Nuclear Safety
NEA/CSNI/R(2018)12
April 2019www.oecd-nea.org

NEA CODAP Project Topical

Report on Basic Principles
of Collecting and Evaluating
Operating Experience Data on
Metallic Passive Components

Nuclear Energy Agency

NEA/CSNI/R(2018)12
Unclassified

English text only
4 April 2019

NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSNEA CODAP PROJECT TOPICAL REPORT ON BASIC PRINCIPLES OF
COLLECTING AND EVALUATING OPERATING EXPERIENCE DATA ON
METALLIC PASSIVE COMPONENTS

Please note that this document is available in PDF format only.

JT03445864

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.

2 │ NEA/CSNI/R(2018)12

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NEA/CSNI/R(2018)12│3

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSThe Committee on the Safety of Nuclear Installations (CSNI) is responsible for the Nuclear Energy
Agency (NEA) programmes and activities that support maintaining and advancing the scientific andtechnical knowledge base of the safety of nuclear installations.
The Committee constitutes 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 has 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 in and abreast of developments in technical safety matters.
The Committee reviews the state of knowledge on important topics of nuclear safety science andtechniques and of safety assessments, and ensures that operating experience is appropriately accounted
for in its activities. It initiates and conducts programmes identified by these reviews and assessments inorder to confirm safety, overcome discrepancies, develop improvements and reach consensus ontechnical issues of common interest. It promotes the co-ordination of work in different member
countries that serve to maintain and enhance competence in nuclear safety matters, including theestablishment of joint undertakings (e.g. joint research and data projects), and assists in the feedback of
the results to participating organisations. The Committee ensures that valuable end-products of thetechnical reviews and analyses are provided to members in a timely manner, and made publicly
available when appropriate, to support broader nuclear safety.
The Committee focuses primarily on the safety aspects of existing power reactors, other nuclear
installations and new power reactors; it also considers the safety implications of scientific and technicaldevelopments of future reactor technologies and designs. Further, the scope for the Committee includes
human and organisational research activities and technical developments that affect nuclear safety.

4 │ NEA/CSNI/R(2018)12

Table of contents

Executive summary...................................................................................................................... 7
List of abbreviations and acronyms ........................................................................................... 9
1. Introduction ............................................................................................................................ 121.1. Data collection methodology............................................................................................. 121.2. Objectives and scope ......................................................................................................... 14
1.3. Nomenclature .................................................................................................................... 15
1.4. Report structure and reading guide.................................................................................... 15
2. Data quality assurance .......................................................................................................... 16
2.1. Principle of data quality .................................................................................................... 16
2.2. QAP scope ......................................................................................................................... 17
2.3. Confidentiality ................................................................................................................... 18
2.4. Coding guideline ............................................................................................................... 19
2.5. Applications handbook ...................................................................................................... 19
2.6. Data completeness ............................................................................................................. 20
2.7. Database capability categorisation .................................................................................... 21
3. Passive component degradation and failure ........................................................................ 24
3.1. Apparent cause and contributing factors ........................................................................... 24
3.2. Metallic passive component degradation/failure manifestations....................................... 24
3.3. Design and construction (D&C) defects............................................................................ 28
3.4. Fatigue of piping components ........................................................................................... 38
3.5. Corrosion fatigue ............................................................................................................... 42
3.6. Flow-assisted degradation ................................................................................................. 43
3.7. Stress corrosion cracking................................................................................................... 45
3.8. High-level database summary ........................................................................................... 48
4. CODAP database structure................................................................................................... 53
4.1. Scope of event database .................................................................................................... 53
4.2. Database submissions ........................................................................................................ 56
4.3. Database user instructions ................................................................................................. 56
4.4. Database field categories ................................................................................................... 61
4.5. Data entry .......................................................................................................................... 61
4.6. Non-through-wall flaw characterisation in CODAP ......................................................... 68
5. Database application facilities............................................................................................... 70
5.1. Records management ........................................................................................................ 70
5.2. Search function .................................................................................................................. 70
5.3. Query function................................................................................................................... 70
5.4. Export function .................................................................................................................. 76
6. Future developments ............................................................................................................. 78
6.1. CODAP improvement plan phase one .............................................................................. 78

NEA/CSNI/R(2018)126.2. CODAP improvement plan phase two .............................................................................. 79
6.3. Advanced database applications ........................................................................................ 81
7. Summary and recommendations .......................................................................................... 90
7.1. Summary ........................................................................................................................... 90
7.2. Next steps and recommendations ...................................................................................... 91
8. References ............................................................................................................................... 93
9. Appendix CODAP database structure ................................................................................ 96
10. Appendix OPDE/CODAP bibliography........................................................................... 137
11. Appendix Glossary of technical terms ............................................................................ 144
List of tables
Table 2.1. CODAP Completeness Index (CI) Definitions ........................................................... 17
Table 4.1. Scope of CODAP Event Database .............................................................................. 55
Table 6.1. OPDE/CODAP Data Submission Summary ............................................................... 79
Table 9.1. CODAP Database Field Definitions & Coding Guidelines ........................................ 97
Table 9.2. Selected Plant System Designators ........................................................................... 116
Table 9.3. Selected Stainless Steel Designations ....................................................................... 117
Table 9.4. Expanded Stainless Steel Cross-Reference Table ..................................................... 118
Table 9.5. Selected Carbon Steel Designations ......................................................................... 123
Table 9.6. ASTM, JIS and DIN Steel Designation .................................................................... 124
Table 9.7. Plant System Cross-Reference Table ........................................................................ 126
Table 9.8. Piping Safety Class Cross Reference Table .............................................................. 128

List of figures
Figure 2.1. Database Capability Categorisation........................................................................... 22
Figure 3.1. Codification of Apparent Cause of Degradation/Failure & Underlying Contributing
Factors ........................................................................................................................ 26
Figure 3.2. Metallic Passive Component Failure Manifestations ................................................ 27
Figure 3.3. D&C event sequence diagram ................................................................................... 30
Figure 3.4. Pipe Failure by Component Type .............................................................................. 31
Figure 3.5. Safety Class 1 Pipe Failures ...................................................................................... 32
Figure 3.6. Operating Experience with Safety Class 1 Field & Shop Welds............................... 33
Figure 3.7. BWR Safety Class IGSCC Failure Data.................................................................... 34
Figure 3.8. Calculated Failure Rates for NPS28 Stainless Steel Welds....................................... 35
Figure 3.9. Calculated Weld Failure Rates Conditional on Crack Depth .................................... 36
Figure 3.10. BWR Reactor Recirculation System Weld Failure Data ......................................... 37
Figure 3.11. Calculated Field Weld & Shop Weld Failure Rates (> 10”) ................................... 38
Figure 3.12. Impact on Plant Operation by EHC & Instrument Air (IA) Pipe Failure ................ 40
Figure 3.13. EHC & IA Piping Degradation Mechanisms........................................................... 41
Figure 3.14. Socket Weld Failure Summary ................................................................................ 42
Figure 3.15. Distinguishing Characteristics of the Different FAD-Mechanisms ......................... 44
Figure 3.16. Summary of Significant PWR Reactor Coolant System Leak Events..................... 49
Figure 3.17. Database Content by Damage / Degradation Mechanism ....................................... 50│56 │ NEA/CSNI/R(2018)12
Figure 3.18. Normalised Piping Degradation Propensity ............................................................ 51
Figure 3.19. Selected PWSCC Operating Experience by Location of Degraded / Failed
Component .............................................................................................................. 52
Figure 4.1. CODAP Online Opening Screen & Main Work Area Screen ................................... 58
Figure 4.2. CODAP Data Entry Form.......................................................................................... 59
Figure 4.3. CODAP Data Entry Form & Work Flow Area ......................................................... 60
Figure 4.4. Event Description – Basic Information ..................................................................... 63
Figure 4.5. Flaw Size Information ............................................................................................... 64
Figure 4.6. In-Service Inspection History .................................................................................... 65
Figure 4.7. Root Cause Analysis Information (Partial Screenshot) ............................................. 66
Figure 5.1. High-Level Data Sort................................................................................................. 72
Figure 5.2. SEARCH for PWSCC Data Records by Event Date, Event Type & System............ 73
Figure 5.3. Query Example Using "STATISTICS" (Partial Screenshot) .................................... 74
Figure 5.4. CSV-File Example..................................................................................................... 75
Figure 5.5. Importing a XML-File to Microsoft® Access Template ............................................ 77
Figure 6.1. Conceptual Web Page / CODAP “Portal” (or “Main Menu”) ................................... 82
Figure 6.2. Conceptual web page/PRG member area .................................................................. 83

NEA/CSNI/R(2018)12

Executive summary
Structural integrity of piping systems is important for plant safety and operability. In
recognition of this, information on degradation and failure of piping components andsystems is collected and evaluated by regulatory agencies, international organisations(e.g. the Nuclear Energy Agency [NEA], the International Atomic Energy Agency[IAEA], the Joint Research Centre [JRC] Operating Experience Clearinghouse) andindustry organisations worldwide. This information is often used to provide systematic
feedback to reactor regulation and research and development programmes associatedwith non-destructive examination (NDE) technology, in-service inspection (ISI)
programmes, leak-before-break evaluations, risk-informed ISI, and probabilistic safetyassessment (PSA) applications involving passive component reliability.
Several NEA member countries have agreed to establish the Component Operational
Experience, Degradation and Ageing Programme (CODAP) to encourage multilateral
co-operation in the collection and analysis of data relating to degradation and failure ofmetallic piping and non-piping metallic passive components in commercial nuclear
power plants. The scope of the data collection includes service-induced wall thinning,part through-wall cracks, through-wall cracks with and without active leakage, andinstances of significant degradation of metallic passive components, including pipingpressure boundary integrity.
The 13 members of CODAP in its third term (2018-2020) are as follows: Canada, theCzech Republic, Finland, France, Germany, Korea, Japan, the Netherlands, the Slovak
Republic, Spain, Switzerland, Chinese Taipei, and the United States. Sigma-Phase Inc.,
from the United States, works as the Operating Agent of the CODAP project.
This fifth CODAP topical report focuses on the CODAP event database structure andthe underlying principles of collecting operating experience data on metallic passive
components. The report represents a summary of the CODAP Operating Procedures, the
CODAP Event Database Coding Guideline, and the CODAP Applications Handbook.
This report documents the CODAP Event Database structure and the underlying
technical considerations to achieve high data quality as well as database
comprehensiveness. Specifically, the report responds to the following frequently asked
questions:••••

How is the quality of the data in the event database controlled and monitored?
What is the level of database completeness and comprehensiveness?
What are the guiding principles of how the database is populated with failure
event information?
Does the database support applications (or, what is its fitness-for-use)?

The report describes the methods to add event data, the internal quality review methodsof data and gives an overview on the number and nature of events in the database.
Providing support for different applications is a continuous target of the CODAP
database development. With respect to the continued database development and│78 │ NEA/CSNI/R(2018)12
maintenance (i.e. data submissions and validation) it is recommended that the following
actions be considered in the ongoing active data submission activities of the CODAP
database project:•••

improve the coding navigation tools;
encourage the PRG Membership to more actively share metallic passive
component operating experience insights, to use the collected data for analysis
and 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 item, national overviewsof recent operational events, including the findings of root cause analyses.

The CODAP PRG faces two important future challenges:••

Firstly, while efforts have been made to promote CODAP and associated data
project products to the nuclear safety community at large, there remain
programmatic issues relative to how to make the restricted CODAP event
database available to PSA practitioners.
Secondly, work remains to be done relative to the development of PSA-centric
database application guidelines and associated analytical infrastructure (i.e.
piping reliability analysis techniques and tools).

Additionally, a proposal has been made for an international benchmark exercise
concerning the use of operating experience data to quantify piping reliability parameters
for input to a standard problem application, e.g. risk informed operability determination.

NEA/CSNI/R(2018)12

List of abbreviations and acronyms
AFNOR
AMP
ASME
ASTM
AKZ
BOP
BSI
BWR
CASS
CC
C-F
CE-OEF
CNRA
CODAP
CRDM
CRP
CSNI
CSV
CUF
D&C
DEGB
DM
DN
DQA
E-C
E/C
EAC
ECCS
ECSCC
EHC
EN
EPRI
ESF
ESWS
FAC
FAD
FIV
FSWOL
FW
FWC
GALL

Association Française de Normalisation (French standards association)
Ageing Management ProgrammeAmerican Society of Mechanical Engineers
American Society for Testing of Materials
Anlagenkennzeichnungssystem (Germany)
Balance-of-plant
British Standards Institution
Boiling water reactor
Cast austenitic stainless steel
Capability category
Corrosion fatigue
Clearinghouse on operating experience feedback (European Union)
Committee on Nuclear Regulatory Activities (NEA)
Component Operational Data and Ageing Programme (NEA)
Control rod drive mechanism
Conditional rupture probability
Committee on the Safety of Nuclear Installations (NEA)
Comma separated values
Cumulative usage factor
Design and construction defectDouble-ended guillotine breaks
Degradation mechanism
Nominal diameter
Data quality assurance
Erosion-cavitation
Erosion/corrosion
Environmental assisted cracking
Emergency core cooling system
External chloride SCC
Electro-hydraulic Control
European standard
Electric Power Research Institute
Emergency safety function
Essential service water system
Flow-accelerated corrosion
Flow-assisted degradation
Flow-induced vibration
Full structural weld overlay
Field weld
Feedwater and condensate
Generic ageing lessons learned│910 │ NEA/CSNI/R(2018)12
GOST
GTAV
HAZ
HC
HCF
HPI
HWC
IA
IAEA
IAGE
IASCC
ICES
ICW
ID
IGSCC
ISI
JIS
JRC
KKS
KWO
LC
LCF
LCO
LDI
LTA
LWR
MIC
MSIP
NC
NDE
NEA
NEI
NMCA
NPP
NPSAG
NRC
NSSS
NTW
NWC
OD
OE
OPDE
PFM
PHWR
PRG
PRIS
PSA
PSI
PWR
PWSCC

Gosudarstvennyy Standart (Russian standards association)
Gas tungsten arc welding
Heat-affected zone
High cycle
High-cycle fatigue
High-pressure injection
Hydrogen water chemistry
Instrument air
International Atomic Energy Agency
CSNI Working Group on Integrity and Ageing of Components and Structures
Irradiation assisted stress corrosion cracking
INPO Consolidated events database system
Intake cooling water
Inside diameter
Intergranular SCC
In-service inspection
Japan Industrial Standard
Joint Research Centre (of the European Commission)
Kernkraft kennzeichnung system (Identification System for NPPs)
Kernkraftwerk Obrigheim
Low cycle
Low cycle fatigue
Limiting condition for operation
Liquid droplet impingement
Less than adequate
Light water reactor
Microbiologically influenced corrosion
Mechanical stress improvement process
National coordinator
Non-destructive examination
Nuclear Energy Agency
Nuclear Energy Institute
Noble metal chemical application.
Nuclear power plant
Nordic PSA Group
Nuclear Regulatory Commission
Nuclear steam supply system
Non-through-wall
Normal water chemistry
Outside diameter
Operating experience
OECD Pipe failure Data Exchange project
Probabilistic fracture mechanics
Pressurised heavy water reactor
Project review group
Power reactor information system
Probabilistic safety assessment
Pre-service inspection
Pressurised water reactor
Primary water SCC

NEA/CSNI/R(2018)12

QA
R&D
RCA
RCP
RCP
RCPB
RCS
RI-ISI
RIM
RIS
RPS
RPV
RR
SAW
SCAP
SCC
SG
SEAS
SFC
S/G
SGTR
SICC
SKI
SLR
SMAW
SQL
SRS
SS
SSM
SSMFS
STP
SW
TAE
TF
TGSCC
TWC
TSO
TTR
UF
UHS
UNE
VGB
WGIAGE
WPS
WWER
XML

Quality assurance
Research and development
Root cause analysis
Rapid crack propagation
Reactor coolant pump
Reactor coolant pressure boundary
Reactor coolant system
Risk-informed in-service inspection
Reliability and integrity management
Radiation induced segregation
Reactor protection system
Reactor pressure vessel
Reactor recirculation
Submerged arc welding
Stress Corrosion and Cable Ageing Project (NEA)
Stress corrosion cracking
Steam generator
Slovenské elektrárne AS (Slovak Republic)
Spent fuel cooling
Steam generator
Steam generator tube rupture
Strain induced corrosion cracking
Statens kärnkraftinspektion (now SSM, Sweden)
Subsequent license renewal
Shielded metal arc welding
Structured query language
Software requirements specification
Stainless steel
Swedish Radiation Safety AuthorityStrålsäkerhetsmyndighetens föreskrifter (Sweden)
Standard temperature and pressure
Depending on the context, either “Service water” or “Shop weld”
Thermal ageing embrittlement
Thermal fatigue
Transgranular stress corrosion cracking
Through-wall crack
Technical support organisation
Time-to-repair
Usage factor
Ultimate heat sink
Spanish standards association
Vereinigung der Großkesselbesitzer (Germany)
Working Group on Integrity and Ageing of Components and Structures (NEA)
Welding process specification
Water-water energetic reactor
eXstensible markup language│ 1112 │ NEA/CSNI/R(2018)12

1. Introduction
Since 2002, the Nuclear Energy Agency (NEA) has operated an event database projectthat collects information on passive metallic component degradation and failures of the
primary system, reactor pressure vessel internals, main process and standby safetysystems, and support systems (i.e. ASME Code Class one, two and three or equivalent),
as well as non-safety-related (non-code) components with significant operational
impact. With an initial focus on piping systems and components (the OPDE Project [1])the scope of the project in 2011 was expanded to also address the rector pressure vessel
and internals as well as certain other metallic passive components that are susceptible
to environmental degradation [2]. In recognition of the expanded scope, the Project
Review Group (PRG) approved the transition of OPDE to a new expanded “Component
Operational Experience, Degradation & Ageing Programme” (CODAP). The CODAP
2011-2014 [3] and 2015-2017 work programmes include tasks to prepare Topical
Reports [4] [5] to foster technical co-operation and to deepen the understanding of
national differences in plant ageing management.
In addition to recognising the intrinsic value of exchanging operating experience data
and related root cause analysis results and insights, an important motivation forsupporting the international collaboration in 2002 was embedded in the then emerging
trend towards implementing risk-informed in-service inspection (RI-ISI) programmes.
An area of specific interest at the time was concerned 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 havebeen explored in multiple database application projects1. The fifth CODAP topicalreport documents the CODAP event database structure and the underlying technicalconsiderations for achievement of data quality, completeness and comprehensiveness,and the principles for how to extract passive component reliability information from thedatabase.

1.1. Data collection methodology
The NEA joint database project CODAP exchanges operating experience data on
metallic passive component degradation and failure, including service-induced wall
thinning, non-through wall cracking, leaking through-wall cracking, pinhole leaks,
leakage, rupture and severance (pipe break caused by external impact). The scope of thedata exchange is articulated in the “Terms and Conditions for Project Operation.” Insummary, for non-through wall cracks the CODAP scope encompasses degradation
exceeding design code allowable for wall thickness or crack depth as well as suchdegradation that could have generic implications regarding the reliability of in-service
inspection (ISI) techniques. The following failure modes are considered:•

Non-through wall defects (e.g. cracks, wall thinning) interpreted as structurally
significant and/or exceeding design code allowable. Unless detected in time,

1. Appendix B includes an OPDE/CODAP bibliography that identifies selected databaseapplications that have been performed or sponsored by the OPDE/CODAP member
organisations.

NEA/CSNI/R(2018)12

•
•
•
•
•
•

these defects are potential precursors to more severe passive component
degradation.
Loss of fracture toughness of cast austenitic stainless steel piping. The loss offracture toughness is attributed to thermal ageing embrittlement (TAE).
Through-wall defects without active leakage (leakage may be detectedfollowing a plant operational mode change involving depressurisation and cooldown, or as part of preparations for non-destructive examination (NDE).
Small leaks (e.g. pinhole leak, drop leakage) resulting in piping repair or
replacement.
Leaks (e.g. leak rates within Technical Specification limits).
Large leaks (e.g. flow rates in excess of Technical Specification limits).
Major structural failure (pressure boundary “breach” or “rupture”).

In other words, the CODAP event database collects data on the full range of degradedconditions, from “precursor events” to major structural failures. The structural integrityof a pressure boundary is determined by multiple and interrelated reliability attributes
and influence factors. Depending on the conjoint requirements for damage anddegradation, certain combinations of material, operating environment, loading
conditions together with applicable design codes and standards, certain passive
components are substantially more resistant to damage and degradation than others. As
an example, for chemically stabilised austenitic stainless steel pressure boundary
components, there are no recorded events involving active, through-wall leakage. By
contrast, for unstabilised (or “lean”) austenitic stainless steel, multiple events involving
through-wall leakage have been recorded, albeit with relatively minor
observed/measured leak rates. Flow-accelerated corrosion (FAC), if unmonitored, is arelatively aggressive degradation mechanism that has produced major structural
failures, including double-ended guillotine breaks (DEGB). The types of passive
component failures that are included in the CODAP event database are:••Event-based failures that are attributed to damage mechanisms and local pipe
stresses. Examples include high-cycle fatigue (e.g. acoustic or flow-induced) in
combination with inadequate or failed pipe support, and hydraulic transient (e.g.
steam or water hammer) acting on a weld flaw (e.g. slag inclusion).
Failures caused by environmental degradation such as stress corrosion cracking
due to combined effects of material properties, operating environment (e.g.
corrosion potential, irradiation) and loading conditions.

The CODAP event database is a web based, relational structured query language (SQL)
database consisting of ca. 100 uniquely defined data fields. It is a blend of free-format
fields for detailed narrative information and fields that are defined by drop-down menus
with key words (or data filters) or related tables. A basic premise of the use of narrative
information is to preserve original event information as recorded in root cause
evaluation reports and reportable occurrence reports. The “related tables” include
information on material, location of damage or degradation, type of damage ordegradation, system name, safety class, etc. The event database structure with databasefield definitions and data input requirements are defined in a coding guideline, which iscentral to the project, including database maintenance, data validation, quality control,and database applications.

│ 13

14 │ NEA/CSNI/R(2018)12

1.2. Objectives and scope
The CODAP project review group (PRG) and its predecessor, the OECD pipe failure
data exchange (OPDE) PRG, have expended considerable resources on the developmentof the CODAP event database, including the database structure and database computer
software. This development has benefitted from a multi-disciplinary approach toestablish a consensus framework for how to record and analyse operating experience(OE) data on metallic passive component degradation and failure. Project participants
with expertise in metallurgy, corrosion science, non-destructive examination (NDE)
technology, nuclear engineering, nuclear regulation, piping design, structural reliability,
root cause analysis, and PSA were actively engaged in formulating the event databasestructure and related database application requirements.
Since its inception in 2002 the CODAP PRG has pursued an outreach effort to informthe nuclear science and nuclear safety communities about its activities including the
progress with the development of the event database. This outreach programme has
consisted of regional workshops, active conference participation, and support toregional R&D activities.
In response to frequently asked questions about the quality of the data in the event
database, the level of database completeness and comprehensiveness, and its fitness forapplication, the PRG during its 12th working group meeting (October 2016) decided toprepare a Topical Report that provides a detailed description of the database structure
and its underlying technical justifications. The specific objectives of this topical report
are to:
•
•
•
•
•
•

Provide a definition of the term “data quality” and the underlying data
qualification process.
Document the technical justifications behind the CODAP database structure.
Describe the database structure in terms of its field definitions and associated“controlling” parameters. That is, the passive component reliability attributes
and influence factors that are translated into uniquely defined key words.
Describe the event database user interface and facilities for interrogating thedatabase.
Describe data completeness and the role and responsibility of respective PRG
Member to ensure an equitable data exchange.
Describe how the event database can support applications.

The CODAP Project places emphasis on data quality, including the completeness andcomprehensiveness of recorded events. Data quality is achieved through a formal
validation process as articulated in a coding guideline. The roles and responsibilities
with respect to data submissions and data validation are defined in the CODAP
operating procedures. This topical report is concerned with the data qualification
process as implemented by the CODAP PRG. The CODAP PRG is fully aware of thefact that the full root cause analysis documentation as prepared by an owner/operator orits subject matter experts is not normally disseminated outside the industry and national
regulators. The CODAP Coding Guideline includes instructions for what “root cause
information” to include in the database. As a guiding principle, the instructions that areprovided state that any relevant information on a cause-consequence relationship is to
be included. Respective national co-ordinator assumes responsibility for the accuracyof the technical information that is input to the event database. Furthermore, the web-

NEA/CSNI/R(2018)12│ 15

based database has provisions for uploading any available supporting information; e.g.
laboratory reports, root cause analysis reports, isometric drawings and photographs.

1.3. Nomenclature
Data quality is a perception or an assessment of a certain data set’s fitness to serve its
purpose in a given context [6] [7] 2 Aspects of data quality include:
•
•
•
•
•
•
•
•

Accuracy.
Completeness; i.e. ensuring that a given event population captures all relevant
OE, and that the process of collecting and evaluating OE data is in compliancewith a well-documented set of definitions.
Update and validation status.
Relevance with respect to the CODAP work scope definition.
Relevance with respect to current ageing management processes.
Consistency across the full set of data records so that any given database query
captures all relevant data.
Robustness with respect to data classification.
Accessibility for users.

Acceptable data quality is crucial to all aspects of the CODAP project infrastructure. Data quality isaffected by the way data is entered, stored and managed. Data quality assurance (DQA) is the processof verifying the accuracy and robustness of data. Maintaining data quality requires going through thedata periodically and “scrubbing” it. Typically this involves updating it, standardising it, and deduplicating records to create a single view of a specific data record. In CODAP “scrubbing” is done
through the implementation of a broad range of database queries. Query results provide a means for
ensuring that all relevant event records are being extracted from the database in response to the datafilters that are being invoked through a query command.

1.4. Report structure and reading guideThis topical report consists of eight sections and three appendices. Section 2 documents
the CODAP data quality assurance (DQA) program. Section 3 includes a primer on
metallic passive component degradation and failures as an introduction to the technicaljustifications for the CODAP event database structure. Section 2 details the databasestructure and data input requirements. A summary of the database application facilities
can be found in Section 5. Future database enhancement plans are documented in
Section 6. The report summary and recommendations are documented in Section 7.
Finally, a list of references is provided in Section 8.
Appendix A includes a listing of all database field definitions and supporting technicalinformation. Appendix B is an OPDE/CODAP bibliography including references todatabase applications performed or sponsored by OPDE/CODAP member organisations
since 2002. Appendix C includes a glossary of terms.

2. For additional perspectives on data quality refer to a white paper entitled “The Six Primary
Dimensions for Data Quality Assessment” (2013) available from www.dqglobal.com.

16 │ NEA/CSNI/R(2018)12

2. Data quality assurance
The usefulness of any component failure data collection depends on the way by whicha stated purpose is translated into database design specifications and requirements fordata input and validation, access rules, support and maintenance, and quality assurance(QA). The objective of the CODAP quality assurance programme (QAP) is to establish
organisational and technical principles and measures for quality assurance (QA) andmonitoring of the work during the operation of the CODAP Project to ensure highquality of its products (CODAP event database and topical reports). The QAP applies
to all activities in the project and is to be followed by all project participants.

2.1. Principle of data quality
To achieve the objectives established for the CODAP event database a coding format
has been developed. This coding format is reflected in a coding guideline. The coding
guideline builds on established pipe failure data analysis practices and routines that
acknowledge the unique aspects of passive component reliability in heavy water reactor
and light water reactor operating environments (e.g. influences by material properties,
water chemistry, temperature, pressure).
For an event to be considered for inclusion in the event database it must undergo an
initial screening for eligibility. 3 An objective of this initial screening is to go beyond theabstracts of event reports to ensure that only events according to the work scope
definition are included in the database. This screening process sometimes is less-thanstraightforward. As one example, a PWR unit in 2016 experienced what initiallyappeared to be a minor reactor coolant pressure boundary (RCPB) leakage on a highpressure safety injection line. On closer evaluation, the leak was located on a seal weldof a threaded small-diameter connection and the leakage path was via the threads andnot through-wall. Therefore, the leakage was not a RCPB leakage per ASME XI
definition. Subsequently this event was not selected for inclusion in the database.
Data quality is affected from the moment the field experience data is recorded at anuclear power plant, interpreted, and finally entered into a database system. The field
experience data is recorded in different types of information systems ranging fromaction requests, work order systems, via ISI databases and outage summary reports, tolicensee event reports or reportable occurrence reports. Consequently, the details of adegradation event or failure tend to be documented to various levels of technical detailin these different information systems. Building a CODAP event database record
containing the full event history often entails extracting information from multiple
sources.
3. Section 4 covers the scope of the database. Furthermore, the passive component types andfailure modes to be considered are detailed in a Microsoft® Excel Workbook in which all
database input parameters are detailed.

