Nuclear Safety
NEA/CSNI/R(2019)13
October 2022
www.oecd-nea.org

NEA CODAP Project

Topical Report

A Review of the “Post-1998”
Experience with Thermal Fatigue
in Heavy Water and Light Water
Reactor Piping Components

Nuclear Energy Agency

NEA/CSNI/R(2019)13
Unclassified

English text only
5 October 2022

NUCLEAR ENERGY AGENCYCOMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSNEA CODAP PROJECT TOPICAL REPORTA REVIEW OF THE “POST-1998” EXPERIENCE WITH THERMAL FATIGUE IN
HEAVY WATER AND LIGHT WATER REACTOR PIPING COMPONENTS

This document is available in PDF format only.

JT03504122
OFDE

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2 │ NEA/CSNI/R(2019)13

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OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 3

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSThe Committee on the Safety of Nuclear Installations (CSNI) addresses Nuclear Energy
Agency (NEA) programmes and activities that support maintaining and advancing thescientific and technical knowledge base of the safety of nuclear installations.
The Committee constitutes a forum for the exchange of technical information and forcollaboration between organisations, which can contribute, from their respective
backgrounds in research, development and engineering, to its activities. It has regard to theexchange of information between member countries and safety R&D programmes ofvarious 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 and techniques and of safety assessments, and ensures that operating experience isappropriately accounted for in its activities. It initiates and conducts programmes identified
by these reviews and assessments in order to confirm safety, overcome discrepancies,
develop improvements and reach consensus on technical issues of common interest. It
promotes the co-ordination of work in different member countries that serve to maintainand enhance competence in nuclear safety matters, including through joint undertakings(e.g. joint research and data projects), and assists in the feedback of the results toparticipating organisations. The Committee ensures that valuable end products of thetechnical reviews and analyses are provided to members in a timely manner, and madepublicly available when appropriate, to support broader nuclear safety.
The Committee focuses primarily on the safety aspects of existing power reactors, othernuclear installations and new power reactors; it also considers the safety implications ofscientific and technical developments of future reactor technologies and designs. Further,
the scope for the Committee includes human and organisational research activities andtechnical developments that affect nuclear safety.

4 │ NEA/CSNI/R(2019)13

List of abbreviations and acronyms
AA
AFW
AMP
AMPP
ANVS
ASME
ASN
BR
BWR
CANDU
CBF
CEA
CEP
CFC
CIPP
CODAP
COG
CP
CRD
CRDRL
CSA
CSN
CSNI
CUF
CVC
CVCS
DHR
EAF
EBS
ECCS
EDF

Asea-atom (Sweden)
Auxiliary feed-water
Ageing management programme
Association for Materials Protection and Performance (formerly NACE
International)
Authority for Nuclear Safety and Radiation Protection (Netherlands)
American Society of Mechanical Engineers
Autorité de Sûreté Nucléaire (Nuclear Safety Authority, France)
Backing ring
Boiling water reactor
Canada Deuterium Uranium
Cycle-based fatigue
Commissariat à l'énergie atomique et aux énergies alternatives(Alternative Energies and Atomic Energy Commission, France)
Concrete encased piping
Corrosion-fatigue cracking
Cured-in-place-pipe
Component operational experience, degradation and ageing programme
CANDU owners group
Cathodic protection
Control rod drive
Control rod drive return line
Canadian Standards Association
Consejo de Seguridad Nuclear (Nuclear Safety Council, Spain)
Committee on the Safety of Nuclear Installations (NEA)
Cumulative usage factor
Chemical and volume control
Chemical and volume control system
Decay heat removal
Environmentally assisted fatigue
Equivalent break size
Emergency core cooling system
Électricité de France

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 5
ENSI
EOL
EPR
EPRI
ESF
FAC
FAMOS
FW
GALL
GRS
HCF
HCTF
HDPE
HPSI
HTS
IAEA
IAGE
ICES
IGALL
IN
INES
INPO
IRS
IRSN
ISI
JSME
KTA
LB
LCF
LCO
LOCA
LPCI
LPSI
LTO
LWR
MB
MB-LOCA
MFL

