of these 4 programmes is the acknowledgement of systematic
1. Detailed information on PMDA, EMDA and GALL are available at www.nrc.gov. The IGALL is summarized in IAEATECDOC-1736 (April 2014), which is available at www.iaea.org.

12

NEA/CSNI/R(2017)3
reviews of the accumulated service experience data as one of several inputs to the development of atechnical basis for practical ageing management of metallic passive components.
In parallel with the efforts to evaluate service experience data and to correlate the occurrence ofmaterial degradation with piping design and operational parameters, initiatives have been presented toestablish an international forum for the systematic collection and exchange of service experience data on
piping. Since May 2002, the NEA has operated the OPDE/SCAP-SCC/CODAP database projects under theco-ordination of the Committee on the Safety of Nuclear Installations (CSNI). The first term of the OPDE
Project covered the years 2002-2005; the second term covered the period 2005-2008; and the final term
covered the period 2008-2011 [1].
In May 2011 the PRG approved the transition of OPDE to a new, expanded NEA CODAP. A first
CODAP National Co-ordinators Meeting was held at NEA Headquarters in November 2011. The CODAP
Project builds on the success of OPDE and a related NEA data project, the SCAP SCC Working Group. In
December 2014 the PRG approved the programme plan for a second term (2015-2017).
In 2006 the Stress Corrosion Cracking and Cable Ageing Project (SCAP) was established under the
auspices of the NEA to assess two subjects: stress corrosion cracking (SCC) and degradation of cableinsulation. The project ran successfully from June 2006 to June 2010. SCAP was financed through avoluntary Japanese contribution to the NEA. Fourteen NEA member countries joined the SCAP in 2006 toshare knowledge and by 2010, seventeen countries had joined the project. The International Atomic
Energy Agency (IAEA) and the European Commission, through its Joint Research Centre in Petten, also
participated as observers. The general objective of the SCAP co-ordinated project was to share corporate
knowledge and operating experience, to understand the failure mechanisms, and to identify effective
techniques and technologies to effectively manage and mitigate active degradation in nuclear power plants.
The specific objectives of the project were to:Establish a database of major ageing phenomena for SCC and degradation of cable insulation
through the collective efforts of NEA members.



Establish a knowledge base of these topics by compiling and evaluating the collected data andinformation systematically.

Perform an assessment of the data and identify the basis for commendable practices to help
regulators and operators enhance ageing management Programme.

The scope of the SCAP SCC Working Group activities covered class 1 and 2 pressure boundary
components, reactor pressure vessel internals and other components, but excluded steam generator tubing.
The entire SCC database consisted of an event database and general information. The general informationconsisted of regulations and codes and standards; inspection, monitoring, and qualification requirements;
preventative maintenance and mitigation strategies, repair and replacement actions; safety assessment; andR&D. This cumulative information comprised the knowledge base [2].
Following the completion of the SCAP project, SCC Working Group participants were interested insome form of continuation and discussions were initiated to explore possible alternatives. It was recognisedthat there were many interests very similar to those existing in OPDE and the concept of a new project wasenvisaged to combine the two projects into the CODAP. The objective of CODAP is to collect information
on passive metallic component degradation and failures of the primary system, reactor pressure vesselinternals, main process and standby safety systems, and support systems (i.e. ASME Code Class 1, 2 and 3,
or equivalent). It also covers non-safety-related (non-Code) components with significant operational
impact [3][4][5].
The CODAP event database resides on a secure server at NEA Headquarters. Provisions exist for
online data submission and database interrogation (e.g. event review, QA, queries) as well as downloading
queries (in CSV- or XML-file format) and selected event records or the entire database (in XML-fileformat) to a local computer or computer network.
13

NEA/CSNI/R(2017)3
The CODAP-PRG membership consists primarily of nuclear regulatory agencies and their technicalsupport organisations (TSOs). Two nuclear utility organisations are also actively engaged in the CODAP
Project work; Électricité de France (EDF) and Slovenské Elektrárne (SEAS).
Apart from recognising the intrinsic value of exchanging operating experience data and related root
cause analysis results and insights, an important motivation for supporting the international databasecollaboration in 2002 was embedded in the then emerging trend towards risk-informed regulation,
including risk-informed in-service inspection (RI-ISI). An area of specific interest at the time wasconcerned with the technical basis for performing pipe failure probability analysis in support of RI-ISI
Programme development. The potential synergies between a comprehensive database such as CODAP and
the development of statistical passive component reliability models have been explored in multiple
database application projects2 . These reliability models include provisions for incorporation of RIM
effectiveness factors that are based on either expert judgement or on observed field experience data
including RIM performance demonstration testing results. The CODAP third topical report specificallyaddresses RIM effectiveness and the qualitative and quantitative insights that are derived from ISI histories
and root cause analysis results as captured in the CODAP event database.
During 2005-2008 the European Commission Joint Research Centre (EC-JRC) Petten and the NEA
organised the co-ordinated risk-informed ISI methodologies benchmark (RISMET) study [6][7]. RISMET
was a unique comparative study of various approaches to set up an ISI Programme. One of several
technical aspects of RI-ISI Programme development addressed by RISMET included the probability ofdetection (POD) and how the different RI-ISI methodologies account for the effects of probability of flaw
detection on an inspection scope.
1.2

Objectives and scope

The third topical report is concerned with the effectiveness of pressure boundary component RIM
processes and lessons learnt from field experience. In this report the term “RIM” accounts for all
Programmatic, organisational and technological elements that exist to ensure the integrity of reactor
components. Specifically, the report addresses current international practices with respect to pressure
testing, leak detection, ISI including risk-informed in-service inspection, NDE, and qualifications tovarious standards to improve the reliability in NDE techniques. RIM Programmes are designed to preventthe occurrence of piping through-wall leaks as well as to monitor passive metallic component degradation.
Through a systematic examination of the operating experience as recorded in the CODAP event
database, the field experience with the different RIM strategies is evaluated in order to draw qualitative andquantitative insights about integrity management effectiveness. The CODAP event database documents theinternational operating experience with metallic passive components. Specifically, the database isconcerned with degradation and failure of metallic passive components, including the implied or observed
safety and operational impacts. As such, the database is concerned with the root cause of degradation andfailure. Less-than-adequate (LTA) implementation of RIM is one important underlying cause of some ofthe observed degradation and failure events. In this report, LTA-RIM is defined as events wheredegradation has progressed beyond acceptable limits in systems, structures or components (SSCs) thathave a RIM Programme. These LTA-RIM events have some safety significance. In this topical report the
LTA-RIM definition is broadened to also include events where a RIM Programme has resulted in a “false
positive”; that is, it has identified degradation that either didn’t exist or was not close to violating
acceptance criteria. While such events needlessly expend resources and could be considered LTA-RIM
from an economic perspective, they do not have any safety significance. Included in this topical report is
2. Appendix A includes an OPDE/CODAP bibliography that identifies selected database applications that have been
performed or sponsored by the OPDE/CODAP member organisations.

14

NEA/CSNI/R(2017)3
an evaluation of the effect of different RIM strategies on pressure boundary component reliability. In
summary, the objectives of the topical report are to:Define passive component RIM and its relationship with existing codes and standards for ISI
and non-destructive examination (NDE) practices (Section 1.3).

Describe the evolution of RIM strategies relative to long-term nuclear power plant operation(Section 2).

Provide a summary of the Programmatic and technological facilities that are in place to ensure
a high degree of passive component structural integrity and the steps taken to prevent throughwall flaws from developing in metallic safety-related and non-safety-related piping
components. Included in this summary is a comparison of the different national practices andregulations with respect to RIM (Section 2).



Perform a systematic evaluation of the field experience data in the CODAP event database to
identify the types of RIM performance deficiencies that have contributed to failures inpreventing relatively benign degraded structural conditions to develop into more serious
structural integrity conditions (Section 3).

Explore the RIM performance data in CODAP with respect to quantitative piping reliability
and operability determinations of degraded conditions. Included in this exploratory aspect of
the topical report is an examination of the technical basis for analysing the effect of different
RIM strategies on the predicted passive component reliability (Section 4).

Provide an introduction to how to evaluate of the effect of different integrity management
strategies on pressure boundary component reliability (Section 5).



Identify potential improvements to the CODAP event database structure to better capture theinfluences of RIM on passive component degradation and failure. Any recommended
improvements will be taken into consideration when implementing future enhancements to the
web-based event database (Section 6.3).

The scope of this study is limited to evaluations of passive component field experience data as
documented in the CODAP event database. Specifically, the evaluations address the role of leak detection,
NDE, visual examination and pressure testing in preventing or mitigating passive component degradationand failure. The CODAP event database includes ISI histories and root cause analysis results from which
RIM performance issues are extracted.
1.3

Nomenclature

RIM involves those aspects of a plant design process that are applied to provide an appropriate level ofreliability of SSCs and a continuing assurance over the life of the plant that such reliability is maintained[8]. These include design features important to reliability performance such as design margins, selection ofmaterials, testing and monitoring, provisions for maintenance, mitigation of degradation processes, repair
and replacement, leak monitoring, pressure and leak testing, and ISI. In the context of the ASME Boiler
and Pressure Vessel Code Section XI, “Rules for In-service Inspection of Nuclear Power Plant
Components” (ASME XI) [9], RIM is an extension of ISI and is performance-based for NDE and online
monitoring of structural integrity using advanced technologies such as acoustic monitoring or guided
ultrasonic waves. In this report the term “RIM” is used broadly to describe all processes that are used tomonitor the structural integrity of reactor components. A database like CODAP provides vital technicalinput to RIM.
The development of a RIM Programme may involve applications of risk-informed passive component
reliability models in which explicit consideration is made of the available body of field experience data on

15

NEA/CSNI/R(2017)3
metallic passive components. Implicit consideration also is given to analytical insights from deterministic
and probabilistic fracture mechanics evaluations that address crack growth (e.g. time for an initial nonthrough-wall crack to grow through-wall) and probability of pipe break given a pre-defined pipe flaw. A
statistical model of piping reliability is expressed by: ix 

  P( R F ) I
ik

x

ik

ik

(1.1)

k

Where:

ix 

ik 
P( Rx Fik ) 

I ik 

Frequency of pipe failure of component type i with break size x, subject to epistemic
uncertainty calculated via Monte Carlo simulation
Failure rate per location-year for pipe component type i due to failure mechanism k,subject to epistemic uncertainty determined by Bayes method
Conditional rupture probability (CRP) of size x given failure of pipe component type i due
to damage or degradation mechanism k, subject to epistemic uncertainty
Integrity management factor for weld type i and degradation mechanism k, subject toepistemic uncertainty which may be determined by Monte Carlo simulation and Markov
model

In the above equation the integrity management factor accounts for the probability of a certain RIM
Programme to successfully identify degradation before the degradation has eroded the safety margin of thepipe to unacceptable levels. The integrity management factor can be determined on the basis of RIM
qualification data, expert judgement or field experience data such as that contained in the CODAP event
database. This model of piping reliability enables a quantitative assessment of the level of risk reductionthat is achievable with ISI.
Section XI (“Rules for In-Service Inspection of Nuclear Power Plant Components”) of the ASME
boiler and pressure vessel code (BPVC) was first issued in 1971 [10]. The philosophy of ASME XI was todeterministically mandate a sufficient number of examinations and pressure tests to provide assurance thatthe original safety that was designed and built into the plant is maintained throughout its service life. The
ASME Code provides requirements for examination, testing, and inspection of components and systems,
and repair/replacement activities in a nuclear power plant. The mandatory Appendix VIII of ASME
Section XI provides requirements for performance demonstration (PD) for ultrasonic examination
procedures, equipment, and personnel used to detect and size flaws. When ASME Section XI was firstissued in 1971, the ultrasonic inspection techniques specified were adapted from manufacturing practices
and were based on previous experience with fatigue cracking. The ultrasonic examination rules wereprescriptive and had not been evaluated on nuclear reactor service-induced degradation mechanisms. The
reliability of the NDE that were specified by Section XI was not quantified via testing. Instead, the initial
premise of Section XI was that reliable ultrasonic testing (UT) could be ensured through detailed rules.
Based on early field experience, the failure of ISI to detect cracks before developing into leaks raised
concerns regarding the effectiveness of UT being conducted at nuclear power plants and showed that
improvements in inspection requirements were needed [11]. In response, the US Nuclear Regulatory
Commission (NRC) sponsored R&D to assess NDE reliability. This research showed that prescriptive
requirements could not be written to sufficiently accommodate the diversity of power plant materials, field
conditions, and material degradation processes typically encountered. It was subsequently decided that aperformance-based testing approach would be the most effective means for achieving the needed
improvements in NDE reliability.
In 1989, ASME Section XI adopted a performance-based philosophy that relied on the qualification of
UT procedures, equipment, and personnel to detect and size flaws in piping and vessels. These newrequirements were incorporated into Section XI in the mandatory Appendix VIII, titled “Performance16

NEA/CSNI/R(2017)3
Demonstration for Ultrasonic Examination Systems.” In 1991 the many of the US utilities and several
international utilities formed the Performance Demonstration Initiative (PDI) to implement the then current
performance demonstration requirements of the ASME Code, Section XI, Appendix VIII “Performance
Demonstration for Ultrasonic Examination Systems” [12]. The PDI designated the Electric Power
Research Institute (EPRI) Non-destructive Examination Centre to be the performance demonstration
Administrator (PDA).
Division 1 of ASME XI applies to existing light water reactor (LWR) designs. Building on insightsfrom the implementation of RI-ISI during 1995 to the present time, as well as on ISI requirements for new
Gen-III+ and Gen-IV reactor designs, a new Division 2 of ASME XI is under development. The new
Division 2 applies a risk-informed approach to the development and implementation of RIM. Of note isthat the new Division 2 of the ASME Code is intended to apply to license extension for operating NPPs
from 60 to 80 years of operation and even for non-nuclear complex engineered systems [13].
1.4Disclaimer

The CODAP Project places strong emphasis on data quality, including the completeness andcomprehensiveness of recorded events. Data quality is achieved through a formal validation process asarticulated in a Coding Guideline. The roles and responsibilities with respect to data submissions and data
validation are defined in the CODAP Operating Procedures. Section 3 of this report includes high-level
database summaries with respect to RIM and its importance relative to mitigation or prevention of passive
component material degradation. In the high-level data processing the term “less-than-adequate RIM”
(LTA-RIM) is used to address deficiencies in RIM processes3 to prevent or mitigate material degradation,or to identify recordable or rejectable flaws before exceeding acceptance criteria for continued operation.
The results of the data processing and analysis that are presented herein are based as much as possible on
the results of root cause analyses. The writing team for this topical report has not performed any reassessments of the findings related to RIM process deficiencies as documented in the root cause analysis
reports that are embedded in the CODAP Event Database.
The CODAP-PRG is fully aware of the fact that the full root cause analysis documentation as
prepared by an owner/operator or its subject matter experts is not normally disseminated outside theindustry. The CODAP Coding Guidelines includes instructions for what “root cause information” toinclude in the database. As a guiding principle, the instructions provided state that any relevant information
on a cause-consequence relationship is to be included. Respective National Coordinator assumes
responsibility for the accuracy on the technical information that is input to the event database. Furthermore,
the web-based database has provisions for uploading (or attaching) any available supporting information;e.g. laboratory reports, root cause analysis reports.
Section 2 of this report includes high-level summaries of selected national RIM practices. The
CODAP-PRG recognises that significant efforts are continuously being directed towards NDE
performance demonstration. It is one of several evolving aspects of RIM. Based on publicly available
information, the purpose of Section 2 is to highlight certain aspects of past and current national
performance demonstration activities.
1.5

Report structure and reading guideThis topical report consists of seven sections and four appendices. Section 2 describes selected codes andstandards for RIM. The section addresses the requirements for ISI and NDE qualification as defined in
ASME XI, Canadian Standards Association (CSA) rules, German KTA rules, French RCC-M Code, andKorean rules. Section 3 documents how structural integrity management influences are recorded in the
3. This includes RIM program definition, implementation and monitoring, application of qualified NDE techniques,
documentation, and evaluation of NDE results.

17

NEA/CSNI/R(2017)3
CODAP event database. Section 4 documents CODAP database insights regarding the effectiveness of
RIM. Key lessons learnt regarding leak detection and RIM performance deficiencies are highlighted. Highlevel database insights are summarised in a suite of charts and tables. Section 5 describes methodologies
for addressing the integrity management factor in Equation 1.1 above and how it relates to field experiencedata. A summary and conclusions are documented in Section 6. Finally, a list of references is provided in
Section 7. Appendix A is an OPDE/CODAP bibliography including references to database applications
performed or sponsored by OPDE/CODAP member organisations. Appendix B is a high-level CODAP
event database status report; Appendix C is a glossary of terms; and Appendix D includes a piping safetyclass cross reference table.

18

NEA/CSNI/R(2017)3

2. RELIABILITY AND INTEGRITY MANAGEMENT PRACTICESThe CODAP event database design is based on a detailed consideration of passive component reliability
attributes and influence factors. For those RIM Programme that rely on non-destructive examination(NDE), NDE qualification techniques continue to play an important role in ensuring a high degree ofstructural integrity of reactor components. The event histories that are included in the database aremanifestations of effective implementation of RIM as well as failures of RIM to identify structural flaws(e.g. non-through-wall cracks) before exceeding code requirements or design limits. The objective of thissection is to summarise current ISI requirements and the associated NDE qualification requirements asformulated in national codes and standards for reactor components. Also included is an overview of ownerdefined (or augmented) in-service inspection programme that go beyond inspection requirements asformulated in existing codes and standards. In some cases, these owner-defined inspection programme
have been formulated on the basis of risk insights from plant-specific probabilistic safety assessment(PSA) studies.
2.1

ASME Section XI, Appendix VIII

The US Nuclear Regulatory Commission (NRC) establishes rules and regulations for the operation ofdomestic nuclear facilities. Section XI of the ASME Code contains the rules for ISI of nuclear plantcomponents [9]. Through Title 10 of the Code of Federal Regulation, Part 50.55a, Codes and Standards(10 CFR 50.55a)1 , the NRC incorporates by reference (thus, mandates the implementation of) the ASME
Code, Section XI, Division 1, Appendix VIII, “Performance Demonstration for Ultrasonic Examination
Systems”. Appendix VIII of ASME XI requires qualification of the procedures, personnel, and equipmentused to detect and size flaws in piping, bolting, and the reactor pressure vessel. Each organisation(e.g. owner or vendor) will be required to have a written programme to insure compliance with therequirements. The general requirements for qualification of non-destructive examination (NDE) personnel
contained in Section XI, IWA-2300, are amended by Section XI, Appendix VIII. Appendix VIII describes
the additional requirements for performance demonstration of ultrasonic examination systems that integrate
personnel, equipment, and procedures into a single entity. ASME Code Section XI, Appendix VIII, also
includes supplements that contain specific instructions for the conduct of performance demonstrations
including test specimen requirements, conduct of performance demonstration and acceptance criteria.
The Performance Demonstration Initiative (PDI) was formed by US utilities in 1991. Its primary
focus was to provide an efficient, cost-effective, and technically sound implementation of Appendix VIII
performance demonstration requirements. The EPRI NDE Centre is the Performance Demonstration
Administrator (PDA) for the programme and provides technical and implementation support to develop
detailed implementation processes and facilities to ensure that all requirements of Appendix VIII of ASME
Section XI and relevant modifications required by 10CFR50.55a are met [14]. Using common protocols
and sample sets, the need for site-specific performance demonstrations is minimised. Piping, bolting, andreactor pressure vessel (RPV) shell weld performance demonstration testing began in 1994 [15]. The
demonstration phase for piping and RPV shell welds has provided significant information relative to the

1. For additional information got www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0055a.html.

19

NEA/CSNI/R(2017)3
reliability of ultrasonic examinations. The numbers of measurements are in excess of 16 000 for piping
detection and length sizing [16].
With the publication of the revised 10 CFR 50.55a rule on 22 September1999, the NRC required
expedited implementation of all Appendix VIII requirements2. As a result, PDI initiated activities for a
Phase II of the Programme. Phase I of the PDI Programme addressed important problems that could havesignificant impacts on the cost and effectiveness of the US performance demonstration approach. These
included the nozzle inner-radius examination from the outside surface, nozzle-to-shell weld examinations,and dissimilar metal welds (DMW).
The PDI Programme addresses the sample fabrication and operational requirements for a unified
performance demonstration programme operated by EPRI. The EPRI Programme contains establishedrequirements for the demonstration test blocks as well as the rules and protocol for the programme
operation.
The samples are a critical ingredient of the programme and represent a substantial portion of the total
cost. The samples are designed to cover the widest practicable range of components in typical boilingwater and pressurised water reactors. Detailed dimensional information was obtained and reviewed todesign samples that span the maximum range according to the rules in Appendix VIII that govern the rangeof qualifications.
2.1.1 Code comparison
The Multinational Design Evaluation Programme (MDEP) Code Comparison Project was initiated in late2006 in response to a request by the MDEP Codes and Standards Working Group (CSWG). The CSWG
facilitated a project to develop comparisons of the examination requirements for Code Class 1 vessels,
piping, pumps and valves. These comparisons addressed multiple code sections and subsections, including
rules for ISI. Published in 2012, Report STP-NU-051-1 [17] documents the differences in ISI examination
requirements between major international nuclear codes and standards for Class 1 equipment; namely thoseof AFCEN (RCC-M), ASME (Section III), CSA (N-285), JSME (S NC1), KEA, Korea Electric Power
Industry Code (KEPIC-MN) and NIKIET (PNAE-G7). Further elaborations are documented below.
2.2

Canadian Standards Association Periodic Inspection Programme

CSA-N285.4 1.1[18] defines requirements for the periodic inspection of primary heat transport andauxiliary pressure-retaining systems, components, and supports that form part of a CANDU nuclear power
plant. Compliance with CSA N285.4 is a licensing requirement for Canadian CANDU plants and has been
adopted for CANDU plants outside of Canada. CSA N285.4 provides criteria to establish the inspection
requirements for system or portions of systems subject to inspection, including:(a) Systems, and systems connected thereto, containing the fluid that, under normal conditions,directly transports heat from nuclear fuel, and other systems whose failure can result in asignificant release of radioactive substances.
(b) Systems essential for the safe shutdown of the reactor and/or the safe cooling of the nuclear fuelin the event of a process system failure.
(c) Systems and equipment whose failure or dislodgement could jeopardise the integrity of systems
in Item (a) or (b), or both.
In addition the CSA establishes specific inspection requirements for fuel channel pressure tubes,
feeder piping and steam generator tubes. Inspection categories in CSA N285.4 are determined based on
failure size, fatigue usage and stress intensity. Inspection categories are A, B, C1 and C2, with category A
2. For additional details, go to www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0055a.html

20

NEA/CSNI/R(2017)3
being the highest and category C2 the lowest (no inspection required). CSA N285.4 provides an implicit
risk-related rationale for selection of inspection locations by:

Selecting high risk areas (high inspection categories) with high failure consequence (failure size)and high failure potential (fatigue usage and stress intensity) for inspection;



Minimising inspection scope, effort and dose by reducing inspections in areas with low failure
potential and low consequence; andIncluding provisions such as exempting systems or portions of systems from inspection if they
are periodically tested or system pressure is continuously monitored.