NEA/CSNI/R(2018)12│ 17

The term “data quality” is an attribute of the processes that have been implemented toensure that any given database record (including all of its constituent elements, or
database fields) can be traced to the source information. The term also encompasses“fitness-for-use”, that is, the database records should contain sufficient technical detailto support database applications.
In CODAP, a “Completeness Index” (CI) is used for database management purposes. It
distinguishes between records for which more information must be sought and thoseconsidered to be complete (Table 2.1). Each record in the database is assigned a CI,
which relates to the completeness and comprehensiveness of the information in the
database relative to the requirements of the Coding Guideline.
Table 2.1. CODAP Completeness Index (CI) Definitions

CompletenessDescription
Index
1
Validated – all source data have been reviewed – no further action is expected
Validated – source data may be missing some non-essential information – no further
action anticipated. The term “non-essential” implies that information about piping
2
layout (including location of a flaw) may not be known exactly but can be inferred
based on other, similar events (at same or similar plant)3
Not validated – validation pending
The “Completeness Index” is also intended as a database filter for determination of the“fitness-for-application.” The range of possible database applications covers advanced
applications (e.g. the study of the effect of different water chemistries on specificdegradation susceptibilities), risk-informed applications (e.g. technical basis for
degradation mechanism assessment in risk-informed ISI programme development, or
statistical parameter estimation in support of internal flooding PSA), and high-levelsummaries of service experience trends and patterns. Advanced database applications
would normally rely on queries that address specific subsets of the overall database
content. By contrast, high-level database applications would draw on information fromthe entire database content.
Completeness also relates to the completeness of the event population in the database.
The operating agent periodically monitors the completeness of the CODAP event
database by comparing how other external data sources capture noteworthy events.

2.2. QAP scope
The QA programme covers all aspects of the CODAP project, including:
•
•
•

•

Confidentiality (see 2.3 for additional information).
Coding guideline (see 2.4 for additional information).
Event database development and maintenance. Any updating of databasestructure or content, including database scope issues, can be performed only by
the operating agent with technical support from NEA-IT Group and must firstbe approved by the PRG.
Data collection and data exchange. Data collection and coding of national datais performed by respective member country’s national co-ordinator (NC) or
persons/organisations to whom/which the NC delegates this responsibility. The

18 │ NEA/CSNI/R(2018)12•••

data submitted to operating agent should be approved by the national
authorities/utilities and ready for data exchange.
Data submittal. The national co-ordinators are responsible for data submissions.
Distribution of information. Official distribution of project documentation takes
place via publication on the password protected project website.
External review. In certain cases, the CODAP PRG may submit a document
containing general information for review by the CSNI or a CSNI working group(e.g. the Working Group on Integrity and Ageing of Components and Structures(WGIAGE), or CNRA Working Group on Operating Experience). Aftercompletion of a review and subsequent comment resolution by the CODAP
PRG, a document containing general information is published as a CSNI report.

2.3. Confidentiality
The CODAP project differentiates between public domain information and confidential
information. There are three levels of confidential information:
1. Level one. Applies to all documentation developed by the project review group.
It is published on the password protected Project website. Selected level one
documents may become available to interested parties via external reporting by
the NEA. Examples of level one documents include the CODAP topical reports.
Such documents must undergo review and approval by the CSNI prior to
publication.
2. Level two. Applies to the online CODAP event database, national data and data
analysis results. This material is kept on the NEA secure server, is password
protected, and can be accessed only by authorised users. It is distributed only
among active PRG members under the “Terms and Conditions for project
operation,” is never published on the project web site or distributed outside of
the PRG.
3. Level three. Applies to proprietary raw data and associated reference material
used in creating database records. This material is kept on the NEA SecureServer. PRG members who are interested in this material shall contact theappropriate national co-ordinator. In the web-based event database, any
attachment containing proprietary information is clearly marked as a “level
three” document.
The CODAP terms and conditions contain statements on the use of data within andoutside the project and on the handling of proprietary information. The event databaseis a restricted database and its access is limited to participating organisations thatprovide input data. The database is available on the internet via a secure server locatedat the NEA headquarters.
It has been recognised by the project review group that many member organisations will
want to pass on the CODAP database to their technical support organisations andconsultants for use in specific projects, and suchlike. For this purpose, a nonconfidential version of the restricted CODAP database will be made available for use
by consultants for a limited period of time. Before supplying a non-confidential version,
the member organisation making the request must provide the national co-ordinator withwritten proof that the intended recipient of the non-confidential version of the databasehas agreed to comply with the confidentiality terms and conditions of the project. As ofthe date of publication of this topical report, the database has been made available to theNEA/CSNI/R(2018)12

Nordic PSA Group (NPSAG), the Swiss nuclear plant operators, the Canadian nuclear
plant operators, and certain members of the VGB PowerTech e.V. in Germany.

2.4. Coding guideline
To achieve the objectives of the CODAP project, a coding format is developed. This
coding format is reflected in the coding guideline, which is a controlled document. The
coding guideline builds on established passive component failure data analysis practices
and routines that acknowledge the unique aspect of passive component reliability (e.g.
influence by material and water chemistry). All database development and data coding
activities are to be based on the coding guideline.

2.5. Applications handbook
The CODAP Applications Handbook (CODAP-AH) includes guidelines for how to
extract specific insights about material degradation through database interrogation, toassess failure trends and to create event population data for input to statistical parameter
estimation tasks. It includes descriptions of the data processing steps that are needed tofacilitate statistical evaluations of operating experience with metallic piping components
and non-piping passive components. Whereas the CODAP coding guideline (CODAPCG) defines database structure and data submission requirements, the CODAP-AH
includes guidelines for creating database queries and associated data processing steps.
The 2004 workshop on OPDE applications (Seoul, Korea) identified a list of potential
database applications. 4 Additional perspectives on NEA data project applications areincluded in the 2013 WGRISK document entitled “Use of OECD Data Project Products
in Probabilistic Safety Assessment”[8].
Since the launch of OPDE in 2002, numerous database applications have been pursued
to address a wide range of quantitative piping reliability analyses; see Appendix B for alisting of summary reports, conference papers and technical reports. Accompanying
these applications has been methods development initiatives to advance the piping
reliability analysis methodology and techniques. Most of these applications have beenin the context of probabilistic safety assessment (PSA) and risk-informed applicationsof PSA models. Some applications have also been pursued to support structural integrity
assessments using probabilistic fracture mechanics (PFM) by providing flaw initiation
data and service experience data analysis results to validate the assumptions used in
PFM. Practical insights from past applications form the basis for the CODAP-AH.
In its present form the online version of the CODAP event database facilitates datasubmissions, various search and sort functions, and database interrogation functions.
The latter are performed in the QUERY area of the online database. In addition, thedatabase may be downloaded to a local computer or computer network via a data “exportfunction”. The export function produces a XML-file 5 that can be converted to access or
excel format for further data processing and analysis.

4. OECD Nuclear Energy Agency, 2005. OECD-NEA Piping Failure Data Exchange Project(OPDE). Workshop on Database Applications, OPDE/SEC(2004)4, Seoul, Korea, 8 December
2004.
5. XML = Extensible Markup Language│ 19

20 │ NEA/CSNI/R(2018)12
Correlating the event population with the relevant plant and component populations that
produced these failure events enables the estimation of reliability parameters for input
to structural reliability models or PSA models. The information contained in the
CODAP event database must be processed according to specific guidelines and rules tosupport reliability parameter estimation. A first step in this data processing involves
querying the event database by applying data filters that address the conjoint
requirements for pipe degradation and failure. The data filters are integral part of thedatabase structure as defined in the CODAP-CG. Specifically, these data filters relate tounique piping reliability attributes and influence factors with respect to piping systemdesign characteristics, design and construction practice, in-service inspection (ISI) andoperating environment. The CODAP-AH consists of five parts:
1. Data exploration in the online version.
2. Exporting the CODAP event database to a local computer and converting theonline database for example into a Microsoft® Access using a template.
3. Overview of basic Microsoft® Access functions.
4. Fundamentals in database query definition. SQL (or Structured Query
Language) is used to manage data in relational databases such as CODAP.
Database queries are defined through SQL statement definition. Allapplications, whether simple or advanced begin by defining queries to extract
specific information from the event database.
5. Basic guidelines for the estimation of piping reliability parameters. These
guidelines build on insights and results from pipe failure database applications(e.g. Appendix A). Included in this section of the CODAP-AH are descriptionsof the statistical analysis tools and techniques that are compatible with theMicrosoft® Access version of the database.

2.6. Data completeness
Most, if not all database applications are concerned with evaluations of event
populations as a function of calendar time, operating time or component age at time of
failure. The technical scope of the evaluations includes determination of trends andpatterns and data homogeneity, and assessment of various statistical parameters ofpassive component reliability. Therefore, an intrinsic aspect of the practical databaseapplications is the completeness and comprehensiveness of an event database. Do theresults of an application correctly reflect the effectiveness of in-service inspection,
ageing management, and/or water chemistry programs? Does the database capture “all”
relevant operational events? In summary:••Completeness is an indication of whether or not all the data necessary to meetcurrent and future analysis demands are available in the database. Essentially,
has the coding guideline been followed in such a way that the SEARCH and
STATISTICS functions of the database produce accurate results?
Comprehensiveness is concerned with how well CODAP captures the full andappropriate range of reliability attributes (e.g. material properties, dimensionaldata) and influence factors (e.g. operating environment, pipe stresses). Thismeans that in addition to using an original event report as a base reference,
additional background information such as a root cause analysis report anddestructive examination results have been utilised in order to create an event
record.

NEA/CSNI/R(2018)12

The completeness of the CODAP database hinges on respective national co-ordinator’sdata submission routines and associated local data management infrastructure. Insights
and results of practical database applications are fed back to the CODAP PRG so that
any identified database weaknesses, omissions or errors can be addressed and corrected
in a timely manner.

2.7. Database capability categorisation
Over the years many different types of pipe failure databases have been developed [12][13]. Relative to the intended use, maintenance/updating routines and QA, a distinction
is made between “failure event database” and “reliability database”. The former is acollection of raw data (or field data) on specified types of piping components or piping
systems with or without database QA programme in place but with direct access tosource data. Usually, a failure event database has a single user (can be a person ororganisation) with sporadic or periodic database maintenance to support high-levelevaluations of failure trends; it is referred to as a category one database. A reliability
database includes processed raw data, is continuously updated and subjected tovalidation for technical accuracy and completeness and is referred to as a category two
database. Some form of independent peer review normally precedes the release of acategory two database for routine application by multiple users and a QA programmeshould be in place. Industry guides and recommendations exist for category 2 databasedevelopment, structure and quality [14]. Chapters 2 and 3 of SSMFS 6 2005:2/2008:13[15] address the need for quality assured failure data in the context of risk-informed inservice inspection (RI-ISI). Invariably, a reliability database has multiple users engaged
in PSA and risk-informed applications or advanced applications (for example expandedrisk-informed application to investigate certain correlations between degradation
mitigation and failure rate).
There is a third type of database, which is referred to as a category zero database; see
Table 2.3 in Reference [9].It is a hybrid database, which includes some of the features
found in category one and category two databases, but it is not intended to exist as astandalone, computerised database for practical use beyond an original relatively
narrowly defined objective. This type of database is typically embedded as extensive
tables in a technical report, sometimes as an appendix, and provides traceable
background to derived piping reliability parameters included in the main body of atechnical report. Historically these published category zero databases have foundwidespread use in risk-informed applications, however. Where this has been the case, adata user’s parameter selections and justifications are rationalised by simply referencing
a table in a public-domain report.
Based on their respective ability to support practical applications, SKI Report 2008:01[9] identifies three database categories (Category zero, category one and category two).
Figure 2.1 shows how these categories compare with the NEI “PSA Peer Review
Guidelines” [10] grading and the ASME PRA Standard “Capability Categories” (CC)[11].

6. “Regulations & General Advice” issued by the Swedish Radiation Safety Authority (SSM).

│ 21

22 │ NEA/CSNI/R(2018)12
Figure 2.1. Database Capability Categorisation

ASME/ANS RA-Sb-2013 (September 2013)
Application Capability Category 7 [11]
CC-I

CC-II

CC-IIINEI 00-02 PSA Peer
Review Guidelines

Grade 1,2

Grade 3

Grade 4

R-Book Database
Categorisation [9]Cat0, Cat1

(Cat1) Cat2

Cat2

In risk-informed applications data quality is particularly important and necessitatesconsiderations for traceability and reproducibility of derived reliability parameters:
including the source data producing database query results and data processing andstatistical analysis of query results. From a user perspective, a category two databaseshould include detailed and correct information on failure events so that database queries
generate relevant and complete results. That is, detailed information with respect toreliability attributes and influence factors. Furthermore, provisions should exist forpooling of different but relevant subsets of failure data to strengthen the statisticalsignificance of obtained parameters. In summary, a minimum set of requirements [11]on a category two database include:
•
•
•
•
•
•
•
•

User-friendly and flexible structure, data input forms should be designed in sucha way as to encourage continuous updating by multiple operators. The structure
should be flexible so that new database fields can be added if so desired.
Clear database field definitions that reflect the attributes and influence factorsthat are unique to pipe degradation and failure.
Input of raw data supported by an extensive, all-inclusive set of roll-down menus
with standardised and complete set of key words.
“All-inclusive” structure in which free-format memo fields for narrativedescriptions support codification and justifications for assumptions if needed.
Support full traceability from field data to processed data so that database users
and independent reviewers have full confidence in the completeness andaccuracy of database field contents.
Configuration control including user access rules.
Use of recognised and proven computer program(s) so that the databasestructure and its content remain impervious to future programme revisions and“upgrades.”Ease of transfer of database query results to external computer programme.

7 The PSA capability categories refer to the level of detail of design information (including OEdata) needed to support PSA development and application. CC-1 refers to a “base-line” PSA
developed to provide relative ranking of contributors to core damage frequency. CC-II refers toplant-specific PSA in which data specialisations are performed to better characterise plant risk
contributors. CC-III refers to advanced PSA applications performed in order to support changes
to plant design-basis.

NEA/CSNI/R(2018)12

•

•
•

•

Data security routines must be established to ensure that all relevant but
potentially sensitive or proprietary failure information is captured in the
database. Also routines must exist for proper sharing of information among
multiple users.
Detailed database documentation including coding guideline to ensure proper
technology transfer.
Approved QA program. To be effective a QA programme should reflect a
consensus perspective on data quality. The prospective database users must havea common understanding of intended usage and steps that are required to ensure
configuration control and validation of database records.
Completeness of database should be ensured through continuous or at least
periodic updating. Completeness is concerned with event populations andassurances that “all” relevant events are captured. It is also concerned withcompleteness of the classification of each database record. Ultimately“completeness” has direct bearing on the statistical significance of derived
reliability parameters.

This “requirements list” for a category two database such as CODAP is not an allinclusive list. Depending on the number of database users and type of application thatis being pursued, additional requirements could be defined. Fundamentally, a databasefor risk-informed applications should be robust in the sense that it must support a broad
range of applications, including repeat applications, and provide analysts with a solid
knowledgebase for database query definition. Ideally a reliability database should be
self-contained so that it includes all facts about the cause-and-consequence of any
degraded condition recorded in it. Why was it recorded in the first place, what were thematerial specifications and operating conditions, and exactly where in a piping system
did the failure occur?

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24 │ NEA/CSNI/R(2018)12

3. Passive component degradation and failure
The metallic passive component physics-of-degradation/failure is determined by metalsenvironment systems, localised loading or stress conditions, and methods offabrication/installation. The interactions between the various controlling parameters areconjoint. That is, all of the individual conditions for the controlling parameters must be met
for a given degree of material degradation. Therefore, the structure of a passive component
failure database like CODAP needs to capture a relatively large number of physicalparameters to correctly characterise the cause and underlying contributing factors of a certain
degraded or failed condition. This section is an overview of metallic passive component
degradation mechanisms and the consequential requirements for a comprehensive operating
experience database structure.

3.1. Apparent cause and contributing factors
In the coding of an event according to the primary degradation mechanism that caused thefailure, the data analyst relies on the information contained in an event report and any
supporting information such as a root cause analysis report that may include results from a
non-destructive examination. The CODAP coding guideline provides basic information thecoding; Figure 3.1. Correct coding of an event relies as much on the available information inan event report as on the analyst’s knowledge of material science, piping design principles,
nuclear plant operations principles, non-destructive examination technologies, and pastcomponent field experience. In many cases the coding can be quite challenging. Sections 3.2
through 3.7 document the basic considerations that enter into the process of ensuring that thecoding process is accurate and complete relative to the apparent cause and contributing
factors.

3.2. Metallic passive component degradation/failure manifestations
The causes of passive component degradation or failure (e.g. loss of structural integrity,
through-wall leak) are attributed to various damage or degradation mechanisms. Passive
component failure occurs due to synergistic effects involving operating environment andloading conditions. CODAP considers two classes of passive component failure types:••

Event-driven failures. These failures are mechanically stress driven and attributed toconditions involving combinations of equipment failures (other than the piping itself;
e.g. loose/failed pipe support, leaking valve) and stress risers or unanticipated loading
conditions (e.g. hydraulic transient or operator error). Examples of event-based
failures include various fatigue failures such as low/high-cycle vibration fatigue andthermal fatigue.
Failures attributed to time-dependent environmental degradation. Environmental
degradation is defined by unique sets of conjoint requirements that include operating
environment, material and loading conditions. These conjoint requirements differ
extensively across different piping designs (material, diameter, wall thickness,

NEA/CSNI/R(2018)12

method of construction/fabrications). Similarly, pipe flaw incubation times and flaw
growth rates differ extensively across the different combinations of degradationsusceptibility and operating environments.
As recorded in CODAP, synthesised in Figure 3.2 is the entire body of field experience withmetallic piping in commercial nuclear power plants. Included in this figure are the many
unique failure manifestations of concern in the codification of field experience data.

│ 25

26 │ NEA/CSNI/R(2018)12
Figure 3.1. Codification of Apparent Cause of Degradation/Failure & Underlying Contributing Factors

NEA/CSNI/R(2018)12

Figure 3.2. Metallic Passive Component Failure Manifestations 8

8. The vertical arrows indicate the presence of potential or actual synergistic effects. For example, thermal fatigue may cause crack initiation, and crack
propagation may occur via IGSCC. The fill-effects in the colourised horizontal bars are commensurate with the observed event populations; i.e. a strong fillcorresponds ‘multiple’ events and a weak fill corresponds ‘relatively few’ major structural failures.

│ 27

28 │ NEA/CSNI/R(2018)12
A “failure” is any degraded condition that necessitates repair or replacement. The“magnitude” of a failure manifestation is assessed through non-destructive/destructive
examination, visual examination or metallographic examination. Through-wall defects
are characterised by the size of a flaw and resulting mass or volumetric leak or flow rate(from perceptible leakage to gross leakage). Some combinations of material, operating
environments have produced “major structural failures” while certain othercombinations have produced only minor flaws. For example, stainless steel piping has
not experienced any major structural failures, while carbon steel in wet steam
environment has experienced major structural failures. According to this failure
synthesis, certain combinations of material, loading, and environmental conditions haveproduced major structural failures. For such combinations direct estimation of pipe
break frequencies is feasible. In contrast, stress corrosion cracking (SCC) induced major
structural failures have not been experienced. Development of a conditional rupture
probability model for SCC requires a probabilistic fracture mechanics approach andwith possible input from an expert elicitation process.
In Figure 3.2 the downward arrows symbolise the potential synergistic effects of various
damage and degradation mechanisms. As one example, various types of weld defects(e.g. lack of fusion, slag inclusions) tend to be a strong contributor to crack initiationsites that ultimately result in stress corrosion cracking (SCC) failure. As anotherexample, thermal fatigue can cause crack initiation while a SCC mechanism can cause
crack propagation in a pipe through-wall direction.
Noteworthy is the fact that certain combinations of metals/environment systems,
localised loading or stress conditions, and methods of fabrication/installation haveproduced major structural failures while other combinations at most have resulted inrelatively minor through-wall flaw. An event database such as CODAP documents
historical information and does not provide predictions about the long-term structural
integrity. However, it does enable the assessment of temporal trends in metallic material
performance. An objective of CODAP is to address ageing and the positive andpotentially negative effects degradation mitigation initiatives.

3.3. Design and construction (D&C) defects
A generic insight from piping failure root cause analyses points to the significance ofhuman error (or organisational factor) contributions. Official process industry incident
statistics show that 20% to 90% of all incidents are indirectly or directly caused byhuman error. Human errors are either latent or active; c.f. Reason [16] and Embrey et al[17]. Effects of a latent error may lie dormant within a system for a long time, only
becoming evident after a period of time when the condition caused by the error combineswith other errors or particular operating conditions. An example of latent error affecting
piping reliability is the design or construction error first revealed, say, several years after
commercial operation began. A root cause of such an error could be lack of design
knowledge. Another example of latent human error affecting piping reliability is themaintenance and ISI-policy that does not acknowledge existing, generic operating
experience with a particular type of piping system. By contrast, effects of an active
human error are felt almost immediately; e.g. water hammer due to improper postmaintenance restoration of a piping system.
Studies have been performed to assess the human error contributions to piping failure[18][19]. Hurst et al. [18] analysed piping failures in the chemical process industry. This

NEA/CSNI/R(2018)12

British study shows that “operating error” was the largest immediate contributor topiping failure (30.9% of all known causes). Overpressure (20.5%) and corrosion(15.6%) were the next largest categories of known immediate causes. The other major
areas of human contribution to immediate causes were human initiated impact (5.6%)
and incorrect installation of equipment (4.5%). The total human contribution to
immediate causes was therefore about 41%. For the underlying causes of piping failure,
maintenance (38.7%) and design (26.7%) were the largest contributors. The largest
potential preventive mechanisms were human factors review (29.5%), hazard study(25.4%) and checking and testing of completed tasks (24.4%). A key conclusion of thestudy was that based on the data analysis, about 90% of all failure events would be
potentially within the control of management to prevent.

3.3.1. D&C event sequence diagram
Service-induced degradation of reactor components results from synergies among
material characteristics, loading (e.g. stress riser), and environmental conditions (e.g.
flow conditions, water chemistry). Through-wall pipe flaws involve initiation andincubation. That is, a pre-existing flaw acting as a stress riser (e.g. an embedded slag
inclusion in a weld) or is exposed to an adverse environment that eventually progresses
to a through-wall defect or non-through-wall defect that is connected to the insidesurface of a pipe. The majority of pipe flaws that result in a corrective action (repair or
replacement) are attributed to a readily identifiable active degradation mechanism or an
off-normal loading condition. An example of the latter can be high-cycle mechanicalfatigue caused by a failed pipe support.
A relatively small subset of all recordable or rejectable pipe failures involves a preexisting defect that over long time grows and is detected through a surface examination(e.g. visual examination or liquid penetrant testing). The D&C event sequence diagramin Figure 3.3 illustrates the classification of weld flaws for which no active degradation
mechanism is present. Code class one welds are subjected to pre-service inspection(PSI) and rejectable flaws are repaired. There is some likelihood that a pre-existing flaw
is not detected, however. Again, an in-service inspection (ISI) may/may not detect aweld defect. If successfully detected, the weld defect is evaluated per ISI program
acceptance standards. Continued operation is possible if repair/replacement isperformed or some degradation mitigation is implemented.
Assuming that a pre-existing flaw is discovered during an ISI but remains unmitigated(i.e. no repair is performed) then crack growth may occur given that it is subjected to an
adverse operating environment (high temperature and corrosive) and the material issusceptible to degradation. The potential of through-wall cracking (TWC) would behigh if the conjoint requirements for degradation are met (“TWC Potential-H” in Figure3.3). In developing an initiating event frequency model, pipe failure rates and rupture
probabilities are derived for all piping components within the evaluation boundaries.
The piping reliability parameter estimation process considers all credible damage anddegradation mechanisms that apply to an evaluation boundary. For locations without
any readily identifiable damage or degradation susceptibility an assumption is made thata pre-existing weld flaw may exist and eventually grow in the through-wall direction.
The flaw growth mechanism is termed “low-cycle fatigue and pressure loading” (LCFAT) and accounts for the effects of normal operation including cool-down and heatcycles. Therefore, a CRP model is needed to resolve the LCF analysis cases. Figure 3.3
displays two event sequence paths that are highlighted in red that represent the highlikelihood of a pre-existing defect growing into a through-wall crack (TWC).

│ 29

30 │ NEA/CSNI/R(2018)123.3.2. Effect of welding process on degradation susceptibility
The operating experience data classification in CODAP differentiates between field
welds and shop welds. High-level database summaries are included in Figures 3.4
through Figure 3.6. In Figure 3.4 the entire database content is organised according thelocation of a flaw; base metal or weld metal/weld-HAZ. Limited to Code class one
piping, Figure 3.5 shows the number of failure records by weld type. Figure 3.6 shows
the results of the following database queries:
field weld/shop weld failures for which no active degradation mechanism
contributed to the flaw discovery;
field weld/shop weld failures attributed to an active degradation mechanism withor without contribution from an initial weld defect.

••

Figure 3.3. D&C event sequence diagram

Preexisting Flaw - Not
Detected by PSI

ISI / NDE Performed?

ASME XI Inspection Rejectable DefectDetected

ASME XI Inspection Recordable DefectDetected

Mitigation Applied /Crack Growth Arrested

YES

Environmental
Degradation - Adverse
Environment

Environmental
Degradation - Susceptible
Material
TWC Potential - H

YESNO
Acceptable for continued
operation

TWC Potential - M

TWC Potential - M

NOAcceptable for continued
operation

TWC Potential - L

TWC Potential - M
Acceptable forContinued Operation

TWC Potential - L

TWC Potential - L

TWC Potential - L

NEA/CSNI/R(2018)12

Figure 3.4. Pipe Failure by Component Type
10000

No. Failure Records

1000

100

10
Through-Wall Flaw
Part Through-Wall Flaw

Weld-HAZ / Weld Metal

Tee

Reducer

Pipe - Downstream of Bend, Elbow, Valve

FlangeFitting

Expansion Joint

ExpanderBend / Elbow

Butt Weld/Socket Weld - Shop Welded

Butt Weld/Socket Weld - Field Welded1Base Metal

Several qualitative insights can be drawn from these query results. For small-bore
piping, only a relatively few failures of shop welds have been reported. This is to be
expected since the majority of the small-bore lines (e.g. drain lines, instrument lines,
vent lines) consist of field welded piping/tubing. The majority (89%) of all reported
shop weld failures involve piping of nominal pipe size greater than DN100 (4”) and withan active degradation mechanism as the “apparent cause of failure.” Further review ofthis larger subpopulation reveals that approximately 79% of the shop weld failures
involve boiling water reactor (BWR) Reactor Recirculation system welds that havefailed due to intergranular stress corrosion cracking (IGSCC); Figure 3.7.

│ 31

32 │ NEA/CSNI/R(2018)12
Figure 3.5. Safety Class 1 Pipe Failures

NEA/CSNI/R(2018)12│ 33

Figure 3.6. Operating Experience with Safety Class 1 Field & Shop Welds

No. Failure Records

1000

100

Weld - Field Welded
Weld - Shop Welded101ø ≤ 2"

2" < ø ≤ 4"

ø > 4"

ø ≤ 2"No Active DM:

2< ø ≤ 4"

ø > 4"

Active DM

Weld - Field Welded

79

17

61

1106

333

1455

Weld - Shop Welded

2

2

26

19

16

277

34 │ NEA/CSNI/R(2018)12

Figure 3.7. BWR Safety Class IGSCC Failure Data
BWR Reactor Recirculation Piping Welds

Shop Weld

Field Weld

0

100

200

300

400

500

600

700

800

No. Failure Records
Non-Through-Wall Flaw

Through-Wall Flaw

In BWR plants with external recirculation loops, approximately 40% of the welds areshop welded. In contrast to the BWR plants and with the exception for pipe-to-safe-end
welds, all code class one pressurised water reactor (PWR) reactor coolant system (RCS)
hot leg, cold leg and surge line welds are shop welded.
The majority of the recorded shop weld failures have occurred in large-bore stainlesssteel piping susceptible to IGSCC. This form of stress corrosion cracking develops as
the result of weld sensitisation of the weld heat-affected zone and weld metal hot
cracking, tensile stresses and corrosive environment. The welding process induces
residual tensile stress that is detrimental to fatigue life. Tensile stresses act to stretch orpull apart the surface of the material. With enough load cycles at a high enough tensilestress, a metal surface will initiate a crack. Weld residual stresses have been measured
in a variety of stainless steel grades and pipe sizes using strain-gage and X-ray methods.
These residual stresses are a major contributor to the overall tensile stresses acting on apipe weldment.
Sensitisation-related remedies include solution heat treatment, corrosion-resistantcladding, and alternative pipe materials. Solution heat treatment is used for shop welds.
It reduces or eliminates weld sensitisation, residual stresses, and the effects ofmachining and grinding.
The welding process affects the final weld microstructure and hence the mechanicalproperties. For shop welds in BWR primary system piping, submerged arc welding(SAW) has been used. Field welding typically has used shielded metal arc welding(SMAW) and gas tungsten arc welding (GTAW). Typically, manual GTAW has beenused for root pass followed by either SMAW or automated GTAW. The best balance ofNEA/CSNI/R(2018)12

strength and toughness is produced by GTAW followed by SMAW. According to someresearchers, the SAW method produces an acceptable but less desirable balance of
mechanical properties.
The weld preparation method influences IGSCC initiation and propagation. In preparing
the piping for welding, the inside surface is machined or ground to match the two pipe
pieces to be welded together. Depending on the pipe diameter, post-weld grinding may
also be used to clean up the weld for inspection. The method of surface preparation willaffect the surface residual stresses. Heavy machining and grinding will result in cold
working, high residual stresses, and areas of stress intensification. These cold-worked
areas will also result in surface re-crystallisation during welding. All of these changes
enhance the initiation process of IGSCC.
To explore the positive effect of a good welding process on the IGSCC susceptibility
the available BWR-specific IGSCC operating experience is explored in further detail.
Extensive service experience data exists on BWR code class one stress corrosion
cracking incidents in large-bore stainless steel piping.
It is noteworthy that reactor coolant purity and corrosion potential can have a marked
effect on the cracking susceptibility. Changes in water conductivity and in oxidisingconditions have changed markedly over time. Early operations under poor water purity
control can have a marked effect on the cracking susceptibility even upon subsequent
improved purity control. Thus, the times to crack detection under these conditions are
not comparable with those obtained where the water purity has been maintained
throughout operations. Calculated prior and posterior weld failure rates are included in
Figure 3.8, and best estimate weld failure rates for different pipe sizes and non-throughwall (NTW) cracks are included in Figure 3.9. These simplified calculations do notaccount for plant-to-plant variability in water purity control, however.
Figure 3.8. Calculated Failure Rates for NPS28 Stainless Steel Welds

Cross-to-reducer

Posterior (After 25 Years of Operation)
Prior (1st 15 Years of Operation)

Pipe-to-tee

Pipe-to-pipe

Pipe-to-pump

Pipe-to-safe-end

Nozzle-to-Safe-end

Pipe-to-valve

Elbow-to-pipe
1.0E-06

1.0E-05

1.0E-04

λ [1/weld-yr.]