Eidgenössische Nuklearsicherheitsinspektorat (Federal Nuclear
Safety Inspectorate, Switzerland)
End of Life
European Pressurised Reactor
Electric Power Research InstituteEngineered safety features
Flow-Accelerated Corrosion
FAtigue-MOnitoring System
Feed-water
Generic Ageing Lessons Learned
Gesellschaft für Anlagen- und Reaktorsicherheit (Germany)
High-cycle fatigue
High-cycle thermal fatigue
High-density polyethylene
High-pressure safety injection
Heat transport system
International Atomic Energy Agency
CSNI Working Group on Integrity and Ageing of Components andStructures
INPO consolidated events database
International GALL
Information noticeInternational Nuclear and Radiological Event Scale
Institute of Nuclear Power Operations (United States)
International Reporting System for Operating Experience
Institut de Radioprotection et de Sûreté Nucléaire (Institute ofRadiological Protection and Nuclear Safety, France)
In-service inspection
Japan Society of Mechanical Engineers
Kerntechnische Ausschuss (Nuclear Safety Standards Commission,Germany)
Large breakLow-cycle fatigue
Limiting condition for operation
Loss-of-coolant-accident
Low-pressure coolant injection
Low-pressure safety injection
Long-term operation
Light-water reactor
Management board
Medium break LOCA
Magnetic flux leakage

6 │ NEA/CSNI/R(2019)13
MHI
MRP
MS
NACE International
NDE
NDT
NEA
NEI
NESC
NGS
NRA
NRC
NSSS
OE
OECD
OPDE
OPG
PHTS
PHWR
PIG
PORV
PrBWS
PRG
PSA
PVP
PWR
PWSCC
PZR
RCIC
RCS
RFM
RHR
RPS
RPV
RCIC
RCPB
RHRS
RWCU
SB
SBF

Mitsubishi Heavy Industries
Materials reliability programme
Main steam
National Association of Corrosion Engineers International (now AMPP)
Non-destructive examination
Non-destructive testing
Nuclear Energy Agency
Nuclear Energy InstituteNetwork for evaluating structural components
Nuclear generating station
Nuclear Regulation Authority (Japan)
Nuclear Regulatory Commission
Nuclear Steam Supply System
Operational experience
Organisation for Economic Co-operation and Development
OECD Pipe failure data exchange project
Ontario power generation
Primary heat transport system (CANDU)
Pressurised heavy water reactor
Pipeline inspection gauge
Pilot-operated relief valve (usually for reduce pressure of PWR
pressuriser)
Pressure bearing water system
Project review group (of CODAP)Probabilistic safety assessment
Pressure Vessels and Piping
Pressurised water reactor
Primary water stress corrosion cracking in dissimilar metal weld locations
Pressuriser
Reactor core isolation cooling
Reactor coolant system
Rain flow method
Residual heat removal (Shutdown cooling)
Reactor protection system
Reactor pressure vessel
Reactor core isolation cooling
Reactor coolant pressure boundary
Reactor Heat Removal System
Reactor water clean-up
Small breakStress-based fatigue

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 7
SEAS
SG
STUK
SÚJB
SVTI
TA
TASCS
TF
TFFG
TGSCC
THERFAT
TLAA
TT
UNS
VT
WANO
WE
WGIAGE
WWER

Slovenské Elektrárne AS (Electric utility company, Slovak Republic)
Steam generator
Radiation and Nuclear Safety Authority of Finland
State Office for Nuclear Safety (Czech Republic)
Schweizerischer Verein für technische Inspektionen (Switzerland)
Volume control system in German´s BWR technology
Thermal stratification, cycling and striping
Thermal fatigue
Thermal Fatigue Focus Group
Transgranular stress corrosion cracking
Thermal fatigue evaluation of piping system tee connections
Time-limited ageing analysis
Thermal transient
Unified numbering system
Visual Testing
World Association of Nuclear Operators
Westinghouse Electric
Working Group on Integrity and Ageing of Components and Structures(NEA)
Water-Water Energetic Reactor (in Russian: VVER)