N285.4 provides guidelines for the classification of components into families based upon design andservice conditions and provides sampling requirements for the number of components within each family
requiring inspection. For multi-unit stations, the sampling criteria also considers the number of components
within a family across all units. The lead unit (typically the unit with the highest number of effective fullpower operating hours) will require the largest sample of components inspected within a family and thenumber of components is reduced for the following units. If service-induced degradation is identified
within a family of components, the sample size must be re-evaluated to ensure that the degradation
mechanism is effectively managed. CSA N285.4 sampling criteria differs from RI-ISI methodologies,
where a percentage of welds from each risk category are selected.
CSA N285.4 allows for the exclusion of systems (or portions of systems) from inspection when thesubject component:

has adequate barriers between fluid boundary and nuclear fuel;

has adequate barriers between fluid boundary and the outside atmosphere;

has acceptable level of dose release in the case of system failure and without operation of thecontainment system; oris dormant and is subject to periodic testing or continuous pressure monitoring all the time.

Because of the unique design of the CANDU Primary Heat Transport System (PHTS), design-specific
NDE qualification programme and NDE technologies have been developed. Special manual and automated
NDE tools have been developed to inspect for pipe wall thinning and cracking at tight radius PHTS feeder
pipe bends 3 [19]. In addition, NDE tools have been developed for the inspection of pressure tubes forservice-induced degradation, irradiation induced wall thickness changes, spacer location and themeasurement of the gap between a pressure tube and a Calandria tube [20].
CSA N285.4 contains a general requirement for licensees to demonstrate the performance of theadequacy of all inspection procedures used to detect and size flaws and demonstrate the proficiency ofinspection personnel. The 2009 edition of the standard introduced specific requirements for inspection
qualification for volumetric inspection methods (typically ultrasonic or eddy-current methods), which
includes:

a specification of the specific inspection requirements for each application of the inspection
method;an inspection procedure that defines the application of the inspection method;a report that documents how the procedure satisfies the specified inspection requirements; anda personnel training and qualification programme.

3. The PHTS consists of a large volume of small-diameter piping (DN50 to DN65 inlet/outlet feeders). Due to the complex
configuration of the PHTS piping, extensive NDE qualification programs have been implemented in conjunction with
automated NDE data collection systems.

21

NEA/CSNI/R(2017)3
To satisfy this inspection qualification requirement of the CSA standard, the Canadian CANDU
industry has chosen to adopt a process based upon the ENIQ approach to inspection qualification and set
up the CANDU Inspection Qualification Bureau (CIQB) to establish industry guidelines for thequalification of inspection procedures and qualification of personnel [21]. The CIQB was establishedthrough the CANDU Owners Group (COG) but operates independently from the Canadian utilities. COG
is also a member of the ENIQ steering committee.
A previously mentioned, the definition of an N285.4 inspection programme contains risk elements
when considering the likelihood and consequences of failure when selecting inspection locations.
However, the Canadian regulator, the Canadian Nuclear Safety Commission (CNSC), and the Canadianindustry have undertaken projects to assess the possibility of adopting risk-informed in-service inspection(RI-ISI) [22] in lieu of the current N285.4 requirements. Information obtained through preliminary workand pilot studies developed by the regulator and industry has been evaluated and it has been mutuallyagreed that the adoption of an RI-ISI methodology would best be implemented through the CSA standards
development process. Work to date has suggested that the EPRI methodology for an RI-ISI Programme[22] would be the most appropriate for Canadian CANDU applications. Update No. 1 for the 2014 editionof the N285.4 standard, which underwent public review in 2015, will incorporate a non-mandatory annex
providing an alternate methodology for identifying components susceptible to fatigue related degradation
mechanisms based upon the EPRI RI-ISI degradation assessment. This will be a first step at theimplementation of RI-ISI based information into that standard.
In addition to conducting inspections on the PHTS and auxiliary systems, Canadian NPP licences
require utilities to implement inspection programme for safety significant balance of plant pressureretaining systems and components. In order to establish a consistent approach to satisfying this licencerequirement, a new standard, CSA N285.7, Periodic inspection of CANDU nuclear power plant balance ofplant systems and components, has been drafted and publication is expected in 2016. This standard will
apply the EPRI RI-ISI methodology to balance of plant pressure boundary systems and components to
identify those having the most significant impact of core damage frequency (CDF) should they fail during
operation due to service-induced degradation. The standard incorporates inspection qualification
requirements for procedures and personnel that are consistent with CSA N285.4.
2.3

French RSE-M code

The RSE-M, “Règles de Surveillance en exploitation des Matériels Mécaniques des Ilots Nucléaires REP” (“InService Inspection Rules for Mechanical Components of PWR Nuclear Islands”) [24] is the French equivalent
to ASME Boiler and Pressure Vessel Code, Section XI. Section A of RSE-M deals with the application ofdestructive and non-destructive testing methods for examining materials or components after their manufactureand during their operation. RSE-M A 4000 defines the requirements for the certification of NDE personnel forin-service components. This section of RSE-M requires personnel to be certified by a third party according to
NF EN ISO 9712 [25] or its other European counterparts. The first edition of the RSE-M code was issued in
1989. Relative to ASME XI, for Code Class 1 components RSE-M differs in several areas with respect toinspection intervals, prescribed NDE technologies, and flaw evaluation processes.
2.4

Nuclear Safety Standards Commission (KTA)

The German safety standard KTA 3201.4 “In-Service Inspections and Operational Monitoring” [26]documents the detailed measures which shall be taken to meet these requirements within the scope of its
application. For this purpose, a large number of standards from conventional engineering, in particular DIN
standards, are also referenced; these are specified in each particular case. The requirements specified under
KTA 3201.4 address the following activities:reliable monitoring of operating conditions;adequate extent of NDEs;
22

NEA/CSNI/R(2017)3

documentation, evaluation, safety-oriented application and up-dating of operating experience;
foreseen identifiability of changes of the as-fabricated condition of the reactor coolant pressure
boundary by means of ISI and operational monitoring;evaluation of the results of in-service expections and operational monitoring.

Section 7.4 of KTA 3201.4 documents requirements regarding NDE personnel. The test personnel
shall have been qualified and certified to satisfy the requirements of DIN 25435-1, DIN 25435-2, DIN
25435-3, DIN 25435-4, DIN 25435-6 and DIN 25435-7 for the test methods to be used. With respect tomanual UT, the test personnel shall fulfil all requirements of DIN 25435-1 Table 2. For eddy-current
testing performed on steam generator tubes the test personnel shall satisfy the conditions of DIN 25435-6
Table 1. The test personnel for evaluation of the general condition shall have the knowledge required toperform their tasks and shall have demonstrated sufficient visual capacity. Finally, the test personnel for
functional tests shall have the knowledge required to perform their tasks.
Section 8 of KTA 3201.4 documents requirements concerning the evaluation of NDE results. As
general principles, “conspicuous findings and peculiarities that influence the result shall be recorded andevaluated.” The test supervisory personnel of the testing agency, the plant owner and the authorised
inspector shall convince themselves and confirm within the recording activities that the tests have been
performed completely to satisfy the requirements and have been evaluated correctly to ensurereconstruction.
The German approach to RIM is the subject of the draft KTA 3206 [27][28][29]. In Germany, theleak-before-break (LBB) concept was implemented for all commercial nuclear power plants in the 1980s.
The technical approach to LBB is based on application of advanced condition monitoring systems
specifically developed to address the major structural integrity influence factors including local loading
conditions. Through the use of these advanced monitoring systems it was possible at the outset to establish
modes of operation that minimise the load on components, to implement out appropriate technical back-fit
measures, and to identify degradation mechanisms and the appropriate degradation mitigation processes.
2.5

Korean Nuclear Safety and Security Commission (NSSC) and KEPIC MI

The Korean NSSC establishes rules and regulations for the operation of domestic nuclear facilities. The
NSSC has established the KEPIC as the approved technical standard used to confirm the safety of nuclear
facilities4. The Notice of the NSSC “Regulation on In-Service Inspection of the Nuclear Reactor Facilities”
requires that MI (ISI) of KEPIC shall apply to the ISI of domestic PWRs. KEPIC MI is the Koreanequivalent to the ASME code Section XI which contains the rules for ISI of nuclear plant components. For
the ISI of the pressurised heavy water reactor (PHWR), CSA N285.4 shall apply.
The NSSC Notice for ISI also requires augmented inspection for the butt welds of class 1 piping (notless than 2-inch nominal pipe size) that is attached to the reactor coolant pressure boundary (RCPB) and is
non-isolable during normal operation. For these piping butt welds in PWRs, volumetric examination in
addition to surface examination required by KEPIC MI shall be performed. In the case of the PHWR, bothvolumetric and surface examination shall be performed for these butt welds which had been excluded fromthe scope of ISI. This augmented requirement was included as a result of two RCPB leak events in primaryheat transport purification system piping welds in a Korea PHWR unit in 2001 and 2002. After the events,
a task force team was formed to develop a comprehensive safety enhancement programme for welds whichbecame the basis of the augmented inspection.
The requirements of performance demonstration of NDE were included in the NSSC Notice in 2004.
The installer or operator shall conduct the performance demonstration of ultrasonic testing (UT) for thesafety-related nuclear components and of eddy-current testing (ECT) for steam generator tubes. The
4. For more information go to www.electricity.or.kr/english/kea/industry%20code.php

23

NEA/CSNI/R(2017)3
technical standards of the UT performance demonstration shall be KEPIC MI and its mandatory
Appendix VIII (Equivalent to ASME Code Section XI, Appendix VIII). The operator shall submit thereport to the NSSC three months prior to application of performance demonstration. The report shalladdress:name and address of the organisation carrying out performance demonstration;

applicable technical standards for performance demonstration and their edition;

test specimen or test data of performance demonstration;

evaluation of performance demonstration;quality assurance of performance demonstration;security related to performance demonstration;

operation of performance demonstration system; andother related items.

The organisation carrying out performance demonstration shall report every year the status of thedemonstration to the NSSC. For the application of KEPIC, limitations defined in the Notice of the NSSC
“Guidelines for Application of Korea Electric Power Industry Code (KEPIC) as Technical Standards of
Nuclear Reactor Facilities” shall be followed.
2.6

European Network for Inspection and Qualification

The European Network for Inspection and Qualification (ENIQ) was formed in 1992 and reflects theimportance of the issue of qualification of NDE inspection procedures used in ISI programme for nuclear
power plants. Driven by European Nuclear Utilities and managed by the European Commission Joint
Research Centre (JRC) in Petten, the Netherlands, ENIQ was intended to be a network in which theavailable resources and expertise could be consistently managed throughout Europe. It was also recognisedthat harmonisation with respect to codes and standards for inspection qualification would represent
important advantages for all parties involved, with the ultimate goal of increasing the safety of European
nuclear power plants.
ENIQ work is carried out by two sub-groups: the Task Group on Qualification (TGQ) focuses on thequalification of ISI systems, while the Task Group on Risk (TGR) focuses on risk-informed in-service
inspection (RI-ISI) issues. The TGR has published the European Framework Document for RI-ISI [30],and is producing more detailed recommended practices and discussion documents on several technicalissues specific to RI-ISI.
Since 2005 ENIQ has published a set of ten recommended practices for the development andimplementation of RIM strategies. These recommended practices address the following topics5 :
1.
2.
3.
4.
5.
6.

NDE qualification;
strategy and recommended contents for technical justifications;recommended contents for the qualifications dossier;guidelines for the design of test pieces and conduct of test piece trials;
the use of modelling in inspection qualification;
recommended general requirements for a body operating qualification of non-destructive tests;

5. The ENIQ Recommended Practice publications are available on the Internet at http://capture.jrc.ec.europa.eu/capture//eniqrecommended-practice24

NEA/CSNI/R(2017)3
7.
8.
9.
10.
2.7

qualification levels and approaches;verification and validation of structural reliability models and associated software to be used
in RI-ISI;
personnel qualifications;guidance on expert panels in RI-ISI.

NDE qualification methodologies

An overview of the evolution of NDE qualification is found in Reference [31]. The field experience with
NDE including failures of NDE to detect flaws prompted the development of procedures and processes toenhance NDE reliability. NDE qualification is concerned with the ability, capability of NDE technology(hardware and software) and the competency of NDE personnel. Where applicable, information about
component-specific NDE qualifications is included in the CODAP event database under “ISI history.” The
NDE technology includes manual and automated flaw detection techniques. In order to determine theability of NDE to reliably detect subsurface or surface flaws, systems have been devised to ensure that
NDE personnel have the proper training, have passed written and practical examinations, and have enoughexperience to properly perform NDE tasks using the applicable test method or technique. Personnel thathave met all three of these requirements may then be certified, which is defined in several different ways
under the various NDE systems. ASME XI Article IWA-2300 defines the NDE personnel qualifications
requirements. In the US, the American Society for Non-destructive Testing (ASNT) has developed
standards and recommended practices for qualification and certification of NDE personnel:
Certification Standard: National or international documents describing the requirements for thequalification and certification of NDE personnel.



Recommended Practice: A formal document that provides nationally or internationallyrecognised guidelines, and describes the qualification and certification process for NDE
personnel. If mandated by governing codes, standards, specifications or contract documents,
these guidelines become requirements for the specified project.

Certification System: The combination of the standard or recommended practice governing thecertification requirements, the third-party certification programme (if applicable) or theemployer's written practice, and additional employer documents used in the administration of
their certification programme.

Certification programme: The documented employer’s or certification body’s procedures andprocesses based on a standard or recommended practice, which defines the requirements of thatspecific programme.

There are multiple NDE certification systems worldwide, but they can generally be divided into two
main types: “utility-based” and “central” certification systems.


Utility-based certification system in which the nuclear power plant owner/operator is responsible
for the administration of the training and the examinations of their own NDE personnel, as wellas the documentation of the required training, examinations and experience in accordance with an
in-house standard or recommended practice. Most utility-based systems do allow theowner/operator to accept training and examination services provided by outside agencies
provided it is properly documented and the employer has determined that the content of thoseservices meet their own company requirements as described in the owner’s/operator’s NDE
procedures. Reference [32] is an example of a utility-based NDE qualification and certification
system.



Upon proof of certification, the owner/operator may issue a certificate, which can be a formal
certificate or in letter format, and can authorise their personnel to perform NDE tasks. In all
25

NEA/CSNI/R(2017)3
utility-based systems, the employer is responsible for authorising their personnel to perform suchwork. Because utility-based certification is usually tailored to plant-specific needs, the resulting
certifications expire when an employee leaves the company that issued the certification.
Central certification systems are systems in which the examinations are administered by anindependent third-party certification body based on a central certification standard. To be eligible to sit forthese examinations, prospective candidates must provide acceptable documentation of their training andexperience to the certification body. Upon successful completion of the third-party examinations, thecertification body will issue a certificate attesting to the fact that the named certificate holder has met therequirements and passed the examinations described in the third-party certification system. The employer
can then choose to accept the third-party certificate(s) as proof of certification. As with employer-based
systems, the employer has the ultimate responsibility to certify (authorise) the certificate holder to perform
NDE tasks. Most certification programme have three levels of NDE qualification: Level I, Level II andLevel III. The descriptions provided here are from the 2006 edition of the ASNT Recommended PracticeNo. SNT-TC-1A [33].


NDE Level I personnel should be qualified to properly perform specific calibrations, specific
NDE and specific evaluations for acceptance or rejection determinations according to written
instructions and to record results. The NDE Level I should receive the necessary instruction andsupervision from a certified NDE Level II or III individual.



NDE Level II personnel should be qualified to set up and calibrate equipment and to interpret andevaluate results with respect to applicable codes, standards and specifications. The NDE Level IIshould be thoroughly familiar with the scope and limitations of the methods for which he isqualified and should exercise assigned responsibility for on-the-job training and guidance of
trainees and NDE Level I personnel. The NDE Level II should be able to organise and report theresults of NDE tests.



NDE Level III personnel should be capable of developing, qualifying and approving procedures,
establishing and approving techniques, interpreting codes, standards, specifications andprocedures, as well as designating the particular NDE methods, techniques and procedures to be
used. The NDE Level III should be responsible for the NDE operations for which he is qualified
and assigned and should be capable of interpreting and evaluating results in terms of existingcodes, standards and specifications. The NDE Level III should have sufficient practicalbackground in applicable materials, fabrication and product technology to establish techniques
and to assist in establishing acceptance criteria when none are otherwise available. The NDE
Level III should have general familiarity with other appropriate NDE methods, as demonstrated
by an ASNT Level III Basic examination or other means. The NDE Level III, in the methods inwhich he is certified, should be capable of training and examining NDE Level I and II personnel
for certification in those methods.



NDE Instructor: The term “NDE Instructor” is used in the ANSI/ASNT Standard CP-189(Qualification and Certification of Non-destructive Testing Personnel) [34], to describe anindividual with the skills and knowledge to plan, organise and present classroom, laboratory,
demonstration, and/or on-the-job NDE instruction, training and/or education programme.

ASME XI, Mandatory Appendices VI, VII and VIII document qualification requirements for NDE
personnel performing visual examination, for personnel performing UT examination, and for performance
demonstration of ultrasonic examination systems, respectively. The US nuclear industry and several
international utilities relies on the EPRI/PDI Appendix VIII Programme to satisfy the ASME XI
requirements as well as the 10 CFR 50.55a rule (refer to Section 2.1 of this report). Similarly, the European
Qualification Methodology Document (EQMD) contains guidelines for the qualification of NDE and it
addresses technical justification, which involves assembling all the supporting evidence for inspection
capability (results of capability evaluation exercises, feedback from site experience, applicable and26

NEA/CSNI/R(2017)3
validated theoretical models, physical reasoning), and may include practical trials using deliberately
defective test pieces.
In 2007 the World Nuclear Association (WNA) established the Co-operation in Reactor Design
Evaluation and Licensing Working Group (CORDEL WG) with the aim of stimulating a dialogue between
the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators on thebenefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs.
The Codes and Standards Task Force (CSTF) of the CORDEL WG was set up in 2010 to collaborate
with the Standards Developing Organisations (SDOs) and the MDEP Codes and Standards Working Group(CSWG) on the international convergence of mechanical codes and standards, related to the design andquality of nuclear power plant components that are important to safety.
The CORDEL WG also encouraged the CSTF to work with the SDOs with the objective ofidentifying areas where a) convergence seemed most feasible; or b) a demonstration of equivalent level ofsafety could be made. On the basis of an internal survey the CSTF established two pilot projects. The first
focused on regulatory requirements: Qualifications for NDE personnel. The other project relates totechnical requirements: Non-Linear Analysis Methodologies Code Requirements. This first report on
Qualification of NDE Personnel looks at and compares requirements in the major nuclear design codes,
presents current international industrial certification practices and presents a recommendation forconvergence in certification requirements of the different codes. A proposal for a harmonised international
alternative for the certification of NDE personnel has been proposed by the CORDEL WG [35].
According to CORDEL, there are three stages that NDE personnel have to undergo before they are
qualified to perform a test on Class 1 components (not required for class 2 and 3 components) in mostcountries. These are Certification, Training and Qualification. The first stage, Certification, is required by
nuclear design codes and national regulators. For example, in the United States this includes an ASNT
SNT-TC-1A type certification under ASME Section III while in France it includes EN ISO 9712 [36] inthe RCC-M. The certification required by each of the design codes investigated is summarised in the
MDEP-CSWG code comparison report. Due to the different requirements, the inability to easily transfer
certification between countries establishes a major obstacle to vendors and operators of existing plants andnew builds. The effect of this on the international markets makes it an issue CORDEL CSTF aims toresolve.
The second stage, Training, is defined as additional in-house, job or site-specific training (proceduretraining, definition of degradation mechanisms, specificities of inspected components). This is often
required by employers, before the employee is authorised to conduct an examination. This is not a code
requirement and is applied at the discretion of the employer based on its needs and its culture. These can beparticularly important when a third-party certification is used to ensure that personnel are fit to follow a
written procedure and perform an NDE test.
The third stage, Qualification, is the highest level. An independent third party is required to validate
the qualification to the certificate holder. This qualification is normally specific to the control of Class 1
components. The ENIQ framework and Section XI requirements are the two major sets of requirements forcertification of NDE personnel for ISI of nuclear-grade components internationally. It is important tounderstand the differences between the personnel certification schemes described by EN ISO 9712:2012 or
ASNT SNT-TC-1A, and qualification of an NDE personnel inspection system in accordance with the
ENIQ Methodology for Qualification (EMQ). ASME Appendix VIII describes the requirements for
qualification of procedures, equipment and personnel.


The certification processes (ISO 9712 and SNT-TC-1A) explain how to qualify the workforce fora basic assessment of the basic knowledge of a method or technique. These certification
processes provide confidence that an operator who is successfully certified in accordance withthe process has broad knowledge of the principles, application and capability of a particular NDE

27

NEA/CSNI/R(2017)3
method for a range of situations. In other words, it is a ‘method’-based certification process for
personnel.
The qualification of the personnel or a procedure according to the ENIQ EMQ or EPRI Appendix
VIII Programme define performance-based requirements. A qualification of an NDE system inaccordance with the ENIQ or Appendix VIII methodologies is a more stringent process that
demonstrates that the combination of inspection system and personnel with an acceptable ofknowledge, obtained via a training dossier, is capable of achieving very specific defect detection
and sizing criteria in a particular situation.