1.0E-03│ 35

36 │ NEA/CSNI/R(2018)12
Figure 3.8. Calculated Weld Failure Rates Conditional on Crack Depth
1.0E-02
NPS12
NPS22
NPS28

λNTWC [1/Weld.Yr]

1.0E-03

1.0E-04

1.0E-05

1.0E-06

> 10%

> 20%

> 30%

> 40%

> 50%

> 60%

> 70%

> 80%

> 90%

a/t-Ratio

For shop welds, two cases are evaluated: 1) “LTA WPS”, and 2) “Screened Data.” A
careful review of the post-1989 service experience data reveals an interesting subpopulation for which detailed root cause analysis results are available. At three multiunit BWR plant sites, a shared, less-than-adequate (LTA) welding process specification(WPS) had been applied. Excessive grinding and polishing of the inside diameter of theweldments generated high residual stresses that contributed to failure of these welds.
Case two, with this subpopulation is screened out, represents a projected weld failure
rate for “IGSCC susceptible environment” and high quality welding process.
The event population data for calculation case #two (posterior weld failure) issummarised in Figure 3.10. The failure data is organised by time period (failures during
1970 through 1988 and failures during 1989 to date) and pipe size. Splitting of the datain two time periods is done to address the impact of IGSCC mitigation strategies post1988 (implementation of U.S. NRC Generic Letter 88-01 [2[20]]). These mitigation
strategies include improved primary water chemistry and transition from Normal Water
Chemistry (NWC) to Hydrogen Water Chemistry (HWC). In modelling the field weld
and shop weld reliability an attempt was made to filter out the effects on IGSCC
susceptibility by water chemistry and weld stresses.

NEA/CSNI/R(2018)12

Figure 3.90. BWR Reactor Recirculation System Weld Failure Data
400

350

No. Failure Records

300

250

200
1970-1988
1989-2013
150

100

50

0

2" < ø ≤ 4"

4" < ø ≤ 10"

ø > 10"2" < ø ≤ 4"

Field Welded4" < ø ≤ 10"

ø > 10"Shop Welded

1970-1988

26

7

364

2

21

103

1989-2013

5

54

154

0

2

65

In calculating the FW and SW failure rates the exposure term consists of respective weld
population multiplied by the number BWR plant years of operation. Only BWR plants
with external recirculation loops are considered. There is plant-to-plant variability in theweld populations. On average, about 40% of the total Reactor Recirculation (RR)
system weld population consists of shop welds. In this analysis, an isometric drawing
review was performed on three plants, resulting in a mid-point value of 71 shop welds
and 123 field welds. At the time of this analysis there were 2,890 reactor operating yearsof external RR-loop plant experience; 936 reactor years for the period 1970-1988 and
1,900 reactor years for the period 1989-2013.
The results of the weld failure calculation are summarised in Figure 3.11. The shop weld
failure population > 10” diameter for the period 1989-2013 includes a significant
subpopulation (32 records) consisting a small group of plants for which the root cause
analyses pointed to a deficient weld procedure specification resulting in excessive cold
working during the fit-up process. Screening out this subpopulation does impact thecalculated weld failure rate.

│ 37

38 │ NEA/CSNI/R(2018)12
Figure 3.11. Calculated Field Weld & Shop Weld Failure Rates (> 10”)

Weld Failure Rate [1/Weld.Year]

1.0E-02

1.0E-03
6.63E-04
4.83E-04

2.46E-04

1.0E-04

1.0E-05

RR-FW
Field Weld

RR-SW - LTA WPS

RR-SW - Screened Data
Shop Weld

3.4. Fatigue of piping components
According to ASTM International 9, fatigue is “the process of progressive localised
permanent structural change occurring in material subjected to conditions which
produce fluctuating stresses and strains at some point or points and which may culminate
in crack or complete fracture after a sufficient number of fluctuations.” The CODAP
event database captures four basic types of fatigue mechanisms: 1) corrosion-fatigue (CF), 2) low-cycle fatigue (LCF), 3) thermal fatigue (TF), and 4) high-cycle fatigue (HCF).
Approximately 80% of the recorded fatigue failures recorded in CODAP are due tovibration fatigue of small-diameter butt welds and socket welds.
Corrosion-fatigue is (C-F) the result of the combined action of alternating or cyclic
stresses and a corrosive environment. The fatigue process is thought to cause rupture of
the protective passive film, upon which corrosion is accelerated. In a corrosive
environment the stress level at which it could be assumed a material has infinite life islowered or non-existent. Contrary to a pure mechanical fatigue, there is no fatigue limit
load in corrosion-assisted fatigue. Much lower failure stresses and much shorter failure
times can occur in a corrosive environment than in a non-corrosive environment. The
fatigue fracture and the cracks are most often transgranular, but not branched. Mitigationof corrosion-fatigue can be accomplished using corrosion resistant materials in
combination with stress reduction strategies, for example by stress redistribution.
In CODAP the corrosion-fatigue event population is small; a total of 27 recorded events
most of which involve recordable/rejectable indications per the definitions of ASME
9. Known until 2001 as the American Society for Testing & Materials

NEA/CSNI/R(2018)12│ 39

Section XI (Rules for Inservice Inspection of Nuclear Power Plant Components). An
example of material degradation induced by corrosion-fatigue is that found throughmetallographic examination of cracked code class two feedwater reducers at the now
decommissioned Trojan nuclear power plant in 1987. The most significant corrosionfatigue failure to date is the major primary coolant leakage that occurred at the Russianplant Kola unit two on 3 March 1994 when a two-inch make-up system pipe ruptured 10.
In the CODAP event database, the term “low-cycle fatigue” is used to characterise crack
growth in the pipe through-wall direction through applied stress and normally occurringcooldown/heatup cycles. An underlying assumption is that of a pre-existing weld flaw
attributed to original construction, fabrication or welding defects missed by pre-service
inspections and/or subsequent in-service inspections. According to the Coding
Guideline, the following conditions must be met for an event to be classified as LCF: 1)
no active environmental degradation mechanism can be identified, and 2) the root cause
evaluation points to presence of a weld flaw such as lack of fusion is one possible cause.
Approximately 8% of the fatigue failures recorded in CODAP are attributed to LCF andof this event population, about 75% involve piping of ≤ DN50.
Thermal fatigue is due to the cyclic stresses that result from changing temperature
conditions in a component or in the piping attached to the component. Thermal fatigue
may involve a relatively low number of cycles at a higher strain (e.g. plant operational
cycles or injection of cold water into a hot nozzle) or due to a high number of cycles at
low stress amplitude (e.g. local leakage effects or cyclic stratification).
High-cycle fatigue (HCF) involves a high number of cycles at relatively low stressamplitudes (typically below the material’s yield strength but above the fatigue
endurance limit of the material). The crack initiation phase is considered to be dominant,since crack growth is usually fairly rapid. High cycle fatigue may be due to vibration orpressure pulses or due to flow-induced vibration (FIV). FIV can induce high-cycle
fatigue (HCF) through interaction with flow adjacent to the component or within thesystem, establishing a cyclic stress response in the component. Power uprate is also ofconcern as an increase in flow may change the vibrational characteristics of the system
and in the worst case excite a HC mode where a resonant frequency is achieved. Of thetotal high-cycle fatigue event population, approximately 40% involve socket weld
failures. The CODAP event database addresses the potential negative influences ofpower uprate on fatigue tolerance. To date only a handful of database records have beenclassified as failures due to FIV and as a direct consequence of power uprates, however.

3.4.1. Fatigue-induced EHC piping failures
The Electro-Hydraulic Control (EHC) system is a non-safety-related system consisting of smalldiameter piping. Failures of EHC piping have resulted in reactor trip and safety system actuation. From
its inception, CODAP has collected operating experience data involving small-diameter (≤ DN25)
piping failure. Turbine trip and reactor trip following an EHC pipe break are not uncommon
occurrences. The observed impact of EHC pipe failure on plant operation is illustrated in Figure 3.12.
In the CODAP Event Database the EHC pipe failure population consists of 83 records [[5]. In Figure
3.13 the EHC pipe failure population is organised by damage and degradation mechanism.

10. International Atomic Energy Agency, Draft Report of a Consultants Meeting on a Primary
Coolant Leak at Kola two NPP Due to the Rupture of a Make-up Pipe. WWER-SC-112, Vienna,
Austria, 1995.

40 │ NEA/CSNI/R(2018)123.4.2. Socket weld integrity management
There have been frequent occurrences of high-cycle fatigue failures of socket welded connections insafety related piping systems; Figure 3.14. NUREG-1801 [[21] documents a technical basis fordetermining the adequacy of ageing management programmes (AMPs) for license renewal. Section
XI.M35 of this reference augments the requirements in ASME Section XI, 2004 Edition (Rules for Inservice Inspection of Nuclear Power Plant Components). According to Table IWB-2500-1 of the ASME
Code, an external surface examination of small-bore class one piping should be included for piping lessthan DN100. Other ASME Code provisions exempt from examination 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 testing of all class one piping.
According to the USNRC [[21], for a one-time inspection to detect cracking resulting from thermal andmechanical loading or intergranular stress corrosion of full-penetration welds, the inspection should be
a volumetric examination. For a one-time inspection to detect cracking in socket welds, the inspection
should be either a volumetric or opportunistic destructive examination. Opportunistic destructive
examination is performed when a weld is removed from service for other considerations, such as plantmodifications. A sampling basis is used if more than one weld is removed. These examinations provide
additional assurance that either ageing of small-bore ASME code class one piping is not occurring or
the ageing is insignificant, such that a plant-specific ageing management program (AMP) is notwarranted and is applicable to small-bore ASME code class one piping and systems less than DN100
and greater than or equal to DN25. The programme includes pipes, fittings, branch connections, and all
full and partial penetration (socket) welds.
Figure 3.12. Impact on Plant Operation by EHC & Instrument Air (IA) Pipe Failure

Manual Shutdown or Power Reduction

LCO Entry / None

Expanded Outage Work Scope

ESF/RPS Actuation

Dual Unit Shutdown

0

10

20

30
IA

EHC

40

50

60

70

NEA/CSNI/R(2018)12│ 41

Figure 3.13. EHC & IA Piping Degradation Mechanisms
Overstressed / Severe Overloading
Low-Cycle Fatigue Pressure Loads
IGSCC

IA

High-Cycle Vibration Fatigue
HF:CONST/INST
Fretting
ErosionCorrosion-External / Pitting
Overstressed / Severe Overloading
Low-Cycle Fatigue Pressure Loads

EHC

High-Cycle Vibration Fatigue
HF:CONST/INST
Fretting
Corrosion-Fatigue
0

10

20

30

40

50

60

No. Failure Records
Part Through-Wall Flaw

Through-Wall Flaw

Major Structural Failure

70

80

90

100

42 │ NEA/CSNI/R(2018)12
Figure 3.14. Socket Weld Failure Summary

3.5. Corrosion fatigue
Corrosion fatigue or “environmentally assisted fatigue” is the behaviour of materials
under cyclic loading conditions and in a corrosive environment. It is considered to bemade up of a region (or life) associated with the formation of an engineering-sized crack
and a region consisting of the growth of this crack up to component failure. One category
relates to the cycling life for the formation of a fatigue crack in a smooth test specimen,
the so-called S-N fatigue properties; (stress versus number of cycles). The second relates
to the growth of a pre-existing crack. Laboratory test have shown that LWR coolant
water can have a detrimental effect on both S-N fatigue properties and fatigue crack
growth. Much lower failure stresses and much shorter failure times can occur in a
corrosive environment compared to the situation where the alternating stress is in a noncorrosive environment.
Corrosion fatigue should not be confused with stress corrosion, which is crack initiationand growth under sustained load or residual stress. Corrosion fatigue is a mostlytransgranular crack growth phenomenon. The corrosion fatigue fracture is brittle andthe transgranular cracks are not branched. The corrosive environment can cause a fastercrack growth and/or crack growth at a lower tension level than in dry air. Even relatively
mild corrosive atmospheres can reduce the fatigue strength of aluminium structuresconsiderably, down to 75 to 25% of the fatigue strength in dry air. No metal is immune
from some reduction of its resistance to cyclic stressing if the metal is in a corrosive

NEA/CSNI/R(2018)12

environment. Control of corrosion fatigue can be accomplished by either lowering thecyclic stresses or by various corrosion control measures.
Results from laboratory tests generally reveal a detrimental effect of BWR and PWR
water environments on the fatigue lives of specimens made from carbon steels, lowalloy steels, austenitic stainless steels and nickel (Ni)-based alloys. The parameters
predominantly affecting the fatigue life of laboratory specimens are strain rate,
temperature, dissolved oxygen concentration in the water and Sulphur content of thematerial, the latter of which is only applicable for carbon steels and low alloy steels.
The detrimental effects of reactor environments on fatigue lives have been known for
more than 30 years. Reactor coolant pressure boundary components exposed to thereactor water environment have exhibited degradation due to environmentally enhanced
fatigue in service. In all these cases, unacceptable component fabrication, material
selection, or plant operation (and combinations of these) were identified as root causes
leading to the degradation. Significant large-scale, generic degradation due toenvironmental fatigue has not been observed in service even though environmental
effects due to the impact of light water reactor (LWR) coolant were not explicitly
considered in current design rules. NRC investigation of the risk associated withcorrosion fatigue in the Fatigue Action Plan concluded that there was no inherent risk
to core damage frequency for operating nuclear reactors, although increased probabilityof leakage indicates this issue requires management for extended plant operation [[22][[23].
Limited observations of cracking due to corrosion fatigue stand in contrast to significant
occurrences of stress corrosion cracking in stainless steels and Ni-based alloys, whichhave been observed more systematically in reactor coolant pressure boundary welds andreactor internals from LWR plant operational experience worldwide.
The lack of significant observed degradation in plant components with regard tocorrosion fatigue is attributed, at least in part, to the generally conservative design
requirements adopted within the ASME code and applicable regulations (e.g. the NRC’s
requirement to keep the cumulative usage factor less than 0.1 for break exclusion
locations). Margins in the design requirements appear to compensate for the detrimental
environmental effects.
Another consideration when comparing the environmental effects between laboratory
and service components is the applied loading associated with pressure and thermaltransients. Laboratory testing typically relies on simple mechanically-controlled loadingtransients (e.g. artificially shaped waves), and may arguably include some amount ofcompensation for the effects of more complex thermal transient loading. Additionally,
plant components are often subjected to thermal transients with long-lasting hold times
at almost constant load or temperature corresponding to steady-state operating
conditions which may lead to some strain recovery within the component. These
differences may affect fatigue lives.

3.6. Flow-assisted degradation
The term “flow-assisted degradation” (FAD) encompasses several phenomena, all ofwhich result in the degradation of piping through material loss. These phenomena
include erosion, erosion-cavitation (E-C), erosion/corrosion (E/C), flow-accelerated
corrosion (FAC), and liquid droplet impingement erosion. The distinguishing│ 43

44 │ NEA/CSNI/R(2018)12
characteristics of the different FAD-mechanisms are illustrated in Figure 3.15. This
figure represents a high-level guidance for event classification and coding.
Figure 3.15. Distinguishing Characteristics of the Different FAD-Mechanisms

3.6.1. Erosion-cavitation
Erosion-cavitation (E-C) is the process of surface deterioration and surface material loss
due to the generation of vapour or gas pockets inside the flow of liquid. These pockets
are formed due to low pressure well below the saturation vapour pressure of the liquidand erosion caused by the bombardment of vapour bubbles on the surface. Erosioncavitation usually involves an attack on the surface by gas or vapour bubbles, creating
a sudden collapse due to a change in pressure near the surface. Low pressure (below thesaturated vapour pressure) is generated hydrodynamically due to various flow
parameters, such as liquid viscosity, temperature, pressure and nature of flow. This
deterioration is initiated by a sudden surge of bubbles hammering the surface, resulting
in deformation, as well as pitting.

3.6.2. Erosion-corrosion & liquid droplet impingement erosion
Erosion-corrosion is a mechanism of material loss by mechanical means due toimpingement, abrasion or impact, etc., resulting from the movement of a liquid or gas
over the surface of a metal coupled with corrosion. This type of degradation ischaracterised by attack like small pits with bright surfaces free from corrosion products.
These pits often have the form of a horseshoe with the nib pointing in the current

NEA/CSNI/R(2018)12

direction. Erosion-corrosion may occur where the velocity of liquid is too high. Most
exposed are places where there are effects of turbulence, e.g. joints, bends etc. The
corrosion rate will accelerate if the liquid contains gas bubbles and/ or solid particles.
Systems susceptible to erosion-corrosion include raw water cooling systems; e.g.
Circulating Water and Service Water systems.
Liquid droplet impingement (LDI) erosion is a subset of erosion-corrosion. Liquid
droplets are often generated in piping that operates in a two-phase flow condition, andis due to the entrainment of liquid water from the upstream and also by the heat transfer
through the pipe wall. In the region behind the orifice and the valve in the pipeline, thevelocity of the droplets is highly accelerated due to the contraction effect. This resultsin the occurrence of high impact pressure on the inner surface of the pipe due to theliquid droplet impingement. The impact pressure of the droplets increases as high as
several hundred MPa, which is beyond the elastic limit of the pipe-wall material, so thatthe pipeline is often damaged by the impact pressure of droplets. This phenomenon iscalled liquid droplet impingement (LDI) erosion. In general, the LDI occurs on thedorsal side (extrados) of a bend or elbow, where the droplets cannot follow the steam
flow due to the inertia of the droplets.

3.6.3. Flow-accelerated corrosion
The first CODAP Topical Report [4] provides details on the flow accelerated corrosion(FAC) mechanism. FAC leads to wall thinning (metal loss) of steel piping exposed toflowing water or wet steam. The wall thinning is the result of the dissolution of thenormally protective oxide layer formed on the surfaces of carbon and low alloy steel
piping. The rate of metal loss depends on a complex interplay of several parameters
including water chemistry, material composition, and hydrodynamics, but based on
operating experience the metal loss can be as high as 3 mm/yr. Carbon steel piping
components that carry wet steam are especially susceptible to FAC. The most dominantvariables are temperature, fluid velocity, fluid pH, the water amine and oxygen content,
steam quality, void fraction of the fluid, piping geometry, and the pipe material
composition.

3.7. Stress corrosion cracking
Stress corrosion cracking (SCC) or environmentally assisted cracking (EAC) is mainly
observed in the weld deposit and heat affected zone and it is considered that it occurs
due to synergistic effect of three factors of material, stress and environments. SCC may
occur when a susceptible material is subjected to stress in a corrosive environment. One
example of a scenario that might lead to SCC is one in which a weldment is (1) sensitised
due to high heat input, (2) subjected to high local stresses such as welding residual
stresses, and (3) the weldment is subjected to a corrosive environment. There are five
types of SCC mechanisms:
•
•
•
•
•

intergranular SCC (IGSCC) of stainless steel;
IGSCC of nickel-base alloys, typically referred to as primary water SCC(PWSCC);
irradiation assisted SCC (IASCC) of stainless steel;transgranular SCC (TGSCC), including external chloride-induced SCC(ECSCC) of stainless steel;strain induced SCC (SICC) of high-strength carbon steel.

│ 45

46 │ NEA/CSNI/R(2018)123.7.1. Intergranular stress corrosion cracking
Intergranular Stress Corrosion Cracking (IGSCC) of stainless steels is a time dependent
type of ageing phenomenon. The IGSCC morphology is associated with thetemperature/time fabrication conditions that gave rise to thermal sensitisation and theformation of chromium carbide precipitation (e.g. M23C6) and chromium depletion at
the grain boundary. The reduction in chromium concentration adjacent to the grainboundary gives rise to a reduction in passivity and makes the material susceptible tointergranular stress corrosion cracking.
Since the late 1970s the importance of water purity control became increasingly
apparent, especially with regard to creviced components (where the geometry andoxidising conditions in the bulk environment could give rise to increased anionicactivity in the creviced region), even though the bulk water purity was acceptable at thattime. This water purity aspect was of importance for environmentally assisted cracking
of stainless steel, low alloy pressure vessel steels and nickel-base alloys.
Subsequent to the introduction of low carbon and stabilised grades of stainless steel,
IGSCC has occurred in these materials that were clearly not in a sensitised condition. It
has been shown that their susceptibility to IGSCC is due to cold work induced during
fabrication. In many cases the initial cracking was found to be initially transgranularthen changing to an intergranular cracking mode. The initial transgranular cracking isoften associated with a surface layer of cold work induced by grinding.
Failures have also occurred where IGSCC is attributed to the presence of either severe
bulk cold-worked material (cold bent piping). The mechanism by which cold work
renders austenitic alloys susceptible to IGSCC in BWR environments is not fully
understood and is still being investigated. It is possible that there is an unfavourable
interaction between deformation-induced martensite, high residual stresses and strains,and localised deformation.

3.7.2. Primary water SCC
Alloy 600 (InconelTM 600), a nickel-base metal, was developed in the 1950s for use asa construction material for nuclear power plants. The material was qualified for use innuclear power plants because of its perceived resistance to SCC; it was viewed as analternative to Type 304 or Type 316 austenitic stainless steels. An early (possibly theearliest) recorded instance of SCC of Alloy 600 material is that of the failed inspection
tubes in the Swedish Ågesta Reactor 11 in September 1964 [24]. According to Reference[25], the materials research in the early 1960s concluded that nickel-base materials withhigh nickel content (> 72%) to be resistant to SCC in chloride and alkaline
environments. The Alloy 600 material was subsequently qualified as structural material
for use in PWR plants. A first Alloy 600 failure in a commercial nuclear power plant
occurred in July 1972 12 when the German plant KWO Obrigheim13 experienced a steam
generator tube through-wall leak attributed to SCC. Due to the SCC failures of Alloy
600 in direct connection with primary water, many PWR and BWR owners have
11. A combined district heating and power reactor sited below-ground near Stockholm, Sweden.
The reactor was permanently shut down in 1974.
12. Der Bundesminister des Innern, Besondere Vorfälle in Kernkraftwerken in der
Bundesrepublik Deutschland, Berichtzeitraum 1965-1976, Bonn, Germany, 1977.
13. A small 2-loop PWR commissioned in 1969 and permanently shut down in 2005.

NEA/CSNI/R(2018)12

replaced it with alternate materials such as Alloy 690, a higher nickel based alloy, orAlloy 800, an iron based alloy. CANDU users now prefer Alloy 800NG.
Nickel base alloys (e.g. Alloy 600, and corresponding weld metals Alloy 82, 132 and182), have proved to be generically susceptible to IGSCC in normal specification PWR
primary water systems (PWSCC). Recent operational experience shows that the
fabrication induced residual stresses have a large influence on PWSCC in alloy 600
weld metal. Examples of components affected include pressuriser, hot leg, cold leg,drain, and reactor coolant pump nozzle-to-safe end dissimilar metal welds, penetrations
welded to the reactor vessel and reactor vessel head and steam generator.
PWSCC in the weld metal grows along the grain boundaries of columnar crystal
dendrite packets. Initiation in the weld metal is often thought to be the result of typicaland non-typical fabrication processes leading to locally high residual stresses, or surface
stresses from, for example, grinding. To date it has been found that the susceptibility to
SCC of nickel-based alloy weld metal is higher than that of the base metal. IGSCC of
ni-base alloys in BWRs is believed to be attributed to Cr depletion at grain boundaries,
similar to IGSCC in thermally sensitised stainless steels.

3.7.3. Irradiation assisted SCC
Irradiation assisted stress corrosion cracking (IASCC) is essentially a time dependent
type of ageing phenomena characterised by the threshold of irradiation level related tosusceptibility of IASCC. There are increasing concerns that it might occur in the highfluence region if no countermeasures could be conducted. IASCC requires stress,
aggressive environment and a susceptible material. However, in the case of IASCC, anormally non-susceptible material can be rendered susceptible by the accumulation ofneutron irradiation and has highly time dependency compared to the other SCC
mechanisms.
IASCC is therefore an ageing mechanism that affects reactor vessel internals in both
BWR and PWR plants. Neutron irradiation effects are primarily thermal but, in the caseof gamma heating of thick section, the higher temperatures generated can have asignificant effect on void swelling. In addition, neutron capture reactions induce
transmutation reactions and hence changes in chemical composition of the material.
Irradiation hardening and radiation induced segregation (RIS), due to chromium
depletion and silicon enrichment at grain boundaries, are considered to be the mostprobable factors leading to IASCC susceptibility.

3.7.4. Transgranular stress corrosion cracking
The earliest indications of cracking in unirradiated austenitic stainless steels occurred in
the late 1960s in components where the temperature was <100OC and this was observed
during storage and fabrication, and operation. The degradation mode was transgranular
stress corrosion cracking (TGSCC) on the outside surface of the pipe. This failure mode
was exacerbated by (a) chloride contamination from humid marine environments orfrom insulation, and (b) the dissolved oxygen (air) in the water or condensate. These
cracking incidents were effectively managed by appropriate control of the chloride│ 47

48 │ NEA/CSNI/R(2018)12
contamination and by taking into account the beneficial effect of soluble silicate
originating from the glass fiber insulation (USNRC, 1973) 14.
As with other types of SCC, TGSCC requires stress, an aggressive environment and asusceptible material. In addition, it requires the presence of chloride contamination orother halide anions such as fluorides, and may occur even in materials in the solution
heat treated condition. It is generally a problem that initiates on the outside surfaces ofcomponents mainly due to lack of attention to adequate cleanliness (also known as
external chloride stress corrosion cracking).
TGSCC has also occurred from inner surfaces, mainly in pipe sections containing
stagnant two phase coolant, where evaporation and concentration of chlorides can occur.
Wetting due to condensation or nearby water leaks allows an aqueous environment toform that leads to TGSCC, usually accompanied by pitting or crevice corrosion. The
stress required for chloride induced TGSCC is relatively modest, the threshold being
close to the proportional yield strength of solution annealed austenitic stainless steels.
Implementation of the known adequate procedures to ensure appropriate surface
cleanliness is a continuing necessity that requires careful management attention at allstages of construction and operation of nuclear power plants. External chloride stresscorrosion cracking (ECSCC) is TGSCC initiated on the outside surface of a component
due to the presence of chloride in sea salt, coatings, etc. attached to the material surfaces
and by perspiration.

3.7.5. Strain induced corrosion cracking
Strain induced corrosion cracking (SICC) [26] is used to refer to those corrosion
situations in which the presence of localised dynamic straining is essential for crack
formation to occur, but in which cyclic loading is either absent or restricted to a very
low number of infrequent events. SICC has been observed in pressurised components
in German NPPs made of higher-strength ferritic carbon steel. This kind of degradation
has caused circumferential cracking in feed water nozzle regions and at welds and axial
cracking in pipe bends but also in straight sections of thin-walled piping in German
BWRs.

3.8. High-level database summary
Examples of the CODAP event database content is summarised in Figures 3.16 throughFigure 3.19. Figure 3.16 is a summary of significant PWR reactor coolant system (RCS)
leak events and includes piping and non-piping RCS components. Figure 3.17 is an
overview of the database content by degradation mechanism. Figure 3.18 shows thedegradation mechanism propensity normalised against IGSCC in a non-mitigated BWR
operating environment. Limited to RCS piping, Figure 3.19 is a summary of the PWSCC
operating experience.

14. Note that the potential substitution of fibrous silicate insulation with mineral wool insulation(an action that would minimise the clogging of sump pump filters during a severe accident)
would reintroduce the danger of chloride-induced TGSCC of stainless steel piping since thebeneficial effect of silicates would be removed.

NEA/CSNI/R(2018)12

Figure 3.16. Summary of Significant PWR Reactor Coolant System Leak Events

15

15. In Figure 3.16 OE data on RCP seal failures or steam generator tube failures (SGTR) are included for reference only. CODAP does notcollect data on these event types.