8 │ NEA/CSNI/R(2019)13

Table of contents
Executive summary ............................................................................................................................... 9
1. Introduction ..................................................................................................................................... 11
1.1. Background and related work ..................................................................................................... 11
1.2. Nomenclature .............................................................................................................................. 11
1.3. Objectives and scope................................................................................................................... 14
1.4. Disclaimer ................................................................................................................................... 15
1.5. Report structure and reading guide ............................................................................................. 16
2. Thermal fatigue phenomena........................................................................................................... 17
2.1. Description of TF phenomena .................................................................................................... 172.2. Thermal stratification .................................................................................................................. 18
2.3. Thermal cycling .......................................................................................................................... 18
2.4. Thermal striping .......................................................................................................................... 19
2.5. Valve in-/out-leakage .................................................................................................................. 19
2.6. Thermal sleeve mixing................................................................................................................ 21
2.7. Thermal transient (TT) fatigue.................................................................................................... 222.8. Significant historical events ........................................................................................................ 22
3. Safety significance of TF failures ................................................................................................... 25
3.1. Consequences of TF failures....................................................................................................... 25
3.2. Considerations of TF in probabilistic safety assessment ............................................................ 26
4. Operating experience data review ................................................................................................. 29
4.1. High-level database summary ..................................................................................................... 29
4.2. Specific TF operating experiences .............................................................................................. 32
5. Conclusions and recommendations................................................................................................ 64
5.1. Conclusions................................................................................................................................. 64
5.2. Recommendations ....................................................................................................................... 65
References ............................................................................................................................................ 66
Annex A OPDE/CODAP Bibliography ............................................................................................. 69
Annex B Glossary of technical terms ................................................................................................. 75

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 9

Executive summary
The implementation of the Nuclear Energy Agency (NEA) Component operational
experience, degradation and ageing programme (CODAP) is the continuation of the 20022011 NEA Pipe failure data exchange project (OPDE) and the 2006-2010 NEA Stresscorrosion cracking and cable ageing project (SCAP). In December 2017, 13 countries andeconomies (Canada, the Czech Republic, Finland, France, Germany, Japan, Korea, theNetherlands, the Slovak Republic, Spain, Switzerland, Chinese Taipei and the UnitedStates) agreed on a third term for CODAP (2018-2020). Sigma-Phase Inc. from the UnitedStates works as the operating agent of CODAP.
Since its inception in 2002, the operating experience with thermal fatigue mechanisms has
been an intrinsic aspect of the technical scope of CODAP, which is a joint database project
within the Committee on the Safety of Nuclear Installations (CSNI). This sixth topicalreport documents the results of thermal fatigue operating experience evaluations performedby the CODAP management board (MB). The technical emphasis is on the lessons learnt
since the publication of the report “Specialist Meeting on Experience with Thermal Fatigue
in light-water reactor (LWR) Piping by Mixing and Stratification” (NEA/CSNI/R(98)8).
This report highlights the need for close co-operation between plant designer and plantowner and, among plant personnel, between maintenance and operation staff to keep therisk of thermal fatigue under control. This sixth topical report addresses the accrued thermal
fatigue operating experience since 1998 and the different programmes to monitor andmitigate thermal fatigue damage in nuclear power plant piping. The CODAP MB identified
254 pipe failures attributed to thermal fatigue. Of the total data subset, 31% (or 80 events)
occurred post-1998. The insights and lessons learnt from evaluating the post-1998
operating experience is the subject of the sixth CODAP topical report.
High-cycle fatigue (HCF) failures are the most prevalent of fatigue events recorded in
CODAP. It is noted that the number of thermal fatigue failures, especially HCF failures,
decreased among CODAP participants after numerous international discussions led toprojects in this area at the end of the 1990s.
The following actions appear to be most effective in preventing thermal fatigue failures:•

Identifying potential locations with a higher risk of stratification based on operating
experiences and research results, especially locations with a risk of HCF.