Convergence of certification requirements would provide vendors with the confidence that thecertified personnel have a broad knowledge of NDE methods. This will allow for transferability of
personnel but will not change the responsibility of the vendor to ensure that the procedure, and the capacityof the personnel to apply the procedure, is adequate.
2.8 Owner-Defined Integrity Management Programme
An owner-defined (or augmented) ISI Programme goes beyond inspection requirements as formulated inexisting codes and standards. There are two types of owner-defined programme: 1) Programme developed
to address specific types of degradation mechanisms, and 2) Programme developed to address plantspecific risk insights on the basis of PSA, especially internal flooding vulnerabilities. Examples of theformer include flow-accelerated corrosion (FAC) inspection programme, inspection programme
responding to the Boiling Water Reactor Vessel and Internals Project (BWRVIP), and inspection
programme that address PWSCC in DMW (see Section 2.9). In the US many utilities have committed to
implement the BWRVIP guidelines. Examples of the second type inspection programme include
volumetric inspection programme for selected Code Class 3 and non-Code piping systems.
The CODAP first topical report [4] documents examples of different national FAC inspection
programme and implementation practices. As one example, in the US, Generic Letter 89-08 [37] requested
that all licensees implement a long-term FAC detection programme to prevent pipe failures in high-energy(single- and two-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 high-energy carbon steel piping systems. As stated in the NRC Inspection
Manual, Procedure 49001, “the long-term programme must be well defined, with clearly documented
results, and must include a complete analysis of the susceptible systems, inspection of the most susceptible
piping components, repair or replacement of damaged piping components, trending of inspection data inorder to determine FAC rates, and continued analysis based on inspection findings.”
An example of a service water piping augmented inspection programme is documented in
Reference [38]. According to this example, the programme activities are subject to inspection during theformal NRC triennial heat sink performance inspection. In addition to inspecting the Indian Point Energy
Centre (IPEC) Programme during the triennial heat sink performance inspections, the NRC Staff reviewed
the programme during the licence renewal ageing management programme audits in 2008. Changes to theprogramme since its inception in response to GL 89-13 [39] were reflected in the programme
documentation reviewed in 2008. There have been improvements and enhancements to the programme
since 2008. Typically, changes to the programme entail the addition of activities that were not required bycommitments to GL 89-13. For example, following a leak in 19956 , the programme was modified to add
valve inspections in addition to the piping inspections that were established to meet GL 89-13
commitments. Furthermore, the IPEC Programme was changed to increase the number and frequency ofinspections of non-safety-related piping [38].

6. CODAP Event Database record #4401.

28

NEA/CSNI/R(2017)3
The established Service Water Integrity Programme targets susceptible locations for wall thickness
inspections and conducts follow-up inspections for trending purposes of piping locations where corrosion
has been previously identified. The programme provides a means to proactively detect and repair areas ofconcern, while also providing a means to schedule and perform future inspections/repairs. The Service
Water Integrity Programme inspects about 20-30 welds per unit prior to each outage (i.e. pre-outage)
which includes both new inspection points as well as follow-up exams of known areas of concern.
Generally, new locations form the majority of the inspection points, and the follow-up inspection points area lesser portion. This is mainly because if areas of concern are found, wall thickness calculations are
performed and repairs are planned based on the projected remaining life. Since 1997, there have been over
600 weld examinations performed at both Indian Point nuclear reactor units, with greater than 90% of theexamined welds meeting the applicable acceptance criteria [38]. Those welds not meeting the establishedacceptance criteria are repaired during subsequent refuelling outages.
Another example of an augmented inspection programme is concerned with BWR core shroud welds.
In 1990, crack indications were reported at core shroud welds located in the beltline region of
Kernkraftwerk Mühleberg (KKM). This reactor had completed approximately 190 months of power
operation before the cracks were discovered. As a result of this discovery, General Electric (GE) issued
Rapid Information Communication Services Information Letter (RICSIL) 054, “Core Support Shroud
Crack Indications”, on 3 October 1990, to all owners of GE BWRs. The RICSIL summarised the cracking
found in the overseas reactor and recommended that at the next refuelling outage plants with high-carbontype 304 stainless steel shrouds perform a visual examination of the accessible areas of the seam welds andassociated HAZ on the inside and outside surfaces of the shroud. In 1994 the US NRC issued Generic
Letter 94-03 [40] to request that each BWR licensee: (1) inspect the core shrouds in their BWR plants nolater than the next scheduled refuelling outage, and perform an appropriate evaluation and/or repair basedon the results of the inspection; and (2) perform a safety analysis supporting continued operation of the
facility until inspections are conducted. Furthermore, this Generic Letter requested that each BWR licensee
develop an inspection plan which addresses: (a) all shroud welds (from support attachments to the vessel tothe top of the shroud) and/or provides a justification for elimination of particular welds from consideration;and (b) examination methods with appropriate consideration given to use of the best available technology
and industry inspection experience (e.g. enhanced VT-1 visual inspections, optimised UT techniques).
Standard methods for inspection of core support structures as specified by the ASME Code, Section XI,
have been shown to be inadequate for consistent detection of Intergranular Stress Corrosion Cracking(IGSCC) in-core shrouds. In 1994, US BWR-owning utilities formed the BWRVIP [41], which is charteredto support a programme addressing the problems of reactor internals, internal attachments, vessel welds,
and vessel nozzles. Most international BWR owners have become members of the BWRVIP.
Based on PSA results, for example internal flooding PSA results, many plants have implemented
NDE Programme for plant systems such as the Circulating Water System, Fire Protection Water Systemand service water system. Under these plant-specific RIM Programme, certain sections of these piping
sections undergo periodic 100% volumetric examinations using UT technique or other types of NDE
techniques. As defined by ASME Code Case N-460, “essentially 100%” means greater than 90% of theexamination volume of each weld where reduction in coverage is due to interference by another component
or part geometry [42].
2.9Code Case N-729-1

ASME Code Case N-729-1 (2006), “Alternate Examination Requirements for PWR Reactor Vessel Upper
Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds” [43] was approved for addition
into the code on March 28, 2006. A majority of this code case deals with the type of inspections to be
performed, i.e. volumetric or surface inspections. The basis for these examination requirements wasdeveloped as part of an industry programme conducted by the Materials Reliability Programme (MRP)
through EPRI. The results of this programme were published in MRP-95 Rev. 1 [44] and document a set of29

NEA/CSNI/R(2017)3
finite element weld residual stress analyses conducted on a variety of upper head penetration nozzles. The
inspection zone selected by the industry was based on the stress where it was assumed that primary water
stress corrosion cracking (PWSCC) would not initiate. As explained in MRP-95 Rev. 1, it has been
illustrated that PWSCC does not occur in the Alloy-600 tube when the stresses are below the yield strengthof that tube.
The Code Case N-729-1 inspection plan for RPV upper heads with Alloy 600/182/82 penetration
nozzles and welds requires periodic bare metal visual (BMV) examinations and periodic nonvisual
examinations using UT, ECT, or dye penetrant testing of the penetration nozzle base metal. BMV
examinations are performed in order to provide indication of any primary coolant leakage based on thepresence of boric acid deposit accumulations. Nonvisual examinations are performed in order to identify
flaws which could lead to leakage or failure of the penetration nozzle.
Such inspections are also required to be performed for RPV upper heads with Alloy 690/152/52
penetration nozzles and welds, but the frequency of inspection is reduced. This reduction is due to theenhanced resistance of these materials against PWSCC.


According to the NRC, Code Case N-729-1 provides an acceptable level of quality and safety as
a long-term inspection plan in lieu of the requirements of the Order7 , subject to the following
limitations and conditions:

Item B4.40 of Code Case N-729-1, Table 1, shall be inspected at least every fourth refuellingoutage or at least every seven calendar years, whichever occurs first, after the first ten-year
inspection interval.



If flaws attributed to PWSCC have been detected, whether acceptable or not for continued
service under Paragraphs – 3130 or – 3140, the re-inspection interval shall be each outage insteadof the re-inspection intervals required by Table 1 of the Code Case.



Instead of the specified ‘examination method’ requirements for volumetric and surface
examinations of Note 6 in Table 1 of the Code Case, the licensee shall perform volumetric and/or
surface examination of essentially 100% of the required volume or equivalent surfaces of thenozzle tube, as identified by Fig. 2 of ASME Code Case N-729-1 [43]. A surface examination
shall be performed on all J-groove welds. If a surface examination is being substituted for a
volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove
weld (Point E on Fig. 2 of ASME Code Case N-729-1 [43]), the surface examination shall be of
the inside and outside wetted surface of the penetration nozzle not examined volumetrically.

7. For details, refer to www.nrc.gov/docs/ML0622/ML062220594.pdf

30

NEA/CSNI/R(2017)3


Appendix 1 of ASME Code Case N-729-1 shall not be implemented without prior NRC approval.



After September 2009, UT-examinations shall be performed using personnel, procedures andequipment that have been qualified by blind demonstration on representative mockups using amethodology that meets the conditions specified below instead of the qualification requirementsof Paragraph – 2500 of ASME Code Case N-729-1:-

The specimen set shall have pipe diameters within ½ in. (13 mm) of the nominal diameter ofthe qualification pipe size and a thickness tolerance of +25%, -40% of the nominal throughwall depth of the qualification pipe thickness. The specimen set shall include geometric andmaterial conditions that normally require discrimination from PWSCC flaws.

-

The specimen set shall have a minimum of ten (10) flaws which provide an acoustic response
similar to PWSCC indications. All flaws shall be greater than 10% of the nominal pipe wall
thickness. A minimum number of 20% of the total flaws shall be outside diameter initiatedand 30% of the total flaws shall be inside diameter initiated. Further, at least 20% of the total
flaws are at each of the depth ranges of 10%-30% and 31%-50% from the inside or outside
diameter, as applicable. At least 30% and no more than 60% of the flaws shall be oriented
axially.

-

Procedures shall identify the equipment and essential variable settings used for thequalification. An essential variable is any variable that has an effect on the results of theexamination. The procedure shall be requalified when an essential variable is changed outside
the demonstration range. Procedure qualification shall include the equivalent of at least three
personnel performance demonstration test sets. Procedure qualification requires at least one
successful personnel performance demonstration.

-

Personnel performance demonstration test acceptance criteria will meet the personnel
performance demonstration detection test acceptance criteria of Table VIII-S10-1 of Section
XI, Appendix VIII, Supplement 10. Examination procedures, equipment, and personnel are
qualified for depth sizing and length sizing when the RMS error of the flaw depth
measurements, as compared to the true flaw depths, do not exceed 1/32-inch (0.8 mm) and theRMS error of the flaw length measurements, as compared to the true flaw lengths, do not
exceed 1/16-inch (1.6 mm), respectfully.

2.10 Code Case N-770-1
At the request of the NRC, ASME developed Code Case N-770-1 [45], which provides inspection
requirements to address PWSCC in Class 1 butt welds containing Alloy 82/182. Code Case N-770-1 has
requirements for inspection of unmitigated as well as mitigated Alloy 82/182 RCS butt welds. As such,specific inspection requirements for welds mitigated by weld overlays (WOLs) are contained in the CodeCase. The NRC incorporated ASME Code Case N-770-1 by reference into §50.55a (76 FR 36232,
p. 36278) in June 2011 [46].
In the US operators of PWRs with dissimilar metal nozzle safe-end welds, which are susceptible to
PWSCC, are required to examine those welds with increased frequency unless mitigating actions are taken.
On 26 January 2009, ASME published Code Case N-770, “Alternative Examination Requirements andAcceptance 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
Section XI, Division 1.” On 25 December 2009, ASME modified ASME Code Case N-770 and issued
version 1 of the code case as N-770-1. Requirements for these examinations are specified in American
Society of Mechanical Engineers (ASME) Section XI, Code Case N-770-1 as modified by 10CFR50.55a.
In June 2011, the US NRC issued rulemaking (76 FR 36278) that implemented ASME CodeCase N-770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping
31

NEA/CSNI/R(2017)3
and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With
or Without the Application of Listed Mitigation Activities, Section XI, Division 1,”
10CFR50.55a(g)(6)(ii)(F)(1), effective date 22 August 2011, states "licensees of existing, operating PWRs
as of 21 July 2011, shall implement the requirements of ASME Code Case N-770-1, subject to theconditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) of this section, by the firstrefuelling outage after 22 August 2011."
According to the Code Case, welds in Inspection Items A-1, A-2, and B are required to be examinedby visual examination and UT. Inspection Item A-1 welds, hot leg welds with an operating temperature
greater than 325°C (i.e. pressuriser nozzle DMWs), are required to be examined visually each refuellingoutage and volumetrically every second refuelling outage. Inspection Item A-2 welds, hot leg welds with
an operating temperature less than or equal to 325°C, are required to be examined visually each refuellingoutage and volumetrically every 5 years. Inspection Item B welds, cold leg welds with an operating
temperature greater than or equal to 274°C and less than or equal to 304°C, are required to be examinedvisually once per interval and volumetrically every second inspection period not to exceed 7 years. For
certain components 100% volumetric coverage is required. Examples of such components include thepressuriser surge line nozzle and the reactor coolant system cold leg and hot leg nozzles. The Code Case
770-1 inspection requirements are summarised in Table 1 below.

32

1

NEA/CSNI/R(2017)3

Table 1: Code Case N-770-1 Inspection Requirements Summary

Mitigated Alloy 82/182 Butt Welds

Unmitigated Alloy
82/182 Butt Welds

Inspection Item
A-1
A-2
B

Inspection Location
Butt welds at Hot Leg operating
temperature > 329°C
Butt welds at Hot Leg operating temperature≤ 329°C
Butt welds at cold leg operating temperature≥ 274°C and < 304°C

C

Uncracked butt welds reinforced by a fullstructural weld overlay (FSWOL) of Alloy
52/152 materialD

Uncracked butt welds mitigated bymechanical stress improvement process(MSIP®)E

Cracked butt welds mitigated by MSIP®

F

Cracked butt welds reinforced mitigated by
a FSWOL of Alloy 52/152 material

Extent and Frequency of Examination
BMV examination each refuelling outage. Volumetric examination every second
refuelling outage.
BMV examination each refuelling outage. Volumetric examination every five
years.
BMV examination once per interval. Essentially 100% volumetric examination foraxial and circumferential flaws in accordance with the applicable requirements of
ASME Section XI, Appendix VIII, every second inspection period not to exceed 7
years. Baseline examinations shall be completed by the end of the next refuellingoutage after 20 January 2012
If volumetric examination is not performed on the weld prior to FSWOL, the weld
shall be assumed cracked and shall be classified Inspection Item F. If cracking is
not observed during post-FSWOL preservice volumetric examination performed
from the outside surface of the overlay, axial and circumferential cracks at least
75% through the original wall thickness are required to be assumed. Inspection
Item C welds are required to be placed in a population to be examined on a 25%
sample basis. A once-per-interval inspection frequency applies to this sample.
No visual inspection requirements. Volumetric examination shall be performed
within 10 years following stress improvement. If the examination volumes of thewelds show no indication of cracking, the welds shall be placed into a population
to be examined on a sample basis. Twenty-five percent of this population shall be
examined once each interval.
No visual inspection requirements. Volumetric examination shall be performed
once during the first or second refuelling outage following stress improvement. If
the examination volumes of the welds show no indication of crack growth or new
cracking, the welds shall be placed into a population to be examined once eachinterval
Inspection Item C welds are required to be placed in a population to be examinedon a 25% sample basis. A once-per-interval inspection frequency applies to thissample. The FSWOL technique results in effective mitigation and that inspectionof FSWOLs serves a defence-in-depth monitoring function rather than adegradation management function

1. For full details on the N-770-1 inspection requirements and NRC conditions for its use, go to www.nrc.gov/reading-rm/doc-collections/cfr/part050/part0500055a.html

33

NEA/CSNI/R(2017)3
Inspection Item
G
H

J
K

Inspection Location
Uncracked butt weld mitigated with an inlayof Alloy 52 or Alloy 152 material
Uncracked butt welds mitigated by a weld
onlay constructed of Alloy 52 material

Cracked butt welds mitigated with an inlayof Alloy 52 or Alloy 152 material
Cracked butt welds mitigated by a weld
onlay constructed of Alloy 52 material

Extent and Frequency of Examination
A volumetric examination shall be performed immediately before application of
inlays or onlays and after application as a preservice baseline inspection.
Volumetric and surface examinations shall be performed no sooner than the shorterof 10 years following the application of the inlay or onlay and the design life of theinlay or onlay. Examination volumes that show no indication of cracking are to beplaced into a population to be examined on a sample basis.
Volumetric and surface examinations shall be performed once during the first or
second refuelling outage following application of the inlay or onlay. Examination
volumes that show no indication of cracking are to be placed into a population to
be examined on a sample basis.

34

NEA/CSNI/R(2017)3
2.11 NDE of small-diameter piping
This section compares and contrasts certain aspects of the French, Korean and US industry practices withrespect to socket weld RIM. In 2002 the French regulatory authority, Autorité de sûreté nucléaire (ASN)
issued a directive concerning socket weld integrity (DGSNR-BCCN/OT/VF No. 020406 [47]). According
to this directive, socket welds not meeting the requirements for weld dimensions and/or weld integrity as
specified by the RCC-M Code [48] “…shall be replaced with butt welds.” In response to the directive by
ASN, EDF implemented systematic socket weld integrity programme for all of its Class 900 MWe PWR
plants [49].
According to Table IWB-2500-1 of ASME XI, an external surface examination of small-bore Class 1
piping should be included for piping less than DN100. Other ASME Code provisions are exempt fromexamination piping of size DN25 and smaller. This programme is augmented to include piping from DN25
to less than DN100. Also, Examination Category B-P requires system leakage test of all Class 1 piping.
The utilities applying for licence renewal have made commitments to address the requirements of the
GALL Report [49]. In the absence of an established US industry standard for NDE of small-diameter
piping, the utilities have implemented their own technical approaches that include the use of phased arrayand conventional shear wave ultrasonic NDE techniques. An overview of selected US NDE practices
applicable to small-bore piping is documented in Reference [50].
In the US, NUREG-1801 (the “GALL Report”) [50] provides a technical basis for determining theadequacy of ageing management programme (AMPs) for licence renewal. Section XI.M35 of NUREG1801 augments the requirements in ASME Section XI, 2004 Edition (Rules for ISI of Nuclear Power Plant
Components). According to NUREG 1801, Gall Revision 2, AMP X1.M35 specifies a one-time inspection
to detect cracking in socket welds; the inspection should be either a volumetric or opportunistic destructive
examination [51]. Opportunistic destructive examination is performed when a weld is removed fromservice for other considerations, such as plant modifications. A sampling basis is used if more than 1 weld
is removed. These examinations provide additional assurance that either ageing of small-bore ASME CodeClass 1 piping is not occurring or the ageing is insignificant, such that a plant-specific AMP is notwarranted.
This programme is applicable to systems that have not experienced cracking of ASME Code Class 1
small-bore piping. This programme can also be used for systems that experienced cracking but have
implemented design changes to effectively mitigate cracking. (Measure of effectiveness includes (1) the
one-time inspection sampling is statistically significant; (2) samples will be selected as described in
Element 5, Monitoring and Trending below; and (3) no repeated failures over an extended period of time.)
For systems that have experienced cracking and operating experience indicates that design changes havenot been implemented to effectively mitigate cracking, periodic inspection is proposed, as managed by aplant-specific AMP. Should evidence of cracking be revealed by a one-time inspection, periodic inspection
is implemented using a plant-specific AMP. If small-bore piping in a particular plant system has
experienced cracking, small-bore piping in all plant systems are evaluated to determine whether the cause
for the cracking affects other systems.
In Korea there is an augmented inspection programme concerned with small-bore piping socket
welds. In 2008, RCPB leakage was reported at a socket weld of 3/4 inch drain line from a steam generator(SG) bowl to a SG drain valve. The root cause of the leak was determined as an internal manufacturingcrack which initiated at the weld root and subsequently had propagated during operation. In the results ofadditional examinations extended to similar socket welds (a total of 37 welds), nine more welds wereidentified to have abnormal indications which were mostly attributed to lack of penetration. The Korea
regulatory body issued an administrative order to request the licensee inspect all class 1 and 2 socket weldsof small-bore piping (less than 2-inch) with volumetric examination (especially radiographic testing) at theearliest next refuelling outage. In response to that, the licensee performed the requested inspections for all

35

NEA/CSNI/R(2017)3
safety-related socket welds and then, developed a small-bore piping (less than 2-inch) socket weld
management programme [52].
The small-bore piping socket weld management Programme was developed based on vibration
measurements at socket welds within the RCPB because the failure mechanism of socket welds is mostly
attributed to service-induced high-cycle fatigue. By considering the measured level of vibration andimportance of welds, all socket welds in the RCPB were categorised into three groups: welds with high,medium, and low risk significance. For the socket welds with high risk significance, both surface andvolumetric examination should be performed at every refuelling outage. Since surface examination has notbeen enough for the earlier detection of flaws that mostly initiated inside the socket weld, volumetricexamination was additionally required. The conventional manual UT, however, has limitations due to thelimited accessibility to the socket weld and difficulties with contact between the ultrasonic probe and theout diameter surface of small-bore piping. In order to overcome the limitations, the Korea licensee
developed phased array UT system and inspection procedures for small-bore piping and socket welds. The
phased array UT has been performed since 2003 under the socket weld management Programme [52].