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50 │ NEA/CSNI/R(2018)12

Figure 3.17. Database Content by Damage / Degradation Mechanism

Fatigue

Corrosion Fatigue
Thermal Fatigue
Low-Cycle Fatigue
High-Cycle Fatigue

Corrosion

General Corrosion
Galvanic Corrosion
Crevice / Pitting Corrosion
Dealloying - Selective Leaching

Flow-Assisted
Degradation

MIC - Microbiologically Influenced Corrosion
Steam Jet Impingement Erosion (SS Material)
Erosion-Corrosion
Erosion-Cavitation
FAC - Flow Accelerated Corrosion (FAC)ECSCC (OD)TGSCCPWSCC - Ni-Base Metals

SCC

SICC - Strain Induced Corrosion Cracking
NI-SCC (CS Material)
IGSCC - Stabilized Austenitic Stainless SteelIGSCC-PWR - Unstabilized Austenitic Stainless SteelIGSCC-BWR - Unstabilized Austenitic Stainless SteelCISCC - Chloride Ion SCC (ID)
0.0%

5.0%

10.0%

15.0%

Percentage of Total Event Population

20.0%

25.0%

NEA/CSNI/R(2018)12

Figure 3.18. Normalised Piping Degradation Propensity

FAC - All BOP Systems - Non-Code

COR+E/C+E-C - Code Class 3 Raw Water Piping

FAC - Extraction SteamIGSCC - BWR-NWC - BWR-3 & BWR-4

PWSCC - Ni-Base Metals

Corrosion-Fatigue
1.0E-2

1.0E-1

1.0E+0Normalized DM Propensity

1.0E+1│ 51

52 │ NEA/CSNI/R(2018)12

Figure 3.19. Selected PWSCC Operating Experience by Location of Degraded / Failed Component 16

16. Superscript ® indicates a repair weld. This chart is a summary of specific PWSCC events. As one example, to date there have been 20 events(one through 20) involving PWSCC in Reactor Coolant System Hot Leg Steam Generator inlet bimetallic welds. Full descriptions are found in
the CODAP event database and by using the following query definition: PWR – RCS Hot Leg – Bi-metallic Weld – PWSCC – Crack Depth.

NEA/CSNI/R(2018)12

4. CODAP database structure
The CODAP Event Database is a web based, relational SQL database consisting of ca.
100 uniquely defined data fields. It is a mixture of free-format fields for detailed
narrative information, fields defined by drop-down menus with key words (or datafilters) 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 ordegradation, system name, safety class, etc. At the end of the second term the CODAP
event database included ca. 4,900 records on degraded and failed metallic piping andnon-piping passive components. Section 4 presents the scope of the event database andsummarises the database structure and main features of the online event database.

4.1. Scope of event database
The event database scope and structure, database field definitions and data input
requirements are defined in the coding guideline, which is central to the project,including database maintenance, data validation and quality control. The database
design has benefitted from a multidisciplinary approach involving chemistry,metallurgy, structural integrity and PSA expertise. The CODAP Event Database collects
service experience data on the full range of degraded conditions, from “precursors” tomajor structural failures involving metallic piping components and non-piping metallic
passive components. According to the IAEA Safety Glossary [[27], a passive
component is defined in the following way:
•
•

•
•

“A passive component is “component whose functioning does not depend on an
external input such as actuation, mechanical movement or supply of power.
A passive component has no moving part, and, for example, only experiences achange in pressure, in temperature or in fluid flow in performing its functions.
In addition, certain components that function with very high reliability based on
irreversible action or change may be assigned to this category.
Examples of passive components are heat exchangers, pipes, vessels, electricalcables and structures. It is emphasised that this definition is necessarily generalin nature, as is the corresponding definition of active component.
Certain components, such as rupture discs, check valves, safety valves, injectors
and some solid state electronic devices, have characteristics which require
special consideration before designation as an active or passive component.”

With the above definition as a basis and building on the OPDE and SCAP-SCC projectexperience, recent operating experience and associated regulatory actions, the projectreview group made further refinements and specialisations to arrive at a scope definition
as summarised in Table 4.1. Consistent with the operating procedures, the scope
definition is revisited and periodically updated. In Table 4.1, the column “Metallic, NonPiping Passive Components” captures the BWR and PWR internals as documented and│ 53

54 │ NEA/CSNI/R(2018)12
evaluated in IAEA-TECDOC-1471 [28] and IAEA-TECDOC-1119 [[29], respectively.
In CODAP the term “failure” covers the full spectrum of degraded conditions, fromrejectable flaws requiring repair or replacement to major structural failures. As an
example, ASME Section XI, Article IWA-3000 (General Requirements) [[30] defines
acceptance standards for flaws that are discovered during non-destructive examinations(NDEs). Flaws determined to be rejectable (i.e. not fit for continued operation)according to relevant NDE code are required to be repaired or replaced.

NEA/CSNI/R(2018)12

Table 4.1. Scope of CODAP Event Database 17
METALLIC PASSIVE COMPONENTSPIPING COMPONENTS
Piping - Below Ground/Concealed
Pipe - Concrete Encased Pipe
‘Bonna’ Pipe
Pipe - External Coating
Ex-RPV - In-Plant Piping (Accessible)
Pipe - Base Metal
Pipe - Cement Lined
Pipe - Epoxy Lined
Pipe - Rubber Lined
Bend
Blind Flange
Branch-Connection - Socket Welded
Branch-Connection - Stub-in Weld
Cap/End-Cap
Elbow
Elbow - Long-Radius
Elbow - 45-Degree
Elbow - 90-Degree
ExpanderExpansion Joint
Fitting
Mixing Tee
ReducerSocket Weld
Tee
Weld - Butt Weld
Weld - Dissimilar Metal Weld
Weld - Girth Weld (Full Penetration Weld)
Weld

NON-PIPING PASSIVE COMPONENTS
Reactor Pressure Vessel (RPV)
Vessel Head Penetration - PWR
Bottom Mounted Instrument (BMI) Nozzle - PWR
RPV Head Thermocoupling (T/C) Housing - PWR
RPV Head T/C Nozzle - PWR
Pressurizer
Pressurizer Heater
Pressurizer Manway Diaphragm Plate
Pressurizer Nozzle
Pressurizer Relief/Safety Valve Nozzle
RPV Internals
Baffle-Former Assembly Bolt - PWRCore Shroud Access Hole Cover Weld
Core Shroud Head Bolt - BWRCore Shroud Weld - BWRCore Shroud Tie Rod - BWRCore Shroud Support - BWRCore Spray Sparger - BWR
In-Core Instrument Tube
Jet Pump Hold-Down Beam
Jet Pump Riser
Jet Pump Support Brace
Steam Dryer - BWR
Pump
Pump Casing
RCP Turning Vane Bolt
Valve
Valve Body

17. Corresponds to drop-down menus, and as currently implemented there is no navigation tool associated with these drop-down menus│ 55

56 │ NEA/CSNI/R(2018)124.2. Database submissions
Respective National Coordinator is responsible for data submissions. The preferred
method for submitting new data to the database is via the web-based interface. Data
submissions may also be handled by e-mail with event information attached in
Microsoft Access, Excel or Word file format. New event information collected by theoperating agent on its own initiative will be included in an Excel file marked “possible
new event” and sent to the appropriate national coordinator for further consideration.
After validation, the national coordinator assumes responsibility for formal datasubmission and maintenance of the national data sets.

4.3. Database user instructions
In CODAP the data entry is managed via input forms, tables, roll down menus anddatabase relationships. The CODAP online opening screen and main work area screen
are displayed in Figure 4.1. The online version is accessible via a secure server at the
Nuclear Energy Agency headquarters. User names and passwords are provided by NEA
IT-Department upon written request by a national coordinator. The online version
includes help menus. In case of need for additional assistance, please contact the NEA
secretariat, NEA-IT and/or the operating agent. The project members work area includes
a FAQ area. Request for new key words to be added to drop-down menus should be sent
to the operating agent.
Consistent with the CODAP security levels, the work flow area of the event databasefacilitates records management including the review and approval of individual failure
records. A single data entry form is used to input failure data. Four data management
commands are included at the bottom of the data entry form; Figure 4.2. Uponcompletion of data entry the user will go to the workflow area; Figure 4.3. A record may
be marked as “draft” if additional technical details are to be added. When a record isready for review the corresponding command is invoked and an e-mail is automaticallysent from the “operator” to the “national coordinator” (NC) prompting the review
process. When a record is ready for QA by the CODAP Operating Agent (CODAP-OA)
the corresponding command is invoked and an e-mail alert is automatically sent fromthe NC to the CODAP-OA with a prompt for final review. In case corrective action isrequired, the CODAP-OA returns the data record to the NC for comment resolution.
Upon completion of the review process the CODAP-OA marks the record as“approved.” In summary:••••

CANCEL. If a record is entered in error, pressing “cancel” deletes information
added from the database.
WORKFLOW. When action is needed (e.g. review or approval), pressing“Workflow” switches screen from data entry mode to workflow area
(Figure 4.3).
SAVE. Whenever data entry is interrupted, pressing the “save” button allows
for continuation of data entry at another time.
FINISH. This is shortcut and returns data entry process to next level. As an
example, if data is uploaded by the OA, pressing “finish” returns status to“approved.” If data is uploaded by an operator, pressing the “Finish” returns
status to “Ready NC Review.”

NEA/CSNI/R(2018)12

•
•
•
•

DRAFT. This indicates work-in-process, and that an operator is in the processof filling out the data entry form. There is no other action pending.
READY FOR REVIEW BY NATIONAL COORDINATOR. Pressing thisbutton results in an e-mail notification to the National Coordinator(s) on record.
Data validation is requested.
READY FOR QA. Pressing this button results in an e-mail notification to the
Operating Agent. Upon final review, additional action by the national
coordinator(s) on record may be requested, or, the record is approved.
RETURN FOR REVIEW. Pressing this button results in an e-mail notification
to the national coordinator(s) on record. Additional data validation is requested.

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58 │ NEA/CSNI/R(2018)12
Figure 4.1. CODAP Online Opening Screen & Main Work Area Screen 18

18. The “HELP” area includes an abbreviated version of the coding guideline.

NEA/CSNI/R(2018)12

Figure 4.2. CODAP Data Entry Form

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60 │ NEA/CSNI/R(2018)12
Figure 4.3. CODAP Data Entry Form & Work Flow Area

NEA/CSNI/R(2018)124.4. Database field categories
It is intended that each CODAP database record provides a complete and unambiguous
description of a specific passive component degradation or failure. Not only shall aselected database record withstand an independent review for technical accuracy andcompleteness, it shall also be fit-for-application, either by direct or indirect databaseapplication. A direct database application implies that query results can be directly
inputted to a calculation. By contrast, an indirect application implies that additional dataprocessing is required to prepare input as specified by a stand-alone application
programme. Irrespective of intended application, database users shall have fullconfidence in passive component failure event interpretations and classifications.
Additionally, users should be able to perform data reinterpretations or reclassifications
to support new application requirements without having to retrieve additional source
data for the database records of interest.

4.5. Data entry
Currently, the CODAP 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 19:
1. General failure data. This area represents the minimum required information(Figure 4.5).
2. Flaw size information. This area is for recording flaw size (depth, length, aspectratio) and orientation (Figure 4.6).
3. In-service inspection (ISI) history. 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 is information regarding ISI program
weaknesses or failures (Figure 4.7).
4. Root cause information. This area records factors or conditions contributing toa degraded condition. Also included in this area is a field for free-format
comments on corrective actions, or other information of relevance to a specific
event (Figure 4.8).
The database is built around event narratives. The screening and classification of eachdatabase record is based on the detailed event narratives. A typical event report includes
the following information:••

Flaw description. This includes details on the date of a discovery, plant
operational state, description of how the flaw was detected, plus a summary of
the preliminary evaluation of the type and extent of the flaw (e.g. a non-throughwall crack or a small through-wall leakage) and the direct (e.g. a reactor trip) or
indirect impact on plant operation.
Non-piping passive component & piping component details. Details on the exact
location of the flaw, with a description of component dimensions, code class,

19. For illustrative purposes, the data entry illustrations are from the Microsoft® Access versionof the database. The current version of the online database uses a single input form. The online
version is currently undergoing a significant program update in which the input format will berevised. The new online version is scheduled for release in the first quarter 2018.

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62 │ NEA/CSNI/R(2018)12•

material and wall thickness (of piping), flaw size, operating temperature andpressure and/or design temperature and pressure.
Root cause determination. A root cause determination involves an evaluation of
NDE results, sometimes in combination with a destructive examination
followed by more detailed metallographic examination. The root cause
determination focuses on the identification of underlying causes of a degradation
or failure.

NEA/CSNI/R(2018)12

Figure 4.4. Event Description – Basic Information│ 6364 │ NEA/CSNI/R(2018)12
Figure 4.5. Flaw Size Information

NEA/CSNI/R(2018)12

Figure 4.6. In-Service Inspection History│ 65

66 │ NEA/CSNI/R(2018)12
Figure 4.7. Root Cause Analysis Information (Partial Screenshot)

NEA/CSNI/R(2018)12•

Results of Augmented ISI. Each country has national guidelines andrequirements for augmented inspections given the detection of a flaw. As one
example, the U.S. NRC Generic Letter 90-05 (Guidance for performing
temporary non-code repair of ASME code class one, two and three piping; 15June 1990) states:
‒

‒

‒
‒
‒
‒

When a flaw has been evaluated and found acceptable (for continued
operation using a temporary repair), the plant owner should perform anaugmented inspection to assess the overall degradation of the affected
system. The augmented inspection, performed within 15 days of detectionof the flaw, which results in a temporary non-code repair, is a part of therelief acceptance criteria of the temporary non-code repair of code classthree piping.
From the root cause determination, the most susceptible locations should
be identified. The extent of the augmented inspection depends on whether
the line is high energy or moderate energy. The failure of a high-energyline may have more severe consequences than the failure of a moderateenergy line because of the energy content. Thus, a more extensive
augmented inspection should be performed for high-energy lines.
The inspection of at least ten most susceptible (and accessible) locations
for high energy lines and at least five most susceptible (and accessible)
locations for moderate energy lines should be performed.
Flaws detected in the augmented inspection should be characterised andevaluated. A review of an augmented inspection report could reveal
additional flaws that result in new database records.
Description of the repair. Details on the type of repair (e.g. weld overlay
repair, application of a mechanical clamp, replacement in kind or
replacement using different material and or design).
Safety significance. The safety significance is based on observed impact(e.g. leak duration, leak rate, range of water/steam jet, spraying/wetting ofsafety equipment, collateral damage) and/or engineering evaluations, aswell as failure potential within or beyond design envelope.

Classifying event reports is oftentimes tedious and time-consuming and can involve
reviews of large volumes of documentation. This is especially so where an initial
discovery of a flaw results in augmented inspections and the discovery of additionalflaws in locations adjacent to the initial discovery, or similar locations but in otherpiping system trains. Once the reporting has been completed by a plant owner/operator
a single set of documents may include detailed technical information on multiple flaws,
where each flaw relates to a uniquely defined component boundary definition. As an
example, plant through-wall flaws were found in emergency core cooling system(ECCS) piping and a condition report (CR-99-0445) was issued after the discovery ofthrough-wall leaks on both ECCS pipe trains (DN600 piping) within the refuelling watertank pipe trench. Initially a single database record was created. The ECCS walk-down
inspection and initial visual examination yielded additional details:•“A” pipe header. Dry, white boric acid crystals on the upper right fillet weld that
attaches the code name plate to the piping spool, approximately 13 mm long
crack. Some “weepage” after plate was removed and area cleaned.

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68 │ NEA/CSNI/R(2018)12•••“B” header. Dry, white boric acid crystals on lower south east lug fillet weld
adjacent to the pipe clamp for support #2407-17, approximately six mm indiameter. No active leakage.
“B” header. Dry, white boric acid crystals on upper north-west lug for support#2407-17, adjacent to the lug fillet weld, less than 6 mm in diameter. No active
leakage. Linear indication approximately 13 mm long.
“B” header. Dry, white boric acid crystals found on a support member just below
support 2407-19. No active leakage (no evidence of leakage or boric acid on thepiping could be found).

Based on the above, the initial database record was modified to address the discovery
of “weepage” on the “A” header. Following the initial discovery the following technicalinformation was obtained:•••

L-99-90 (Augmented Inspections dated 7 April 1999) with the attachment PSLENG-SEMS-98-102.
Calculation No. AES-C-3566-1 (Evaluation of Corrosion Degradation of 24inch ECCS Piping at St. Lucie, Unit 2).
US NRC (24 June 1999): Relief from ASME Code Requirements Related to theInterim Relief Request No. 26 for Emergency Core Cooling System Piping for
St. Lucie Plant, unit two.

The report on the augmented inspections provided details on a total of 32 recordable
crack indications in the ECCS A- and B-train. Of these indications, two were throughwall flaws adjacent to field welds FW-3 and FW-4, respectively; both located in the
train B and two new records were added to the database. Repair of the train A throughwall flaw was completed on 7 April 1999 while repair of the train B flaws was completed
on 16 April 1999. In CODAP, all three records have 6 April 1999 as the event date. The“MER Check Box” is check-marked for the three records to ensure that a future databaseuser is made aware of the fact that the flaw discoveries are related.

4.6. Non-through-wall flaw characterisation in CODAP
The flaw characterisation area of the OE data input form consists of 34 fields. Use “enterkey” or “arrow keys” to move from one field to another. The data entry requirements
are defined below:
•

•
•
•

Flaw Description is a free-format memo field. For through-wall flaws,
information about size (e.g. equivalent diameter) is included in this field. For
part through-wall flaws, this field includes
information on flaw depth (a) and length (l), andorientation. For multiple flaws, the number offlaws and their lengths are recorded in thedesignated fields.
Check if multiple circumferential flaws (in weld
or weld-HAZ).
nCF (number of circumferential flaws) is the total
number of flaws in an affected weld.
D#-## is the distance, in [mm], between adjacentcircumferential flaws. For example, D0-1 is the
distance from the zero-degree position (top deadcentre for a horizontal pipe per crack profile in theNEA/CSNI/R(2018)12

•
•
•
•
•
•

figure) to flaw #1, and D2-3 is the distance between flaw #2 and flaw #3, etc. A
blank field indicates that no information on the spacing is available. (for a
vertical pipe, the zero-degree position must be clearly defined).
CF-# is the length of circumferential flaw “#” [mm]. The flaw number is relative
to the 0-degree position; CF-1 is the first circumferential flaw from the reference
position.
Crack Depth [%] is the ratio of crack depth to pipe wall thickness.
nAF (number of axial flaws) is the total number of axial flaws in an affectedarea.
Axial length [mm]: this field relates to the flaw description.
Ratio of crack length to circumference (relative to the inside pipe circumference.
Aspect ratio. The ratio of the crack depth to the total crack length.

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70 │ NEA/CSNI/R(2018)12

5. Database application facilities
The CODAP event database is an internet based (or online) relational (SQL) database.
In its present form the online version facilitates data submissions, various search andsort functions, and database interrogation functions. The latter are performed in the“Statistics” area of the database. This section of the report addresses the four application
facilities: 1) Records Management, 2) SEARCH, 3) Database Query Function, and 4)
Export Function. The export function of the Online Version of CODAP produces aXML-file 20 that can be converted to Access or Excel format for further data processing
and analysis.

5.1. Records management
The RECORDS tab includes a listing of all data base records. In its current format, thedatabase content can be sorted by “Status” (i.e. “Draft,” “Ready for Review by NC,”
“Ready for QA,” or “Approved”), Country, Plant Name, and/or Year of Event.
Figure 5.1 shows the results of a data sort by country (the United States) and the calendar
year (2017) that an event occurred.

5.2. Search function
The SEARCH tab includes two areas: 1) Search criteria, and 2) Result column. To
demonstrate the SEARCH tab functionality, a search is made for all records that address
PWSCC. This is done as follows. In the criteria field, select the event database field“Damage/Degradation mechanism.” Next, add a search criterion and select “PWSCC.”
The programme returns a total of 362 records. Using the “Result Column,” a sort ismade by country, event date, event type and system. Placing the cursor on the “Event
Date Column” provides a sort in ascending or descending order. The example in Figure
5.2 shows the earliest recorded PWSCC event to have occurred (or been discovered
through in-service inspection) on 27 February 1986.

5.3. Query function
The “STATISTICS” tab supports basic database queries. To demonstrate the“SEARCH” tab functionality, a search is made for failures that are attributed to ECSCC.
In the criteria field, select “Damage/Degradation mechanism” and add the criteria“ECSCC.” Next, under “Field” select “Diameter Class [mm].” Organised by pipe size,
this query returns the number of ECSCC records in database. Since check marks areplaced in the “Table” and “Chart” check boxes, tabular and graphic results are displayed
once the “Refresh” button is pressed, Figure 5.3. According to this query, the current
version of the database includes 155 failure records, of which 14 records address non20. XML = Extensible markup language

NEA/CSNI/R(2018)12

piping passive components. Further data processing may be performed by exporting thequery results as a CSV-file 21 (Figure 5.4). When working within the “STATISTICS”
tab:
•
•

Press “Refresh” to launch a database query.
Press “Finish” to return to “Records.”

21. A “comma-separated-values” (CSV) file stores tabular data (numbers and text) in plain text.
Files in the CSV-format can be imported to Microsoft® Excel.

│ 71

72 │ NEA/CSNI/R(2018)12
Figure 5.1. High-Level Data Sort

NEA/CSNI/R(2018)12

Figure 5.2. SEARCH for PWSCC Data Records by Event Date, Event Type & System│ 73

74 │ NEA/CSNI/R(2018)12
Figure 5.3. Query Example Using "STATISTICS" (Partial Screenshot)

NEA/CSNI/R(2018)12

Figure 5.4. CSV-File Example│ 75

76 │ NEA/CSNI/R(2018)12

5.4. Export function
The online version of the CODAP event database is a central repository of event records
and supporting documents (e.g. root cause analysis reports, isometric drawings). The
database includes provisions for conducting simple queries. In its current form,
advanced database applications should be performed on a local computer or computer
network, however. The “export function” of the online version facilitates the transfer of
selected data records or the entire database to a local computer or computer network.
Downloading records from the online version is straightforward. Pressing the “Export”
button returns a listing of all records. Selected records or the entire database can beexported to a local computer. The online version creates a zip-file (“Export” file) thatcan be opened or saved to a local disc. The data records are converted to a XML fileformat (Extensible mark-up language) that is compatible with Microsoft® Officeprograms (e.g. Access, Excel, Word). Note that the full CODAP event database is amajor collection of information organised in a structured manner. It is a relational
database and all data relationships must be retained (or enhanced) in order to support
advanced applications. The Microsoft® Access software platform is recommended forthe transfer of data from the online version to a local computer or computer network.
An Access template is available from the CODAP work area 22. Other equivalent
database software platforms may also be considered.
The Microsoft® Office products are pre-programmed for XML file formats. XML files
can be uploaded and formatted using an existing template (e.g. database platform). For
illustrative purposes, this section uses an Access database template to facilitate thetransfer of data from the online version to a local computer environment. A successfultransfer of data involves the following steps:
1. In the online version, go to “export”.
2. At the bottom of the screen three options are listed; “select all”, “clear
selection”, and “continue with selected records.” Invoking “continue withselected records” returns a new screen with “export records” at the bottom of
the screen.
3. Invoking “export records” returns a “file download” window. At this point,
define the path for the data transfer (from oecd-nea.org to local disc and folder).
4. Open access database template and go to the “external data” folder and then to“Import XML file.”
5. Specify the data source, and once defined press “ok” and check “append data toexisting table(s); Figure 5.5.
6. When using the template (which is available from the CODAP work area), thiscompletes the data download process. The XML-file is automatically imported
into the “CODAP failure data” table.
7. In case the template is not used, data format validation must be performed toensure full database functionality. Note that in converting the XML file all
database fields are assigned data type “text.” Therefore, a data validation mustbe performed consistent with database field definitions as documented in thecoding guideline.
22. The template preserves all built-in data relationships. If template is not used, then the analyst
must manually restore data relationships as needed.

NEA/CSNI/R(2018)12

Figure 5.5. Importing a XML-File to Microsoft® Access Template│ 77

78 │ NEA/CSNI/R(2018)12

6. Future developments
At its eleventh working group meeting (23-24 February 2016), the CODAP projectreview group (PRG) approved a database improvement plan to be implemented in twophases over an 18-month period. Phase one involves certain subtle modifications to theexisting software to improve the user friendliness. Phase two involves a significant
programming effort to produce an advanced, state-of-the-art database user interface. A
software requirements specification (SRS) has been developed to establish the basis foran agreement between the database users (i.e. the CODAP PRG) and the developer(NEA-IT) on what the software product is expected to do.

6.1. CODAP improvement plan phase one
The purpose of phase one is to implement certain subtle software changes to the existing
CODAP event database. First, the database structure will be simplified23. Second, data
input will be via three input forms 24 instead of the current single input form. When fully
implemented, the ultimate objective of the proposed changes is to make the practical
database usage more intuitive.
In its current form, database queries are performed via the statistics page. As currently
implemented, the current “statistics” functions are sub-optimal and do not support thedefinition of multi-attribute cross-tab queries. In option one, the statistics page will
include a set of standardised queries that are invoked by pressing a corresponding “querybutton”:
•
•
•
•
•
•
•
•

primary water SCC operating experience (OE) summaryintergranular SCC OE summary
fatigue OE summary (Low-cycle fatigue, high-cycle fatigue and thermal
fatigue)buried pipe OE summary
socket weld OE summary
WWER OE summary
flow-accelerated corrosion (FAC) OE summary
cast austenitic stainless steel (CASS) OE summary

23. The current database structure consists of 102 fields plus lookup tables. The new databasestructure will have ca. 60 field plus lookup tables and without any loss of information.
24. The three forms are titled “Event information”, “Flaw size & NDE information”, and “Rootcause evaluation”.

NEA/CSNI/R(2018)12│ 79

6.2. CODAP improvement plan phase two
The project review group (PRG) has worked extensively towards making the CODAP
event database both user-friendly and applications-oriented. Full achievement of afrequently used and fully recognised (in an organisational sense) international event
database has not yet been reached, however. Over the years, the PRG has identified
numerous impediments regarding an active and timely data exchange. These
impediments centre around three aspects of CODAP: 1) overly complex databasestructure, 2) tedious data entry process, and 3) non-optimum search and query functions.
Each of the three “impediments” shall be addressed in phase two of the improvementplan.
In the context of nuclear plant ageing management, structural integrity assessments andprobabilistic safety assessment (PSA), an objective of an event database such as
CODAP is to provide complete and comprehensive information on the operating
experience so that independent and realistic “measurements” of material performancecan be obtained. Therefore, phase two has two “target success criteria.” First, the newsoftware should motivate project member organisations to actively participate in dataexchange. Second, the new software should encourage PRG members to actively utilise
the database. The CSNI project review group in 2014 recommended that the CODAP
project put in place operating procedures and processes whereby future national datasubmissions are commensurate with the number of operating reactors. By addressing
the three impediments described above, the database modifications in phase two areintended to greatly facilitate data entry and ultimately lead to a greater percentage of
events being recorded within the database by each country. A summary of the events
recorded in the database for each country is provided in Table 6.1.
Table 6.1. OPDE/CODAP Data Submission Summary

MemberCountry

Validation Status as of September 2017
Ready
Ready forApproved
Draft
for QA Review
by NC

Total
No.
Records[% of Total]

Comment

BE - Belgium

8

--

--

--

8 [< 1%]CA - Canada
CH - Switzerland
CZ – Czech Republic
DE - Germany
ES - Spain

211
91
31
354
54

6
-----

1
7
-1
--

2
1
-2
1

220 [4.5%]
98 [2%]
31 [< 1%]357 [7.3%]
55 [1%]

Participated in OPDE
1st term only

FI - Finland

55

--

2

--

57 [1.1%]

FR - France
JP - JapanKR – Korea

140
288
78

----

27
-5

----

167 [3.4%]
288 [5.9%]
83 [1.7%]

SE - Sweden

365

1

--

366 [7.5%]\

SK – Slovak Republic
TW – Chinese Taipei
US – United States ofAmerica

2
21

--

10
4

---

12 [< 1%]25 [< 1%]

3146

--

3

--

3149 [64%]2002-2014
MemberPRG

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

80 │ NEA/CSNI/R(2018)12
Validation Status as of September 2017
Ready
Ready forApproved
Draft
for QA Review
by NC
4848
6
61
6

MemberCountry
ALL

Total
No.
Records[% of Total]

Comment

4921

The completeness and comprehensiveness of the database are key factors in motivating
materials and nuclear safety specialists to use the database. That is, the ability of thedatabase to capture all key events within respective PRG country.

6.2.1. Phase two success criteria
A main objective of CODAP phase two is to re-design the web based work space inorder to address the three “impediments” as defined above. The original databasestructure was defined by the PRG to ensure that all known material degradation
mechanism conjoint requirements were being addressed. This database structure did not
sufficiently differentiate between passive component reliability attributes and influence
factors, however. Event reports that provide the fundamental input to CODAP typicallyaddress all relevant reliability attributes (e.g. dimensional data, material type andmaterial designation). Information on influence factors (e.g. water chemistry,mechanical properties, material chemical composition, and irradiation dose) typicallymust be derived from information sources other than event reports. Hence, data input
involves a substantial amount of additional processing in order to fill in all data fields.
On the basis of past experiences in working with the existing CODAP event database
the PRG must precisely define how to address the three impediments; that is, 1) overly
complex database structure, 2) tedious data entry process, and 3) non-optimum searchand query functions. This is a critical task in the CODAP phase two work plan.