•

Implementing design modifications to decrease thermal fatigue degradation in thehigh-risk locations. Design modifications will vary depending on reactor designand could include flow circulation changes to avoid stratification, material changes
to avoid quick deterioration, and process and procedure changes to avoid hightemperature differences. Most participants mentioned the optimisation of theoperating procedures based on operating experiences as an effective method to
avoid fatigue events.

10 │ NEA/CSNI/R(2019)13•

Conducting online monitoring of critical locations during operation based on
detection and screening of critical locations. The monitoring results should becontinuously analysed together with the results of periodic inspections.

•

Developing more precise NDE methods, techniques and procedures for periodic
inspections.

The following international activities could be undertaken to support activities in eachcountry to prevent thermal fatigue events:•

Developing more precise and consistent NDE methods, techniques and proceduresfor periodic inspections. An international benchmark to establish the reliability ofdetecting existing cracks could support this objective. The benchmark could
include, for example, a comparison of the detection techniques, calibratingprocedures of the equipment for the plant’s environment, and ultrasonicexamination procedures.

•

Understanding the design modifications and online monitoring systems used tomitigate thermal fatigue and stratification. This information could be sharedthrough international workshops on fatigue during the next five years, and then leadto the development of a consistent set of recommendations on best practices tomanage thermal fatigue.

In summary, thermal fatigue damage continues to be of concern with respect to the safe
long-term operation of nuclear power plants. The intended audience for this report includes
ageing management programme (AMP) managers, in-service inspection (ISI) specialists,
structural integrity engineers and nuclear safety specialists.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 11

1. Introduction
Since 2002, the NEA has operated a joint event database project that collects information
on passive metallic component degradation and failures. The scope of the database includes
primary system piping components, reactor pressure vessel internals (“reactor
components”), main process and standby safety systems, and support system (i.e. ASME
code class one, two and three, or equivalent) components, as well as non-safety-related(non-code) components whose degradation or failure can have significant operational
impact. With an initial focus on piping system components (the OPDE Project) the scopeof the project in 2011 was expanded to also address the reactor pressure vessel and internals
as well as certain other metallic passive components that are susceptible to damage ordegradation. In recognition of the expanded scope, the MB approved the transition of
OPDE to a new, expanded NEA component operational experience, a degradation andageing programme (CODAP). The 2018-2020 work programme includes the preparationof topical reports to foster technical co-operation and to deepen the understanding ofdifferences among participants in plant ageing management. The sixth CODAP topicalreport is concerned with the operating experience of thermal fatigue in heavy water and
LWR piping systems. The focus of this report is on the post-1998 thermal fatigue operating
experience.

1.1. Background and related work
Fatigue mechanisms have caused significant challenges to nuclear power plant operability,
including forced reactor shutdowns and extended outages. Since the 1970s, significant
research and development efforts have been devoted to better understanding the underlying
physical phenomena in order to prevent reoccurrences and to the development of fatigue
management programmes that ensure the long-term structural integrity of piping pressure
boundary components. Recent operating experience data point to new technical challenges
and the need for additional research. The new fatigue-related challenges include, but are
not limited to, evidence of thermal fatigue in locations believed to have been mitigated 20
or 30 years ago through piping design solutions implemented to resolve potential cracking
and through-wall leakages. Also among the new challenges are potential limitations or“gaps” in existing fatigue management guidelines that have resulted in “out-of-scope”
inspection locations degrading and failing due to fatigue mechanisms.
High-level summaries of the fatigue-specific operating experience as recorded in CODAP
are included in Figures 1.1 through 1.3. In numerous cases, piping failures attributed tofatigue mechanisms have had significant operational impact.

1.2. Nomenclature
Metallic fatigue is “the process of progressive localised permanent structural change
occurring in material subjected to conditions which produce fluctuating stresses and strainsat some point or points and which may culminate in crack or complete fracture after asufficient 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12 │ NEA/CSNI/R(2019)13
fatigue (TF); and 4) high-cycle fatigue (HCF). Approximately 80% of the recorded fatigue
failures recorded in CODAP are due to HCF of small-diameter butt welds and socket welds.
Figure 1.1. Fatigue of metallic piping components as documented in CODAP

Figure 1.2. Operational impact of fatigue-induced pipe failures 1

1.