36

NEA/CSNI/R(2017)3

3. RIM INFLUENCES IN CODAP

The CODAP Event Database is a web-based SQL database (i.e. relational database) consisting of
ca. 100 uniquely defined data fields for each database record. It is a mixture of free-format fields fordetailed narrative information, fields defined by drop-down menus with key words (or data filters) or
related tables, and hyperlinks to additional background information (e.g. photographs, root cause
evaluation reports). The related tables include information on material, location of damage or degradation,
type of damage or degradation, system name, safety class, etc. The “Online Version” facilitates data input,
search and query routines and data export to a local computer. On a local computer, the database can beconverted into a Microsoft® Access database format or any other user-defined format. In this section, allfigures are based on the Microsoft® Access database version of the database.
3.1 Database structure and coding guideline
The CODAP event database is populated by the National Coordinators (NCs) of the member countries. In
accordance with the Operating Procedures [53] and QA Plan [54], validation of data submissions is
performed by the respective NC and the Operating Agent. To achieve the objectives established for the
CODAP event database a coding format has been developed. This coding format is reflected in the Coding
Guideline (CG) [55]. The Coding Guideline builds on established pipe failure data analysis practices androutines that acknowledge the unique aspects of passive component reliability in heavy water reactor andlight water reactor operating environments (e.g. influences by material and water chemistry).
Data quality is affected from the moment the field experience data is recorded at a nuclear power
plant, interpreted, and finally entered into a database system. The field experience data is recorded indifferent types of information systems ranging from action requests, work order systems, ISI databases,
outage summary reports, licensee event reports, and reportable occurrence reports. Consequently, thedetails of a degradation event or failure event tend to be documented to various levels of technical detail inthese different information systems. Building a CODAP event database record containing the full event
history often entails extracting information from multiple sources.
The term “data quality” is an attribute of the processes that have been implemented to ensure that any
given database record (including all of its constituent elements, or database fields) can be traced to thesource information. The term also encompasses “fitness-for-use”, that is, the database records should
contain sufficient technical detail to support database applications.
As one of several-steps to ensure data integrity, all relevant source data are retained within the
database. As one example, the narrative portions of the database retain all of the original event
descriptions. Furthermore, a provision exists to attach supporting documents to each event record insupport of independent validation of event classifications.
In CODAP, a Completeness Index (CI) is used for database management purposes. It distinguishes
between records for which more information must be sought and those considered to be complete. Each
record in the database is assigned a CI, which relates to the completeness and comprehensiveness of theinformation in the database relative to the requirements of the Coding Guideline.

37

NEA/CSNI/R(2017)3
The database structure consists of a single data entry form. The data entry form is organised to capture
essential passive component failure information together with supporting information. The data entry form
consists of four areas:
General failure data: This area represents the minimum required information to characterise an event
and it includes an event narrative together with details on the affected component (e.g. diameter and wall
thickness) and system, impact on plant operation, observed through-wall leak rate, and Code Class.
11. Flaw size information: This area is for recording flaw size (depth, length, and aspect ratio),orientation, location of the flaw (e.g. within weld metal, weld heat affected zone or base metal),and number of flaws within a specified component boundary. For multiple flaws within a
specified component boundary, the distance between respective flaws is indicated.
12. In-service inspection (ISI) information: This area is used to record any relevant information about
ISI performed in the past (e.g. date of most recent inspection). Also documented here isinformation regarding ISI Programme weaknesses or failures. Included in this field are results ofexpanded NDEs in response to the identification of a rejectable flaw or to an identified throughwall flaw. Information on component-specific NDE qualification is also included in this databasearea.
13. Root cause information: This area records factors or conditions contributing to a degradedcondition. Also included in this area is a field for free-format comments on corrective actions, orother information of relevance to a specific event. The method and technique of flaw
identification and sizing is recorded in this area of the database.
As one example of validation of data integrity, the Operating Agent on a regular basis performs abroad set of database queries to identify records with missing information or records that contain errors inspecific database fields. When errors are identified, the Online Version of the database, via its Workflow
Area enables a QA Communication directly between the Operating Agent and the National Coordinator.
3.2 Method of detecting degradation
The CODAP event database differentiates between the method of flaw detection and the technique of flaw
detection. The former relates to programme used to detect the flaw while the latter relates to the specificprocess or technology used to detect subsurface and surface flaws. In the online event database, drop-downmenus include lists of flaw detection methods and techniques. The user of the database may makeadditions to the pre-formatted drop-down menus.
Illustrated in Figure 1 is an example of how method of flaw detection is classified in the database. The
given example is concerned with the discovery in 1994 of a degraded pressurised water reactor (PWR)
pressure vessel head penetration during a scheduled bare metal visual examination (BMV).
Illustrated in Figure 2 is an example of how technique of flaw detection is classified in the database.
The given example is concerned with the discovery in 2013 of a degraded PWR pressure vessel head
penetration during a scheduled BMV examination using Time-of-Flight Diffraction (TOFD)
UT technology.

38

NEA/CSNI/R(2017)3

Figure 1: Method of flaw detection

39

NEA/CSNI/R(2017)3

Figure 2: Technique of flaw detection

40

NEA/CSNI/R(2017)3
3.3

High-level CODAP database summary

A high-level database summary by method-of-detection and plant operational state is given in Figures 3
through 8. The pipe failure database content is summarised by safety class and plant operational state in
Figures 3 through 7. The plant operational state refers to the plant status at the time of the discovery of adegraded or failed passive metallic component. Summarised in Figure 8 is the database content withrespect to BWR core shroud weld flaws and PWR reactor pressure vessel (RPV) head penetration (VHP)
flaws as a function of method and technique of flaw detection. All charts are based on database queries andthe following nomenclature is used:Control room indication. This term applies to any remote sensing and annunciation of sumpin-leakage indicating an active pressure boundary leak whether it is in the containment,
Drywell, Auxiliary Building, Reactor Building or any other area of a nuclear power plant.
Methods for leak detection include [56]:- humidity sensors at specific locations;- online surveillance via video camera;- containment atmosphere humidity, pressure, temperature sensing;- airborne gaseous radioactivity;- level or flow into sumps or tanks.



Ultrasonic testing examination. This term applies to all UT examination technologies,
including encoded and non-encoded scanning [57].

Visual inspection is used for any informal or formalised visual inspection (e.g. per ASME XI)of a pressure boundary for leakage. The inspection may be performed while at-power or duringa scheduled maintenance or refuelling outage [58].

WOR/IHSI. In CODAP this term applies when a through-wall defect is identified while
performing weld preparation before applying a weld overlay (WOR). The term also applies toevent involving the discovery of a through-wall defect while applying an induction heat stress
improvement (IHSI) process.

Part through-wall flaw. Any Code rejectable flaw indication that results in a repair or
replacement. This term also applies to carbon steel piping for which the measured wall
thickness during inspection cannot meet 87.5% of the nominal wall thickness specified in the
ASME Code Case N-480.

Pressure boundary failure:- Small Leak. The measured through-wall flow rate is ≤ 1 gpm.
- Leak. The measured though-wall flow rate (ν) is 1 < ν ≤ 5 gpm.
- Large Leak. The measured though-wall flow rate (FR) is 5 < ν ≤ 50 gpm.
- Very Large Leak. 50 < ν ≤ 500 gpm.
- Rupture corresponds to a major structural failure up to and including a double-ended
guillotine pipe break (DEGB).



Pressure testing. Detection of a latent failure through application of a pressure test process.

The event data sets that are summarised in Figures 3 through 8 consist of self-revealing surface andflaws and sub-surface flaws. Summarised in Figure 7 is the CODAP data set on fire protection water
system pipe failures. It is limited to the US service experience. In the US, NDE of this particular class of41

NEA/CSNI/R(2017)3
piping is governed by the National Fire Protection Association (NFPA) Standard NFPA-25 [59]. In
addition, many plant owners have implemented owner-defined NDE Programme that are based on risk
insights from plant-specific internal flooding PSA studies.
Figures 3 through 8 were generated by invoking cross-tab queries. Respective vertical axis indicates
pre-defined keywords in the database drop-down menus, and the horizontal axis represents event counts.
The displayed query results represent the results of an initial step to organise the database for more indepth analysis, if so required. In CODAP, for an event to be attributed to LTA-RIM, the root cause
analysis must positively refer to the causal factors that contributed to a RIM performance deficiency.
Examples of such causal factors include the following:

A volumetric examination of the failed component was performed during the last scheduled ISI
but failed to detect and/or record a pre-existing flaw that, based on a subsequent metallographic
examination must have been present at the time of the ISI.



Application of too coarse UT-grid fails to identify a thinned area prior to failure.

Additional examples are given in Section 4.2.

The through-wall leaks are detected by various types of sensors, by plant personnel performing
routine walk down inspections, or various types of leak or pressure tests. The subsurface flaws are detectedby the application of various types of non-destructive examination (NDE) techniques. Stress corrosion andfatigue failure mechanisms involve incubation, crack initiation and crack growth. The objective of NDE isto detect flaws before crack growth exceeds a pre-determined value. Leak events in the database thatinvolve safety-related components could be interpreted as being manifestations of LTA-RIM. For example,
ISI performed during past inspection intervals could potentially have missed detecting a pre-existing flaw.
On the other hand, crack initiation could occur between intervals.
The CODAP event database selectively tracks ISI histories. For detailed ISI histories to be developed
it is necessary to research past ISI summary reports to determine when a certain weld flaw was first
identified and whether crack growth beyond a certain wall depth could have been prevented. It is aformidable task to establish complete and comprehensive ISI histories for all events in the database and iswell beyond the current scope of the PRG work plan. Nevertheless, where information on failures of leakdetection or LTA-RIM is readily available, for example, via root cause analysis reports, then that
information is recorded in the database. In the current version of the database about 4.6% of all events haveenough information to establish that they are attributed to LTA-RIM to properly detect a flaw before
exceeding some pre-determined value. One way of interpreting this database insight would be to say thatan upper bound (95th percentile) probability of detection is on the order of 95% based on field experience
data alone. While outside the scope of work for this topical report, further data processing is required toreach more in-depth qualitative and quantitative insights about RIM reliability. Two examples of how
LTA-RIM is recorded in CODAP are given in Figures 9 and 10:
Figure 9. During an ISI of a dissimilar metal weld in a reactor coolant system hot leg inlet to a
SG nozzle at a US PWR, several axially oriented flaws went undetected by the manual
conventional ultrasonic testing (UT) technique. The flaws were subsequently detected as aresult of outside diameter (OD) surface machining in preparation for a weld overlay.

Figure 10. During normal power operation a US PWR experienced a moisture separator reheaterpipe rupture. The ruptured pipe was included in the owner’s FAC Programme. However, anincorrect orifice opening size had been entered into the FAC prediction software, which caused
the software to incorrectly calculate the pipe wall wear rate. Based on the calculated results thefailed pipe had not been subject to any volumetric examination.

42

NEA/CSNI/R(2017)3
Power Operation

Walkdown Inspection
Visual Inspection
Video Camera
UT-examination
Surveillance Testing
Control Room Indication

HSB / HSD

Walkdown Inspection
Visual Inspection
Surveillance Testing
Pressure Testing / Leak Test
Control Room Indication
Walkdown Inspection
Visual Inspection
UT-examination
Surveillance Testing

CSD

Routine MaintenanceRadiographic Testing
Pressure Testing / Leak Test
Pre-emptive WOR /IHSI
Liquid Penetrant Testing
Eddy Current Testing
Control Room Indication
0.1

1

10

100

1000

10000

No. Failure Records
RuptureVery Large Leak

Large Leak

Leak

Small Leak

Part Through-Wall
1

Figure 3: Code class 1 pipe failures by plant operational state and method of flaw detection

1. HSB = Hot Standby; HSD = Hot Shutdown

43

CSD

HSB / HSD

Power Operation

NEA/CSNI/R(2017)3
Walkdown Inspection
Visual Inspection
Video Camera
UT-examination
Surveillance TestingRoutine MaintenanceRadiographic Testing
Pressure Testing - Liquid
Liquid Penetrant Testing
Hydrotesting
Control Room Indication
Walkdown Inspection
Visual Inspection
UT-examination
Surveillance TestingRoutine MaintenanceRadiographic Testing
Pressure Test
Leak Test
Control Room Indication
Walkdown Inspection
Visual Inspection
UT-examination
Surveillance TestingRoutine MaintenanceRadiographic Testing
Pressure Test
Liquid Penetrant Testing
Leak Test
Eddy Current Testing
Control Room Indication
0.1

1

10

100

No. Failure Records
RuptureVery Large Leak

Large Leak

Leak

Small Leak

Part Through-Wall

Figure 4: Code class 2 pipe failures by plant operational state and method of flaw detection

44

1000

CSD

HSB / HSD

Power Operation

NEA/CSNI/R(2017)3
Weld Preparation for Overlay
Walkdown Inspection
Visual Inspection
UT-examination
Surveillance TestingRoutine MaintenanceRadiographic Testing
Maintenance - Freeze Seal
Liquid Penetrant Testing
Pressure Test
Groundwater SamplingFire Detection Alarm
Control Room Indication
Walkdown Inspection
UT-examination
Surveillance Testing
Pressure Test
Walkdown Inspection
Visual Inspection
UT-TOFDUT-examination
Surveillance Testing
Remote Robotic Crawler
Radiographic Testing
Maintenance - Freeze Seal
Liquid Penetrant Testing
Pressure Test
Eddy Current Testing
Control Room Indication
0.1

1

10

100

1000

No. Failure Records
RuptureVery Large Leak

Large Leak

Leak

Small Leak

Part Through-Wall

Figure 5: Code class 3 pipe failures by plant operational state and method of flaw detection

45

10000

NEA/CSNI/R(2017)3
Weld Preparation for Overlay
Walkdown Inspection
Visual Inspection

Power Operation

UT-examination
Surveillance TestingRoutine MaintenanceRadiographic Testing
Maintenance - Freeze Seal
Liquid Penetrant Testing
Leak-Inspection
Leak-detection
Hydrotesting
Control Room Indication

Weld Preparation for Overlay
Walkdown Inspection
Visual Inspection
UT-examination

CSD

Surveillance TestingRoutine MaintenanceRadiographic Testing
Liquid Penetrant Testing
Hydrotesting
Eddy Current Testing
Control Room Indication
0.1

1

10

100

1000

No. Failure Records
RuptureVery Large Leak

Large Leak

Leak

Small Leak

Part Through-Wall

Figure 6: Non-code pipe failures by plant operational state and method of flaw detection

46

10000

NEA/CSNI/R(2017)3
Walkdown Inspection

Power Operation

Visual Inspection
Video Camera

UT-examination
Surveillance Testing
Control Room Indication
Boroscopic Examination
Walkdown Inspection

CSD

Visual Inspection
UT-examination
Surveillance TestingRadiographic Testing
0

20

40

60

80

100

No. Failure Records
RuptureVery Large Leak

Large Leak

Leak

Small Leak

Part Through-Wall

Figure 7: Fire protection water system pipe failures by plant operational state and method of flaw detection

47

120

NEA/CSNI/R(2017)3

UT-Examination

135

6

BMV

VHP

Liquid Penetrant Testing

Hydrostatic Pressure Test

1

IVVI

Core Shroud Weld

Eddy Current Testing

9

UT-Examination

50

0

20

40

60

80
100
No. Database Records

120

140

160

Figure 8: Core shroud weld cracks and vessel head penetration cracks by inspection method and technique48

NEA/CSNI/R(2017)3

Figure 9: An example of failure of UT examination to detect a flaw

49

NEA/CSNI/R(2017)3

Figure 10: An example of FAC programme deficiency

50

NEA/CSNI/R(2017)3

4. FAILURE OF RIM TO DETECT FLAWSThe CODAP event database captures events that are attributed to less-than-adequate (LTA)
implementation of RIM. Mainly, the instances of LTA-RIM have involved deficiencies in an inspection
process to detect pre-existing flaws before exceeding acceptance criteria. In the database passive
component failure information is recorded in a tiered manner. First, basic failure information is recorded toaddress the most fundamental information about an event and this includes a free-format event narrativethat describes the sequence of events, including plant response, consequence, in-plant location of failed
component, dimensional data, and component type. This is followed by recording the known ISI history,
including date when the failed component was last inspected, method of NDE qualification if a qualified
method had been used, and any NDE performance deficiencies or failures. This section summarises
insights about RIM deficiencies as recorded in the CODAP event database. Also included are examples ofrecent RIM deficiencies, such as failure to select the proper NDE locations, inadequate implementation of
NDE procedures, and failures in calibration and using the NDE tools. The event histories include thefindings of associated root cause analyses as recorded in CODAP.
4.1

An overview of RIM failures in CODAP

The ASME Boiler and Pressure Vessel Code Section XI, “Rules for In-service Inspection of Nuclear
Power Plant Components” [9] provides requirements for examination, testing, and inspection ofcomponents and systems, and repair/replacement activities in a nuclear power plant. The mandatory
Appendix VIII of ASME Section XI provides requirements for performance demonstration (PD) for
ultrasonic examination procedures, equipment, and personnel used to detect and size flaws. The ultrasonicinspection techniques that were in use in the 1970s had been adapted from the manufacturing industry andmodified using lessons learnt from fatigue cracking service experience. The ultrasonic examination rules
were prescriptive and were not based on the degradation mechanisms. The non-destructive examinations(NDE) were not quantified via testing. The early field experience showed that improvements in inspection
requirements were needed. Therefore, the NRC sponsored R&D to assess non-destructive examination(NDE) reliability. The results indicated that a performance-based testing approach would be the mosteffective means for achieving the needed improvements in reliability [10].
As already stated, the CODAP event database captures instances of LTA-RIM, that is events where
pre-existing flaws have not been identified before exceeding acceptance criteria. In the database passive
component failure information is recorded in a tiered manner. First, basic failure information is recorded toaddress the most fundamental information about an event and this includes a free-format event narrativethat describes the sequence of events, including plant response, consequence, in-plant location of failed
component, dimensional data, and component type. This is followed by recording the known ISI history,
including date when the failed component was last inspected, method of NDE qualification if a qualified
method had been used, and any NDE performance deficiencies or failures. Finally, details about the service
environment (e.g. water chemistry, stresses, pressure and temperature) are recorded as a lead-in to details
from root cause evaluations (flaw data, chemical composition of material, results of metallographic
examinations, apparent and underlying causes of material degradation).When RIM fails to detect flaws one
or more of the following factors are often present:51

NEA/CSNI/R(2017)3
Accepting a rejectable flaw indication for continued operation. This could be due tomisinterpretation of NDE results.

Misinterpreting flaw signals by qualified personnel due to complexity of the examination (forexample, when several embedded flaws are stacked together and appear to be connected).



Rationalising away detected defects.

Using an improperly qualified NDE technique.



Poorly implementing an owner-defined inspection programme.

Missing a flaw with a qualified procedure.

Allowing human factors to influence results.[60]- time pressure;- tedium;- perceptional (or in attentional) blindness.

The CODAP database content with respect to piping-related LTA-RIM events is summarised in
Figures 11 and 12. In Figure 11, the LTA-RIM data is organised by time period. The absolute number of
instances of LTA-RIM has remained largely the same over the past three decades. This observation is to becontrasted with the fact that the overall number of inspections have increased significantly over this time
period.
In Figure 12, the LTA-RIM data is organised by failure mechanism and mode of failure. The term
“flow-assisted degradation” includes the following degradation mechanisms: erosion-cavitation, erosioncorrosion and flow-accelerated corrosion.
Figure 13 is a summary of pipe failures that have been discovered through application of a formalpressure test process, including hydro testing. The through-wall flaws in Code Class 1 and components thathave been detected through a pressure test are manifestations of latent failures that were not previously
discovered through liquid penetrant testing or UT examination.

52

NEA/CSNI/R(2017)3

50

45

43
40

40

39

40

35

33

No. NDE Failures

30
27
25

20

15
11
10

5

0
Pre-1985

1985-89

1990-94

1995-99

2000-04

2005-092010-15

Figure 11: LTA-RIM events by time period

Weld DefectVibration-Fatigue

Thermal Fatigue

SCC

Hydraulic Transient

Low-Cycle Fatigue Pressure Loads

Flow-Assisted Degradation

Corrosion Mechanisms

Corrosion-Fatigue

0

10

20

30

40

50

60

70

No. NDE Failures
Part Through-Wall

Small Leak

Leak

Large Leak

Very Large Leak

RuptureFigure 12: LTA-RIM by failure mechanism and failure mode

53

80

90

NEA/CSNI/R(2017)3

60
55

No. Through-Wall Leaks Discovered through Pressure Test

50

40
35

30

20
17
15
13
10

0
≤ DN25

DN25 < Ø ≤ DN50

DN50 < Ø ≤ DN100

DN100 < Ø ≤ DN250

Ø > DN250

Nominal Diameter [mm]Figure 13: Through-wall class 1 pipe flaws discovered through pressure test

4.2

Selected (1999-2015) RIM failures

Despite the great progress that has been made with non-destructive examination, the detection andassessment of crack indications continue to pose challenges, some of which are directly related to the NDE
techniques, personnel and equipment deployed and some of which are not. Many of the reviewed events
that involve failure of RIM to prevent passive component material degradation to exceed acceptancecriteria are attributed to human factors issues. Examples of (1999-2015) RIM failures include the following
selected events1:
German PWR Unit, May 2015. At the conclusion of the annual refuelling outage plant
personnel performed a scheduled operability test of a block valve in the Main Steam pressure
relief system. The reactor was in hot shutdown at the time of the test. Shortly after the testcommenced control room personnel received high-temperature alarms and fire alarm indicating
a sudden steam release. Subsequent inspections noted a DN50 drain line off of one of the relief
had ruptured. During power operation the drain line is not pressurised. The root cause analysis
noted that the failed line had never been subjected to NDE.

US PWR Unit, May 2015. The control room operators heard the sounds of rushing air or steam
in the turbine building and began an investigation. The operators observed steam on the turbine
deck and began a controlled power decrease. At approximately 30% reactor power the operators
initiated a manual reactor trip in accordance with plant procedures. Once the plant was shut
down workers isolated the condenser and cooled the reactor using the atmospheric relief valves.
The cause of steam release into the turbine building was a ruptured 4-inch diameter pipe. The
pipe rupture caused the fire suppression system to actuate. The failed elbow in the 4-inch piping
was 10.25 inches downstream of a restricting orifice. This orifice, which restricts the amount of

1. These descriptions have been extracted from the CODAP event database.

54

NEA/CSNI/R(2017)3
scavenging steam drawn from the second stage reheater tube bundle outlet header, wasmodelled in the FAC software as having an opening 3.0 inches in diameter when vendor
drawings showed an orifice opening of 0.859 inches. The incorrect orifice opening size wasentered in the FAC software in the 1987 to 1989 time frame during initial FAC software
database development, and caused the FAC software to incorrectly calculate the wear rate of theelbow. Because of this incorrectly calculated wear rate, no wall thickness measurements were
performed directly on the failed elbow.