6.2.2. Option two conceptual user interface
Conceptual CODAP phase two web pages are shown in Figures 6.1 and 6.2. Figure 6.1
is intended to represent the CODAP event database portal (or “Main Menu”). It is
divided into five areas. General project information and recent updates of generalinterest are to be displayed in the main area. The “public area” has links to project reports(e.g. status summaries and topical reports). Access to the “project information tools” isrestricted to CODAP PRG members and other authorised users such as TSOs. In the“database user area” authorised users have limited access to the event database; searchesand queries may be performed and results saved & downloaded. Finally, in entering the“PRG member area” complete access to the database is obtained.
The conceptual web page for the “PRG member area” is displayed in Figure 6.2. The
intent of the “YYYY data submission status” is to provide automatically updated,current information on data submission status and data validation status. The work area
consists of three fields:•

New data submission. This area provides links to the data input area. The user
will have the option of performing a pre-screening of database fields to be
invoked by the software. For example checking the “piping checkbox” implies
that the program will request piping dimensional data. When checking the“reactor internals checkbox” those database fields specific to piping components
will not be invoked.

NEA/CSNI/R(2018)12••

Updates/edits. In this area the user can search for records that are not yet
approved. As is the case with the current database, the software responds bylisting records for which additional information is needed and with links to thedata input form.
The “download area” shall be equivalent to the current version of the database.
Additional options for database conversions may be considered; i.e. databaseconversion formats in addition to that based on the current Access template
format.

6.3. Advanced database applications
The future development includes the consideration of “advanced database applications.”
This implies using a novel approach to data analysis of location-specific, materialspecific, and degradation mechanism (DM) specific structural reliability parameters. A
methodology for obtaining structural reliability parameters such as rate of degradation
conditional on material, pipe size and operating environment and conditional pipe
failure probability given a certain degraded state builds on established statistical models
and includes full recognition of the different sources of uncertainty. Additionally and in
the context of probabilistic safety assessment (PSA), a practical application needs to be
fully risk-informed, which implies that optimum use is made of the best available
information about structural integrity analysis, relevant operating experience data, inservice inspection practices, and degradation mitigation practices.
“Data specialisation” is an important aspect part of PSA applications. Data
specialisation involves updating generic, industry-wide data parameters with plantspecific data. Typically, the data updating is accomplished using a Bayesian frameworkin which well qualified generic data is represented by a prior distribution. In piping
reliability analysis, data specialisation includes the following tasks:│ 81

82 │ NEA/CSNI/R(2018)12
Figure 6.1. Conceptual Web Page / CODAP “Portal” (or “Main Menu”)

Project History• Background• Project Objectives• Project Reports• Technical Reports• PPT Presentations• Conference Papers
CODAP Information Tools• Applications Handbook• Coding Guideline• Knowledge Base
PRG Member Area
• Create – New Data Entry• Search• Query• Apply• Database Tools
Database Users• Search• Query• Create Reports
PublicWelcome to CODAP. CODAP is one of four OECD Nuclear Energy Agency
database projects (CADAK, CODAP, FIRE & ICDE); for additional information,
go to http://www.oecd-nea.org/jointproj/). Established in 2002 as a multilateral cooperative project to collect and evaluate operating experience dataon degradation and failure of metallic piping components and selected
metallic non-piping passive components, CODAP has produced an extensive
web based event database. Eleven countries have signed the current (20152017) “Terms and Conditions.” The CODAP event database is restricted to PRG
Members.
How to Become a Member. Project membership is open to any nuclear
industry organization with the proviso that it can supply operating experience
data in accordance with the CODAP Terms and Conditions. For additional
information, please contact the Project Secretary olli.nevander@oecd.org
Project News

Upcoming Meetings

Restricted – PRG & Database Users

PRG Members Only

Database Users Only(e.g. TSOs, PRG Contractors)

NEA/CSNI/R(2018)12│ 83

Figure 6.2. Conceptual web page/PRG member area

PRG
DATABASE MAINTENANCE
New Data Submission (new window)
Basic Information
 Check if Piping Check if Rx Internals
 Check if “Other”
Optional Information

MEMBER

2015 Data Submission Status
The
National
Coordinatorsareresponsible for data submissions. The
preferred method for submitting new
data to the database is via the web-based
interface. Data submissions may also be
handled by e-mail with event information
attached in Microsoft Access, Excel or
Word file format. The Operating Agent
will ensure that data submissions via email be uploaded to the Online Versionof the event database. Respective NC
remains responsible for data validation.
Country
CA
CH
CZ
DE
ES
FI
FR
JP
KR
SE
SK
TW
US
Totals:Approved
205
91
25
351
48
55
131
288
70
365
2
12
3041
4684

Status of Data Validation
Ready for
Review byNC

Ready for QA

4

2
2
22
1
1
5
2
23
62

Draft
6
1
1

CODAP DATA SUBMISSIONSPRG Member
CanadaChinese Taipei
Czech Republic
FinlandFranceGermanyJapan
Korea (Republic of)
Slovak Republic
Spain
SwedenSwitzerland
USA

CODAP TOR
CODAP-OP
KR - CODAP08
SK - CODAP07
JP - CODAP03

5
13

AREA

CY 2012CY 2011No data submittals the work scopefocused on finalizing
DB structure,developing &implementing the
Web-Based Event
Database

Totals:

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

CY 2013
25
6
1
--4
-17
-3
-5
61
122
Notes

CY 2014

CY 2015

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

18
--1
5
2
-2
--1
1
30
60

CY 2016

CY 2017

N/A

N/A

N/A

N/A

Total as ofToday's Date
51
15
4
10
22
24
1
20
5
5
0
8
180
347

"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 TOR) above""The data exchange is carried out through signatories in the participating countries, with the possibility of delegating to otherorganizations."
36 legacy records to be input to database
Screening of Bohunice-3/4 Events→ 29 events selected for CODAP. Also note, the CY2014 records include event data & plant name but noother information
Presentation of "Recent Experiences Related to Pipe Wall Thinning in Japan" - 59 event summaries included in presentation material.

84 │ NEA/CSNI/R(2018)12
Updates/Edits
Download Area
StatusCountry
Plant
Event Year
SelectSearch for Pending Actions
Country
StatusNew Events of Potential Interest

Link to Pending Actions (this window provides list with link)

NEA/CSNI/R(2018)12

•
•

•

•

•

•

Update of existing piping reliability parameter estimates by using new service
experience data (“routine” or ordinary data specialisation).
Modifying generic piping reliability parameter estimates to account for impact
on reliability by changes to an inspection programme, or DM mitigation such as
full structural weld overlay (FSWOL), mechanical stress improvement process(MSIP®), and use of DM-resistant material.
Derivation of DM-centred pipe failure rates and rupture frequencies. Includedin this task is development of conditional rupture probability (CRP) models thatare conditional on the presence of a specific active or assumed inactive
degradation mechanism.
Derivation of piping reliability parameters for new reactor designs on the basisof existing industry-wide service experience data. This involves informed
application of lessons learned from the GenI, GenII and GenIII reactor operating
experience.
For some PSA applications pipe rupture frequencies have been developed for
different through-wall flow rate categories. For example, “spray events” (≤5 kg/s), “general flooding” (between 5 kg/s and 100 kg/s) and “major flooding”(> 100 kg/s). To remove conservatism a refined treatment of flow rate ranges toparse the pipe rupture frequency for flow rate ranges of varying sizes may be
warranted.
The quality of a data specialisation task is a function of the analyst’s knowledge
and experience and how a parameter estimation task is structured to adequately
address a specific application requirement. Guidelines and best practices forpiping reliability are developed that address:‒ Knowledge Base. A fundamental basis for a qualified piping reliability
analysis rests on a deep understanding of how, the typically robust metallic
piping systems degrade and fail or sustain damage due to different offnormal operating environments. Also of importance is a deep
understanding of piping system design principles, including the different
piping construction/fabrication practices.
‒ Service Experience Data. Under what conditions can service experience
data support quantitative piping reliability analysis? The completeness andcomprehensiveness of a database are essential characteristics for a
database to support the derivation of "robust" reliability parameter
estimates.
‒ Qualitative Analysis Requirements. Query functions are defined to extract
event population and exposure term data from a comprehensive relational
database. Oftentimes, a query definition must address a complex set of
reliability attributes and influence factors. The characterisation of aleatory
and epistemic uncertainties depends on the intrinsic qualities of a query
definition.
‒ Quantitative Analysis Requirements. Pipe failure rate calculation is basedon event populations that reflect different piping designs. Therefore, anestablished practice is to apply a Monte Carlo posterior weighting
technique to synthesise the variability in weld counts and DM
susceptibility. Pipe rupture frequencies are calculated for well-defined
break sizes and resulting through-wall flow rates. CRP models are required
for a pre-defined set of break size ranges.

│ 85

86 │ NEA/CSNI/R(2018)12‒

Special Considerations. Certain follow-up (or sensitivity) studies may
have to be performed once a base case set of reliability parameters havebeen obtained.

Five types of metrics are considered in quantitative piping reliability analysis in supportof PSA: 1) failure rate, 2) conditional failure probability, 3) inspection effectiveness, 4)
DM mitigation effectiveness, and 5) ageing factors. A pipe failure event database cannot
support failure rate estimation, unless the database also includes extensive piping systemdesign information that yield information on the total piping component population thathas produced the failure observations; i.e. exposure term data. Relative measures ofpiping reliability such as conditional failure probabilities can be generated by querying
an event database. The statistical robustness of such relative measures is correlated withthe completeness of the event population.
Completeness and comprehensiveness of a service experience database should beensured through a sustained and systematic maintenance and update process.
Completeness is an indication of whether or not all the data necessary to meet current
and future analysis demands are available in the database. The comprehensiveness of aservice experience database is concerned with how well its structure and content
correctly capture piping reliability attributes and influence factors. A clear basis should
be included for the identification of events as failures.
The inherent latency in structured data collection efforts is on the order of five years.
This means that ca. five years could elapse before achievement of high confidence indata completeness. In other words, around 2020 the data mining for the previous ten
years (2005-2015) would be expected to approach “saturation” (as in high confidence
in completeness of a database). Could “cliff-edge-effects” (e.g. small change in input
parameter resulting in large results variation) affect an analysis due to databaseinfrastructure factors? It depends on the maturity of inspection programmes and ourstate-of-knowledge concerning certain degradation mechanisms. Considerations about
the use of up-to-date failure data is intrinsically assumed to be factored into an analysis
task.
The design and infrastructure associated with a service experience database should becommensurate with application demands and evolving application requirements. In
PSA, the completeness of a relevant event population should be validated, either
independently or assured through a sustained maintenance effort. To achieve theobjectives defined for a database, a coding format should be established and documented
in a coding guideline. Such a guideline is built on recognised pipe failure data analysis
practices and routines that acknowledge the unique aspects of piping reliability incommercial nuclear power plant operating environments. For an event to be considered
for inclusion in the database it must undergo an initial screening for eligibility. An
objective of this initial screening is to go beyond abstracts of event reports to ensure that
only pipe degradation and failures according to a certain work scope definition are
included in the database. As stated, the knowledge and experience of the analyst is a key
to performing well-qualified piping reliability analysis.
Correlating an event population with the relevant plant and component populations that
produced these failure events enables the estimation of reliability parameters for input
to a calculation case. The information contained in a database must be processed
according to specific guidelines and rules to support reliability parameter estimation. A
first step in this data processing involves querying the event database by applying datafilters that address the conjoint requirements for pipe degradation and failure. These data

NEA/CSNI/R(2018)12

filters are integral part of a database structure. Specifically, these data filters relate tounique piping reliability attributes and influence factors with respect to piping systemdesign characteristics, design and construction practice, in-service inspection (ISI) andoperating environment. A qualitative analysis of service experience data is concernedwith establishing the unique sets of calculation cases that are needed to accomplish theoverall analysis objectives and the corresponding event populations and exposure terms.
Most, if not all database applications are concerned with evaluations of event
populations as a function of calendar time, operating time or component age at time of
failure. The technical scope of the evaluations includes determination of trends andpatterns and data homogeneity, and assessment of various statistical parameters of
piping reliability. Therefore, an intrinsic aspect of practical database applications is thecompleteness and quality of an event database. Do the results of an application correctlyreflect the effectiveness of in-service inspection, ageing management, and/or waterchemistry programmes?
Before commencing with a statistical parameter estimation task it is essential to develop
a thorough understanding of the range of influence factors that act on metallic piping
components. Database “exploration” (or data reduction) should be an integral part of allqualitative analysis steps to ensure that the defined evaluation boundary is associatedwith the most relevant event population data and exposure term data. It entails theidentification of unique event sub-populations, time trends/temporal changes anddependencies.
The technical approach to estimating pipe failure rates and rupture frequencies is basedon the model expressed by Equations (1) and (2) for estimating the frequency of a pipe
break of a given magnitude. Typically, the magnitude is expressed by an equivalent
break size (EBS) and corresponding through-wall flow rate. The parameter x is treatedas a discrete variable representing different equivalent break-size ranges.

F(IE x ) =ρ ix =

∑m ρ
i

i ix

∑ λ P(R F )I
k

ik

x

ik

(1)
ik

(2)

Where:
F(IE x ) =
mi =

ρix =
λ ik =

Frequency of pipe break of size x, per reactor operating-year,
subject to epistemic (or state-of-knowledge) uncertainty.
Number of pipe welds (or fittings, segments or inspection
locations of type i; each type determined by pipe size, weld
type, applicable damage or degradation mechanisms, andinspection status (leak test and non-destructive examination).
While not explicitly addressed in the given example, for theburied ESW piping the parameter mi corresponds to the total
length of piping being analysed.
Frequency of rupture of component type i with break size x,
subject to epistemic uncertainty.
Failure rate per “location-year” for pipe component type i due
to failure mechanism k, subject to epistemic uncertainty,
equation three below. In this analysis the failure rate is│ 87

88 │ NEA/CSNI/R(2018)12P(R x Fik ) =I ik =

calculated on the basis of per linear meter and reactor
operating year.
Conditional rupture probability (CRP) of size x given failureof pipe component type i due to damage or degradation
mechanism k, subject to epistemic uncertainty. Thisparameter may be determined on the basis of probabilistic
fracture mechanics, expert elicitation or service experience
insights.
Integrity (RIM) management factor for weld type i and failure
mechanism k, subject to epistemic uncertainty determined byMonte Carlo simulation and Markov modelling. Thisparameter is not explicitly addressed in this example.

For a point estimate of the failure rate of piping component type i and degradation
mechanism k:nn ik
λ ik = ik =
τ ik f ik N i Ti

(3)

Where:
n ik =

τ ik =
f ik =Ni =Ti =Number of failures in pipe component of type i due todegradation mechanism k. The component boundary used indefining exposure terms is a function of the susceptibility tocertain damage or degradation mechanisms. A CODAP
database query provides this number.
Component exposure population (in component years) for
welds of type i susceptible to degradation mechanism k.
CODAP does not include any exposure term data.
Estimate of the fraction of the component exposure population
for piping component type i that is susceptible to degradation
mechanism k, estimated from results of a formal degradation
mechanism evaluation.
Estimate of the average number of pipe components of type i
per reactor in the reactor operating years of exposure for thedata query used to determine nik. Determined from reviews ofisometric drawings (fabrication isometrics or ISI isometrics)
and ISI programme plans for a representative population ofplants and combined with expert knowledge of degradation
mechanisms.
Total exposure in reactor-years for the data collection forcomponent type i. CODAP event database provides thenumber of reactor operating years that produced the operating
experience data.

For a Bayes’ estimate, a prior distribution for the failure rate is updated using nik and τikwith a poisson likelihood function. The formulation of equation two enables thequantification of conditional failure rates, given the known susceptibility to the given
damage or degradation mechanism. When the parameter fik is applied, the units of the
failure rate are failures per piping component susceptible to the degradation mechanismof concern.

NEA/CSNI/R(2018)12

Applying the above seemingly simple relationships invariably results in significant
analysis efforts, however. First, the failure event population(s) must fully match aselected evaluation boundary; i.e. piping system of certain material and in a specificoperating environment. The exposure term definition involves extensive reviews ofisometric drawing information to correctly address plant-to-plant piping system design
variability, which is essential in correctly matching event populations and exposureterms.
For a Bayes’ estimate, a prior distribution for the failure rate is updated using nik and τikwith a Poisson likelihood function. The formulation of equation three enables thequantification of conditional failure rates, given the known susceptibility to the given
damage or degradation mechanism. When the parameter fik is applied, the units of the
failure rate are failures per welds susceptible to the damage or degradation mechanism.
This formulation of the failure rate estimate is done because the susceptible damage ordegradation mechanisms typically are known from the results of a previously performed
degradation mechanism analyses. If the parameter fik is set to 1.0, the failure rates
become unconditional failure rates, i.e. independent of any knowledge about thesusceptibility of damage or degradation mechanism, or alternatively that 100% of thecomponents in the population exposure estimate are known to be susceptible to a certain
damage or degradation mechanism.
The likelihood of a pipe flaw propagating to a significant structural failure (SF) isexpressed by the conditional failure probability P(Rx\Fik) where Fik represents degradedcondition. When there are limited or no SFs in the database to support a direct statisticalestimation of the conditional probability, the assessment can be based on probabilistic
fracture mechanics (PFM), expert judgment, and/or service experience data insights.
Different PFM algorithms have been developed, but with a focus on fatigue growth andstress corrosion cracking.
There remain issues of dispute with respect to reconciliation of results obtained throughstatistical estimation versus the physical models of PFM, however. Results from studies
to benchmark PFM calculations against field experience have shown PFM computercodes to over-predict pipe failure rates by more than an order magnitude relative tostatistical estimates of field experience data. In general, the results obtained with PFM
computer codes are quite sensitive to assumptions about weld residual stresses, crack
growth rates, and correlations of crack initiation times and growth rates. Also, PFM
calculations are invariably done for very specific geometries that may or may not apply
to a broader set of evaluation boundaries under consideration in PSA.

│ 89

90 │ NEA/CSNI/R(2018)12

7. Summary and recommendations
This report is the fifth CODAP topical report and focuses on the CODAP event databasestructure and the underlying principles of collecting operating experience data on
metallic passive components. The report includes a summary of the CODAP Operating
Procedures, the CODAP event database Coding Guideline, and the CODAP
Applications Handbook.

7.1. Summary
Since May 2002, the NEA has operated an event database on passive component
degradation and failure. During 2002-2011 the project, referred to as OPDE, focused on
piping component failures. In May 2011, the project review group approved thetransition of OPDE to a new, expanded “NEA Component Operational Experience,
Degradation and Ageing Programme” (CODAP).
The objective of CODAP is to collect information on passive metallic component
degradation and failures of the primary system, reactor pressure vessel internals, mainprocess and safety systems, and support systems. It also covers non-safety-related
components with significant operational impact. At the present time, 11 NEA member
countries participate in the database project. An effort is underway to systematicallyevaluate the database content and to make a series of database insight reports available
to material scientists as well as risk management practitioners. Data exchange among
participating organisations promotes understanding of the different national practices
regarding reliability and integrity management of passive components.
The CODAP Event Database is a web-based, relational SQL database consisting ofapproximately 100 uniquely defined data fields. It is a blend of free-format fields fordetailed narrative information and fields defined by drop-down menus with key words(or data filters) or related tables. A basic premise of the use of narrative information isto preserve original event information as recorded in root cause evaluation reports andreportable occurrence reports. The “related tables” include information on material,
location of damage or degradation, type of damage or degradation, system name, safetyclass, etc. The event database structure, database field definitions and data input
requirements are defined in a coding guideline, which is central to the project, including
database maintenance, data validation and quality control. The database design has
benefitted from a multidisciplinary approach involving chemistry, metallurgy, nondestructive examination, structural integrity and probabilistic safety assessment (PSA).
The database structure has evolved over a period of fifteen years.
At its eleventh working group meeting (23-24 February 2016), the CODAP projectreview group (PRG) approved a database improvement plan to be implemented in twophases over an 18-month period. Specifically, the CODAP database improvement plan
encompasses two phases. The first involves certain subtle modifications to the existing
software to improve the user friendliness. The second involves a significant

NEA/CSNI/R(2018)12

programming effort to produce an advanced, state-of-the-art database user interface. A
software requirements specification (SRS) has been developed to establish the basis foran agreement between the database users (i.e. the CODAP PRG) and the developer(NEA-IT) on what the software product is expected to do. The complete description of
the functions to be performed by the software specified in the SRS will assist thepotential users to determine if the software specified meets their needs or how thesoftware must be modified to meet their needs.

7.2. Next steps and recommendations
The project review group (PRG) has worked extensively towards making the CODAP
event database both user-friendly and applications-oriented. Full realisation of afrequently used and fully recognised (in an organisational sense) international event
database has not yet been achieved, however. Over the years, the PRG has identified
numerous impediments to an active, comprehensive and timely data exchange. These
impediments centre around three aspects of CODAP: 1) overly complex databasestructure 25, 2) tedious data entry process, and 3) non-optimum search and query
functions. CODAP phase two sets out to provide a more user-friendly work space fordata entry, data analysis and advanced database applications. Each of the three“impediments” will be addressed in CODAP phase two.
In the context of nuclear plant ageing management, structural integrity assessments andprobabilistic safety assessment (PSA), a fundamental objective of an event databasesuch as CODAP is to provide complete and comprehensive information on the field
experience so that independent and accurate “measurements” of material performancecan be obtained, including the identification of adverse trends. Therefore, phase two of
the CODAP database enhancement project has two target success criteria. First, the newsoftware must motivate project member organisations to actively participate in dataexchange. Second, the new software must encourage PRG members to actively utilise
the database. The CSNI project review group in 2014 recommended that the CODAP
project puts in place operating procedures and processes whereby future national datasubmissions are commensurate with the number of operating reactors. CODAP phase
two is intended to be one step towards achieving of a more “balanced” event database.
With respect to the continued database development and maintenance (i.e. datasubmissions and validation) it is recommended that the following items be consideredin the ongoing active data submission activities by the CODAP PRG Members as wellas in the current programme for an enhanced version of the online database (“CODAP
option two” Project)26:••

Implement a coding navigation tool that, for example, links the drop-down menu
for “passive component category” with the drop-down menu for “passive
component type”.
Encourage the PRG Membership to more actively share metallic passive
component operating experience insights. As a standing item, future working

25. The current database structure was approved by the project review group during the first yearof the first term (2011-2014) of the CODAP project.
26. Approved by the CODAP PRG at its eleventh working group meeting (May 2015), “CODAP
option two” entails the development of software requirements specifications for an enhanced
web-based database.

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92 │ NEA/CSNI/R(2018)12•

group meetings should focus on technical discussions on how to utilise 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 item on national overviewsof recent operational events, including the findings of root cause analyses.

The CODAP PRG faces two important future challenges. Firstly, while efforts havebeen made to promote CODAP and associated data project products to the nuclear safety
community at large, there remain programmatic issues relative to how to make therestricted CODAP event database available to PSA practitioners. Secondly, work
remains to be done to develop PSA-centric database application guidelines and theassociated analytical infrastructure (i.e. piping reliability analysis techniques and tools).
Two initiatives are under consideration by the PRG to address the stated challenges. The
working group on risk assessment (WGRISK) of the Committee on the Safety of
Nuclear Installations (CSNI) is planning the “joint workshop on use of NEA data project
operating experience data for probabilistic risk assessment.” The CODAP PRG intends
to actively support this joint workshop. Additionally, a proposal has been made for aninternational benchmark exercise concerning the use of operating experience data toquantify piping reliability parameters for input to a standard problem application, e.g.
risk-informed operability determination. 27

27. The topic of an international benchmark exercise has been under discussion since theinception of the OPDE/CODAP project.

NEA/CSNI/R(2018)12│ 93

8. References[1]

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

[2]

Nuclear Energy Agency, Technical Basis for Commendable Practices on Ageing Management– SCC and Cable Ageing Project (SCAP), Final Report, NEA/CSNI/R(2010)5, BoulogneBillancourt, France, 2010.

[3]

Nuclear Energy Agency, NEA Component Operational Experience, Degradation & Aging
Program (CODAP): First Term (2011-2014) Status Report, NEA/CSNI/R(2015)7, BoulogneBillancourt, France, 2015.

[4]

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

[5]

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, BoulogneBillancourt, France, 2015.

[6]

Moosemiller, B. and Weber, B., Guidelines for Improving Plant Reliability through Data
Collection and Analysis, Centre for Chemical Process Safety (CCPS) of the American Instituteof Chemical Engineers, New York, NY, 1998.

[7]

ESReDA, Handbook on Quality of Reliability Data, ISBN 82-515-02535, Det Norske Veritas,
Høvik, Noway, 1999.

[8]

Nuclear Energy Agency, Working Group on Risk Assessment (WGRISK) Report on the Useof NEA Data Project Products in Probabilistic Safety Assessment, NEA/CSNI/R(2014)2,
Boulogne-Billancourt, France, 2014.

[9]

Lydell, B. and Olsson, A., Reliability Data for Piping Components in Nordic Nuclear Power
Plants “R-Book” Project Phase I, SKI Report 2008:01, Swedish Radiation Safety Authority,
Stockholm, Sweden.

[10]

Nuclear Energy Institute, Probabilistic Risk Assessment Peer Review Process Guideline, NEI
00-02, Washington (DC), 2000.

[11]

The American Society of Mechanical Engineers, ASME RA-Sb-2013 Standard for
Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sb-2005(Addenda to ASME RA-S-2008), New York (NY), 2013.

[12]

Nyman, R. et al, Reliability of Piping System Components. Volume 1: A Resource Documentfor PSA Applications, SKI Report 95:58, Swedish Nuclear Power Inspectorate, Stockholm(Sweden), 1995.

[13]

Nyman, R. et al, Reliability of Piping System Components. Framework for Estimating Failure
Parameters from Service Data, SKI Report 97:26 (3rd Edition), Swedish Nuclear PowerInspectorate, Stockholm (Sweden), 2005.

94 │ NEA/CSNI/R(2018)12[14]

Stevens, B. (Editor), Guide to Reliability Data Collection and Management, EuReDatA Project
Report No. 3, Commission of the European Communities, Joint Research Centre, Ispra
Establishment, Ispra (Italy), 1986.

[15]

Swedish Radiation Safety Authority, Föreskrifter om mekaniska anordningar i vissa
kärntekniska anläggningar, (Regulations Concerning Mechanical Equipment) SSMFS 2008:13,
Stockholm, Sweden, 2008.

[16]

Reason, J., Human Error, Cambridge University Press, Cambridge, UK, ISBN 0-521-31419-6,
pp 173-188.

[17]

Embrey, D. et al, Guidelines for Preventing Human Error in Process Safety, Center for
Chemical Process Safety, American Institute of Chemical Engineers, New York, NY, ISBN 08169-0461-8, pp 41-44.

[18]

Hurst, N.W. et al, “A Classification Scheme for Pipework Failures to Include Human andSociotechnical Errors and Their Contribution to Pipework Failure Frequencies,” J. Hazardous
Materials, 26:159-186, 1991.

[19]

Barnes, R.W. and Cooper, G.D., Failure Rates in Piping Manufactured to Different Standards,
INFO-0607, Atomic Energy Control Board 28, Ottawa, Canada, 1995.

[20]

U.S. Nuclear Regulatory Commission, NRC Position on IGSCC in BWR Austenitic StainlessSteel Piping, Generic Letter 88-01, Washington, DC, January 1988.

[21]

U.S. Nuclear Regulatory Commission, Generic Aging Lessons Learned (GALL) Report,
NUREG-1801, Revision 2, Washington, DC, 2010.

[22]

Taylor, J.M., Completion of the Fatigue Action Plan, SECY-95-245, U.S. Nuclear Regulatory
Commission, Washington, DC, 1995.

[23]

Khaleel, M.A. et al, Fatigue Analysis of Components for 60-Year Plant Life, NUREG/CR6674, U.S. Nuclear Regulatory Commission, Washington, DC, 2000.

[24]

Grövall, B. et al, Intercrystalline Stress Corrosion Cracking of Inconel 600 Inspection Tubes inthe Ågesta Reactor, AE-245, Aktiebolaget Atomenergi, Stockholm, Sweden, 1964.

[25]

Combrade, P., Ford, P. and Nordmann, F., Key Results from Recent Conferences on Structural
Materials Degradation in Water Cooled Reactors, LCC6 Special Report, Advanced Nuclear
Technology International, Mölnlycke, Sweden, 2010.

[26]

Hickling, J., “Strain-Induced Corrosion Cracking of Low Alloy Steels in LWR Systems – Case
Histories and Identification of Conditions Leading to Susceptibility,” Nuclear Engineering and
Design, 91:305-330, 1986.

[27]

IAEA, Safety Glossary. Terminology Used in Nuclear Safety and Radiation Protection,
STI/PUB/1290, Vienna, Austria, 2007.

[28]

IAEA, Assessment and Management of Ageing of Major Nuclear Power Plant Components
Important to Safety: BWR Pressure Vessel Internals, IAEA-TECDOC-1471, Vienna, Austria,
2005.