Note the logarithmic scale.

OECD/NEA CODAP PROJECT TOPICAL REPORT:CF = Corrosion-Fatigue; COR = Corrosion ; D&C = Construction / Fabrication Flaw; E-C = Erosion-Cavitation; TF = Thermal Fatigue; E/C = Flow
Assisted Corrosion (incl. Erosion-Corrosion, Flow-Accelerated Corrosion, Liquid Droplet Impingement Erosion); HCF = High-Cycle Fatigue;
LCF = Low-Cycle Fatigue; SCC = Stress Corrosion Cracking.

Figure 1.3. Fatigue mechanisms versus “other” degradation mechanisms

NEA/CSNI/R(2019)13 │ 13

14 │ NEA/CSNI/R(2019)13
Corrosion-fatigue, or “environmentally assisted fatigue”, is the behaviour of materials
under cyclic loading conditions and in a corrosive environment. It is considered to be madeup of a region (or life) associated with the formation of an engineering-sized crack and aregion 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 socalled S-N fatigue properties (stress versus number of cycles). The second relates to thegrowth of a pre-existing crack. Laboratory test have shown that LWR coolant water canhave a detrimental effect on both S-N fatigue properties and fatigue crack growth. Muchlower failure stresses and much shorter failure times can occur in a corrosive environment
compared to a situation where the alternating stress is in a non-corrosive environment.
In CODAP, the corrosion-fatigue event population is small, with 52 recorded events, mostof which involve recordable/rejectable indications per ASME XI definitions. An example
of material degradation induced by corrosion-fatigue is that found through metallographic
examination of cracked code class two feed-water reducers at the now decommissioned
Trojan nuclear power plant in 1987. Although not applicable to the CODAP work scope,
the most significant corrosion-fatigue failure to date is the major primary coolant leakage
that occurred at the Russian plant Kola unit 2 on 3 March 1994, when a 2-inch make-up
system pipe ruptured. 2
TF is due to the cyclic stresses that result from changing temperature conditions in acomponent or in the piping attached to the component. TF 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).

1.3. Objectives and scope
On 8-11 June 1998, France’s Institut de Radioprotection et de Sûreté Nucléaire (IRSN)
hosted the “Specialist Meeting on Experience with TF in LWR Piping by Mixing andStratification” [NEA/CSNI/R(98)8] [1]. The meeting was prompted by numerous
significant operational events and the lessons learnt from the root cause evaluations and therelated R&D to better understand the underlying thermal-hydraulic phenomena. The
meeting was co-sponsored by the World Association of Nuclear Operators (WANO) andthe NEA Working Group on Integrity and Ageing of Components and Structures(WGIAGE) of the CSNI. This workshop was a precursor to multiple international
initiatives to better monitor and understand TF and a starting point for a series ofconferences on fatigue in reactor components as well as work programmes under the
auspices of the WGIAGE. Please see the following documents for any additional details:

2.

•“Proceedings of the International Conference on Fatigue of Reactor Components”,
August 2000, Napa, California, United States, NEA/CSNI/R(2000)24 [2].

•“Proceedings of the EPRI/USNRC/OECD International Conference on Fatigue ofReactor Components”, July 2002, Snowbird, Utah, United States,
NEA/CSNI/R(2003)2 [3].

International Atomic Energy Agency (IAEA), “Draft Report of a Consultants Meeting on a
Primary Coolant Leak at Kola 2 NPP Due to the Rupture of a Make-up Pipe”, WWER-SC-112,
Vienna, Austria, 1995.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 15•

“Proceedings of the Third International Conference on the Fatigue of Reactor
Components”, 3-6 October 2004, Seville, Spain, NEA/CSNI/R(2004)21 [4].