US PWR Unit, April 2014. While performing planned inspections of reactor coolant systempiping welds, two flaw indications were identified on a 1.5 inch High Pressure Safety Injection
Line connection to the reactor coolant system piping (cold leg). After confirmatory inspections
and evaluation, one of the two flaw indications was determined to not meet the acceptancecriteria specified in ASME Section XI. This flaw was missed during the previous refuellingoutage as a result of a probable skill-based human performance error. The flaw was detectedduring normal inspections required by the NDE augmented examination programme, which isdriven by EPRI MRP-146 ("Thermal Fatigue in Normally Stagnant Non-Isolable reactor coolant
system Branch Lines") [61]. Non-isolable branch lines connected to the Reactor Coolant System
are susceptible to high-cycle thermal fatigue if exposed to specific operational conditions andconfigurations. Examples of susceptible locations are horizontal lines where in-leakage past avalve is present and lines that see turbulent swirl penetration from adjacent piping flow. Initial
MRP-146 inspections were performed in 2008 and did not include the area where the flaw was
discovered.



US PWR Unit, November 2013. Following a controlled shutdown in response to a RCPB leak,visual inspection confirmed that the leak was located on a High Pressure Injection Line. The
root cause evaluation determined that a crack-like indication in the failed location had existed
since 2011. Furthermore, the root cause evaluation concluded that inadequate procedural
guidance existed for the conduct of augmented examinations and appropriate disposition of UT
examination results where conditions limited the weld volume that could be examined. A UTlimitation is defined as any obstruction or condition that limits the extent of angle beamscanning or limits the extent of required coverage using straight beam scanning. When adequate
weld volumes could not be examined on the nozzle safe end-to-pipe butt weld, no procedural
guidance provided weld volume acceptance criteria or directed these limitations to be enteredinto the corrective action programme for evaluation. During the root cause investigation,
metallurgical analysis documented that the crack propagated over several operating cycles.
Historical NDE data revealed that the crack was visible in existing radiographs. Had the failed
weld volume been adequately interrogated, the crack would have been identified before
propagating through wall. The licensee reviewed the results of the previous UT examination
performed in 2012 using a NDE procedure and found no reportable indications. However, in
2011, the licensee performed a radiographic examination specifically to check the condition andposition of a thermal sleeve. The focus of the review was limited to that area; however, the safeend area containing cracked weld was incidentally visible on the film. Following the current
event, the licensee re-reviewed the 2011 radiographic film and a crack-like indication was
identified in the side wall image of the weld at approximately the same location ascorresponding to the current crack location. From the re-review of the film, this crack-likeindication appeared to be approximately 50% through-wall.



US PWR Unit, March 2012. During an ISI of a dissimilar metal weld in an inlet (hot leg) SG
nozzle, several axially oriented flaws went undetected by the licensee’s manual conventional
ultrasonic testing (UT) technique. The flaws were subsequently detected as a result of outside
diameter (OD) surface machining in preparation for a full structural weld overlay. The
machining operation uncovered the existence of two through-wall flaws, based on the55

NEA/CSNI/R(2017)3
observance of primary coolant water leaking from the DMW. Further ultrasonic tests were thenperformed, and a total of five axially oriented flaws, classified as PWSCC, were detected invaried locations around the weld circumference. The particular hot leg safe end-to-nozzle weld
configuration has an approximate 11-degree OD taper from the thinner austenitic piping side, up
to the thicker carbon steel nozzle, and is typical of a DMW created during SG replacement at
Westinghouse-designed nuclear power plants. However, the level of OD taper exhibited by thisparticular design was not included as a blind performance demonstration mock-up used by the
PDI. In response to this event, the US industry established the “NDE Improvement Focus
Group” (NIFG) [62].
Korea PWR Unit, January 2011. During full power operation, an operator found large amountof steam leak due to a pipe rupture. The ruptured location was a drain line tee-end from the firststage of a moisture separator reheater (MSR) to a high pressure feed water heater. The ruptureof the tee-end was attributed to erosion due to liquid droplet impingement of two-phase flow.
Although the MSR system is subjected to the licensee’s FAC management programme, thefailed location had been exempted from the programme because it satisfied the exempt
condition of stagnant flow during normal operation. After the event, similar locations have been
included in the FAC management programme.

Korea PWR Unit, May 2011. Another unit also experienced a pipe rupture at a vent pipe of
MSR first stage reheater to a high pressure feed water heater during full power operation on
May 2011. The ruptured length was estimated about 15 cm on the straight line below a45-elbow. The ruptured location had not been subjected to the licensee’s FAC Programme,though an adjacent elbow has been inspected. Another cause of the failure was attributed to thechange of operating conditions due to the replacement of the MSR internals (May 1996) and thepower uprate (2009). The changes accelerated FAC at the failed location. Locations of FAC
inspection should have been re-determined by taking account of the changed operating
conditions.



German PWR Unit, December 2010. An automated ultrasonic ISI was performed on the nozzleof the reactor coolant system Hot Leg connecting to the surge line [63]. This examination
revealed a circumferential indication in the connecting area of the thermal sleeve at the upper
end of the radius at the safe end of the nozzle. The indication was detected by means of theultrasonic examination technique 45° ET2 (single-transducer method with transverse waves),
pointing to the outside wall of the safe end. The indication was fully circumferential; in thecircumferential region of approximately 330° to 60° beyond the 12 o'clock position, the echo
amplitude exceeded the recording limit by 5 dB in some cases.
- A verification measurement with the “52 SET” testing method (dual search unit; transmitterreceiver with transverse waves) showed the same indication structure, but the echo
amplitudes were completely below the recording limit. The indication depth was checked byrun-time analyses using a phased-array probe. The indication depth determined this way was
about 2.7 mm in scanning direction and about 2.2 mm in the opposite direction. All
ultrasonic testing methods have been qualified using a similar reference test block.
- Ultrasonic examinations were again carried out during the 2011 outage and the results werecompared with the previous examination results of 2010. The same examination technique
as in 2010 was applied. The results showed again a fully circumferential indication of thesame extent. However, the echo amplitude was 3 dB less compared with the 2010
measurements. The extent to which the recording threshold was exceeded also corresponded
to the extent of the 2010 examination. Using the same ultrasonic analysis method as in 2010,
the indication depth measured this way was again 2.7 mm.

56

NEA/CSNI/R(2017)3- Owing to the change identified in the amplitude of the indication, the examination technique
was to be reviewed and newly qualified. In addition, an examination technique fordetermining the amplitude of flaws was to be qualified [63]. Said qualification was
performed on a test body of similar geometry and material. Grooves were applied as flaws to
be used for comparison. The choice of grooves was guided by the relevant regulations andwas to ensure that the examination technique chosen was capable of detecting the different
crack depths and assess them with a certain degree of tolerance. In order to be able tosimulate branching and inclined cracks, too, twisted and tilted grooves were also included.
The smallest reflector size was 0.5 mm. The evaluation of the qualification measurements on
the test body showed that the examination technique chosen can detect all reflectors,
including inclined reflectors, with a sufficient signal-to-noise ratio. However, it was only
possible to determine the groove depth with sufficient accuracy from a depth of ≥ 2 mm. In
the case of smaller reflectors, the groove tip was no longer clearly detectable, so that it was
only possible to determine the extension of the depth of smaller reflectors by estimation.
- Subsequently, a re-assessment of the (real) indication was carried out with the newly
qualified analysis technique, and the indication was assessed as being 2 mm deep at themost. Since a possible crack growth could not be excluded, the safe end was removed, cut
and subjected to a metallographic examination. This examination revealed a minutecircumferential crack which had apparently formed through fatigue as a result of changing
thermal loads. Its location corresponded with the results of the ultrasonic examination. The
actual maximum depth of this minute crack was, however, clearly below the crack depth
determined by means of the ultrasonic examination. The maximum depth determined wasapprox. 0.3 mm in 0° position.
- A visual inspection revealed that the actual geometry of the connecting area of the thermal
sleeve (radius) deviated clearly from the specifications of the drawing. Instead of a radius of
2.5 mm, an almost rectangular groove cross-section was found, superposed by geometricdiscontinuities, turning grooves, and chatter marks from manufacture.
Korea PWR Unit, April 2007. While performing planned visual inspection and UT thickness
measurement of the secondary sides of three SGs, a through-wall flaw was identified on thefeed water ring connection to the J-nozzle No. 2 in both the SG A and B. In addition, somesignificant wall thinning was also found in the similar area in all three SGs. The flaws and wall
thinning were attributed to FAC due to the single phase flow of the feed water at the branches ofthe feed water ring to the J-nozzle. Based on the UT thickness measurement results, affectedparts of the feed water ring which did not meet the minimum thickness criterion were replaced.
Although the UT thickness measurement of the feed water ring has been performed inaccordance with the SG management programme, the failed locations (around feed water ring toJ-nozzle welds), however, had never been inspected by UT.



US BWR Unit, February 2007. During the refuelling outage NDE of dissimilar metal welds inthe Reactor Recirculation system found surface breaking cracks in two welds’ 55% and 74%
through-wall. The two welds had been previously inspected in 1999 using an un-qualified
inspection procedure. Evidence of flaws was present but not called out by the inspection
personnel. In 2005, the two weld had been re-inspected using a qualified procedure and theweld crowns were removed and no flaws were detected. Unacceptable contact was obtained in
the location of the flaw due to gravity. The flaws were on the bottom of a horizontal pipe.

Canadian CANDU Units (multiple instances). Feeder pipes are part of the CANDU (Canadian
Deuterium Uranium) PHT (Primary Heat Transport) system. Essential function of the feeder
pipe is to transports heavy water (D2O) coolant to and from the fuel channels and the inlet andoutlet headers in order to cool the fuel bundles in the pressure tubes. The feeder pipe is57

NEA/CSNI/R(2017)3
considered a Nuclear Class 1 piping component and governed by CSA N285.4 on periodic
inspection of CANDU nuclear power plant components because of containing fluid that directlytransports heat from nuclear fuel. Two types of cracking are considered as degradation
mechanisms in feeders: first one being cracking due to stress corrosion cracking (SCC) on thetight radius bend/elbow and repaired welds, and the other is cracking due to primary water
stress corrosion cracking (PWSCC) in Dissimilar Metal Welds (DMWs). In 1997, a throughwall crack was detected in the tight radius bend of outlet feeder S08 at CANDU Unit 1. After
this initial finding, bend cracking in 16 feeders were reliably detected by ultrasonic NDE andleak detection system up to the beginning of the refurbishment of CANDU Unit 1. However, in
2003, a through-wall crack in the repaired field weld of the feeder G09 at CANDU Unit 2 was
identified by leak detection system but not by NDE. It should be noted that susceptibility to
cracking on partially repaired welds in the Grayloc region would be higher than the otherrepaired regions (i.e. pipe to header welds and field welds) due to high local residual stressalong with high operating stress on the repaired Grayloc region. After cracks in these twonuclear power plant units were observed, many cracking inspection have been conducted as per
the inspection programme to capture crack indication on the tight radius bend/elbow, theGrayloc weld and repaired Grayloc weld. ISI results indicate that no further crack indications
have been reported since then. During the refurbishment of CANDU Unit 1, the entire portionof feeder piping from the Grayloc hub up to header nozzle was replaced with feeders fabricatedfrom improved material SA-106 Grade C, having a low free nitrogen concentration, which isless susceptible to cracking and wall thinning. In addition, improved bend fabrication andinstallation method were used to reduce residual stress on the tight radius bend and welds for
replacement feeders.


US PWR Unit, February 2002. In response to Bulletin 2001-01, the plant began a refuellingoutage with the intent to perform work that included remotely inspecting the VHP nozzles fromunderneath the head focusing on the CRDM. The licensee found that three CRDM nozzles had
indications of through-wall axial cracking. Specifically, the licensee found these indications in
CRDM nozzles 1, 2, and 3, which are located near the top of the reactor pressure vessel (RPV)
head. The licensee attempted to repair these VHP nozzles remotely by approaching fromunderneath the pressure vessel head. On March 6, 2002, the licensee terminated the repair
process on CRDM nozzle 3 to determine the cause of unusual equipment operation andremoved the machining apparatus from the nozzle. During the removal, the nozzle tipped in thedownhill direction until it rested against an adjacent CRDM. If the surrounding steel had been
structurally sound, it should have held the nozzle in position. The licensee investigated thecondition of the RPV head surrounding CRDM nozzle 3. The investigation included removing
the CRDM nozzle and removing large boric acid deposits from the top of the RPV head. Uponcompleting the boric acid removal on March 8, 2002, the licensee conducted a visual
examination of the area and identified a large cavity in the RPV head on the downhill side of
CRDM nozzle 3. The corrosion was caused by borated water that leaked from the reactor
coolant system onto the vessel head through cracks in the nozzle and the weld that attachednozzle 3 to the RPV head. The licensee discovered the remaining thickness of the RPV head in
the wastage area to be about 9.5 mm (3/8 inch). This thickness consisted of only stainless steel
cladding on the inside surface of the RPV head, which is nominally 9.5 mm (3/8 inch) thick.
The stainless steel cladding is resistant to corrosion by boric acid, but it is not intended toprovide structural integrity to the vessel. Failure of the stainless steel cladding would haveresulted in a loss-of-coolant accident (LOCA).



Korea PWR Unit, March 1999. During the refuelling outage, wall thinning that did not meet theminimum thickness requirement was found in main feed water straight piping. The thinned area
was located near the containment penetration area. All weld points of main feed water pipingfrom the containment penetration up to the main isolation valve should be 100% inspected
58

NEA/CSNI/R(2017)3
every 10 years according to the commitment in FSAR chapter 3.6 of the unit. Although all
welds near the thinned area had been inspected by UT and MT, the examination had been
focused to fining flaws in the welds, missing the wall thinning. Other Korean PWR units also
experienced the same event on main feed water piping near containment penetrations. The
detailed investigation showed that the wall thinning of the main feed water straight lines was
attributed to a deficiency in material design.
German Experience with Qualification of UT Technique for Nozzle Weld Seams. Owing to a
stipulation in the regulations, an examination technique was to be developed for examinationsof austenitic nozzle weld seams for operationally-induced deficiencies [64]. This examination
technique was to be capable of finding the stipulated reflectors and assessing them according to
the amplitude criterion. The technique was optimised using a test body with applied grooves
and artificial reflectors. All reflectors in the test body could be detected and assessed with therequired accuracy.
When the examination technique was used for the first time on a real test object, reflectors weredetected whose flaw pattern did not indicate cracks. Subsequently, the magnitude of the detectedreflectors was to be determined, and it was to be clarified whether there was a connection of thereflector to the outer surface. The magnitude determined initially lay within the lower range ofthe comparative reflectors.
In order to be able to make a better statement on the magnitude of the reflector, the examination
technique was optimised by various measures. Among other things, efforts were undertaken tofocus on the area of the position of the reflector. In order to achieve this, computerised focusingwith the help of the SAFT algorithm (see Glossary of Technical Terms) was used. This
optimisation was carried out on the test body with the artificially applied reflectors. By reassessing the measured data using the SAFT algorithm, it was possible to put the indications in
concrete terms and achieve an improvement of the accuracy when assessing the reflectors present
in the calibration block. The smallest reflector examined in this way had a height amplitude of
3 mm and was located at a distance of 2 mm from the outer surface.
Despite the improvements that were made, it was still not possible to assess the indications
detected on site clearly. To clarify the facts nevertheless, comparative measurements were carriedout on a comparable real part in the Gemeinschaftskernkraftwerk Tullnerfeld (GKT nuclear
power plant)2 . By carrying out measurements with the same ultrasonic examination technique it
was possible to identify indications that were absolutely comparable with the indications on sitewith respect to their pattern. The assessment of one of these indications with the help of SAFT
showed a height amplitude of 2.5 mm. Metallographic examinations of the removed sample
clearly characterised the reflector as a manufacturing flaw (weld imperfection). The height
amplitude determined by metallographic examination was 0.35 mm.
The discrepancy between the actual height amplitude and the one determined by means of SAFT
was explained by the fact that the examination technique was optimised for the assessment offlaws with a height amplitude of more than 2.0 mm. No comparative values existed for smallerreflectors. If one wants to assess even smaller reflectors, correspondingly smaller comparativevalues are needed for the qualification of the examination technique. Here, there is a limit as it isvery difficult to generate clearly defined smaller artificial reflectors.

2. Also referred to as Kernkraftwerk Zwentendorf; a BWR Type 69 designed by AEG-KWU, completed in 1978 but
it never obtained an operating license. The BWR unit is located in Niederösterreich, Austria.

59

NEA/CSNI/R(2017)3
4.3

Data analysis insights regarding LTA-RIM

A review of the CODAP event data shows that LTA-RIM occurs when detected flaws area rationalised(i.e. the flaws are not seen as recordable) or RIM procedures are not implemented in an optimum manner.
Based on the operating experience data:

No inspection procedure can succeed if a detected indication is rationalised away.

When qualified procedures are used appropriately, ultrasonic inspections have shown to have ahigh reliability.



The reliability of inspection procedures declines sharply when the procedures deviate from thequalified procedures.



If the regulatory and industry goal is to prevent leakage RIM will need to be implemented in
new areas.

The service experience data in CODAP indicates that RIM reliability challenges continue to exist.
New inspection requirements are implemented that apply to visually inaccessible piping or materials(e.g. cast austenitic stainless steel) for which qualified NDE technology is being developed.
The examples in Section 4.2 show the limits of today's performance capability of RIM. Current UTtechnology enables the detection of very minor discontinuities. With respect to the NDE qualification
process, it is concluded that the validity limits within which an examination technique will produce reliable
results are clearly defined. In practice, attempts are made (as shown in the examples above) to assess
indications that may be outside the validity scope. This can lead to a marked overestimation of smaller
flaws.
The examinations have shown that even the smallest reflectors can be detected with a large signal-tonoise ratio. From a safety-related point of view, overestimating the size of flaws means that one is on thesafe side. However, in the case of the “German PWR Unit, December 2010” it has been shown that a
considerable overestimation of the magnitude of the indications may lead to measures that may be counterproductive. Besides the already explained differences between the reflectivity of artificial flaws (grooves)and natural faults (cracks), other possible sources for wrong assessments may also be differences in thegeometry of the calibration block and the real component.

60

NEA/CSNI/R(2017)3

5. ASSESSMENT OF INFLUENCE OF RIM ON PASSIVE COMPONENT RELIABILITY

A database like CODAP provides qualitative and quantitative insights regarding the effectiveness of RIM
Programmes. The effectiveness of RIM can be characterised by the reliability of leak detection and nondestructive examination (NDE) programme implementation. The latter factor is defined by the probabilityof detecting (POD) a subsurface flaw of certain dimension (e.g. length and depth). POD is not an easily
measured quantity. It is usually inferred from experiments such as performance demonstration data. In
theory, POD values can be derived from operating experience data but a statistical analysis requires
detailed information on the total number of inspection performed across the total plant population. Thissection documents basic principles of how to account for RIM effectiveness factors in structural reliability
analysis. The focus is on risk-informed principles and the application of statistical models of structural
integrity. Also included is a summary of POD assessment results.
The statistical models and tools presented in this chapter could be used in analysing the event data of
CODAP database or other similar event database for finding interactions between degradation and damage
mechanisms, for defining the probability that flaw is found in inspection, and for identifying effective
detection and repair strategies which help mitigating the progression of cracks or leaks to major structural
failure. The models and tools presented in this chapter are examples of how to establish an analytical“bridge” between the CODAP database and passive component reliability analysis.
5.1

Problem statement

Extensive service experience data exists on passive component degradation and failure. Given that theavailable data is recorded and classified in a systematic and comprehensive manner, possibilities exist tocorrelate the failure data with the RIM strategies that were present at the time of failure. With detailed
knowledge of the relationships between different leak detection technologies and RIM strategies
opportunities present themselves for determining quantitatively the achievable level of risk reduction given
the current state-of-practice in NDE qualification and RI-ISI.
The effect of implementing a RI-ISI Programme is to alter the number and effectiveness of locations
that are inspected in relation to existing programme such as ASME Section XI. The impact of theinspection strategies on a specific location could include adding locations to the inspection programme,
removing locations, or changing the effectiveness of the programme. When an element is selected forinspection the effectiveness of the programme is expected to increase because of the knowledge gained
from the RI-ISI evaluation to determine which degradation mechanisms are most likely to be present atthat location. Hence when the strategy is switched from an existing inspection programme to RI-ISI, there
are three possibilities for each location to change its rupture frequency:If the location was inspected in the existing programme and the effectiveness of the programme
is improved, the change in the pipe rupture frequency, if any, would be to reduce the pipe
rupture frequency due to the capability to inspect for the most probable degradation
mechanisms.

61

NEA/CSNI/R(2017)3


If the location was not inspected in the existing programme but added to the RI-ISI Programme,
as frequently occurs in medium or high risk segments, the pipe rupture frequency will decrease.



If the location was inspected in the existing programme and not retained in the RI-ISI
Programme, there may be increases in pipe rupture frequency.

The integrity management factor Iik as defined in Section 1.3 accounts for the probability of a certain
RIM to successfully identify a flaw POD and the effectiveness of a leak detection system to identify aminor through-wall pipe flaw before exceeding any plant operational limits. The integrity management
factor can be determined on the basis of NDE qualification data, expert judgement or field experience datasuch as that contained in the CODAP event database. This model of piping reliability enables a quantitative
assessment of the level of risk reduction that is achievable with RI-ISI. Markov modelling may be used todetermine the integrity management factor for different types of ISI.
5.2

Models of RIM reliability

Markov modelling enables the analysis of interactions between degradation and damage mechanisms thatcause pipe failure, and the inspection, detection and repair strategies that can reduce the probability that
failure occurs, or that cracks or leaks will progress to major structural failure before being detected andrepaired. This Markov modelling technique starts with a representation of a “system” in a set of discreteand mutually exclusive states. The states refer to various degrees of piping system degradation; that is, theexistence of flaws, leaks or major structural failure. The flaws can be pipe wall thinning or circumferential
cracking of a weld heat affected zone. Figure 14 is a representation of a general four-state Markov modelof piping reliability.
Piping Reliability States:
S=
Success (or undamaged state);
C=
Crack (non-through wall flaw);F=
Leaking through-wall flaw (leak rate is within
Technical Specification limit);
L=
Large leak (leak rate in excess, or well in excess ofTechnical Specification limit).