[29]

IAEA, Assessment and Management of Ageing of Major Nuclear Power Plant Components
Important to Safety: PWR Pressure Vessel Internals, IAEA-TECDOC-1119, Vienna, Austria,
1999.
28. Now the Canadian Nuclear Safety Commission (CNSC).

NEA/CSNI/R(2018)12

[30]

│ 95

American Society of Mechanical Engineers, 2007 ASME Boiler & Pressure Vessel Code,
Section XI, Rules for In-service Inspection of Nuclear Power Plant Components, 2008a
Addenda, pp 14-16, New York, NY, 2008.

96 │ NEA/CSNI/R(2018)12

9. Appendix
CODAP database structure

NEA/CSNI/R(2018)12│ 97

Table 9.1. CODAP Database Field Definitions & Coding Guidelines

Item
No.
001a

Field NameType

REPRESENTATIVE
Yes/NoEVENT (RE)001b

REFERENCE
CASE ID

001c

MER

Yes/No 29

002

COMPLETENESS
INDEX

Number003

EVENT DATE

Date

004

PLANT NAME

Text

Text

Description

User Instruction/Note

FORM #1 – EVENT NARRATIVE – MANDATORY
Indicates that detailed information is available. All
Identifies a record considered representative of otherrelevant information are to be provided for the RE
similar events
only.
Unique ID to indicate that a record is related to aCC-## where CC is country code and ## is number“Representative Event”
Some events result in augmented inspection of other,similar locations. The “MER” box is checked if
additional flaws are found and if these additional
Multiple Events Reportflaws can be uniquely defined. Add new databaserecords for each additional, uniquely defineddegraded component. CODAP-CG includes
additional details.
This field supports database management activities
Roll-down menu with options & definitions.
and applications. This index is assigned by Operating
Agent.
Event date or date of discovery
Online version uses the format YYYY-MM-DD
Organised by country, the Online Version includes a
Roll-down menu with list of all commercial nuclear
NPP information library that includes plant type,power plants
design type, NSSS vendor and name of constructor. 30

29. “Yes” = True; “No” = False. Depending on the context, a check box without check mark implies either “No”, “Not Available”, “Not Applicable”, or“Unknown/Pending.”
30. This is a “database-within-the-database” and builds on information from the IAEA power reactor information system (PRIS).

98 │ NEA/CSNI/R(2018)12
Item
No.

005

006a
006b
006c
007a
007b

Field NamePLANT
OPERATIONALSTATEREFERENCE PRIMARYPRIMARYREFERENCE IN
PUBLIC DOMAIN?
PDF1
REFERENCE SECONDARYSECONDARYREFERENCE IN
PUBLIC DOMAIN?

Type

Text

Text
Yes/NoHyperlink
Text
Yes/NoDescription

User Instruction/Note

Plant operational state (at the time of discovery) per
generally accepted or standard nomenclature. Rolldown menu31 with the following options:• CSD – Cold Shutdown (other than Refuelling
Outage)• HSD – Hot Shutdown• HSB – Hot Standby• Power Operation• Low Power Operation• Refuelling• Shutting Down
• Starting Up

This field is used as a data filter when defining a
database query. For example, it allows the user toquickly differentiate between events with anoperational impact (e.g. forced shutdown) and those
events discovered through scheduled or augmented
inspections.

Primary reference
Upload original reference as a PDF or TIF. Except for
Check box to indicate if primary reference istitle, no translation into English is required. The
restricted (confidential) or not – mark box if YES
original language should be included as well as if theProvision for uploading original primary reference as material is confidential or not. CODAP operatingprocedures address handling of confidential
PDF
information. Also, refer to item n° 16;
Secondary (or supplemental) reference
PDF1/PDF2/PDF3/PDF4 could include isometricdrawings, photographs, etc.
Check box to indicate if secondary reference isrestricted (confidential) or not – mark box if YES

31. Contact the Operating Agent for any change requests concerning the roll-down menus.

NEA/CSNI/R(2018)12

Item
No.

Field NameType

Description

007c

PDF2

Hyperlink

Provision for uploading original secondary reference
as PDF

008a
008b
008c
009a
009b
009c

REFERENCE TERTIARYTERTIARY
REFERENCE IN
PUBLIC DOMAIN?
PDF3REFERENCE QUARTIARYQUARTIARY
REFERENCE IN
PUBLIC DOMAIN?
PDF4

Text

Tertiary (or supplemental) reference

Yes/NoCheck box to indicate if tertiary reference is restricted(confidential) or not – mark box if YES

Hyperlink

Provision for uploading original tertiary reference as
PDF

Text

Quartiary (or supplemental) reference

Yes/NoCheck box to indicate if quartiary reference isrestricted (confidential) or not – mark box if YES

Hyperlink

Provision for uploading original quartiary reference
as PDF

User Instruction/Note│ 99

100 │ NEA/CSNI/R(2018)12
Item
No.

Field NameType

Description
•
•
•
•
•
•

010

EVENT TYPE

Text

•
•
•
•
•
•

011

Crack Morphology

Text

•
•

Roll-down menu with the following options:
Wall Thinning
Crack-Full (through-wall crack w/o active
leakage)
Crack-Part (part through-wall crack)
Leak (leak rate within Tech. Spec. limit)
Large Leak (leak/flow rate above Tech. Spec.
limit)
P/H-leak (pinhole leak)
Recordable indication
Reportable Indication
Rupture (large flow rate, loss of structural
integrity)
Severance
Small Leak (leak rate well below Tech. Spec.
limit)
Intergranular/InterdendricTransgranular

User Instruction/Note

This field is used as a database filter for user-definedqueries. “Tech. Spec.” refers to the technicalspecifications established for plant operation andincludes the limiting conditions for operation. The
term “severance” relates to complete structural
failure of small-diameter piping

This field applies to stress corrosion cracking (SCC)
eventsNEA/CSNI/R(2018)12

Item
No.

Field Name

012

EVENT
CATEGORY

013

COLLATERALDAMAGE

Type

Text

Text

Description
Roll-down menu with the following options:• CCI-Precursor• Containment Bypass• CCI (Common Cause Initiator)• Internal Flooding• Internal Flooding Precursor• IS-LOCA• IS-LOCA Precursor• LOCA• LOCA Precursor• RCPB Leak
• RCP Seal LOCA
• RCP Seal LOCA Precursor• System Degraded• System Disabled• Train Degraded• Train Disabled
Roll-down menu with the following options:• N/A – None
• Adjacent Line Damaged• Flooding of Equipment Area
• Jet Stream Impact• Pipe Whip – Adjacent Line Damaged• Spray Impact on Adjacent Equipment• Spurious Fire Protection System Actuation• Loss of Supported Function│ 101

User Instruction/Note

This field is used as a data filter

This field relates to operational events involving
active through-wall leakage on ex-RPV passive
component. An active leak in Auxiliary
Building/Reactor Building could potentially generatesufficient heat load to activate a Fire Protection Water
Systemresultinginconsequential
wetting/spraying/floodingofsafety-related
equipment.

102 │ NEA/CSNI/R(2018)12
Item
No.

Field Name

014

IMPACT ON
PLANT
OPERATION

Text

015

Time-to-Repair
TTR-ClassNumber016

EVENT
NARRATIVE

Memo

Type

Description
Roll-down menu with the following options:• Cooldown (from Hot Shutdown)• Cooldown (from Hot Standby)• Expanded Outage Work Scope• LCO Entry (LCO = Limited Condition for
Operation)• Manual Shutdown• Multi-Unit Shutdown• N/A – None
• Power Reduction• ESF/RPS Actuation (automatic Rx trip)• System Train Inoperable• Tritium Release to Environment• Turbine Trip / Reactor Trip
• Unplanned Outage Work• Shutdown of Normal Letdown (PWR)
Roll-down menu with the following options:• 1: TTR ≤ 8 hours• 2: 8 < TTR ≤ 24 hours;• 3: 24 < TTR ≤ 96 hours;• 4: 96 < TTR ≤168 hours• 5: TTR > 168 hours.
Description of plant condition prior to event and plant
response during event, method of detection,corrective action plan.

User Instruction/Note

Mainly, this field relates to operational events
involving active through-wall leakage on ex-RPV
passive component. Definition of “LCO” is found in
the plant specific Technical Specification document

TTR = Time-to-Repair; this field is used as a datafilterThis free format field should include sufficient
information to support independent verification of theevent classification. ENGLISH LANGUAGE ONLYNEA/CSNI/R(2018)12

Item
No.

017

018

019
020
021

Field Name

LOCATION OF
FAILURE

PLANT
LOCATIONQUANTITYRELEASEDDURATION OF
LEAKAGE / SPILLLEAK RATE

│ 103

Type

Description

User Instruction/Note

Memo

Location of crack/thinned area/leak/rupture;description of where a degradation or failure
occurred. Include sufficient detail to support theconsequence evaluation and event classification.

Text

Roll-down menu with the following options:• Auxiliary Building• Cooling Tower (UHS)• Containment (PWR)• Drywell (BWR)• EDG Building• Fuel Transfer Canal• Owner Controlled Area
• Radwaste Treatment Building• Reactor Building• Secondary Containment (Annulus)• Steam Tunnel• Torus Area (BWR)• Turbine Building• Wet Well (BWR)

Each component should be uniquely identified using
an identifier as indicated on an isometric drawing orin plant reference documentation. This is particularly
important in case a single event report relates tomultiple degraded or failed passive components.
Utilise hyperlinks as needed.

Text

Quantity of process medium released [kg]This may be based on assumptions using information
included in the event narrative or based on calculation

Text

Estimated time of leakage/release/spill

Indicate time in minutes[kg/s]

Assumed or measured

NumberIn-plant location of degraded/failed component.
Forward change requests to the operating agent

104 │ NEA/CSNI/R(2018)12
Item
No.

Field Name

Type

022

LEAK RATEASSUMED OR
MEASURED?

Yes/No023

SYSTEM NAME

Text

024

SYSTEM GROUP

Text

025
026
027

P-C-C
(Passive Component
Category)
P-C-T(Passive Component
Type)
WELD
POSITION

Text

Description
Check box

User Instruction/Note
Mark check box if Leak/Flow Rate is measured

See Table B-1 for details; it is not an all-inclusive list
System name; this field is supported by a roll-down
and user-defined entries are permitted/encouraged.
menu
Forward all change requests to the operating agent.
Supported by a roll-down menu:• AUXC – Auxiliary Cooling System• CS – Containment Spray System• FWC – Main & Auxiliary Feedwater &Condensate
See Table B-2 for details. This field is a database• RAS – Reactor Auxiliary Systemfilter. Forward all change requests to the operating• RCPB – Reactor Coolant Pressure Boundaryagent• RPV – Reactor Pressure Vessel• SIR – Safety Injection & Recirculation
System• SG – Steam Generator System• STEAM – Steam Systems
Passive Component Category – supported by a roll- This field is a database filter, which supports userdown menu with current options as indicated in defined database queries. Change requests should be
forwarded to the Operating Agent.
Table A-3

Text

Passive Component Type Part. Roll down menu with Change requests should be forwarded to the operating
current options as indicated in Table A-3.
agent.

Text

Roll down menu

This field is a database filter, and it applies to piping
only. Change requests should be forwarded to theoperating agent

NEA/CSNI/R(2018)12

Item
No.

Field NameType

Description

028

SAFETY CLASS /CODE CLASS

NumberRoll-down menu with the following options:
• 1
• 2
• 3
• 4 (= Non-Code, non-safety related)

029

DIMENSIONS

Text

Dimension(s) of non-piping passive component

030

031
032
033
034

│ 105

User Instruction/Note
This field is a database filter, which supports userdefined database queries. Classification ofcomponents differs among participating countries. At
the end of the trial period a “classification crossreference table” will be prepared to show what thesedifferences are.
Dimension of bolting should be given as:“DIAMETER × LENGTH.” CODAP-KB includes aComponent Catalogue with dimensional data As an
example, dimension of bolting should be given as:“DIAMETER × LENGTH.”, .

DIAMETER
Unit
WALL
THICKNESS[mm]

NumberYes/NoRoll-down menu with the following options (basedon nominal diameter):• 1 – ∅ ≤ 15 mm• 2 – 15 < ∅ ≤ 25 mm
Database filter for piping components• 3 – 25 < ≤ 50 mm• 4 – 50 < ∅ ≤ 100 mm• 5 – 100 ∅ ≤ 250 mm• 6 - ∅ > 250 mm
Measured diameter
Check Box
Mark check box if unit is [inch]

NumberMeasured wall thickness [mm]

PIPE SCHEDULE

NumberPipe schedule according to standard commercial pipe For definition, see “Glossary of Terms” (Appendix
sizes (U.S. practice)D)

DIAMETERCLASS(PIPING)

Number

106 │ NEA/CSNI/R(2018)12

035

AGE

036

ACTUALOPERATING TIME
OF COMPONENTFORM #2 – ROOT CAUSE INFORMATION – MANDATORY
Note that the current version of the database platform
automatically generates a default component age,which is calculated as:
In-service component age at time of failure [hours]; AGE = TEvent_Date – TRx-Crit
respective NC is responsible for ensuring that a given If the affected component has a repair/replacement
Numberin-service age accounts for repair/replacement history, this should be reflected in the estimated age.
If possible (and relevant), the component age should
history(ies)exclude time in other than power operating modes.
Note that some systems are required to be operational
during non-power operating modes of operation
Number037

METHOD OF
FLAW
DETECTION

Text

038

TECHNIQUE OF
FLAW
DETECTION

Text

If available, this should be in terms of “Effective Full- This should be exclusive of time in other than power
Power Years” (EFPY)
operating modes.
Roll-down menu with the following options:• Containment/Drywell Inspection• Control Room Indication/Alarm• ISI – Inservice Inspection• Leak Detection• Routine Maintenance• Maintenance on demand• Periodic Testing• Walkdown Inspection
Roll-down menu with the following options:• Eddy Current Testing• Hydrotesting• Inservice Leak Test
• Liquid Penetrant Testing• Radiographic Testing

Forward change requests to the operating agent

Forward change requests to the operating agent

NEA/CSNI/R(2018)12039

MATERIAL(Base Metal)

Text

040

MATERIAL
DESIGNATION(Base Metal)

Text

041

PROCESSMEDIUM at time ofdetection

Text• UT-Examination• Video Camera• Visual Inspection/Testing
List of check boxes for the following options:• Duplex Steel• High-Performance Stainless Steel• Stainless Steel• Ni-base Alloy• SS/Carbon Steel (stainless steel clad carbon
steel)• Carbon Steel• Low Alloy Steel• Other│ 107

User-defined entry is permitted. If it is a DM weld,
there are two types of base metals. If information isavailable, include information on the heat treatment
method; e.g. MA = mill-annealed, TT = thermal
treatedIt is not an all-inclusive list and user-defined entries
Roll-down menu with options as listed in Tables A-4
are permitted and encouraged. Forward change
through A-6
requests to the Operating Agent
Roll-down menu with the following options:• Borated Water• BWR Primary Water – NWC• BWR Primary Water – HWC• BWR Primary Water – HWC+NMCA• Coordinated chemistry• High pH
Forward change requests to the Operating Agent• PWR primary side
• Chemical treatment• Condensate• Demineralised Water• Feedwater (Conditioned)• Heavy Water – D2O

108 │ NEA/CSNI/R(2018)12••Radioactive Waste Water
Steam

Damage /Degradation
Mechanism

Text

Drop-down menu supported by a roll-down menu

043

CONTRIBUTING
FACTOR -1(ENVIRONMENT)

Text

044

CONTRIBUTING
FACTOR – 2(MATERIAL)

Roll-down menu with the following options:• Boric Acid/Stagnant Flow Conditions• Contamination – ID (e.g., chlorides, sulfides)• Contamination – OD (e.g., chlorides,
sulfides)• Cyclic loading• Demineralizer Breakdown – pH Decrease• HF: Construction/Installation Error• HF: Repair/Maintenance Error
HF = Human Factor (includes safety culture issues).
• HF: Welding Error
Forward change requests to the operating agent• High Residual Stresses• Hydrogen Concentration• Increased Concentration of Inclusions• ISI Programmatic Deficiency• Oxygen Containing Water• Stress Riser due to Root Notch• Thermal Fatigue• Transient Stress During Start-up• Use of Higher-Strength Material

Text

Roll-down menu to be developed

042

Change requests should be forwarded to operating
agent

Forward change requests to the operating agent

NEA/CSNI/R(2018)12

045
046
047

048

CONTRIBUTING
FACTOR – 3(STRESS FACTOR)CONTRIBUTING
FACTOR – 4(OTHER)
ROOT CAUSEANALYSIS

CORRECTIVE
ACTION

│ 109

Text

Roll-down menu to be developed

Indicate whether there is evidence of repair(s)performed on component prior to discovery of flaw
(e.g. repair performed during plant construction)

Text

Roll-down menu to be developed

Forward change requests to the operating agent

Memo

Narrative of root cause analysis results

Text

Roll-down menu with the following options:• Abandon – Cap• Base Metal Repair• Code Repair• Electro-Discharge Machining• Evaluation – Accepted for Continued
Operation• Half-nozzle Technique• IHSI Process (Induction Heat Stress
Improvement)• Isolate – Cap• Isolate – Temporary• L-SIP – Laser Stress Improvement Process• Left As-Is• Mechanical Clamp• MNSA – Mechanical Nozzle Seal AssemblyInstalled• MSIP – Mechanical Stress Improvement
Process• Repair• Repair – Freeze Seal

The text should include information on the type of
examinations performed and the results.

The term “temporary repair” implies that a permanentcode repair be made at next outage (given it is ofsufficient duration) or next scheduled refuellingoutage.
Forward change requests to the operating agent

110 │ NEA/CSNI/R(2018)12
Replace – In-Kind
Replace – In Kind – Below T-min (FAC)
Replace – New Design
Replace – New Material
Replace – New Material – Below T-min(FAC)• Replace – New Weld Configuration• RPV Head Replacement• Temporary Repair• Temporary Repair – Furmanite Leak Seal
Enclosure• Temporary Repair – Welded Plate• Temporary Repair – Mechanical Clamp• Weld Overlay• Weld Overlay – Full Structural• Weld Repair
Description of weld repair(s) performed prior tocomponent being taken into operation
•
•
•
•
•

049

Weld Repair
Description

Memo

050

ADDITIONALCOMMENTMemo

The purpose of this free-format database field is to
Any other information of relevance to understanding
facilitate future applications, for example, byof underlying causal factors. Also include
codifying the information on passive component
information on the extent of repair/replacement.
replacements.

Memo

Free format text field

Memo

Free format text field

051
052

FOLLOW-UP
INSPECTION
SPECIFICREGULATORYACTIONNEA/CSNI/R(2018)12

053

FLAW
DESCRIPTION

054

LOCATION OFFLAW
INITIATION

055

MATERIAL AT
LOCATION OFFLAW
INITIATION

056

DIRECTION OF
FLAW
PROPAGATIONFORM #3 – FLAW CHARACTERISATION – MANDATORY
Narrative description of flaw; orientation and Include any relevant information with respect toMemosize/geometry of crack or fracture
actions taken subsequent to discovery
Roll-down menu with the following options:• Base Metal – Heat Affected Zone
• Base Metal – Not Heat Affected Zone
Text
Forward change requests to operating agent• Buttering• Cladding• Weld Metal
Roll-down menu with the following options:• Stainless Steel• Ni-based Alloy• SS/Carbon Steel (stainless steel clad carbon
Text
Forward change requests to operating agent
steel)• Carbon Steel• Low Alloy Steel• Other
Roll-down menu with the following options:• Axial Direction
Text
Direction as determined by NDE• Circumferential Direction• Oblique Direction• Multiple Directions
Yes/NoMultiple flaws in weld-HAZ?

058

MULTIPLE
FLAWS
nF

Number059

D#-##/CF#

Number060

FLAW-DEPTH

NumberNumber of flaws
Distance between flaws and length of each flaw.
Database fields provided for up to ten flaws
Flaw depth – percentage of nominal wall thickness

057

│ 111

Mark check box if “Yes”

112 │ NEA/CSNI/R(2018)12
061
062
063

FLAW-DEPTH
FLAW-LENGTHLONGEST CRACK

064

ASPECT-RATIO

065

ISI DEFICIENCY

066

067
068

ISI HISTORY /AGINGMANAGEMENTPROGRAMQUALIFIEDINSPECTIONTECHNIQUENDEQUALIFICATIONBACKGROUND

NumberNumberNumberDepth of deepest flaw
Depth of longest flaw (if deepest flaw is not longest)
Length of deepest flaw
Length of longest flaw (if deepest flaw is not longest)
Ratio of length of deepest crack to depth of deepest The aspect ratio is used in probabilistic fracture
Numbercrack
mechanics.
FORM #4 – IN-SERVICE INSPECTION (ISI) INFORMATION - OPTIONAL
Example (FitzPatrick, 5/4/1990): “The weld waspreviously inspected during the 1988 refuellingoutage. At that time a “Level III” inspector inspectedMark check box if the pipe failure is attributed to a the weld. Indications were noted with maximum
Yes/Noweakness in ISI program plan, or if the affected amplitude at 5.5-degrees clockwise, 6.65-degrees
counter- clockwise, and 21.5-degrees clockwise.
component is not included in ISI program.
These indications were misinterpreted to be root
geometry, since the indications were noted on the
opposite side of the weld (beam reflection).”Memo

Include any relevant information about the ISI history The narrative text should differentiate between(e.g. time of most recent inspection and findings) or “scheduled”, “augmented”, and “mandatory” (bymaintenance history
regulatory order/request)

Yes/NoMark check box if the flaw was detected using a If “Yes”, provide relevant background information in
database field 049
qualified inspection technique

Memo

Include details on the standard used to qualify the This information should only be included if it isinspection techniquecomponent specificNEA/CSNI/R(2018)12

069

070
071

METHOD OF
FABRICATION

Text

FORM #5 – SERVICE ENVIRONMENT – OPTIONAL
Roll-down menu with the following options:• Casting• Forging• Cold Bending
Forward change requests to the operating agent• Hot Bending• Seamless Pipe
• Seam Welded Pipe
If “yes”, provide details in field #71 below
PWHT = Post Weld Heat Treatment

PWHT
PWHT-DETAILSCOMPONENTTEMPERATUREGAMMAHEATING
INCLUDED?

Yes/NoMemo074

T-DESIGN

Yes/No075

OPERATING
PRESSURE

Text

Operating pressure [MPa]076

P-DESIGN

Yes/NoMark check box if pressure as given in “060” isdesign value

072
073

077
078

ALLOYING
ELEMENTS(Base Metal)
ALLOYING
ELEMENTS(Details)

Text
Yes/NoComponent temperature [º C] (fluid temperature)
This field applies to reactor internals only. “No”
Mark check box if gamma heating is included in
implies “does not apply” in the case of ex-RPV“095” (IRRADIATION DOSE)
components.
Mark check box if temperature as given in “074” isdesign value

Text
Yes/No│ 113

Of relevance to degradation mechanism
Mark check box if detailed information is available

Supply information separately as PDF and attach torecord in designated field (utilise Fields 006, 007,
008, 009 as needed).

114 │ NEA/CSNI/R(2018)12

079

080
081
082
083
084
085
086

087

WELDINGMETHODWELDINGTECHNIQUE
WELD MATERIALALLOYING
ELEMENTS(Weld Metal)
STRESS AT
LOCATIONSTRESS DETAILSAVAILABLE
MECHANICALPROPERTIESSURFACEFINISHCHEMISTRYHISTORY

Text

Roll-down menu with the following options:• Shop weld – GMAW
• Shop weld – GTAW
• Shop weld – SAW• Shop weld – SMAW
• Field weld – GMAW
• Field weld – GTAW
• Field weld – SAW• Field weld – SMAW
• Repair weld

Check box for manual vs. automatic welding.

Yes/NoCheck box for manual vs. automatic welding.

Mark check box if automatic welding

Text

Roll-down menu.

Forward change requests to operating agent
Supply information separately as PDF and attach torecord in designated field of online version

Text
Text
Yes/NoDifferentiate between residual versus operational
stresses
Mark check box if detailed information is available Supply information separately as PDF and attach toand append as PDF
record in designated field of online version

Estimated or measured

Text

For example yield strength, hardness

Indicate if per standard or specification.

Text

Roll-down menu to be developed by PRG

Enter surface finish as text

Narrative description of chemistry programme

Include dates of program changes (e.g. start ofchemical addition). Include information on method
for pH control. Indicate major changes in chemistry,
such as transition from NWC to HWC , chemicaltransients, condenser in-leakage

MemoNEA/CSNI/R(2018)12

088
089
090
091
092
093
094
095
096

pH
MAX-LITHIUM
CONCENTRATION
MAX-BORON
CONCENTRATIONSTAGNANTPROCESSMEDIUMCONDUCTIVITY AVERAGE
DISSOLVEDOXYGENCONTENTDISSOLVED
HYDROGEN
CONTENTIRRADIATIONDOSE
REPEAT
EVENT

│ 115

For PWR primary system address whether primarysystem pH is within or outside limits

Text
Text

Maximum lithium content in [ppb] (PWR)

Text

Maximum boron content in [ppb] (PWR)(Yes/No)A check mark indicates normally stagnant process
Should address recent history
mediumNumberConductivity [µS/cm] (BWR) – average

Number

Dissolved oxygen concentration in [ppm]

Number

Dissolved hydrogen concentration [cm3 STP/kg
H2O] (PWR)

Text

Maximum neutron fluence [neutrons/cm2]

Yes/NoPrevious event at this location?

Method of calculating fluence to be determined116 │ NEA/CSNI/R(2018)12
Table 9.2. Selected Plant System Designators

Designator
Definition
CRD
Control Rod Drive (BWR Hydraulic Scram)CS
Containment Spray
CVC
Chemical & Volume Control
ECCS
Emergency Core Cooling
Flux Detector
Neutron Flux Detector inside the core (WWER)
HPCS
High Pressure Core Spray (BWR)
HPSIHigh Pressure Safety Injection (PWR)
ICS
Isolation Condenser (BWR)
LPCS
Low Pressure Core Spray (BWR)
LPSI
Low Pressure Safety Injection (PWR)
MS
Main Stem
PHTS
Primary Heat Transport System (CANDU)
RCIC
Reactor Core Isolation Cooling (BWR)
RCS
Reactor Coolant System (PWR)
RHR
Residual Heat Removal
RPV
Reactor Pressure Vessel
RR
Reactor Recirculation (BWR)
RVLIS
Reactor Vessel Level Indication System
RWCU
Reactor Water Cleanup (BWR)
SFC
Spent Fuel Pool Cooling
S/G-System
Steam Generator (S/G) incl. S/G Blowdown
SLC
Standby Liquid Control (BWR)
See IEEE Std 805-1984 (IEEE Recommended Practice for System Identification in Nuclear Power Plants andRelated Facilities) for information on system boundary definitions and system descriptions.