•“Thermal Cycling in LWR Components in NEA Member Countries”, CSNIWGIAGE, NEA/CSNI/R(2005)8 [4].

•“Proceedings of the Fourth International Conference on the Fatigue of Reactor
Components”, 28 September-1 October 2015, Seville, Spain,
NEA/CSNI/R(2017)2 [6].

Additionally, in 2003 the Network for Evaluation of Structural Components (NESC) set up
a project involving both utilities and R&D organisations. Co-ordinated by the Institute forEnergy at the European Commission Joint Research Centre in Petten (Netherlands), the
NESC aims to produce a consensus methodology for assessing high-cycle thermal fatigue
in piping components, with special attention to turbulent mixing phenomena at mixing teesin LWR systems. Completed in 2007, the work focused on two main aspects: 1) creating a
database of service and mock-up data for improved understanding of TF damage
mechanisms, and 2) developing a European multi-level TF damage assessment procedurethat accounts for thermal-hydraulics, material properties, strain evaluation through finite
element analysis, operating experience, fracture mechanics and ISI effectiveness. At theconclusion of the project, the NESC TF database included about 40 TF events [7].
This topical report documents the results of evaluations of the accrued TF operating
experience since 1998 and the different programmes to monitor and mitigate TF damage
in nuclear power plant piping. The CODAP MB has identified about 250 pipe failures
attributed to TF. Of the total data subset, about 32% (or 80 events) occurred post-1998. The
specific objectives of this report are:•

To provide a brief overview of significant historical events, i.e. events that occurredprior to 1999 and resulted in industry and regulatory actions.

•

Based on the information stored in the CODAP event database, to summarise theinsights and results of root cause evaluations of TF events that have occurred since
the publication of NEA/CSNI/R(98)8 [1].

•Document recent examples of specific operational events attributed to TF, with anemphasis of remedial actions by industry and regulator.

•

Determine the potential consequences of TF failures and any impacts on long-term
operation.

•

Provide examples of current TF management strategies, including the related nondestructive examination (NDE) procedures and NDE programme implementations.

•

Demonstrate how to use the TF failure data in CODAP to draw qualitative andquantitative insights about structural reliability and integrity management.

•

Provide some international-level conclusions and recommendations to help
participants decrease the number of fatigue failures and the risks to safe operation.

1.4. Disclaimer
The CODAP project places strong emphasis on data quality, including the completenessand comprehensiveness of recorded events. Data quality is achieved through a formal
validation process as articulated in a coding guideline. The roles and responsibilities with

16 │ NEA/CSNI/R(2019)13
respect to data submissions and data validation are defined in the CODAP operatingprocedures. The CODAP event database relies on information provided by respective MB
members.
The results of the data processing and analysis that are presented herein are based as muchas possible on the results of root cause analyses. However, the writing team for this topicalreport has not performed any reassessments of the findings related to the degraded passive
components as documented in the root cause analysis reports that are embedded in the
CODAP event database.
The CODAP MB is fully aware of the fact that the full root cause analysis documentation
as prepared by an owner/operator or its subject matter experts is not normally disseminatedbeyond the industry and regulators. The CODAP coding guideline includes instructions forwhat “root cause information” to include in the database. As a guiding principle, theinstructions provided state that any relevant information on a cause-consequence
relationship is to be included. The co-ordinators in the respective countries/economies
assume responsibility for the accuracy of the technical information that is input to the event
database. Furthermore, the web-based database has provisions for uploading (or attaching)
any available supporting information; e.g. laboratory reports, root cause analysis reports.