SState Transitions:

S
S

C

C

C

F
F

L





Occurrence of non-through wall flaw

S

Occurrence of small leak given an undamaged state (‘S’)

C

Occurrence of small leak given a flaw (‘C’)

S

Occurrence of large leak given no flaw

C

Occurrence of large leak given a non-through wall flaw

F

Occurrence of large leak given a small leak

Detect and repair a through-wall flaw

Inspect and repair a non-through wall flaw

Figure 14: Markov model of piping reliability

The state transition parameters of the Markov model can be estimated directly from service data. The
model can be used to investigate the time dependence of pipe failure frequencies and the impact ofalternative ISI and leak inspection strategies. With the integrity management factor associated with the new62

NEA/CSNI/R(2017)3(i.e. RI-ISI) programme and old programme denoted as respectively Ii,New and Ii,Old the following definitions
are applied:I i , New 

h40 {RIISI}
h40 {noinsp}(5.1)

I i ,Old 

h40 {SecXI}
h40 {noinsp}

(5.2)

Where
h40{RIISI} = hazard rate (or time-dependent rupture frequency) at the end of a 40-year design life for
weld subjected to the RI-ISI inspection strategyh40{Original Pgm} = hazard rate for weld subjected to the Section XI inspection strategyh40{noinsp} = hazard rate for weld subjected to no ISI
These hazard rates are a function of time and the parameters of the Markov model in Figure 14.
According to this figure, there are six parameters that are associated with the Markov model, an occurrence
rate for non-through-wall flaws, ϕ, failure rate for leaks given the existence of a flaw, λF, two rupture
frequencies including one from the initial state of a flaw ρC, and one from the initial state of a leak, ρF, arepair rate for detectable flaws, ω and a repair rate for leaks, μ. The latter two parameters dealing withrepair are further developed by the following simple models.


PFI  POD
(TFI  TR )

(5.3)



PLD
(TLI  TR )

(5.4)

Where
PFI = Probability that a piping element with a flaw will be inspected per inspection interval. Thisparameter has a value of 0 if it is not in the inspection programme and 1 if it is in the inspection
Programme.
POD = Probability that a flaw will be detected given the weld of concern is inspected. This value isbased on engineering judgement, NDE qualification data or on field experience data.
TFI =Mean time between inspections for flaws (“inspection interval”)

TR =Mean time to repair once detected. There is an assumption that any significant flaw that is
detected will be repaired. Depending on the location of the weld to be repaired, the weld repair could takeon the order of several days to a week. Since this term is always combined with TFI, and TFI is 10 years in
ASME XI, in practice the results are insensitive to assumptions regarding TR
PLD =periodProbability that the leak in the element will be detected per leak inspection or detection

63

NEA/CSNI/R(2017)3
TLI = Mean time between inspections for leaks. For RCPB piping, the time interval between leaks
can be essentially instantaneous if the leak is picked up by radiation alarms, to as long as the time period
between leak tests performed on the system. All ASME Class 1, 2, and 3 piping must be tested for leaks atleast once per refuelling outage.
TR = The minimum of the actual repair time and the time associated with any limiting condition for
operation if the leak rate exceeds technical specification requirements.
The root parameters of the Markov model in Figure 14 are obtainable directly from the CODAP event
database (i.e. parameters ϕ, λ and ρ) or are based on the particulars of an ISI Programme. Reference [65]includes an example of how to estimate the Markov model parameters. The results of an application of theMarkov model are displayed in Figure 15. This example involves a BWR Reactor Recirculation systemand shows how different assumptions about the POD affect the calculated time-dependent small LOCA
frequency.

Small Loss-of-Coolant Accident Frequency [1/ROY]

1.0E-5

1.0E-6

1.0E-7

1.0E-8
0

10

20

30

40

50

60

Reactor Operating Years [ROY]
No ISI

POD = 0.5

POD = 0.9

3

Figure 15: Influence of POD on structural integrity

5.3

Probability of detection

Probabilities of detection (POD) curves are used to quantify the reliability of NDE systems. These curves
are developed from performance demonstration data and a typical statistical model used for processing thedata is the logistic model [16][66][67][68][69], which expresses the POD a flaw of size x as:
POD(x) = [exp(β1 + β2x)]/(1+[exp(β1 + β2x)])(5.5)

3. This analysis was originally performed in support of NUREG-1829 (“Estimation Loss-of-Coolant-Accident
Frequency Through the Elicitation Process,” 2008).

64

NEA/CSNI/R(2017)3
The two parameters β1 and β2 are estimated by a fitting procedure to determine the POD curve. Twotypes of POD-curves are used: 1) POD related to a specific NDE technique, and 2) POD related to a
specific NDE procedure. A NDE procedure describes the process for how to perform a specific type of
NDE, and it also includes documentation of the underlying qualification. Due to variability in theinspection system and other system related properties, an inspection system can only detect a certain range
of defects. The POD curve for the inspection procedure shows the probability of detection when using
certain inspection techniques. In general the detection probability for the inspection procedure will belower than for the inspection technique, since the POD curve for the inspection technique assumes an ideal
defect resolution. Various types of human factors influences are incorporated in the inspection procedure
related POD curve.
In risk-informed piping reliability applications efforts are made to apply real-life data pertaining tooccurrence rates of flaws and observed NDE and leak detection performance measures to obtain realistic
assessments of piping reliability. What, if any, role do POD-curves have in risk-informed piping
reliability? For specific, well defined analysis cases POD-curves provide good indications of the variability
in projected NDE performance. In other words, the POD-curves enable an uncertainty treatment ofdifferent sets of NDE techniques and processes. Next, the qualitative and quantitative information on NDE
performance embedded in CODAP could be applied in a Bayesian analysis framework to account for theuncertainties in the integrity management factor.

65

NEA/CSNI/R(2017)3

6. SUMMARY AND CONCLUSIONS

6.1

Summary

Effectiveness of reliability and integrity management (RIM) practices was selected as the subject of thethird CODAP topical report. The report addresses selected international practices with respect to pressure
testing, leak detection, in-service inspection (ISI) including non-destructive examination (NDE), andperformance demonstration initiatives to improve the reliability of NDE techniques. The purpose of RIM is
to prevent the occurrence of piping through-wall leaks as well as to monitor passive metallic component
degradation. RIM Programme can utilise risk insights to augment or enhance existing deterministic
integrity management programme. Through a systematic examination of the operating experience as
recorded in the CODAP event database, the field experience with the different RIM strategies has been
evaluated in order to primarily draw qualitative insights about integrity management reliability.
6.2

Conclusions

The third topical report documents how RIM strategies are accounted for in the CODAP event database.
According to the CODAP Coding Guideline that has been prepared by and adopted by the PRG, for eachrecord an evaluation is performed of the various RIM influences that have played a role in preventing orcontributing to a structural failure. Hence, the database includes a significant volume of in-service
inspection (ISI) information from which valuable insights about RIM performance issues can be drawn. It
is quite clear that RIM very significantly contributes to a high-level of structural integrity. The operating
experience insights also point to RIM implementation challenges.
The CODAP event database captures instances of LTA-RIM, including failures in detecting preexisting flaws before exceeding acceptance criteria. CODAP uses a broad definition of LTA-RIM” in thatthe term is defined as events where degradation has progressed beyond acceptable limits in systems,
structures or components (SSCs) that have a RIM Programme. These LTA-RIM events have some safetysignificance. In this topical report the LTA-RIM definition is broadened to also include events where a
RIM Programme has resulted in a “false positive”; that is, it has identified degradation that either didn’t
exist or was not close to violating acceptance criteria. While such events needlessly expend resources andcould be considered LTA-RIM from an economic perspective, they do not have any safety significance. In
the database passive component failure information is recorded in a tiered manner. All data submissions
undergo verification for technical accuracy and completeness in accordance with procedures and protocols
established by the CODAP-PRG. First, basic failure information is recorded to address the mostfundamental information about an event and this includes a free-format event narrative that describes thesequence of events, including plant response, consequence, in-plant location of failed component,
dimensional data, and component type. This is followed by recording the known ISI history, including datewhen the failed component was last inspected, method of NDE qualification if a qualified method had been
used, and any NDE performance deficiencies or failures. Finally, details about the service environment(e.g. water chemistry, stresses, pressure and temperature) are recorded as a lead-in to details from root
cause evaluations (flaw data, chemical composition of material, results of metallographic examinations,apparent and underlying causes of material degradation). When RIM fails, one or more of the following
factors are often present:66

NEA/CSNI/R(2017)3
Accepting a rejectable flaw indication for continued operation. This could be due tomisinterpretation of NDE results.



Rationalising away detected defects.

Using improperly qualified or modified NDE techniques or not selecting the correct procedure
to implement.



Poorly implementing qualified procedures.

Poorly implementing owner-defined inspection programme.



Not identifying the correct location to inspect.

Missing a flaw with a qualified procedure. A procedure may not sufficiently document the basis
for the examination details used to inspect for a specific, previously observed degradation
mechanism. There may also be inadequate administrative controls for augmented inspections
and disposition of inspection results.

Experience from examinations in the field allows the conclusion that the sensitivity of advanced
UT examination techniques is so high that the detection of material flaws, among them also
crack-like flaws, does not pose a problem in general. The difficulty lies rather more in thecharacterisation and assessment of the indications, especially if these are actually outside thevalidity scope that has been defined for the examination techniques by their qualification.
Examples from volumetric examinations performed in the filed show that there can be atendency to put greater emphasis on assessing the examining technician's results in the“impermissible scope”, which may then lead to large discrepancies between the determined andthe actual dimensions of the flaw.

According to the high-level data analysis, the number of instances of LTA-RIM has remained largely
the same over the past three decades. The rate of LTA-RIM is a function of the number of such instances
versus the overall number of examinations that have been performed, however. There has been a verysignificant evolution in RIM practices and requirements, and therefore qualified statistical insights
concerning the reliability of RIM Programme necessitates an in-depth analysis of the field experience dataas collected by CODAP.
6.3

Recommendations

With respect to the continued database development and maintenance (i.e. data submissions and validation)
it is recommended that the following items be considered in the ongoing active data submission activities
by the CODAP-PRG Members as well as in the current programme for an enhanced version of the online
database (“CODAP Option 2” Project) :Encourage the PRG Membership to more actively share RIM experience insights. As a standing
action, future Working Group meetings should focus on technical discussions regarding how toutilise CODAP and how to share data analysis insights with the nuclear safety community.



Expand the sharing of operating experience data within the PRG. Future Working Group
Meetings should include as a standing action, national overviews of recent operational events,
including the findings of root cause analyses.



For the PRG Member States that have implemented RI-ISI, add appropriate database fields thatindicate events that involve reactor components that are included in a RI-ISI Programme.
Having access to this information would be highly beneficial to future database applications sothat the database content and inspection programme can be correlated. It is noted that theEuropean Network for Inspection and Qualification (ENIQ) has undertaken an evaluation of67

NEA/CSNI/R(2017)3
lessons learnt from the application of risk-informed ISI (RI-ISI) to European nuclear power
plants. The PRG Membership is encouraged to review this ENIQ effort and to determine how
conclusion by ENIQ corresponds to the field experience data as recorded in CODAP.


Similarly, add appropriate database fields that indicate presence of an augmented inspection
programme. The basis for this recommendation is as follows:- Embedded in the database are examples where an augmented inspection programme is inplace with the provision that a 100% volumetric examination of a given component
boundary is to be performed. Yet, through-wall defects have occurred. The underlying
contributing factors include use of non-qualified NDE technique, or application of toocoarse UT-scanning matrix. Having the ability to quickly and reliably identify such eventsin the database would greatly enhance the level of user friendliness.



Based on the results of the evaluations of the CODAP database content, the number of throughwall leakages could be decreased by the following actions:-

Periodic review and independent validation of UT-scanning matrices used in inspection
piping components;-

RIM Programme optimisation on the basis of probabilistic and risk-informed
methodologies.

The CODAP software is undergoing revision and upgrade. As part of ongoing Programming
work, the PRG Membership is encouraged to review and modify as needed the definitions of
Method of Flaw Detection and Technique of Flaw Detection.

68

NEA/CSNI/R(2017)3
7. REFERENCES[1]

NEA (2012), Piping Failure Data Exchange Project (OECD/NEA OPDE), Final Report,
NEA/CSNI/R(2012)16, OECD, Paris, 2012.

[2]

NEA (2010), Technical Basis for Commendable Practices on Ageing Management – SCC and
Cable Ageing Project (SCAP), Final Report, NEA/CSNI/R(2010)5, OECD, Paris.

[3]

NEA (2015), OECD/NEA Component Operational Experience, Degradation and Ageing
Programme (CODAP): First Term (2011-2014) Status Report, NEA/CSNI/R(2015)7, OECD, Paris.

[4]

NEA (2014), CODAP Topical Report on Flow-Accelerated Corrosion (FAC) of Carbon Steel and
Low Alloy Steel Piping in Commercial Nuclear Power Plants, NEA/CSNI/R(2014)/6, OECD, Paris.

[5]

NAE (2015), CODAP Topical Report on Operating Experience Insights Into Pipe Failures inElectro-Hydraulic and Instrument Air Systems, NEA/CSNI/R(2015)6, OECD, Paris.

[6]

NEA (2008), CSNI Workshop on Risk-Informed Piping Integrity Management, CSNI Working
Group on Integrity of Components and Structures (IAGE), NEA/SEN/SIN/IAGE(2008)6, OECD,
Paris.

[7]

NEA (2011), EC-JRC/OECD-NEA Benchmark Study on Risk-Informed In-Service Inspection
Methodologies (RISMET), NEA/CSNI/R(2010)13, OECD, Paris.

[8]

Idaho National Laboratory (2011), Next Generation Nuclear Plant Probabilistic Risk Assessment
White Paper, INL/EXT-1121270, NRC Project No. 0748, Idaho Falls, ID.

[9]

American Society of Mechanical Engineers (2008), 2007 ASME Boiler and Pressure Vessel Code,
Section XI: Rules for In-service Inspection of Nuclear Power Plant Components, 2008a Addenda,
New York, NY, July 2008.

[10] Hedden, O.F. (2000), “Evolution of Section XI of the ASME Boiler and Pressure Vessel Code,”
Journal of Pressure Vessel Technology, 122:234-241.
[11] Ammirato, F. (1998), ASME-IAEA Regional Workshop on Basic Elements of ISI Program Planning,
SP-108471, EPRI Non-destructive Evaluation Centre, Charlotte, NC.
[12] Becker, L. (1997), Experience with Inspection Qualifications for Austenitic Piping, SP-108338,
EPRI Non-destructive Evaluation Centre, Charlotte, NC.
[13] Swayne, R. (2012), “Rewrite of Section XI, Division 2, Using Risk-Informed Methodology,”
Presentation at ASME/NRC Semi-Annual Management Meeting, Accession No. ML12250A317,
United States Nuclear Regulatory Commission, Washington, DC.
[14] Latiolais, C. and Orihuela, M. (2010), Performance Demonstration Initiative (PDI) Program
Description, Revision 4, 1020593, Electric Power Research Institute, Palo Alto, CA.
[15] Gray, H. et al. (1996), The Ultrasonic Testing (UT) Performance Demonstration Initiative (PDI) of
the US Nuclear Industry as Currently Implemented by the Electric Power Research Institute (EPRI)
Non-destructive Examination (NDE) Centre as the Performance Demonstration Administrator(PDA), Report No. 999 01288/95-01, United States Nuclear Regulatory Commission, Washington,
DC.
[16] Gosselin, S.R., Simonen, F.A., Heasler, P.G. and Doctor, S.R. (2007), Fatigue Crack Flaw
Tolerance in Nuclear Power Plant Piping. A Basis for Improvements to ASME Code Section XI
Appendix L, NUREG/CR-6934, United States Nuclear Regulatory Commission, Washington, DC.

69

NEA/CSNI/R(2017)3[17] ASME Standards Technology LLC (2012), Code Comparison Report for Class 1 Nuclear Power
Plant Components, STP-NU-051-1, New York, NY.
[18] Canadian Standards Association (2014), Periodic Inspection of CANDU Nuclear Power Plant
Components, N285.4-14, Toronto, Canada.
[19] Marcotte, O., Rousseau, G. and Rochefort, E. -2012), “Overview of Improvements in Work
Practices and Instrumentation for CANDU Primary Heat Transport Feeders In-Service Inspections,”
Proc. 4th International CANDU In-Service Inspection Workshop and NDT in Canada 2012
Workshop, Toronto, Canada.
[20] Trelinski, M. (2008), “Inspection of CANDU Reactor Pressure Tubes Using Ultrasonics,” Proc. 17th
World Conference on Non-destructive Testing, 25-28 Octobre 2008, Shanghai China.
[21] Baron, J. A. (2014), “Qualification of Inspection Systems in the CANDU Nuclear Industry”, CINDE
Journal, Vol. 35, Jan-Feb, 2014, pp. 10-12, 14.
[22] Izdebska, K. (2010), “Overview of CANDU RI-ISI,” Proc. 3rd International CANDU In-Service
Inspection Workshop and NDT in Canada 2010, Toronto, Canada.
[23] Rezaie-Manesh et al. (2016), “Development of a Comprehensive Risk-Informed Categorization
Process for CANDU Conventional Components and Components,” Proc. 11th International
Conference on NDE in Relation to Structural Components for Nuclear and Pressurized
Components, EUR 27790 EN, Institute for Energy and Transport, Petten, The Netherlands..
[24] AFCEN (2014), In-Service Inspection Rules for Mechanical Components of PWR Nuclear Islands,
RSE M-2010 Edition, September 2014 Addendum, La Défense, France.
[25] International Organisation for Standardisation (2012), Non-Destructive Testing – Qualification andCertification of NDT Personnel, ISO 9712:2012, Geneva, Switzerland.
[26] Kerntechnische Ausschuss (2010), Components of Light Water Reactors, Part 4: In-service
Inspections and Operational Monitoring, KTA 3201.4, Salzgitter, Germany.
[27] Fuchs, M. (2014), “An Overview of In-Service Inspection Practices at German Nuclear Power
Plants,” Presentation in German at Sicherheitskonferenz zum Thema Alterung undErmüdungsverhalten sowie wiederkehrende Prüfungen von Komponenten der Atomkraftwerke inder Restlaufzeit, Niedersächsisches Ministerium für Umwelt, Energie und Klimaschutz (NUMEK),
Hameln, Germany, December 2014.4[28] Kerntechnische Ausschuss (2015), Verification Analysis for Rupture Preclusion for PressureRetaining Components in Nuclear Power Plants, KTA 3206:2014, Salzgitter, Germany.
[29] Elmas, M. et al. (2013), The Effectiveness of Measures to Maintain the As-built Quality of thePressure-Retaining Components in German Nuclear Power Plants, GRS-A-3700 (in German),Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Cologne, Germany.
[30] European Network for Inspection Qualification (2005), European Framework Document for RiskInformed In-Service Inspection, ENIQ-23, Institute for Energy, Joint Research Centre, Petten, The
Netherlands.
[31] Doctor, S.R., Cumblidge, S.E., Taylor, T.T. and Anderson, M.T. (2013), The Technical Basis
Supporting ASME Code, Section XI, Appendix VIII: Performance Demonstration for UltrasonicExamination, NUREG/CR-7165, United States Nuclear Regulatory Commission, Washington, DC.

4.

For access to the proceedings go to
http://www.umwelt.niedersachsen.de/aktuelles/veranstaltungen/sicherheitskonferenz-129055.html

70NEA/CSNI/R(2017)3[32] Exelon Nuclear (2006), Qualification and Certification of Non-Destructive (NDE) Personnel,
Procedure TQ-AA-122, Revision 3, Warrenville, IL.5
[33] American Society for Non-Destructive Testing (2016), Recommended Practice No. SNT-TC-1A,
2016 Edition, and ASNT Standard Topical Outlines for Qualification of Non-destructive Testing
Personnel (ANSI/ASNT CP-105-2011), Columbus, OH, 2016.
[34] American Society for Non-Destructive Testing (2016), ASNT Standard for Qualification andCertification of Non-Destructive Testing Personnel (ANSI/ASNT CP-189-2016), Columbus, OH.
[35] World Nuclear Association (2014), Certification of NDE Personnel. Harmonization of International
Code Requirements, WNA Report No. 2014/003, London, United Kingdom.
[36] International Organisation for Standardisation (2012), Non-Destructive Testing – Qualification andCertification of NDT Personnel, ISO 9712-2012, Geneva, Switzerland.
[37] United States Nuclear Regulatory Commission (1989), Erosion/Corrosion-Induced Pipe WallThinning, Generic Letter 1989-08, Washington, DC.
[38] Entergy Nuclear Northeast (2015), Reply to Request for Additional Information for the Review of the
Indian Point Nuclear Generating Station Units 2 and 3 License Renewal Application, NL-15-092,
Buchanan, NY.
[39] United States Nuclear Regulatory Commission (1989), Service Water System Problems Affecting
Safety-Related Equipment, Generic Letter 89-13, Washington, DC.
[40] United States Nuclear Regulatory Commission (1994), Intergranular Stress Corrosion Cracking ofCore Shrouds in Boiling Water Reactors, Generic Letter 94-03, Washington, DC.
[41] BWRVIP Inspection Focus Group (2013), BWR Vessel and Internals Project, Reactor Pressure
Vessel and Internals Examination Guidelines, TR-105696-R15 (BWRVIP-03NP), Electric Power
Research Institute, Palo Alto, CA.
[42] American Society of Mechanical Engineers (1988), Case N-460, Alternative Examination Coverage
for Class 1 and Class 2 Welds, Approval Date: 27 July 1988, New York, NY.
[43] American Society of Mechanical Engineers (2006), Case N-729-1, Alternative Examination
Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining
Partial-Penetration Welds, Approval Date 28 March 2006, New York, NY.
[44] Riccardella, P. (2004), Materials Reliability Program: Generic Evaluation of Examination Coverage
Requirements for Reactor Pressure Vessel Head Penetration Nozzles, MRP-95R1NP, Electric Power
Research Institute, Palo Alto, CA.
[45] Doctor, S.R et al. (2013), The Technical Basis Supporting ASME Code, Section XI, Appendix VIII:
Performance Demonstration for Ultrasonic Examination, NUREG/CR-7165, United States Nuclear
Regulatory Commission, Washington, DC, May 2013.
[46] Federal Register (2011), “10 CFR Part 50 American Society of Mechanical Engineers (ASME)
Codes and New and Revised ASME Code Cases; Final Rule,” Volume 76, No. 119, United States
Government Printing Office, Washington, DC.
[47] Direction Générale de la Sûreté Nucléaire et de la Radioprotection (2002), Assemblages “Socket
Welding” du Circuit Primaire Principal des REP, Décision DGSNR-BCCN/OT/VF No. 020406,
Paris, France, 10 October2002.