NEA/CSNI/R(2018)12

Table 9.3. Selected Stainless Steel Designations

National Standard
Chemical Composition [Weight %] - Typical
SS (Sweden)
NF (France)
DIN
C
Cr
Ni
Mo
Other
2333
Z6 CN 18.09
1.4301
0.04
17.0
8.5
--(2352)-(1.4306)
0.02
18.0
8.0
--2337Z6 CNT 18.10
1.4541
0.04
17.0
9.0
-Ti
2352
Z2 CN 18.10
1.4306
0.02
18.0
10.0
--(2347)-(1.4401)
0.04
16.5
10.0
2.0
-(2348)-(1.4404)
0.02
16.5
10.0
2.0
-2350
(Z6 CNDT
1.4571
0.04
16.5
10.0
2.0
Ti
17.12)
1.4436316
2343
Z6 CND 17.11(1.4436)
0.04
16.5
10.5
2.5
-1.4432 316L
2343-(1.4435)
0.02
16.5
10.5
2.5
-1.4435 316L
2353
Z2 CND 17.13
1.4435
0.02
17.0
12.5
2.5
-Old SS and DIN-designations within brackets specify a slightly higher Ni-content, which is insignificant for the corrosion
resistance. The NF designation within brackets specifies a slightly higher Ni-content than the stated typical value.
EN
1.43011.43071.45411.43061.44011.4404
1.4571

ASTM
(304)
304L
321
304L
316
316L
317Ti

│ 117

118 │ NEA/CSNI/R(2018)12
Table 9.4. Expanded Stainless Steel Cross-Reference Table

GermanyUSA
AISI

FranceAFNOR

DIN EN 10027-1

DIN
EN
100272201
301

Z 12 CN 1707

301

X 12 CrNi 17 7

1.4310X3CrNiN17-8

1.4319302

Z 10 CN 1809

X9CrNi18-9

1.4325303

Z 10 CNF 1809

X8CrNiS18-9

1.4305303 SeZ 10 CNF 1809304

Z 6 CN 18-09

X5CrNi18-10

1.4301304 H
Z 2 CN 18-10

SUS
201
SUS
301
SUS
301
SUS
302

X2CrNi19-11

1.4306

RussiaGOST

12KH18N9

SUS
303
SUS
303 Se

304 N

304 L

Japan
JIS

SUS
304
SUS
304N1
SUS F
304H
SUS
304L

12KH18N10E
08KH18N10
06KH18N1103KH18N11

Spain
UNE

SwedenSIS

UK
BSI

European UnionEN

X 12 CrNi
17-07

23 31

301S21

X 12 CrNi 17 7

301S26

X3CrNiN17-8

23 31

302S25X9CrNi18-9

23 46

303S22

X8CrNiS18-9

X 10 CrNi
18-09
X 10
CrNiS 1809
X 10
CrNiS 1809
X 6 CrNi
19-10
X 6 CrNi
19-10
X 2 CrNi
19-10

303S41
23 32

304S15 304S16

X5CrNi18-10

23 52

304S11X2CrNi19-11

NEA/CSNI/R(2018)12

GermanyUSA
AISI

France
AFNOR

304 L

DIN EN 10027-1

DIN
EN
100272

X2CrNi18-9

1.4307304 LNZ 2 CN 1810-Az

X 2 CrNiN 18 10

1.4311305

Z 8 CN 18-12

X4CrNi18-12

1.4303309

Z 6 CNU 1810
Z 17 CNS 2012
Z 15 CN 2313

X15CrNiSi20-12

1.4828

X7CrNi23-14

1.4833

310

X15CrNiSi25-20

1.4841310 S

X12CrNi25-20

1.4842

X8CrNi25-21

1.4845

X 15 CrNiSi 25-20

1.4841

X 5 CrNiMo 17-12-2

1.4401

309 S

310 S
314
316

Z 12 CN 2621
Z 12 CNS 2520
Z 6 CND 1711

Japan
JIS
SUS
304L
SUS
304LN
SUS
305
SUS
XM7
SUH
309
SUS
309S
SUH
310
SUS
310S
SUS
310S

SUS
316

Russia
GOST

Spain
UNE

Sweden
SIS

UK
BSI

│ 119

European UnionEN
X2CrNi18-9

X 8 CrNi
18-12

23 71

304S6

23 33

305S19/305S17

X4CrNi18-12
X 6 CrNiCu 18
10 4 Kd

309S24

X15CrNiSi20-12

20KH23N18

309S24

X7CrNi23-14

10KH23N18

310S24

X15CrNiSi25-20
X12CrNi25-20

20KH25N20S2

23 61

X6
CrNiMo17-12-03

X8CrNi25-21
X 15 CrNiSi 25
20

23 47

316S31 316S17

X 5 CrNiMo 17
12 2

120 │ NEA/CSNI/R(2018)12
GermanyUSA
AISI

France
AFNOR

316

Z 6 CND 1712

316 LN
316 LN316 F?

Z 2 CND 1713-Az

DIN EN 10027-1

DIN
EN
100272

X3CrNiMo17-13-3

1.4436

X2CrNiMoN17-11-2

1.4406

X2CrNiMoN17-13-3

1.4429

X 12 CrNiMoS 18 11

1.4427

X6CrNiMo17-13

1.4919316 N?
316 H

Japan
JIS

Russia
GOST

SUS
316
SUS
316LN
SUS
316LN

Z 2 CND 1712

X2CrNiMo17-12-2

1.4404316 L

Z 2 CND 1713

X2CrNiMo 18-14-3

1.4435316 Ti

Z6 CNDT 1712

X6CrNiMoTi17-12-2

1.4571UK
BSI

European UnionEN

X6
CrNiMo17-12-03

23 43

316S33 316S19

X3CrNiMo17-133

08KH17N13M2T
10KH17N13M2T

SUS F
316H
03KH17N14M2

316 L

SwedenSIS

316S61SUS
316N

316 H?

Spain
UNE

SUS
316L

08KH17N13M2T
10KH17N13M2T

23 75

X5
CrNiMo17-12
X6
CrNiMo17-12-03
X2
CrNiMo17-12-03
X2
CrNiMo17-12-03
X6
CrNiMoTi
17-12-03

316S63

316S50X6CrNiMo17-13

23 48

316S11X2CrNiMo17-122

23 53

316S13

X2CrNiMo18-143

23 50

320S18 320S31

X6CrNiMoTi1712-2

NEA/CSNI/R(2018)12

Germany
DIN EN 10027-1

DIN
EN
100272316 Ti

X10CrNiMoTi18-12

1.4573317

X3CrNiMo18-12-3

1.4449X2CrNiMo18-15-4

1.4438

X12NiCrSi35-16

1.4864

X6CrNiTi18-10

1.4541X8CrNiTi18-10

1.4878X3CrNiMoN27-5-2

1.4460X6CrNiNb18-10

1.4550

USA
AISI

317 L
330

FranceAFNOR

Z 2 CND 1915
Z 20 NCS 3316

321
321(H)

Z 6 CNT 1810

329
347

Z 6 CNNb
18-10

SUS
317
SUS
317L
SUH
330
SUS
321
SUS
321
SUS
329J1
SUS
347
SUS F
347H

347 H?
904L

Japan
JIS

Z2NCDU 252

X2NiCrMoCu 25-205

1.4539Russia
GOST

Spain
UNE

SwedenSIS

X6
08KH16N13M2B CrNiMoTi
17-12-03

UK
BSI

European Union
EN

320S33

X10CrNiMoTi1812

23 66

317S1623 67

317S1208KH18N10T
12KH18N10T│ 121

X3CrNiMo18-123
X2CrNiMo18-154
X12NiCrSi35-12

X6
CrNiTi
18-11

23 37

321S12

X6CrNiTi18-10

321 S3108KH18N12B

X3CrNiMoN275-2

23 24
X6
CrNiNb18-11
X7
CrNiNb
18-11

23 38

2562

347S20 347S31

X6CrNiNb18-10

X2NiCrMoCu
25-20-5

122 │ NEA/CSNI/R(2018)12
GermanyUSA
AISI

FranceAFNOR
Z12CNDV1202

UNS31803

DIN EN 10027-1

DIN
EN
100272

X12CrNiMoV12-3

1.4939

X15CrNiSi25-4

1.4821X2CrNiMoN22-5-3

1.4462

UNS32760

Z 3 CND 2506Az

X2CrNiMoCuWN257-4

1.4501403

Z 12 C 13

X6Cr131.4000405

Z 6 CA 13

X 6 CrAl 13

1.4002

Z 8 CA 7
Z 13 C 13
Z 12 CAS 18

X 10 CrAl 7
X 10 CrAl 13
X 10 CrAl 18

1.47131.47241.4742Z 6 CT 12

X2CrTi12

1.4512

409
410
420

Z 12 C 13 M

X12Cr131.4006

X 15 Cr 13

1.4024

Japan
JIS

Russia
GOST

Spain
UNE

Sweden
SIS

UK
BSI

318S

European UnionEN

X15CrNiSi25X2CrNiMoN225-3

12Kh13
SUS
403
SUS
405

SUH
409
SUS
410

10Kh13SYu
15Kh18SYu

X 6 Cr 13

23 01?

403S17

X6Cr13

X 6 CrAl
13

23 01

405S17X 6 CrAl 13
X 10 CrAl 7
X 10 CrAl 18

409S19X2CrTi12

12Kh13
410S21
08Kh13?

X 12 Cr
13?

23 02?

X15Cr13

NEA/CSNI/R(2018)12

Table 9.5. Selected Carbon Steel Designations

ASTM
105

National Standard
CZ
SS
EN Grade 32

106 Gr. A
106 Gr. B
106 Gr. C
53 Gr. A
53 Gr. B
334 WP22335 P12335 P22B179
170.1

12022.1

123306143505P235GH-TR1/2(St 35.8)
P265GH-TR1/2(St 45.8)123305143405

P235TR1/2(St. 34.2)
P265TR1/2(St. 37.2)

Chemical Composition [Weight%] – Max.
C
Mn
P
Cr
Mo
0.35
1.05
0.040
--0.17

--

0.30

1.06

0.048

0.30

0.12

0.35

1.20

0.050

0.40

0.15

--

--

0.30

1.20

0.05

--

--

0.20
0.15
0.15
0.15

0.60
0.60
0.61
0.60

0.04
0.04
0.045
0.030

0.25
2.60
1.25
2.60

-1.13
0.65
1.13

--

0.55

--

0.40

13 CrMo 44
10 CrMo 9 10
17 MnMoV 6 4
0.21
1.80
0.035
(WB 35)15 NiCuMoNb 5 S
1
0.17
0.016
(WB 36)
ASTM A 105 mainly used for forged fittings (elbows, flanges)
ASTM A 106 is for high-temperature service (e.g. feedwater and steam piping)
ASTM A 333 is for low temperature service
ASTM A 335 is for high-temperature service (more resistant to FAC than A 106 steel)
32. Designation per DIN Standard in parenthesis.

--

│ 123

124 │ NEA/CSNI/R(2018)12
Table 9.6. ASTM, JIS and DIN Steel Designation

Steel Composition

ASTM Grade 33

JIS Grade

Carbon Steel

A120

SGP

Carbon Steel

A53-B

STPG38Carbon Steel

A53-B

STPG42

Carbon Steel

A106-A

STPT38Carbon SteelCarbon SteelCarbon Steel3 1/2% Ni Steel

A106-B
A106-C
A333 and A334-6
A333 and A334-3

STPT42
STPT49STPL39STPL46

Carbon-Molybdenum Steel

A335-P1

STPA12

1% Cr-1/2% Molybdenum Steel

A335-P12

STPA22

1 ¼% Cr-1/2% Molybdenum Steel

A335-P11

STPA232 ¼% Cr-1% Molybdenum Steel

A335-P22

STPA24

5% Cr-1/2% Molybdenum Steel
7% Cr-1/2% Molybdenum Steel
9% Cr-1% Molybdenum Steel

A335-P5
A335-P7
A335-P9

STPA25
STPA26

European Standard
EN Grade 34
10255 / S195(2440-ST33-1) 35
10217-1/P235TR1 P235TR2(1626-ST37)
10216-2/P235GHTC1 P235GHTC2 (17175-St 35.8)
17175-St 45.8
10216-2/16Mo3(17175-15 Mo3)10216-2/13CrMo4-5(17175-13CrMo44)10216-2/10CrMoVNb9-1(17175-10CrMo910)-

33. TP stands for tube or piping
34. DIN Standard designation in parenthesis.
35. Norm DIN 2440 was for threaded tubing. The norm EN 10255 is for "Non-Alloy steel tubes suitable for welding, threading and other joining methods"

NEA/CSNI/R(2018)12

Steel Composition

ASTM Grade 33

JIS Grade

18%Cr -8% Ni SteelA312-TP304

SUS304TP

18%Cr -8% Ni(0.04-0.10)% C SteelA312-TP304H

SUS304HTP

18%Cr -8% Ni - 0.035% C SteelA312-TP304L

SUS304LTP

22%Cr - 12% Ni Steel
25%Cr - 20% Ni SteelA312-TP309
A312-TP310

SUS309STP
SUS310STP

25%Cr - 8% Ni-Cb+Ta SteelA312-TP347

SUS347TP

18%Cr -8% Ni-Mo SteelA312-TP316

SUS316TP

18%Cr -8% Ni-Mo-(0.04-0.10)% C SteelA312-TP316H

SUS316HTP

18%Cr -8% Ni-Mo-0.035% C SteelA312-TP316L

SUS316LTP

18%Cr -8% Ni-Ti SteelA312-TP321

SUS321TP

18%Cr -8% Ni-Ti-(0.04-0.10)% C Steel
18%Cr -8% Ni-Cb+Ta-(0.04-0.10)% C SteelA312-TP321H
A312-TP347H

SUS321HTPSUS347HTP│ 125

European Standard
EN Grade 34
10216-5/X5CrNi18-10(17440-X5CrNi189)
10215-5/X2CrNi19-11(17440-X2CrNi189)
10216-5/X6CrNiNb18-1(17440-X10CrNiNb189)
10216-5/X5CrNiMo17-12-2,
X3CrNiMo17-13-3(17440-X5CrNiMo1810 )
10216-5/X2CrNiMo18-14-3(17440-X2CrNiMo1810)
10216-5/X6CrNiTi18-10(17440-X10CrNiMo189)-

126 │ NEA/CSNI/R(2018)12
Table 9.7. Plant System Cross-Reference Table

CODAP
Generic1
ADS
AFW
CC
COND
CRD
CS
CVC
CW
EHC
EXT
FPS
FW
HPCS
HPSI
IA
LPCS
LPSI
MS
MSR
RCS
RHR
RR
RPV-HC
RVLIS
RWCU

Description
BWR Primary Depressurisation System (BWR)
Auxiliary Feedwater System
Component Cooling Water System
Condensate System
Control Rod Drive (Insert/Removal/Crud Removal)
Containment Spray System
Chemical & Volume Control System (PWR)
Circulating Water System / Intake Cooling Water
Electro Hydraulic Control System
Steam Extraction System
Fire Protection Water System
Main Feedwater System
High Pressure Core Spray (BWR)High Pressure Safety Injection (PWR)
Instrument Air System
Low Pressure Core Spray (BWR)
Low Pressure Safety Injection (PWR)
Main Steam System
Moisture Separator Reheater System
Reactor Coolant System (PWR)
Residual Heat Removal System
Reactor Recirculation System (BWR)
RPV Head Cooling System (BWR)
Reactor Vessel Level Indication System (BWR)
Reactor Water Cleanup System (BWR)

Czech
Republic-TF

France
-ASG
RRI

-TQ

RGL
EAS
RCV

C-52

JPx
ARE
-RIS

-TJ
US
-TH

(2)
-----

-RIS
VVP
RCP
RRA
-----

Germany (7)AKZ
KKS
TK, RA
RQTF
LA
RM, RNLC
TA, TC,

KB

RL
TJ
TH

LA

TK, TMTH
RA
RB
YA, YB,
TH

JN
JN
LB
LB
JA, JE
JN

TC
TC

KB

Sweden314
327
711/712414/430 (4)
354
322
334
443
442
419/423762
312/415 (5)
--484
323
-311/411 (6)
422
313
321
313
326
536
331

NEA/CSNI/R(2018)12

CODAP
Generic1

Description

Czech
Republic
TL
TG

FranceGermany (7)
AKZ
KKS
TL
KL
TG
FA
RS
LA

│ 127

SwedenSA
Service Air System
753
SFC
Spent Fuel Pool Cooling System
PTR
324
S/G Blowdown
Steam Generator Blowdown System (PWR)
APG
337
SLC
Standby Liquid Control System (BWR)--351
SW
Service Water System (3)
VF
SEC
VE
PE
712/715
Notes:
See IEEE Std 805-1984 (IEEE Recommended Practice for System Identification in Nuclear Power Plants and Related Facilities) for information on systemboundary definitions and system descriptions.
No dedicated RHR system in WWER-440 (decay heat removal is through natural circulation)
It is common practice in the US to use different system IDs for safety-related and non-safety related SW systems; e.g. ESW vs. PSW, or SX vs. WS,
respectively for SX for Code Class 3 piping and non-Code piping
414 for F1/F2/R1/R2/R3/R4 and 430 for O1/O2/O3
312 for O1/O2/O3 and 415 for F1/F2/R1/R2/R3/R4. Also note that 312 is the designation for steam generators in Ringhals-2/3/4
311 for O1/O2/O3 411 for F1/F2/R1/R2/R3/R4
AKZ = Anlagen Kennzeichnungs System, KKS = Kraftwerk Kennzeichnungs System.

128 │ NEA/CSNI/R(2018)12
Table 9.8. Piping Safety Class Cross Reference Table 36
USA / ASME Section III 37
Class
1

Definition
Piping that forms
the reactor coolant
pressure boundary(RCPB). That is, allpiping componentsthat are part of thereactor coolant
system RCS), or
connected to the
RCS up to andincluding any or allof the following:a) the outermost
primary
containment
isolation valvein piping thatpenetrates theprimary
containment;b) the second of
two valvesnormally
closed during
normal reactor
operation insystem piping

Canada(CSA N285.0-95)
Class
Definition
1
Pipe diameter >DN20; Sectionsof systems, or
systems
connected
thereto, whichcontain fluid that
directlytransports heatfrom nuclear
fuel, and whose
failure would
cause a loss of
coolant accident
as defined in thesafety report.

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
1
Equipment of the
RCPB, except
equipment whose
rupture would result ina leakage of amagnitude within thecapacity of the normal
coolant make-up
system.

France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class
1

Class
1

Class
SK-1

Definition
RCS and its
connecting lines
with inside
diameter greaterthan 10.6 mm forwater or greaterthan 21.9 mm forsteam, up to andincluding the tworeactor coolant
isolation valves.
class one piping
also includes thepressurizer
letdown line up to
and including therelief and safetyvalves.

Definition
For PWRs:a) RPV;b) Primary side of
the S/Gs, thesecondary shellof the S/Gs incl.
the FW-inlet andMS-exit nozzles
up to the pipe
connecting
welds (excl.
small-diameter
fittings);c) pressuriser;d) RCP casing;e) connecting pipes
between theabove
components andthe valvecasings of any
type containedin the piping
system;f) pipes branching
off from theabove

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

Definition
Pressure retaining
boundary of thereactor coolingsystem up to thesecond isolation
valve or safetyvalve, including
small-diameter
piping and pressure
retaining parts of
instrumentation.

NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

Definition
that does not
penetrate
primary
containment;or,c) the RCS
safety andrelief valves.

2

Systems or portionsof systems│ 129

Canada(CSA N285.0-95)
Class
Definition

Czech Republic(Regulation 214/1997 Sb)
Class
Definition

France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Definition

Class

Class

Definition

2

2

2

Equipment andcomponents of

2&3

SK-2

a) reactor
cooling andPipe diameter >DN20; Sections

a)components
creating theDefinition
components andtheir connecting
pipes including
the valve bodies
up to andincluding thefirst shut-off
valve;g) control rod
drives and thein-coreinstrumentation.
For BWR:a) RPV;b) pipingconnected to the
RPV including
the valve bodies
incl. first shutoff valve,pipework
penetrating thecontainmentshell incl. thelast shut-off
valve locatedoutside thecontainmentshell;c) control rod drive
and in-core
instrumentation.
Piping and piping
components that are not130 │ NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

Definition
important to safetythat are designedfor:a) emergency
core cooling;b) post-accident
containment
heat removal,orc) post-accident
fissionproduct
removal.

Canada(CSA N285.0-95)
Class
Definitionof processsystems thatpenetrate thecontainment
structure andform part of thecontainment
boundary.

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
RCPB, that are
not ranked asclass one;b) components forthe reactor
shutdown during
the abnormal
operation during
the states which
could lead to theaccident
conditions, andfor the reactor
shutdown withthe aim to
mitigate theconsequences ofaccident
conditions;c) components
necessary toretain the coolant
inventory
sufficient for thecore cooling
during theaccident
conditions whenno damage of thereactor coolant
pressure system
has occurred, andafter these
conditions;France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Class

Class

Definition
systems carrying
reactor coolantthat are not safetyclass one and toequipment andcomponents
required to ensure
containment ofradioactivity in theevent of a loss of
coolant accident.
This includes:a) equipment andcomponentsthat are notsafety classone;b) mainequipment andcomponents ofthe following
systems:
RHRS, CVCS,
ECCS, CSS;c) equipment andcomponentsthat constitutethe third
barrier: thereactorcontainment
and theassociatedisolation

Definition
part of the RCPB but
have a certain
significance withrespect to reactor
safety:a) The component isrequired for the
mitigation of
DBAs withrespect to shut
down, long-term
maintenance ofsubcriticality, anddecay heatremoval.
b) Requirements
regardingcomponents ofsystems which
only indirectly
serve in residual
heat removal –these are the nonradioactivity
retaining closed
cooling water
systems andservice water
systems – shall be
specified on aplant-specificbasis taking thedesign redundancy(e.g. redundancy,

Definition
emergencycooling;b) residual heatremoval fromreactor,containment,
and steam
generators;c) cooling of
RCS in thecolddepressurised
state;d) all reactor shut
downfunctions andfunctions tomaintainsubcriticality;e) safety
functions ofprimary
containment
systems;f) components tomaintainsubcriticality
in the fuelelement
storage;g) BWR: Main
steam andfeed water line
between

NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

Definition

Canada(CSA N285.0-95)
Class
Definition

Czech Republic(Regulation 214/1997 Sb)
Class
Definitiond) components
necessary to
remove the coreheat, when thereactor coolant
pressure system
is damaged, withthe aim to limit
the fuel damage;e) components of
the residual heatremoval system
during thenormal andabnormaloperation andunder theaccident
conditions,
without the lossof the RCPB’s
integrity;f) components
necessary for theprevention ofradioactive
leakage from thecontainment
during theaccident andpost-accident
conditions;g) components
necessary to limit│ 131

France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Class

Class

Definition
systems,
portions ofsecondarysystems inside
the reactor
building up to
and including
the first
isolation valve
located outside
the reactor
building,
containmenthydrogencontrol system,
equipment andcomponents of
the in-core
instrumentation
system up to
and including
the manualisolation valve.

c)

d)

e)

Definition
diversity) into
consideration.
Large energies are
released in case of
failure of the plantcomponent and no
mitigating
measures such as
structural
measures, spatial
separation or othersafety measures
are available tokeep the effects ofthe failure to an
acceptable limit
with respect tonuclear safety.
A failure of theplant component
could either
directly or
indirectly througha chain ofassumedsequential events,
lead to a DBA.
Systems andcomponents towhich none ofcriteria a) throughc) apply, thefailure of which,however, would

Definition
secondisolation valve
and next
remote control
isolation
valve;h) PWR:
Secondaryside of steam
generator up
to isolation
valve outsideprimary
containment;i) componentsthat could
cause a dose
limit violationaccording to
Article 94
Paragraph 4 ofRadiation
Protection
Ordinance
SR814.501.

132 │ NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

3

Definition

Cooling water andauxiliary feedwater
systems or portionsof these systems
important to safetythat are designedfor:a) emergency
core cooling;b) post-accident
containment
heat removal;c) post-accident
containment
atmospherecleanup, orCanada(CSA N285.0-95)
Class
Definition

3

Pipe diameter >DN20; Sectionsof systems, notclassified as classone or two, that
containradioactive
substances with atritium
concentration
exceeding 0.4
TBq/kg (0.01
Ci/g), or anenergy-weighted
activity
concentration ofCzech Republic(Regulation 214/1997 Sb)
Class
Definition
the ionisingradiation
penetration
outside thecontainment,
during and after
the accident
conditions;h) components
necessary toaccomplish thesafety functions
for the power
supply or for thecontrol of othercomponents
ranked as thesafety class two.
3a) components
necessary to
prevent theunallowabletransient processes
connected with thereactivity changes;b) components
necessary tomaintain the
nuclear reactor in
the safe shutdown
conditions;c) components
necessary tomaintain sufficientFrance (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Class

Class

3

Definition

Safety Class 3
includes:a) CVCS
equipment andcomponents
required forthepurification of
the reactor
coolant waterand the boron
makeupsystem andequipment;b) S/G AFWS
equipment andDefinition
lead to major plantinternal damage.

See “2 & 3” above

SK-3

Definition

a)

b)
c)

d)

systems forleakage andseal water inthe primarycontainment;cooling of fuelelement
storage pool;systems forgaseousradioactive
media;RWCU ofBWR(typically SK1 and 3),

NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

Definitiond) residual heatremoval fromreactor andfrom the spent
fuel storage
pool(including
primary andsecondarycoolingsystems).
Portions of thesesystems that are
required for theirsafety functions
and that:a) do not operate
during any
mode ofnormal reactor
operation andb) cannot be
testedadequately
should beclassified asclass two.

Canada(CSA N285.0-95)
Class
Definition
radionuclides
exceeding that of
0.4 TBg/kg of
Tritium.

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
reactor coolant
inventory for thecore cooling during
the normal andabnormaloperation;d) components
necessary to
remove heat fromthe safety systems
to the firstaccumulating
volume, which is
sufficient from theviewpoint of
performance ofsafety functions;e) components
necessary tomaintain theradiation exposureof population andof nuclear
installation
personnel below
the establishedlimits, during theaccident conditions
connected with theleakage ofradioactive
substances andionizing radiationfrom the sources│ 133

France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Class

Class

Definition
components
located outside
reactorcontainment;c) CCWS and
ESWS
equipment andcomponents;d) reactor cavity
and spent fuelpit cooling andtreatment
systemequipment andcomponents;e) radioactive
waste
treatment
systems
equipment andcomponents
whose failure
could cause
release ofradioactive
gases normallystored for
decay.

Definitione)f)Definition
CVCS of
PWR(typically SK2 and 3);
auxiliary
systems for
SK-1 through3 components
and 1E
classified
electricalequipment;
systems for
accident
mitigation.

134 │ NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

Definition

Canada(CSA N285.0-95)
Class
Definition

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
located outside thecontainment, andafter suchconditions;f) components
requisite tomaintain suchenvironmental
conditions inside
the nuclear
installation that arenecessary for theoperation of safetysystems and for theaccess of thepersonnel toperform theimportant activities
related to safety;g) components
necessary to
prevent theradioactive leakage
from the irradiatedfuel that istransported orstored within the
nuclear installation,out of the corecooling system
during all states ofnormal andabnormaloperation;

France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Class

Class

Definition

Definition

Definition

NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

Definition

Canada(CSA N285.0-95)
Class
Definition

Czech Republic(Regulation 214/1997 Sb)
Class
Definitionh) components
necessary to
remove fission heatfrom the irradiatedfuel stored within
the nuclear
installation, out of
the core coolingsystem;i) components
requisite tomaintain sufficient
sub-criticality offuel stored within
the nuclear
installation, out of
the core coolingsystem;j) components
requisite to limit
the effluents or the
leakage of solid,
liquid or gaseousradioactive
substances andionizing radiation
below theestablished limiting
values during allstates of normal
and abnormal
operation;k) components
requisite to│ 135

France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Class

Class

Definition

Definition

Definition

136 │ NEA/CSNI/R(2018)12
USA / ASME Section III 37
Class

Definition

Canada(CSA N285.0-95)
Class
Definition

Czech Republic(Regulation 214/1997 Sb)
Class
Definition
perform the safetyfunctions related tothe power supply
or to the control ofother components
ranked as thesafety class 3;l) components
requisite toperform the safetyfunctions for theassurance offunctional
capability of othercomponents ranked
as the safetyclasses 1, 2 and 3,
that are not related
to the control or to
the power supply;m) components
necessary forprevention or
mitigation of theconsequences offailures of the othercomponents orconstructions ofsafety systems
ranked as thesafety classes one,
two and three.

France (RRC-P 900 R.4)

Germany (KTA 3201/3211)

Switzerland (ENSI-G01/d)

Class

Class

Class

Definition

Definition

Definition

NEA/CSNI/R(2018)12│ 137

10. Appendix
OPDE/CODAP bibliography

Unclassified

138 │ NEA/CSNI/R(2018)12
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,” Key Engineering Materials, 270-273:
1731-1736, 2004.

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 Conference on 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,” Proc. International Conference on Nuclear
Engineering, pp. 65-70, ICONE12-49273, Arlington, Virginia USA, 25-29 April 2004.

OPDE-6.

Sun, Y.C., Young, H.C., Yeon, K.C. “Application of a Piping Failure Database to Some
Nuclear Safety Issues in Korea,” Proc. International Conference on Nuclear
Engineering, pp. 175-179, ICONE12-49572, Arlington, Virginia USA, 25-29 April
2004.

OPDE-7.

Fleming, K. N. and Lydell, B. O. Y, (2004), “Use of Markov Piping Reliability Models
to Evaluate Time Dependent Frequencies of Loss of Coolant Accidents,” Proc.
ICONE-12, the 12th International Conference on Nuclear Engineering. 25-29 April
2004, Arlington, Virginia, U.S.A.

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,”
Proc. 2004 ASME Pressure Vessels and Piping (PVP) Conference, San Diego, CA, 2529 July 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.

Nuclear Energy Agency, The OECD/NEA Piping Failure Data Exchange Project,
Workshop on Database Applications, OPDE/SEC(2004)4, Boulogne-Billancourt,
France, 2005.

OPDE-11.

Choi, S.Y., Yang, J.E. and Choi, Y.H., “Piping Service Experience Related to Agingin OECD Pipe Failure Data Exchange (OPDE) Project,” Int. Workshop on Practical
Applications of Age-Dependent Reliability Models and Analysis of Operational Data,
Fontenay aux Roses, France 2005.

OPDE-12.

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

OPDE-13.

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

OPDE-14.

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

NEA/CSNI/R(2018)12│ 139

OPDE-15.

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-16.

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, 14-19 May
2006.

OPDE-17.

Olsson, A., Lydell, B. and Kochenhauer, M., “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, 2006.

OPDE-18.

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-19.

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-20.

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-21.

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-22.

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,”PVP2007-26373, Proc. 2007 ASME Pressure Vessel and Piping Division Conference,
San Antonio, Texas, 2007.

OPDE-23.

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-24.

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

OPDE-25.

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

OPDE-26.

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-27.

Pandey, M., Piping Failure Frequency Analysis Using OECD/NEA Data, Universityof Waterloo, Final Report, CNSC Research and Support Program. Report RSP-0236,
February 2008.

OPDE-28.

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

Unclassified

140 │ NEA/CSNI/R(2018)12
OPDE-29.

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

OPDE-30.

Yuan, X.X., Pandey, M.D., Riznic, J.R., (2008), “A Point Process Model for Piping
Failure Frequency Analysis using OPDE Data,” ICONE16-48078, Proc. 16th
International Conference on Nuclear Engineering ICONE-16, Orlando, Florida, 11-15
May 2008, American Society of Mechanical Engineers.