5. This document is available from the NRC Public Document Room (PDR). The accession no. is ML070600393.

71

NEA/CSNI/R(2017)3[48] AFCN (2012), Design and Construction Rules for Mechanical Components of the PWR Nuclear
Islands, RRC-M, Paris La Defense, France.
[49] Economou, J., Thebault, Y. and Costes, P-E. (2011), “Small-Bore Branch Connection Fatigue,”
PVP2011-57983, Proc. ASME 2011 Pressure Vessels and Piping Division Conference, American
Society of Mechanical Engineers, New York, NY.
[50] United States Nuclear Regulatory Commission (2010), Generic Ageing Lessons Learned (GALL)
Report, NUREG-1801, Revision 2, Washington, DC.
[51] Lara, P., Flesner, B. and Dennis, M. (2007), Non-destructive Evaluation: Examination of Small-Bore
Piping Welds – Phase 1, 1015155, Electric Power Research Institute, Palo Alto, CA.
[52] Byungsik, Y., Yongsik, K. and Jeongseok, L. (2013), ”Development of Inspection Technique forSocket Weld in Small-Bore Piping in Nuclear Power Plant,” Proc. Korean Nuclear Society 2013
Spring Meeting, Daejon, Korea.
[53] NEA (2011), CODAP Operating Procedures, CODAP PR03 (Restricted), OECD, Paris.
[54] NEA (2012), CODAP Quality Assurance Programme (CODAP-QAP), CODAP PR02 (Restricted),OECD, Paris[55] NEA (2013), CODAP Coding Guideline and User Manual(CODAP-CG), CODAP PR01
(Restricted), OECD, Paris.
[56] Chu, S. (2015), Materials Reliability Program: Survey of On-Line PWR Primary Coolant Leak
Detection Technologies (MRP-187), 1012947, Electric Power Research Institute, Palo Alto.
[57] Kramb, V. et al. (2007), Advances in Phased Array Ultrasonic Technology Applications, Olympus
NDT, Waltham, MA, ISBN 0-9735933-4-2.
[58] International Atomic Energy Agency (2012), Training Guidelines in Non-Destructive Testing
Techniques: Leak Testing at Level 2, Training Course Series No. 52, Vienna, Austria.
[59] National Fire Protection Association (2002), Standard for the Inspection, Testing, and Maintenanceof Water-Based Fire Protection Systems, NFPA-25, Quincy, MA.
[60] Bertović, M. (2016), Human Factors in Non-Destructive Testing (NDT): Risks and Challenges in
Mechanized NDT, BAM-Dissertations Reihe Band 145, Bundesanstalt fürMaterialforschung- undprüfung (BAM), Berlin, Germany.
[61] Electric Power Research Institute (2011), Management of Thermal Fatigue in Normally Stagnant
Non-Isolable Reactor Coolant System Branch Lines, MRP-146 Revision 1, Pressurized Water
Reactor Materials Reliability Program, Palo Alto, CA.
[62] Ashwin, P., Dyle, R. and Wirtz, C., (2013) Non-destructive Evaluation Improvement Focus Group
Extent of Condition Actions in Response to North Anna Dissimilar Metal Weld Operating
Experience: Revision 1, 3002000041, Electric Power Research Institute, Palo Alto, CA.
[63] Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (2011), Weiterleitungsnachricht:
Anzeigen im Vorschuhende des Stutzens der Hauptkühlmittelleitung zur Volumenausgleichsleitung
im Kernkraftwerk Grafenrheinfeld. Weiterleitungsnachricht zu Ereignissen in Kernkraftwerken des
Auslandes, WLN 2011/05, Cologne, Germany.
[64] Mohr, F. (2011), “Überprüfung der Aussagekraft der Ultraschallprüfung bei der Reflektorgrößenbestimmung - Beispiele aus Qualifikationen und von realen Bauteilen,” Proc. Seminar desFachausschusses Ultraschallprüfung: Verbesserung der Prüfaussage für spezielle Prüfaufgaben,
Deutsche Gesellschaft für Zerstörungsfreie Prüfung e.V.

72

NEA/CSNI/R(2017)3[65] Arkadov, G.V., Getman, A.F. and Rodinov, A.N. (2012), “Application of the Markov Model for
Forecasting Reliability of PWR Pipelines,” Appendix 2 in Probabilistic Safety Assessment for
Optimum Nuclear Power Plant Life Management (PLiM), Woodhead Publishing Ltd., Cambridge,
United Kingdom, ISBN 978-0-85709-398-1.
[66] Gandossi, L. and Simola, K. (2007), Sensitivity of Risk Reduction to Probability of Detection Curves(POD) Level of Detail, EUR 21902 EN, Institute for Energy, Petten, The Netherlands.
[67] Ammirato, F. (2009), Materials Reliability Program: Development of Probability of Detection
Curves for Ultrasonic Examination of Dissimilar Metal Welds (MRP-262, Revision 1), TypicalPWR Leak-before-Break Line Locations, 1020451, Electric Power Research Institute, Palo Alto,CA.
[68] Meyer, M.E., Crawford, S.L., Lareau, J.P. and Anderson, M.T. (2014), Review of Literature forModel Assisted Probability of Detection, PNNL-23714, Pacific Northwest National Laboratory,
Richland, WA.
[69] Georgiou, G.A. (2006), Probability of Detection (POD) Curves. Derivation, Applications andLimitations, HSE Research Report 454, Health and Safety Executive, Sudbury, United Kingdom.

73

NEA/CSNI/R(2017)3
APPENDIX A

OPDE/CODAP BIBLIOGRAPHY

OPDE-1.

Lydell, B., Mathet, E. and Gott, K., “OECD Pipe Failure Data Exchange (OPDE) Project:
A Framework for International Cooperation in Piping Reliability,” Proc. American
Nuclear Society International Topical Meeting on Probabilistic Safety Assessment,
Detroit, Michigan, 2002.

OPDE-2.

Colligan, A., Lojk, R., Riznic, J., Lydell, B., “The OECD Pipe Failure Data Exchange
Project - Canadian Contribution on Data Validation”, Canadian Nuclear Society Bulletin,Vol. 24, No. 3, August 2003.

OPDE-3.

Choi, S.Y., Choi, Y.H., Piping Failure Analysis for the Korean Nuclear Power Piping
including the effect of In-Service Inspection, Asian Pacific Conference on NDT, APC
NDT 2003, Jeju, Korea, 2003.

OPDE-4.

Lydell, B., Mathet, E. and Gott, K., “OECD Pipe Failure Data Exchange Project (OPDE) –2003 Progress Report,” ICONE12-49217, Proc. ICONE-12: 12th International Conferenceon Nuclear Engineering, Arlington, Virginia, 2004.

OPDE-5.

Colligan, A., Lojk, R., Riznic, J. “The OECD pipe failure data exchange project - Data
validation on Canadian plants”, Proceedings of the International Conference on Nuclear
Engineering, pp. 65-70, ICONE12-49273, Arlington, Virginia USA, April 25-29 2004.

OPDE-6.

Sun, Y.C., Young, H.C., Yeon, K.C. “Application of a piping failure database to somenuclear safety issues in Korea”, Proceedings of the International Conference on Nuclear
Engineering, pp. 175-179, ICONE12-49572, Arlington, Virginia USA, April 25-29 2004.

OPDE-7.

Fleming, K. N. and Lydell, B. O. Y, (2004), Use of Markov piping reliability models toevaluate time dependent frequencies of loss of coolant accidents, Proceedings of ICONE12, the 12th International Conference on Nuclear Engineering. April 25-29, 2004,
Arlington, Virginia, United States.

OPDE-8.

Lydell, B., Mathet, E., Gott, K., “Piping Service Life Experience in Commercial Nuclear
Power Plants: Progress with the OECD Pipe Failure Data Exchange Project”, 2004 ASME
Pressure Vessels and Piping (PVP) Conference, San Diego, CA, July 25-29, 2004.

OPDE-9.

Choi, Y.H. et al, “Safety Evaluation of Socket Weld Integrity in Nuclear Piping,” KeyEngineering Materials, 270-273:1725-1730, 2004.

OPDE-10.

OECD/NEA, OECD/NEA Piping Failure Data Exchange Project, Workshop on Database
Applications, OPDE/SEC(2004)4, 2005.

OPDE-11.

Choi, S. Y., Choi, Y. H., 2005, “Evaluation of Nuclear Piping Failure Frequency in
Korean Pressurized Water Reactors”, Key Engineering Materials, Vols. 297-300, pp.16451651.

OPDE-12.

NEA, 2006a. OPDE 2006:2 Coding Guideline (OPDE-CG), OPDE PR01, OECD Nuclear
Energy Agency, Issy-les-Moulineaux, France.

74

NEA/CSNI/R(2017)3
OPDE-13.

NEA, 2006b. OPDE Quality Assurance Program (OPDE-QA), OPDE-PR-02, OECD
Nuclear Energy Agency, Issy-les-Moulineaux, France.

OPDE-14.

Vglasky, T, A. Blahoianu, B. Lydell, J. Riznic, “The OECD Pipe Failure Data Exchange
Project – Validation of Canadian Data,” ICONE14-89176, Proc. ICONE-14: 14th
International Conference on Nuclear Engineering, Miami, Florida, 2006.

OPDE-15.

Lydell, B.O.Y., Mathet, E., Gott, K., “OECD Pipe Failure Data Exchange Project: First
Term (2002-2005) Results and Insights”, International Conference on Probabilistic Safety
Assessment and Management- PSAM 8, New Orleans, Louisiana, May, 14-19, 2006.

OPDE-16.

Olsson, A., Lydell, B. and Kochenhauer, M., 2006. „Development of a Reliability Data
Handbook for Piping Components in Nordic Nuclear Power Plants”, Proc. Int. 8thInternational Probabilistic Safety Assessment and Management Conference, ASME, New
York, NY.

OPDE-17.

Reck, H. and Schulz, S., “Internationale Betriebserfahrung mit Rohrleitungen in
Kernkraftwerken: Fortschritt des OECD Pipe Failure Data Exchange (OPDE) Datenbank
Projektes, ” MPA Seminar 32, Stuttgart, Germany, 2006.

OPDE-18.

Choi, Y.H. and Choi, S.Y., “Socket Weld Integrity in Nuclear Piping under Fatigue
Loading Conditions,” Nuclear Engineering and Design, 237:213-218, 2007.

OPDE-19.

Lydell, B., “The Probability of Pipe Failure on the Basis of Operating Experience,
PVP2007-26281, Proc. 2007 ASME Pressure Vessel and Piping Division Conference, San
Antonio, Texas, 2007.

OPDE-20.

Lydell, B., Huerta, A. and Gott, K., “Progress with the International Pipe Failure Data
Exchange Project,: PVP2007-26278, Proc. 2007 ASME Pressure Vessel and Piping
Division Conference, San Antonio, Texas, 2007.

OPDE-21.

Simonen, F.A., Gosselin, S.R., Lydell, B.O.Y. and Rudland, D.L., “Application of Failure
Event Data to Benchmark Probabilistic Fracture Mechanics Computer Codes,” PVP200726373, Proc. 2007 ASME Pressure Vessel and Piping Division Conference, San Antonio,
Texas, 2007.

OPDE-22.

Lydell, B., Anders, O., “Reliability Data for Piping Components in Nordic Nuclear Power
Plants “R-Book” Project Phase I” SKI Report 2008:01, Swedish Nuclear PowerInspectorate, Stockholm, Sweden, January 2008.

OPDE-23.

OECD/NEA, “OECD Piping Failure Data Exchange (OPDE) Project, Terms andConditions for Project Operation 2008-2011”, Paris, France, 2008.

OPDE-24.

Nuclear Energy Agency, “OPDE 2008:1 Coding Guideline (OPDE-CG) and User’s
Guide”, OPDE PR01, Version 05, Issy-les-Moulineaux, France, 2008.

OPDE-25.

OECD Nuclear Energy Agency, CSNI Workshop on Risk-Informed Piping Integrity
Management, CSNI Working Group on Integrity of Components and Structures (IAGE),
NEA/SEN/SIN/IAGE(2008)6, Boulogne-Billancourt, France, 2008.

OPDE-26.

Pandey, M., “Piping Failure Frequency Analysis using OECD/NEA Data”, University ofWaterloo, Final Report, CNSC Research and Support Program. Report RSP-0236,
February 2008.

OPDE-27.

Olsson, A., Lydell, B. “Development of a Reliability Data Handbook for Piping
Components in Nordic Nuclear Power Plants - Part II”, 9th International Conference on
Probabilistic Safety Assessment and Management 2008, pp. 8-15, 2008.

75

NEA/CSNI/R(2017)3
OPDE-28.

Lydell, B., Huerta, A. and Gott, K., “Characteristics of Damage and Degradation
Mechanisms in Nuclear Power Plant Piping Systems,” PVP2008-61914, Proc. 2008
ASME Pressure Vessel and Piping Division Conference, Chicago, Illinois, 2008.

OPDE-29.

Yuan, X.X., Pandey, M.D., Riznic,J.R., (2008), “A Point Process Model for Piping Failure
Frequency Analysis using OPDE Data”, ICONE16-48078, Proceedings of the 16th
International Conference on Nuclear Engineering ICONE-16, Orlando, Florida, May 1115, 2008, American Society of Mechanical Engineers.

OPDE-30.

Lydell, B. and Riznic, J., “OPDE – The International Pipe Failure Data Exchange Project,”
Nuclear Engineering and Design, 238:2115-2123, 2008.

OPDE-31.

Yuan, X. X., Pandey, M. D., Riznic, J. R., (2009), A Stochastic Model for Piping Failure
Frequency Analysis using OPDE Data, Journal of Engineering for Gas Turbine andPower, 131, (5), art. 052901, Sept. 2009.

OPDE-32.

Riznic, J. “OECD/NEA Pipe Failure Data Exchange (OPDE) 2002-2008 Status Report”,
Canadian Nuclear Safety Commission, 2009.

OPDE-33.

Choi, Y.H. and Choi, S.Y., “Assessment of Socket Weld Integrity in Piping,” J. Loss
Prevention in the Process Industries, 22:850-853, 2009.

OPDE-34.

Olsson, A., Hedtjärn Swaling, V., Lydell, B. “Reliability Data Handbook for Piping
Components in Nordic Nuclear Power Plants - Part III”, 10th International Conference on
Probabilistic Safety Assessment and Management 2010, pp. 2337-2348, 2010.

OPDE-35.

OECD Nuclear Energy Agency, EC-JRC/OECD-NEA Benchmark Study on Risk
Informed In Service Inspection Methodologies (RISMET), NEA/CSNI/R(2010)13,
Boulogne-Billancourt, France, 2011.

OPDE-36.

Lydell, B., Huerta, A., Gott, K. “Insights from PSA Applications of the OECD Nuclear
Energy Agency (OECD/NEA) OPDE Database”, International Topical Meeting on
Probabilistic Safety Assessment and Analysis 2011, pp. 1696-1706, 2011.

OPDE-37.

Olsson A., Swaling V. H., Lydell B., “Experiences from Implementation of Updated
Reliability Data for Piping Components Using the R-Book”, ANS PSA 2011 International
Topical Meeting on Probabilistic Safety Assessment and Analysis, American Nuclear
Society, Wilmington, NC, March 13-17, 2011.

OPDE-38.

Lydell B., Mosleh A., Chrun D., “Enhanced Piping Reliability Models for Use in Internal
Flooding PSA”, ANS PSA 2011 International Topical Meeting on Probabilistic Safety
Assessment and Analysis, American Nuclear Society, Wilmington, NC, March 13-17,
2011.

OPDE-39.

Cronvall, O., Männistö, I., Kaunisto, K. “On Applications Concerning OECD Pipe Failure
Database OPDE”, 11th International Probabilistic Safety Assessment and Management
Conference and the Annual European Safety and Reliability Conference 2012, pp. 28912901, 2012.

OPDE-40.

Wood, J., Gonzalez, M., Harris, C., Tobin, M., Coyne, K. “Estimating Conditional Failure
Probabilities of Observed Piping Degradation”, 11th International Probabilistic Safety
Assessment and Management Conference and the Annual European Safety and Reliability
Conference 2012, pp. 2870-2875, 2012.

OPDE-41.

Park, J.S., Choi, Y.H. “Application of Piping Failure Database to Nuclear Safety Issues in
Korea”, International Journal of Pressure Vessels and Piping, 90-91, pp. 56-60, 2012.

76

NEA/CSNI/R(2017)3
OPDE-42.

OECD Nuclear Energy Agency, OECD/NEA Piping Failure Data Exchange Project(OECD/NEA OPDE), Final Report, NEA/CSNI/R(2012)16, Boulogne-Billancourt,
France, 2012.

CODAP-1.

OECD/NEA, “Component Operational Experience, Degradation and Ageing Programme(CODAP), Terms and Conditions for Project Operation, August 2011.

CODAP-2.

Lydell, B., Huerta, A., Gott, K., Riznic, J. “OECD-NEA CODAP Event Data Project on
Passive Component Degradation and Failures in Commercial Nuclear Power Plants,”
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, pp.
663-672, 2013.

CODAP-3.

Coyne, K., Tobin, M., Siu, N., Roewekamp, M. “Use of OECD Data Project Products in
Probabilistic Safety Assessment,” International Topical Meeting on Probabilistic Safety
Assessment and Analysis 2013, pp. 558-568, 2013.

CODAP-4.

Lydell, B., “Incorporation of FAC Considerations in Probabilistic Safety Assessment,”
Proc. FAC2013: Conference of Flow Accelerated Corrosion, May 21-24, 2013, Avignon,
France.

CODAP-5.

OECD Nuclear Energy Agency, CODAP Topical Report on Flow-Accelerated Corrosion(FAC) of Carbon Steel and Low Alloy Steel Piping in Commercial Nuclear Power Plants,
NEA/CSNI/R(2014)/6, Boulogne-Billancourt, France, 2014.

CODAP-6.

Lydell, B., Nevander, O., Gott, K., Riznic, J. “CODAP Project on International
Cooperation in the Area of Structural Integrity of NPP”, ASME International MechanicalEngineering Congress and Exposition, 2014.

CODAP-7.

Breest, A., Gott, K., Lydell, B. and Riznic, J., “OECD/NEA Multilateral Cooperation in
the Area of Structural Integrity and Ageing Management,” Paper PBNC2014-025, Proc.
19th Pacific Basin Nuclear Conference, August 24-28, 2014, Vancouver, British Columbia,
Canada.

CODAP-8.

Nevander, O., Riznic, J., Gott, K. and Lydell, B., “OECD/NEA Component Operating
Experience, Ageing and Degradation Project,” Paper O-T05-163, Proc. Fontevraud 8
Conference, September 15-18, 2014, Avignon, France.

CODAP-9.

OECD Nuclear Energy Agency, OECD/NEA Component Operational Experience,
Degradation and Ageing Program (CODAP): First Term (2011-2014) Status Report,
NEA/CSNI/R(2015)7, Boulogne-Billancourt, France, 2015.

CODAP-10.

Lydell, B., Castelao, C. “CODAP Project Operating Experience Insights Related to
Fatigue Mechanisms”, 4th International Conference on Fatigue of Nuclear Reactor
Components, 28th September – 1st October 2015, Seville, Spain, 2015.

CODAP-11.

Dragea, T., Riznic, J. “The Component Operational Experience Degradation and Ageing
Program (CODAP): Review and Lessons Learned (2011-2014)”, 23th International
Conference on Nuclear Engineering, ICONE23-1001, Chiba, Japan, May 17-21 2015.

CODAP-12.

Rivet, M., Riznic, J., (2015), The Component Operational Experience Degradation andAgeing Program (CODAP) 2014-2015 Status Report RSP R 144.6, Operational
Engineering Assessment Division Report, CNSC, Ottawa, e-docs#4904008, December
2015.

CODAP-13.

OECD Nuclear Energy Agency, CODAP Topical Report on Operating Experience Insights
Into Pipe Failures in Electro-Hydraulic and Instrument Air Systems,
NEA/CSNI/R(2015)6, Boulogne-Billancourt, France, 2015.

77

NEA/CSNI/R(2017)3
CODAP-14.

Rivet, M., Riznic, J., (2015), The Component Operational Experience Degradation andAgeing Program (CODAP) 2014-2015 Status Report- RSP R 144.6, Operational
Engineering Assessment Division, CNSC, e-docs# 4904008, December 2015.

CODAP-15.

NEA/CSNI, “NEA/CSNI/R(2016)NN: OECD/NEA CODAP Project Topical Report on
Operating Experience Insights into Pressure Boundary Component Reliability andIntegrity Management,” Boulogne-Billancourt, France, May 2016.

CODAP-16.

Lydell, B., Nevander, O. and Riznic, C., “Piping Corrosion Risk Management on the Basisof OECD/NEA CODAP Project Database,” RISK16-8327, Proc. NACE International
Corrosion Risk Management Conference, Houston, TX, May 23-25, 2016.

CODAP-17.

Rivet, M., Cormier, K., Riznic, J., “The Component Experience, Degradation and Ageing
Programme (CODAP) Project; Canada’s Contributions and Benefits to the CanadianIndustry,” submitted to the 40th Annual CNS/CNA Student Conference, Toronto, Canada,
June 19-22 2016.

CODAP-18.

Lydell, B., Nevander, O. and Riznic, J., “The OECD/NEA CODAP Project and Its
Contribution to Plant Ageing Management and Probabilistic Safety Assessment, Paper
062, Proc. 13th International Conference on Probabilistic Safety Assessment andManagement (PSAM 13), Seoul, Korea, October 2-7, 2016.

CODAP-19.

Lydell, B., “A Review of the Progress with Statistical Models of Passive Component
Reliability,” Paper 063, Proc. 13th International Conference on Probabilistic Safety
Assessment and Management (PSAM 13), Seoul, Korea, 2-7 October 2016.

78

NEA/CSNI/R(2017)3

APPENDIX B
CODAP 2016 EVENT DATABASE STATUS REPORT79

NEA/CSNI/R(2017)3

CODAP DATA SUBMISSIONS BY CALENDAR YEAR (CY)
PRG Member
CanadaChinese Taipei
Czech Republic
FinlandFranceGermanyJapan
Korea (Republic of)
Slovak Republic
Spain
SwedenSwitzerland
USA

CY 2011

CY 2012No data submittals the work scopefocused on finalizing
DB structure,developing &implementing the
Web-Based Event
Database

Totals:CODAP TORCY 2013

1
----8
--1
--1
33
44

25
6
1
--4
1
17
-3
-5
61
123
Notes

CY 2014

CY 2015

CY 2016

CY 2017

8
9
3
9
17
10
1
1
4
2
-1
56
121

18
3
-1
5
5
-2
--1
1
45
81

6
4
3
N/A
8
1
-7
5
3
N/A
-33
70

---N/A------N/A---

Total as of 31October-2016
57
22
7
10
30
28
2
27
10
8
1
8
228
439

"Each Participating Country shall submit data and general information on component degradation and failure events in English through its
National Coordinator referred to in Paragraph 16 of the Terms of Reference.
"The data exchange is carried out through signatories in the participating countries, with the possibility of delegating to otherorganizations."

CODAP-OP

A-1: Database submissions by calendar year and PRG Member
CODAP 2nd Term
CODAP 1st Term
OPDE

CODAP
Member
CA
CH
CZ
DE
ES
FI
FR
JP
KR
SE
SK
TW
US

2002

2003

2004

-1
1
17
1
-9
111
6
1
1
-56
204

8
4
1
9
4
1
3
39
1
7
1
1
74
153

3
2
21
3
-6
14
-4
1
-63
117

Number of Event Data Record Submittals by Event Date - As of 31-October-2016
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014

2015

2016

Total

4
6
-21
8
-2
6
1
1
1
-68
118

5
-2
2
-N/A
2
--N/A
-2
30
43

1

92
36
12
144
20
12
49
215
32
16
9
17
582
1236

14
3
1
19
1
1
6
14
-1
1
1
29
91

22
5
-20
--6
10
3
1
--22
89

10
1
1
7
-4
3
13
2
--1
25
67

3
7
-17
1
-2
1
2
--2
31
66

9
2
1
4
-1
2
4
2
-2
-26
53

6
2
-3
1
1
4
1
4
1
-4
27
54

-4
1
2
-3
3
1
2
--3
36
55

2
1
2
1
-1
1
4
-2
2
36
52

5
1
1
--1
--5
--1
37
51

22
23

SCAP-SCC

A-2: Database Submissions by Project Review Group members and event date80

NEA/CSNI/R(2017)3
Table A-2: OPDE/CODAP data Submission Summary
Validation Status as of 30-April-2016
Country

Approved

BE

8

CA
CH
CZ
DE
ES

202
91
25
351
48

FI

Ready for
QA

Ready for
Review by
NC

Draft

Total No.
Records

Comment

8

Participated in OPDE 1st
Term

2

202
96
25
354
50

55

2

57

FR
JP
KR

131
288
70

23
1

154
288
71

SE

365

1

366

SK
TW
US

2
13
3065
4716

5
4
24
66

4

1
3

--

1
5

81

7
18
3089
4787

2002-2014 PRG
Member

2002-2014 PRG
MemberJoined project in 2011
Joined project in 2011

NEA/CSNI/R(2017)3

APPENDIX CGLOSSARY OF TECHNICAL TERMS

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 that the flaw
length and crack mouth opening displacement meet the resolution requirements of the observation
technique. The EVT-1 specification augments the VT-i requirements to provide more rigorous inspection
standards for stress corrosion cracking (SCC). EVT-1 is also conducted in accordance with therequirements 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 take into account potential embrittlement due to thermal ageing or
neutron irradiation. Acceptance criteria methodologies to support plant-specific augmented examinations
are documented in WCAP-17096-NP1 .
Full Structural Weld Overlay (FSWOL). A structural reinforcement and stress corrosion cracking(SCC) mitigation technique through application of a SCC-resistant material layer around the entire
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 the dissimilar metal weld to be treated,
where R is the outer radius of the item and t is the nominal thickness of the item.
Grayloc Region. 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 mechanicalconnector that provides metal-to-metal seals in piping systems.
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 is provided.
Inclusion. An “inclusion” is a nonmetallic 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 nonmetallic
inclusions or “stringers”. These stringers may be surface or subsurface and are usually short in length andparallel to the grain flow.
Indication. The definition of the term “indication” as it applies to NDE is: “A response or evidence ofa response disclosed through NDE that requires further evaluation to determine its true significance.”
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 toa visible flaw (e.g. through-wall crack, active leakage).
1.

Westinghouse Electric Company LLC, Reactor Internals Acceptance Criteria Methodology and Data
Requirements, WCAP-17096-NP, Cranberry Township, PA, December 200982

NEA/CSNI/R(2017)3
Leak Detection System. Instrumentation and controls that use various temperature, pressure, level andflow sensors to detect water and steam leakages in selected reactor systems and to initiate annunciation andprovide 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.
Less-than-Adequate (LTA). In the context of a root cause analysis of RIM Programme deficiencies,
the term “LTA” is used to characterise a procedure that lacks something essential to successfully toperform an activity.
Liquid Penetrant Examination. Liquid Penetrant Examination (LPT) uses liquids to detect cracks inmaterials. In the mid and late 1930's, 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 obtainablewith the visible dyes. The fluorescent method was quickly accepted by the military for aircraft partexamination. Since then, the use of both colour-contrast and fluorescent penetrants has spread topractically all fields of manufacturing, and new 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 are calculated
and video displayed in real-time, high resolution, 3-D colour graphics that can be saved for further data
analysis or permanent record archiving Very low-frequency magnetic signals are not affected by iron oxideor any non-magnetic surface deposits which allows for accurate testing on base metals in piping.
LTA-NDE. As used in this report the term “Less-than-Adequate NDE” implies that deficiencies in theimplementation of a qualified NDE process have contributed to a reportable or rejectable flaw remaining
undetected for a certain period.
LTA-RIM. In this report, LTA-RIM is defined as events where degradation has progressed beyond
acceptable limits in systems, structures or components (SSCs) that have a RIM Programme. These LTARIM events have some safety significance. In this topical report the LTA-RIM definition is broadened toalso include events where a RIM Programme has resulted in a “false positive”; that is, it has identified
degradation that either didn’t exist or was not close to violating acceptance criteria. While such eventsneedlessly expend resources and could be considered LTA-RIM from an economic perspective, they do nothave any safety significance.
Mechanical Stress Improvement Process (MSIP®). A patented process that was invented, developed
and first used in 1986 by NuVision Engineering Inc. for mitigating SCC in nuclear plant weldments.
MSIP® works by using a hydraulically operated clamp which contracts the pipe on one side of theweldment. A typical tool design consists of a specially designed hydraulic box press for bringing the clamp
halves together. By contracting the pipe on one side of 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.
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 lessthan that of a full structural weld overlay in order allow completion in the time available in a typicalrefuelling outage for the larger geometries.

83

NEA/CSNI/R(2017)3
Probability of Detection (POD). It is the probability that a flaw of a certain size will be detected and it
is conditional on factors such as wall thickness, NDE personnel qualifications, and flaw orientation.
Project Review Group (PRG) of the CODAP Project. According to the CODAP Operating
Procedures, the PRG runs the Project, with assistance from the NEA Project Secretary and the Operating
Agent. The PRG meets at least once per year. The PRG responsibilities include but are not limited to thefollowing types of decisions:

Secure the financial and technical resources necessary to carry out the Project;Nominate the CODAP Project chairperson;

Define the information flow (public information and confidential information);

Approve the admittance of new members;

Nominate project task leaders (lead countries) and key persons for the PRG tasks;Define the priority of the task activities;

Monitor the progress of the project and task activities;

Monitor the work of the Operating Agent and quality assurance.

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) topenetrate various materials.
Reliability and Integrity Management (RIM). Those aspects of the plant design and operational phasethat are applied to provide an appropriate level of reliability of SSCs and a continuing assurance over thelife of the plant that such reliability is maintained.
Root Mean Square (RMS) Evaluation. NDE qualifications include the use of an ultrasonic sizing
procedure which should be developed and qualified for equipment, technique, and sizing examination
personnel. At least 10 flawed specimens should be used in the performance demonstration. A RMS
evaluation should be used to demonstrate adequate sizing performance. This is given by the formula:
RMS= √((T-U) 2/N)
Where
T = Truth or actual flaw depth
U = Ultrasonic flaw depth estimate
N = Number of test specimens or flaws sized
Acceptable flaw sizing performance demonstration is achieved when the RMS is 12.5% or less. Thisis comparable to the Appendix VIII criteria proposed in ASME Code Section XI. Accordingly, it wasdemonstrated that at an RMS of 15% or less, acceptable sizing performance is achieved comparable to thecurrent EPRI NDE Centre IGSCC, sizing programme. The advanced ultrasonic sizing techniques describedin this handbook have been developed in accordance with recommended guidelines of the EPRI NDE
Centre Ultrasonic Planar Flaw Sizing of IGSCC. Variations or modifications of the techniques have been
incorporated to improve accuracy of flaw depth sizing of stress corrosion, thermal fatigue and mechanicalfatigue cracks.
SAFT. Synthetic Aperture Focusing Technique is a signal processing technique which takes advantageof the movement of a small conventional transducer to simulate, in effect, a phased array that is extremely
long. This allows high resolution at long range, with relatively small transducers. SAFT in ultrasonics has
been around for over 20 years but the amount of processing required has meant that it has had to wait for84

NEA/CSNI/R(2017)3
developments in computing technology before it can be readily applied. Phased-array techniques havedeveloped at a faster pace than SAFT, however.
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 PWSCC-susceptible material(Alloy 82/182) from the primary reactor coolant.
Unified Numbering System (UNS). An alloy designation system in use in North America. It consists ofa prefix letter and five digits designating a material composition. For example, a prefix of S indicatesstainless steel, C indicates copper, brass or bronze alloys.
VT-1 Examination. A limited visual examination specific to ASME Section XI which is theobservation of exposed surfaces of a part, component, or weld to determine its physical condition including
such irregularities as cracks, wear, erosion, corrosion, or physical damage.
VT-2 Examination. Per ASME XI, a visual surface examination to locate evidence of leakage frompressure-retaining components.
VT-3 Examination. A limited visual examination specific to ASME Section XI which is theobservation to determine the general mechanical and structural condition of components and their supports,such as the verification of clearances, settings, physical displacements, loose or missing parts, debris,
corrosion, wear, erosion, or the loss of integrity at bolted or welded connections. The VT-3 examinations
shall include examinations for conditions that could affect operability of functional adequacy of snubbers,
and constant load and spring type supports. The VT-3 examination is intended to identify individualcomponents with significant levels of existing degradation. As the VT-3 examination is not intended todetect the early stages of component cracking or other incipient degradation effects, it should not be used
when failure of an individual component could threaten either plant safety or operational stability. The VT3 examination may be appropriate for inspecting highly redundant components (such as baffle-edge bolts),
where a single failure does not compromise the function or integrity of the critical assembly.
Visual Examination. The oldest and most commonly 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 with various
magnifying devices, without changing, altering, or destroying the object being examined”. Per ASME XI,
there are three different VT methods; VT-1, VT-2 and VT-3.

85

NEA/CSNI/R(2017)3

APPENDIX DPIPING SAFETY CLASS CROSS REFERENCE TABLE86

NEA/CSNI/R(2017)3

1

Table D-1: Piping Safety Class Cross Reference Table
USA / ASME Section III2
Class
1

Definition
Piping that forms the
RCPB. That is, allpiping componentsthat are part of thereactorcoolant
system RCS), or
connected to the
RCS up to andincluding any or all of
the following:a) the outermost
primary
containment
isolation valve inpiping thatpenetrates theprimary
containment;b) the second of twovalves normallyclosed during
normal reactor
operation insystem pipingthat does not
penetrate primarycontainment; or,c) the RCS safety
and relief valves

Canada(CSA N285.0-95)
Class
Definition
1
Pipe diameter > DN20;
Sections of systems, or
systems
connected
thereto, which contain
fluid
that
directlytransports heat from
nuclear fuel, and whose
failure would cause a
LOCA as defined in thesafety report.

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
1
Equipment of the RCPB,
except equipment whose
rupture would result in aleakage of a magnitude
within the capacity of thenormal coolant make-up
system.

France (RRC-P 900 R.4)
Class
1

Definition
RCS and its connecting lines
with inside diameter greaterthan 10.6 mm for water orgreater than 21.9 mm forsteam, up to and including thetwo reactor coolant isolation
valves. Class 1 piping also
includesthepressuriser
letdown line up to and including
the relief and safety valves.

1. This table was prepared by the OPDE-PRG in 2005. It is reproduced from the OPDE/CODAP Coding Guideline/
2. The ASME III classification is explained in NRC Regulatory Guide R.G 1.26 (Revision 4, 2007).

87

Germany (KTA 3201/3211)
Class
1

Definition
For PWRs:a) RPV;b) Primary side of the
S/Gs, the secondary
shell of the S/Gs incl.
the FW-inlet and MSexit nozzles up to thepipe connecting welds(excl. small-diameter
fittings) ,c) Pressuriser,d) RCP casing,e) Connecting pipes
between the above
components and thevalve casings of any
type contained in thepiping system,f) Pipes branching off fromthe above components
and their connecting
pipes including thevalve bodies up to andincluding the first shutoff valve,g) Control rod drives andthe in-core
instrumentation,
For BWR:a) RPV,b) Piping connected to the
RPV including the valve
bodies incl. first shut-off
valve, pipework
penetrating thecontainment shell incl.
the last shut-off valve
located outside thecontainment shell,c) Control rod drive and incore instrumentation,

Switzerland (ENSI-G01/d)
Class
SK-1

Definition
Pressure-retaining
boundary of the reactor
cooling system up to
the second isolation
valve or safety valve,
includingsmalldiameter piping andpressure-retaining
partsof
instrumentation.

NEA/CSNI/R(2017)3
USA / ASME Section III2
Class
2

Definition
Systems or portionsof systems importantto safety that aredesigned for1) emergency corecooling,2) post-accident
containment heatremoval, or3) post-accident
fission productremovalCanada(CSA N285.0-95)
Class
Definition
2
Pipe diameter > DN20;
Sections of processsystems that penetratethecontainment
structure and form partof the containment
boundary.

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
2a) Components creating the
RCPB, that are notranked as Class 1,b) Components for thereactor shutdown during
the abnormal operation
during the states which
could lead to theaccident conditions, andfor the reactor shutdownwith the aim to mitigate
the consequences ofaccident conditions,c) Components necessary
to retain the coolant
inventory sufficient forthe core cooling during
the accident conditions
when no damage of thereactor coolant pressure
system has occurred,
and after these
conditions,d) Components necessary
to remove the core heat,
when the reactor coolant
pressure system isdamaged, with the aim to
limit the fuel damage,e) Components of theresidual heat removal
system during the normal
and abnormal operation
and under the accident
conditions, without theloss of the RCPB’s
integrity,f) Components necessary
for the prevention ofradioactive leakage fromthe containment during
the accident and postaccident conditions,g) Components necessary
to limit the ionisingradiation penetration
outside the containment,
during and after theFrance (RRC-P 900 R.4)
Class
2

88

Definition
Equipment and components ofsystems
carrying
reactorcoolant that are not safety class
1 and to equipment andcomponents required to ensure
containment of radioactivity inthe event of a LOCA. This
includes:a) Equipment andcomponents that are notsafety class 1;b) Main equipment andcomponents of the following
systems: RHRS, CVCS,
ECCS, CSS;c) Equipment andcomponents that constitutethe third barrier: the reactor
containment and theassociated isolation
systems, portions ofsecondary systems inside
the reactor building up to
and including the firstisolation valve locatedoutside the reactor building,containment hydrogen
control system, equipment
and components of the incore instrumentation
system up to and including
the manual isolation valve.

Germany (KTA 3201/3211)
Class
2 and
3

Definition
Pipingandpipingcomponents that are notpart of the RCPB but have acertain significance withrespect to reactor safety:a) The component isrequired for the
mitigation of DBAs withrespect to shut down,
long-term maintenanceof subcriticality, anddecay heat removal.
Requirements
regardingcomponents of systems
which only indirectly servein residual heat removal –these are the nonradioactivity retaining closed
cooling water systems andservice water systems –shall be specified on aplant-specific basis taking
the design redundancy (e.g.
redundancy, diversity) into
consideration.
b) Large energies are
released in case of
failure of the plantcomponent and no
mitigating measures
such as structural
measures, spatial
separation or othersafety measures areavailable to keep theeffects of the failure toan acceptable limit withrespect to nuclear
safety.
c) A failure of the plantcomponent could either
directly or indirectly
through a chain ofassumed sequential
events, lead to a DBA.
d) Systems andcomponents to which
none of criteria a)

Switzerland (ENSI-G01/d)
Class
SK-2

Definition
 Reactor coolingand emergency
cooling; Residual heatremoval fromreactor,containment, andSGs; Cooling of RCS in
the cold
depressurised
state;
 All reactor shut
down functions andfunctions tomaintainsubcriticality; Safety functions ofprimary
containment
systems;
 Components tomaintainsubcriticality in thefuel element
storage; BWR: Main steam
and feed water line
between second
isolation valve andnext remote control
isolation valve; PWR: Secondaryside of SG up toisolation valve
outside primarycontainment; Components that
could cause a dose
limit violationaccording to HSKR-11

NEA/CSNI/R(2017)3
USA / ASME Section III2
Class

Definition

3

Cooling water andauxiliary feedwater
systems or portionsof these systems
important to safetythat are designed for1) emergency corecooling,2) post-accident
containment heatremoval,3) post-accident
containment
atmosphereclean-up, or4) residual heatremoval fromreactor and fromthe spent fuelstorage pool(including primaryand secondary
cooling systems).
Portions of thesesystems that are
required for theirsafety functions andthat1) do not operate
during any modeof normal reactor
operation and2) cannot be tested
adequately
should beclassified as
Class 2.

Canada(CSA N285.0-95)
Class
Definition

3

Pipe diameter > DN20;
Sections of systems,
not classified as Class
1 or 2, that containradioactive substances
withatritium
concentration
exceeding 0.4 TBq/kg(0.01 Ci/g), or anenergy-weighted
activity concentration ofradionuclides
exceeding that of 0.4
TBg/kg of Tritium.

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
accident conditions,h) Components necessary
to accomplish the safetyfunctions for the power
supply or for the controlof other components
ranked as the safetyclass 2,3a) Components necessary
to prevent theunallowable transient
processes connectedwith the reactivity
changes,b) Components necessary
to maintain the nuclear
reactor in the safe
shutdown conditions,c) Components necessary
to maintain sufficient
reactor coolant inventory
for the core cooling
during the normal andabnormal operation,d) Components necessary
to remove heat from thesafety systems to thefirst accumulating
volume, which is
sufficient from theviewpoint of performanceof safety functions,e) Components necessary
to maintain the radiation
exposure of populationand of nuclear
installation personnel
below the establishedlimits, during theaccident conditions
connected with theleakage of radioactive
substances and ionisingradiation from thesources located outside
the containment, andafter such conditions,France (RRC-P 900 R.4)
Class

Definition

3

Safety Class 3 includes:a) CVCS equipment andcomponents required forthe purification of thereactor coolant water andthe boron make-up systemand equipment;b) S/G AFWS equipment andcomponents locatedoutside reactor
containment;c) CCWS and ESWS
equipment andcomponents;d) Reactor cavity and spent
fuel pit cooling andtreatment systemequipment andcomponents;e) Radioactive waste
treatment systems
equipment and components
whose failure could cause
release of radioactive
gases normally stored for
decay.

89

Germany (KTA 3201/3211)
Class

Switzerland (ENSI-G01/d)

Definition
through c) apply, thefailure of which,however, would lead tomajor plant internal
damage.

Class

Definition

See “2 and 3” above

SK-3

 Systems forleakage and sealwater in the
primary
containment; Cooling of fuelelement storagepool;
 Systems forgaseousradioactive media;
 RWCU of BWR(typically SK-1 and3), CVCS of PWR(typically SK-2 and3); Auxiliary systems
for SK-1 through 3
components and1E classified
electricalequipment; Systems for
accident mitigation

NEA/CSNI/R(2017)3
USA / ASME Section III2
Class

Definition

Canada(CSA N285.0-95)
Class
Definition

Czech Republic(Regulation 214/1997 Sb)
Class
Definitionf) Components requisite tomaintain suchenvironmental conditions
inside the nuclear
installation that arenecessary for theoperation of safetysystems and for theaccess of the personnel
to perform the important
activities related tosafety,g) Components necessary
to prevent the radioactive
leakage from theirradiated fuel that istransported or stored
within the nuclear
installation, out of thecore cooling system
during all states ofnormal and abnormal
operation,h) Components necessary
to remove fission heatfrom the irradiated fuelstored within the nuclear
installation, out of thecore cooling system,i) Components requisite tomaintain sufficient
subcriticality of fuelstored within the nuclear
installation, out of thecore cooling system,j) Components requisite tolimit the effluents or the
leakage of solid, liquid or
gaseous radioactive
substances and ionisingradiation below theestablished limiting
values during all statesof normal and abnormal
operation,k) Components requisite toperform the safety

France (RRC-P 900 R.4)
Class

90

Definition

Germany (KTA 3201/3211)
Class

Definition

Switzerland (ENSI-G01/d)
Class

Definition

NEA/CSNI/R(2017)3
USA / ASME Section III2
Class

Definition

Canada(CSA N285.0-95)
Class
Definition

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
functions related to thepower supply or to thecontrol of othercomponents ranked as
the safety class 3,l) Components requisite toperform the safetyfunctions for theassurance of functional
capability of othercomponents ranked as
the safety classes 1, 2
and 3, that are notrelated to the control orto the power supply,m) Components necessary
for prevention or
mitigation of theconsequences of failuresof the other components
or constructions of safetysystems ranked as thesafety classes 1, 2 and3.

France (RRC-P 900 R.4)
Class

91

Definition

Germany (KTA 3201/3211)
Class

Definition

Switzerland (ENSI-G01/d)
Class

Definition