OPDE-31.

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

OPDE-32.

Reck, H., et al, Bewertung des Potentials für unentdeckten Schadens-fortschritt an
druckführenden Komponenten, GRS-A-3640, Gesellschaft für Anlagen- undReaktorsicherheit (GRS) mbH, Cologne, Germany, 2009.

OPDE-33.

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 and Power, 131, (5), art. 052901, Sept. 2009.

OPDE-34.

Koriyama, T. et al, “Study on Risk-Informed In-Service Inspection for BWR Piping,”
Journal of Nuclear Science and Technology, 46:846-873, 2009.

OPDE-35.

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

OPDE-36.

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-37.

Pandey, M.D., Yuan, X.X. and Riznic, J., “An Evaluation of Impact of Ageing in
OPDE Piping Failure Data,’ Proc. Risk-Informed Integrity Management Workshop,
NEA/CSNI/R(2009)2, Nuclear Energy Agency, Boulogne-Billancourt, France, 2009,
pp 245-258. https://www.oecd-nea.org/nsd/docs/2009/csni-r2009-2.pdf

OPDE-38.

Nordic PSA Group, R-Book: Reliability Data Handbook for Piping Components in
Nordic Nuclear Power Plants, Stockholm, Sweden, 2010. 38

OPDE-39.

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

OPDE-40.

Grebner, H. et al., Development and Test Applications of Methods and Tools for
Probabilistic Safety Analyses, GRS-A-3555, Gesellschaft für Analagen und
Reaktorsicherheit, Cologne, Germany, 2010.

OPDE-41.

Uwe-Klügel, J., Dinu, I.P. and Nitoi, M., Evaluation of LOCA Frequencies Using
Alternative Methods, EUR 24646 EN, Institute for Energy, EU Joint Research Centre,
Petten, The Netherlands, 2010.

38. Electronic version only on CD (includes methodology description, data input/output files andresult summaries). For information on the terms and conditions for how to access the restricted
R-Book, contact the NPSAG Secretariat; www.npsag.org.

NEA/CSNI/R(2018)12│ 141

OPDE-42.

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-43.

Lydell, B., Huerta, A., Gott, K. “Insights from PSA Applications of the OECD nuclear
Energy Agency (OECD/NEA) OPDE Database,” Paper #144, International Topical
Meeting on Probabilistic Safety Assessment and Analysis 2011, pp. 1696-1706, 2011.

OPDE-44.

Olsson A., Swaling V. H., Lydell B., “Experiences from Implementation of Updated
Reliability Data for Piping Components Using the R-Book,” Paper #056, Proc. ANS
PSA 2011 International Topical Meeting on Probabilistic Safety Assessment andAnalysis, American Nuclear Society, Wilmington, NC, 13-17 March 2011.

OPDE-45.

Lydell B., Mosleh A., Chrun D., “Enhanced Piping Reliability Models for Use inInternal Flooding PSA,” Paper #145, Proc. ANS PSA 2011 International Topical
Meeting on Probabilistic Safety Assessment and Analysis, American Nuclear Society,Wilmington, NC, 13-17 March 2011.

OPDE-46.

Cronvall, O., Männistö, I., Kaunisto, K. “On Applications Concerning OECD Pipe
Failure Database OPDE,” Proc. 11th International Probabilistic Safety Assessment andManagement Conference and the Annual European Safety and Reliability Conference
2012, pp. 2891-2901, 2012.

OPDE-47.

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

OPDE-48.

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.

OPDE-49.

OECD Nuclear Energy Agency, OECD/NEA Piping Failure Data Exchange Project(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 & Failures in Commercial Nuclear Power Plants”,
Proc. Int. Topical Meeting on Probabilistic Safety Assessment and Analysis 2013, pp.
663-672, 2013.

CODAP-3.

Coyne, K., Tobin, M., Siu, N., Röwekamp, 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, 21-24 May 2013,
Avignon, France.

CODAP-5.

OECD Nuclear Energy Agency, CODAP Topical Report on Flow-Accelerated
Corrosion (FAC) of Carbon Steel & 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
Mechanical Engineering Congress & Exposition, 2014.

Unclassified

142 │ NEA/CSNI/R(2018)12
CODAP-7.

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

CODAP-8.

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

CODAP-9.

OECD Nuclear Energy Agency, OECD/NEA Component Operational Experience,
Degradation & Aging 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, 28 September-1 October 2015, Seville, Spain, 2015.

CODAP-11.

Dragea, T., Riznic, J. “The Component Operational Experience Degradation andAgeing Program (CODAP): Review and Lessons Learned (2011-2014)”, 23th
International Conference on Nuclear Engineering, ICONE23-1001, Chiba, Japan, 1721 May 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.

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.

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

CODAP-16.

Lydell, B., Nevander, O. and Riznic, J., “The OECD/NEA CODAP Project & Its
Contribution to Plant Aging Management and Probabilistic Safety Assessment,” Paper
062, Proc. 13th International Conference on Probabilistic Safety Assessment andManagement (PSAM13), Seoul, Korea, 2016.

CODAP-17.

OECD Nuclear Energy Agency, CODAP Project Topical Report on Operating
Experience Insights into Pressure Boundary Component Reliability & Integrity
Management,” NEA/CSNI/R(2017)3, Boulogne-Billancourt, France, 2017.

CODAP-18.

Lydell, B., “A Review of the Progress with Statistical Models of Passive Component
Reliability,” Nuclear Engineering and Technology, 49:340-359, 2017.

CODAP-19.

Nevander, O., Riznic, J. and Lydell, B., “The OECD/NEA Contribution to Nuclear
Materials Performance Knowledge Preservation,” Paper C2018-10530, Proc. NACE
Corrosion 2018 Conference, NACE, Houston, TX, 2018.

CODAP-20.

Lydell, B., Fleming, K.N. and Roy, J-F., “Analysis of Possible

NEA/CSNI/R(2018)12

CODAP-21.

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Trends in the Estimation of Piping System Failure Rates for Internal Flooding PRA,”
Proc. International Conference on Probabilistic Safety Assessment and Management(PSAM 14), Los Angeles, CA, 16-21 September 2018.

Unclassified

144 │ NEA/CSNI/R(2018)12

11. Appendix
Glossary of technical terms

NEA/CSNI/R(2018)12

Austenitic Alloy Steel. Also high-alloy steels with the main alloying elements being
chromium (Cr) and nickel (Ni). Some high-alloy steels include niobium (Nb) to improve
welding properties, or titanium (Ti) to prevent intergranular corrosion and weld decay.
Below Ground Piping. See “underground piping” below.
BONNA® Pipe. A thin steel pipe embedded in reinforced concrete. It has rebar or a heavy
wire mesh embedded in the OD concrete.
Buried Piping. Piping that is below grade and in direct contact with soil. Buried piping isprovided with corrosion protection such as coating and cathodic protection.
Cathodic Protection (CP). A corrosion protection technique in which the potential
difference is applied to buried piping from an external power source or a more anodic
material (sacrificial anode) for the purpose of making the piping behave in a cathodic
manner. Through the use of CP, the corrosion rate is normally reduced to an acceptablelevel.
Component Boundary. Defines the physical boundary of a component required for system
operation. A component boundary definition should be consistent with the parameter
database supporting PRA model quantification. For piping components, the component
boundary is established through degradation mechanism evaluations (see below).
Concrete Encased Piping (CEP). Below ground piping that is embedded in concrete. The
piping is not easily extracted nor is the interior pipe surface readily accessible forinspection. The CEP category also includes piping recessed in plant building floors.
Corrosion Fatigue. The behaviour of materials under cyclic loading conditions iscommonly considered as consisting of two broad categories of material properties. One
category relates to cyclic life for the formation of a fatigue crack in a smooth test specimen,
the so-called S-N fatigue properties. The second relates to the growth of a pre-existing
crack. Laboratory test have shown that LWR coolant water can have a detrimental effecton both S-N fatigue properties and fatigue crack growth.
Crevice Corrosion. Crevice corrosion occurs in a wetted or buried environment when acrevice or area of stagnant or low flow exists that allows a corrosive environment to develop
in a component. It occurs most frequently in joints and connections, or points of contact
between metals and nonmetals, such as gasket surfaces, lap joints, and under bolt heads.
Carbon steel, cast iron, low alloy steels, stainless steel, copper, and nickel base alloys are
all susceptible to crevice corrosion. Steel can be subject to crevice corrosion in some cases
after lining/cladding degradation.
Cumulative Usage Factor (CUF). The cumulative usage factor (CUF) is the sum of theindividual usage factors (UFs; see below), and ASME Code Section III requires that the
CUF at each location must not exceed one.
Cured-in-Place-Pipe (CIPP). A BP temporary repair method. A resin-saturated felt tube
made of polyester, fiberglass cloth or a number of other materials suitable for resin
impregnation, is inverted or pulled into a damaged pipe. It is usually done from the
upstream access point (manhole or excavation).
Damage Mechanism. Excessive internal or external loading conditions that cause physicaldamage to a component pressure boundary. Pressure shocks from a water hammer might
damage pipe hangers and snubbers, or distort a piping section.

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Degradation Mechanism. Phenomena or processes that attack (wear, erode, crack, etc.) thepressure-retaining material over time and might result in a reduction of component pressure
boundary integrity. It should be noted that damage mechanisms (e.g. a sudden hydraulicpressure transient) and degradation mechanisms could interact to cause major, catastrophic
passive component failures.
Degradation Mechanism Evaluation. The identification of degradation mechanisms in apipe segment by comparing actual piping design and operating conditions to a well-defined
set of material and environmental attributes. The evaluation considers plant-specific service
experience involving cracking and leakage.
Discontinuity. A lack of continuity or cohesion; an interruption in the normal physicalstructure of material or a product.
Double-Walled Pipe. A double-walled pipe is a secondary contained piping system. It is apipe-within-a-pipe, or encased in an outer covering, with an annulus (interstitial space)between the two diameters. The inner pipe is the primary or carrier pipe and the outer pipe
is called the secondary or containment pipe.
Enhanced Visual Examination (EVT-1).The EVT-1 method is intended for the visual
examination of surface breaking flaws. Any visual inspection for cracking requires areasonable expectation that the flaw length and crack mouth opening displacement meetthe resolution requirements of the observation technique. The EVT-1 specification
augments the VT-i requirements to provide more rigorous inspection standards for stresscorrosion cracking. EVT-1 is also conducted in accordance with the requirements describedfor visual examination (i.e. VT-1) with additional requirements (such as camera scanningspeed). Any recommendation for EVT-1 inspection will require additional analysis toestablish flaw-tolerance criteria, which must take into account potential embrittlement due
to thermal aging or neutron irradiation. Acceptance criteria methodologies to support plantspecific augmented examinations are documented in WCAP-17096-NP 39.
Erosion Cavitation (E-C). This phenomenon occurs downstream of a directional change orin the presence of an eddy. Evidence can be seen by round pits in the base metal and isoften misdiagnosed as FAC (see below). Like erosion, E-C involves fluids acceleratingover the surface of a material; however, unlike erosion, the actual fluid is not doing thedamage. Rather, cavitation results from small bubbles in a liquid striking a surface. Suchbubbles form when the pressure of a fluid drops below the vapour pressure, the pressure atwhich a liquid becomes a gas. When these bubbles strike the surface, they collapse, or
implode. Although a single bubble imploding does not carry much force, over time, thesmall damage caused by each bubble accumulates. The repeated impact of these implosions
results in the formation of pits. Also, like erosion, the presence of chemical corrosion
enhances the damage and rate of material removal. E-C has been observed in PWR decay
heat removal and charging systems.
Erosion/Corrosion (E/C): “Erosion” is the destruction of metals by the abrasive action ofmoving fluids, usually accelerated by the presence of solid particles or matter insuspension. When corrosion occurs simultaneously, the term erosion-corrosion is used. In
the OPDE database the term “erosion/corrosion” applies only to moderate energy carbon
steel piping (e.g. raw water piping).

39. Westinghouse Electric Company LLC, Reactor Internals Acceptance Criteria Methodology andData Requirements, WCAP-17096-NP, Cranberry Township, PA, December 2009.

NEA/CSNI/R(2018)12

Ferritic Alloy Steel. Also low-alloy steels, which have a carbon content less than 0.2% andcontain a total of < 12% alloying elements (e.g., Cr, MN, Mo, Ni).
Fillet Weld. A weld of approximately triangular cross section joining two surfaces
approximately at right angles to each other in a lap joint, tee joint, corner joint, or socket
weld.
Flaw. An imperfection or unintentional discontinuity that is detectable by non-destructive
examination (NDE).
Flaw Aspect Ratio. Ratio of the length of the deepest crack to the depth of the deepest
crack.
Flow Accelerated (or Assisted) Corrosion (FAC). EPRI defines FAC as “a process whereby
the normally protective oxide layer on carbon or low-alloy steel dissolves into a stream offlowing water or water-steam mixture.” It can occur in single phase and in two phaseregions. According to EPRI, the cause of FAC is a specific set of water chemistry
conditions (e.g. pH, level of dissolved oxygen), and absent a mechanical contribution to
the dissolution of the normally protective iron oxide (magnetite) layer on the inside pipe
wall.
Full Structural Weld Overlay (FSWOL). A structural reinforcement and stress corrosion
cracking (SCC) mitigation technique through application of a SCC-resistant material layer
around 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.
General Corrosion. An approximately uniform wastage of a surface of a component,
through chemical or electrochemical action, free of deep pits or cracks.
High Energy Piping. A piping system for which the maximum operating temperature
exceeds 200 ºF (94.33 ºC) or the maximum operating pressure exceeds 275 psig (1.896
MPa).
Holiday in Pipe Coating. A holiday is a hole or void in the coating film which exposes theburied piping to corrosion.
Hydrostatic Pressure Test. A pressure test conducted during a plant or system shutdown at
a pressure above nominal operating pressure or system pressure for which overpressure
protection is provided.
Inclusion. An “inclusion” is a non-metallic impurity such as slag, oxide, and sulphide thatis present in the original ingot. During rolling of billets into bar stock, impurities are rolled
in a lengthwise direction. These direction-oriented inclusions in the finished product are
generally referred to as non-metallic inclusions or “stringers”. These stringers may besurface or subsurface and are usually short in length and parallel to the grain flow.
Indication. The definition of the term “indication” as it applies to NDE is: “A response orevidence of a response disclosed through NDE that requires further evaluation to determine
its true significance.”
Intergranular Stress Corrosion Cracking (IGSCC). IGSCC is associated in particular witha sensitised material (e.g. sensitised austenitic stainless steels are susceptible to IGSCC inan oxidizing environment). Sensitisation of unstabilised austenitic stainless steels is│ 147

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characterised by a precipitation of a network of chromium carbides with depletion ofchromium at the grain boundaries, making these boundaries vulnerable to corrosive attack.
Irradiation Assisted Stress Corrosion Cracking (IASCC). IASCC refers to intergranular
cracking of materials exposed to ionising radiation. As with SCC, IASCC requires stress,
aggressive environment and a susceptible material. However, in the case of IASCC, anormally non-susceptible material is rendered susceptible by exposure to neutron
irradiation. IASCC is a plausible ageing mechanism, in particular for PWR internal
components (e.g. baffle bolts).
In-service Pressure Test. A system pressure test conducted to perform visual examination
VT-2 while the system is in service under operating pressure.
Isometric Drawing. In the context of piping design, an isometric drawing is a threedimensional representation of a piping system showing the length, depth and width in asingle view. Piping isometrics are often used by designers prior to a stress analysis and arealso used by draftsmen to produce shop fabrication spool drawings. In-service inspection(ISI) engineers use isometric drawings to define and identify ISI locations.
JRC Operating Experience Clearinghouse (CE-OEF). Located in Petten, the Netherlands,
the Clearinghouse gathers nuclear safety experts performing the following technical tasks
in support to the EU member states:
•
•
•
•
•
•
•

“Topical Studies” providing in-depth assessment of either particularly significant
events or either families of events. These studies are drafted by experts on the topic
and based on an analysis of usually hundreds of event reports.
Trend analysis of events in order to identify priority areas.
Improvement of the quality of event reports submitted by the EU Member States tothe international reporting system jointly operated by the OECD-NEA and the
IAEA.
Reporting every three months the main events having occurred in NPPs.
Database: a European central OE repository being developed in order to ensure
long term storage of OE and to facilitate information retrieval.
Further to these activities, the EU clearinghouse is participating to several
international cooperation projects on OE, mainly through the IAEA and the OECDNEA working groups.
https://clearinghouse-oef.jrc.ec.europa.eu/

Latent Failure. A degraded material condition that may lie dormant for a long period before
leading to a visible flaw (e.g. through-wall crack, active leakage).
Leak Detection System. Instrumentation and controls that use various temperature,pressure, level and flow sensors to detect water and steam leakages in selected reactor
systems and to initiate annunciation and provide isolation signal (in certain cases) to limit
leakage from the reactor coolant pressure boundary when limiting leakage conditions
exists.
Leakage Pressure Test. A system pressure test conducted during operation at nominal
operating pressure, or when pressurised to nominal operating pressure and temperature.
Limiting Condition for Operation (LCO). According to the Technical Specifications 40, a
LCO is the lowest functional capability or performance level of a piece of equipment
40. “Betriebshandbuch” in German.

NEA/CSNI/R(2018)12

required for safe operation of a nuclear plant. When a LCO cannot be met, the reactor must
be shut down or the licensee must follow any remedial action permitted by the Technical
Specifications until the condition can be met.
Liquid Penetrant Examination. Liquid Penetrant Examination (LPT) uses liquids to detectcracks in materials. In the mid and late 1930's, Robert and Joseph Switzer worked withprocesses incorporating visible coloured dyes in the penetrant to give better contrast. In
1941 they introduced processes using fluorescent dyes which, when viewed under a black
light, produced contrasts superior to those obtainable with the visible dyes. The fluorescentmethod was quickly accepted by the military for aircraft part examination. Since then, theuse of both colour-contrast and fluorescent penetrants has spread to practically all fields ofmanufacturing, and new and improved PT products are constantly being developed.
Low-Frequency Electromagnetic Testing (LFET). This technique measures the changes inelectro-magnetic 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 oxide or 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 the implementation of a qualified NDE process have contributed to areportable or rejectable flaw remaining undetected for a certain period.
LTA-RIM. In this report, LTA-RIM is defined as events where degradation has progressedbeyond acceptable limits in systems, structures or components (SSCs) that have a RIM
program. 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 program 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 safetysignificance.
Mechanical Stress Improvement Process (MSIP®). A patented process that was invented,
developed and first used in 1986 by NuVision Engineering Inc. for mitigating stresscorrosion cracking in nuclear plant weldments. MSIP® works by using a hydraulicallyoperated clamp which contracts the pipe on one side of the weldment. A typical tool design
consists of a specially designed hydraulic box press for bringing the clamp halves together.
By contracting the pipe on one side of the weldment, the residual tensile stresses are
replaced with compressive stresses.
Moderate Energy Piping. A piping system for which the maximum operating temperatureis less than 200 ºF (94.33 ºC) or the maximum operating pressure is less than 275 psig
(1.896 MPa).
NDE Qualification. In the context of NDE, qualification includes technical justification,
which involves assembling all the supporting evidence for inspection capability (results ofcapability evaluation exercises, feedback from site experience, applicable and validated
theoretical models, physical reasoning), and may include practical trials using deliberately
defective test pieces.
Nondestructive Examination (NDE). An examination by the visual, surface, or volumetricmethod.

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Operating Agent. According to the CODAP operating procedures, to assure consistency ofthe data contributed by the national coordinators the project operates through an operating
agent. The operating agent verifies whether the event information provided by the national
coordinators complies with the CODAP coding guidelines (CG). It also verifies thecompleteness and accuracy of the data and assigns the quality index jointly with respective
national coordinator who has provided such data.
Optimized Weld Overlay (OWOL). A subset of the full structural weld overlay (FSWOL)
process. It has been developed for larger geometries (e.g. RCS Hot and Cold Leg nozzles)
where FSWOL application becomes too time consuming for a typical refuelling outage.
The optimised weld overlay thickness is less than that of a full structural weld overlay inorder allow completion in the time available in a typical refuelling outage for the larger
geometries.
Pattern Recognition. Pattern recognition is applied to the interpretation of event reports
with scarce (or unclear) details on failure location and root cause. In the context of event
data analysis, “pattern recognition” is a structured process of determining the cause ofdegradation using known failure patterns for similar piping systems. Data analysis andclassification builds on the retrieval of data on similar events, and performing acomparative analysis to determine the nature of apparent similarities between industry data
and the specific event.
Pipe Schedule Designation. The schedule number (SN) is defined as SN = 1000 × (P/SE),
where P is operating pressure in lb/in2 and SE is allowable stress range multiplied by joint
efficiency in lb/in2. Most US pipe failure reports include pipe schedule information.
Primary Water Stress Corrosion Cracking (PWSCC). PWSCC is a form of IGSCC and isdefined as intergranular cracking in primary water within specification limits (i.e. no needfor additional aggressive species – for example, IGSCC of Alloy 600 in primary water).
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.
Radiographic Examination. A non-destructive testing (NDE) method of inspecting
materials for hidden flaws by using the ability of short wavelength electromagnetic
radiation (high energy photons) to penetrate various materials.
Reliability and Integrity Management (RIM). Those aspects of the plant design andoperational phase that are applied to provide an appropriate level of reliability of SSCs anda continuing assurance over the life of the plant that such reliability is maintained.
Root Mean Square (RMS) Evaluation. NDE qualifications include the use of an ultrasonicsizing procedure which should be developed and qualified for equipment, technique, andsizing examination personnel. At least ten flawed specimens should be used in theperformance demonstration. A Root Mean Square (RMS) evaluation should be used todemonstrate adequate sizing performance. This is given by the formula:Where

𝑅𝑅𝑅𝑅𝑅𝑅 = �(𝑇𝑇 − 𝑈𝑈)2 /𝑁𝑁

T = Truth or actual flaw depth
U = Ultrasonic flaw depth estimate
N = Number of test specimens or flaws sized

NEA/CSNI/R(2018)12

Acceptable flaw sizing performance demonstration is achieved when the RMS is 12.5% or
less. This is comparable to the Appendix VIII criteria proposed in ASME Code Section XI.
Accordingly, it was demonstrated that at an RMS of 15% or less, acceptable sizing
performance is achieved comparable to the current EPRI NDE centre intergranular stresscorrosion cracking (IGSCC), sizing programme. The advanced ultrasonic sizing techniques
described in this handbook have been developed in accordance with recommended
guidelines of the EPRI NDE centre ultrasonic planar flaw sizing of IGSCC. Variations ormodifications of the techniques have been incorporated to improve accuracy of flaw depth
sizing of stress corrosion, thermal fatigue and mechanical fatigue cracks.
Root Cause Analysis (RCA). RCA is a collective term that describes a wide range ofsystematic approaches and techniques used to uncover causes of problems. Root cause is afactor that causes a non-conformance (e.g. structural degradation or failure) and should bepermanently eliminated through process improvement.
SAFT. Synthetic Aperture Focusing Technique is a signal processing technique which takes
advantage of the movement of a small conventional transducer to simulate, in effect, aphased array that is extremely long. This allows high resolution at long range, withrelatively small transducers. SAFT in ultrasonics has been around for over twenty years but
the amount of processing required has meant that it has had to wait for developments incomputing technology before it can be readily applied. Phased array techniques havedeveloped at a faster pace than SAFT, however.
Seal Weld. A fillet weld used on a pipe joint primarily to obtain fluid tightness as opposed
to mechanical strength. It is usually used in connection with a threaded joint.
Selective Leaching. Also referred to as dealloying, demetalification, parting and selective
corrosion, is a corrosion type in some solid solution alloys, when in suitable conditions acomponent of the alloys is preferentially leached from the material. The less noble metal isremoved from the alloy by a microscopic-scale galvanic corrosion mechanism. The mostsusceptible alloys are the ones containing metals with high distance between each other inthe galvanic series, e.g. copper and zinc in brass.
Socket Weld. Fillet-type weld used to join pipe to valves and fittings or to other sections ofpipe. Generally used for piping whose nominal diameter is 50 mm or smaller.
Strain Induced Corrosion Cracking (SICC). SICC is used to refer to those corrosion
situations in which the presence of localised dynamic straining is essential for crack
formation (i.e. initiation and propagation) to occur, but in which cyclic loading is either
absent or restricted to a very low number of infrequent events. SICC has been observed inparticular in pressurised components in German nuclear power plants made of higherstrength carbon steel and low-alloy steel.
Stress Corrosion Cracking (SCC). SCC is a localised non-ductile failure which occurs only
under the combination of three factors: 1) tensile stress, 2) aggressive environment, and 3)
susceptible material. The SCC failure mode can be intergranular (IGSCC), or transgranular(TGSCC). In a nuclear power plant operating environment, primary water SCC (PWSCC),
and irradiation assisted SCC (IASCC) are also defined.
Structured Query Language (SQL). A standard computer language for relational databasemanagement and data manipulation. SQL is used to query, insert, update and modify data.
Sweepolet (Weldolet). Tradename for a contoured, integrally reinforced, butt-weldedbranch connection.

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TECHITE® Pipe. Fiberglass (or Fiber Reinforced Polymer) reinforced mortar pipe. This
type of piping has found very limited use in cooling tower blowdown/discharge
applications. This material can be affected by the environment, becoming brittle or soft,and breaking or leaking.
Thermal Stratification. Hot water can flow above cold water in horizontal runs of piping
when the flow (hot water into a cold pipe or cold water into a hot pipe) does not haveenough velocity to flush the fluid in the pipe. The temperature profiles in the pipe where
the top of the pipe is hotter than the bottom causes the pipe to bow along with the normal
expansion at the average temperature.
Transgranular Stress Corrosion Cracking (TGSCC). TGSCC is caused by aggressive
chemical species especially if coupled with oxygen and combined with high stresses.
Tritium. Tritium is a naturally occurring radioactive form of hydrogen that is produced in
the atmosphere when cosmic rays collide with air molecules. As a result, tritium is foundin very small or trace amounts in groundwater throughout the world. It is also a byproductof the production of electricity by nuclear power plants. Tritium emits a weak form ofradiation, a low-energy beta particle similar to an electron. The tritium radiation does nottravel very far in air and cannot penetrate the skin.
Tritium in Nuclear Power Plants. Most of the tritium produced in nuclear power plants
stems from a chemical, known as boron, absorbing neutrons from the plant's chain reaction.
Nuclear reactors use boron, a good neutron absorber, to help control the chain reaction.
Toward that end, boron either is added directly to the coolant water or is used in the control
rods to control the chain reaction. Much smaller amounts of tritium can also be produced
from the splitting of Uranium-235 in the reactor core, or when other chemicals (e.g. lithiumor heavy water) in the coolant water absorb neutrons. Like normal hydrogen, tritium canbond with oxygen to form water. When this happens, the resulting “tritiated” water isradioactive. Tritiated water (not to be confused with heavy water) is chemically identicalto normal water and the tritium cannot be filtered out of the water.
Underground Piping. Piping that is below grade, but is contained within a tunnel or vault
such that it is contact with air and is located where access for inspection is restricted.
Unified Numbering System (UNS). An alloy designation system in use in North America.
It consists of a prefix letter and five digits designating a material composition. For example,
a prefix of S indicates stainless steel, C indicates copper, brass or bronze alloys.
Usage Factor (UF). Cyclic loadings on a structural component occur because of changes
in mechanical and thermal loadings as the system goes from one load set (e.g. pressure,temperature, moment, and force loading) to another. For each load set, an individual fatigue
usage factor (UF) is determined by the ratio of the number of cycles anticipated during the
lifetime of the component to the allowable cycles.
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 physicalcondition including such irregularities as cracks, wear, erosion, corrosion, or physicaldamage.
VT-2 Examination. Per ASME XI, a visual surface examination to locate evidence ofleakage from pressure-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

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and their supports, such as the verification of clearances, settings, physical displacements,
lose or missing parts, debris, corrosion, wear, erosion, or the loss of integrity at bolted orwelded 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 individual components withsignificant levels of existing degradation. As the VT-3 examination is not intended to detect
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 VT-3 examination may be appropriate for inspecting highly
redundant components (such as baffle-edge bolts), where a single failure does notcompromise 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 orin 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, VT2 and VT-3.
Water Hammer. If the velocity of water or other liquid flowing in a pipe is suddenly
reduced, a pressure wave results, which travels up and down the pipe system at the speedof sound in the liquid. Water hammer occurs in systems that are subject to rapid changes
in fluid flow rate, including systems with rapidly actuated valves, fast-starting pumps, andcheck valves.
WEKO-SEAL®. A flexible rubber leak clamp that ensures a non-corrodible, bottle-tight sealaround the full inside circumference of the pipe-joint area. The design incorporates a seriesof proprietary lip seals that create a leak proof fit on either side of the joint.
Weld Inlay. A mitigation technique defined as application of PWSCC-resistant material(Alloy 52/52M) to the inside diameter of a dissimilar metal weld that isolates the PWSCCsusceptible material (Alloy 82/182) from the primary reactor coolant.
Weldolet. The most common of all branch connections, and is welded onto a largerdiameter pipe. The ends are bevelled to facilitate this process, and therefore the “weldolet”
is considered a butt-weld fitting. Weldolets are designed to minimise stress concentrations
and provide integral reinforcement.

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