1.5. Report structure and reading guideThis topical report consists of six sections and three appendices. Section 2 includes asummary of the key TF phenomena. Also included in Section 2 is a summary of significant
historical TF events. Section 3 explains the reasons why TF operating experience data iscollected and analysed. Section 3 also addresses the potential safety significance of TF
failures. Section 4 summarises the specific operating experiences in countries andeconomies. The summary and conclusions are documented in Section 5. Finally, a list ofreferences is provided in Section 6.
Annex A is an OPDE/CODAP bibliography including references to database applications
performed or sponsored by OPDE/CODAP member organisations since 2002. Annex B isa glossary of terms.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 17

2. Thermal fatigue phenomena
Thermal fatigue (TF) is due to the cyclic stresses that result from changing temperature
conditions in a component or in the piping attached to the component. TF may involve a
relatively low number of cycles at a higher strain (e.g. plant operational cycles or injectionof 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). According to ASTM International 3 ,fatigue is “the process of progressive localised permanent structural change occurring inmaterial subjected to conditions which produce fluctuating stresses and strains at somepoint or points, and which may culminate in crack or complete fracture after a sufficientnumber of fluctuations.”

2.1. Description of TF phenomena
TF can occur by a number of mechanisms. When a piping system is exposed to a thermaltransient, through-wall thermal gradient stresses occur, which is a function of the rate at
which the pipe cross-section responds to sudden changes in temperature. These stresses are
considered in the ASME code design stress analysis and, although they can cause a fatigue
failure, they have been accounted for in the design. Piping thermal expansion, if constrained
and occurring with sufficient cycles, can also cause fatigue damage and is considered.
However, there are several fatigue phenomena that are not considered in the ASME code
design stress analysis, including: 1) thermal stratification, 2) thermal cycling, 3) thermal
striping, 4) valve in- /out-leakage, and 5) thermal sleeve mixing.
TF cracks initiate and grow because of alternating thermal loads. The loads cause
tensional/compressive stresses due to gradients in thermal expansion. The micromechanism for crack initiation and growth by thermal loads is basically the same as formechanical loads but the resulting crack morphology, in some respects, deviates from thatof mechanical fatigue. A major difference between mechanical and TF is the crack patternon the surface of the component. With few exceptions, TF cracking results in a large
number of cracks. In addition, the crack orientation is distributed more randomly in TF
cracking compared to mechanical fatigue due to the absence of a well-defined major stressdirection under thermal loads. The cracks often occur in a fabric-like pattern known as“cobble stone” pattern. In the through-wall direction, TF cracks are characterised by a lowtendency for crack branching and transgranular crack growth. The shape in the throughwall direction is in most cases straight, although bent and winding cracks have been
observed. TF cracking in ferritic low-alloy steels and stainless steels at high temperaturescauses oxidation of the crack surfaces and at the crack tip [8].
Adapted from References [9][10], the following sections provide additional details on thefive TF phenomena, including examples of operating experience to help illustrate howthese phenomena evolve.

3.

Known until 2001 as the American Society for Testing and Materials.

18 │ NEA/CSNI/R(2019)13

2.2. Thermal stratification
Thermal stratification is a condition in which two streams of fluid of different temperatures
flow in separate layers without appreciable mixing. When the flow rate is low, turbulence
is low and the potential for mixing is minimal. The density of water varies with temperature.
When streams of water at different temperatures meet, the warmer fluid, which is lessdense, tends to seek the upper portion of the pipe, while the cooler fluid remains at thebottom. The two layers can remain in thermal equilibrium without mixing under a wide
range of flow conditions. An indicator of whether stratification will occur is the Richardson
number (Ri) 4:
Ri = (g×Δρ×ID)/(ρ×υ2)
Whereg= Standard acceleration due to gravity
ρ= Density of fluidΔρ= Difference in density of hot and cold fluidυ= Flow velocity
ID= Pipe Inside diameter
The Richardson number does not address the effects of the piping configuration such as
pipe length and slope or the mixing effects of elbows and tees. Nonetheless, the Richardson
number is used as a screening criterion to assess the potential for thermal stratification. As
a rule of thumb, when Ri > 4.0 a susceptibility exists for thermal stratification.
The effect of thermal stratification on the state of stress in the pipe is manifested in two
ways. The difference in temperature between the top and bottom of the pipe causes greater
thermal expansion at the top, resulting in bowing of the pipe. If the bowing is restrained,
global bending moment stresses result. In addition, the interface between the two fluidlayers causes a local stress in the pipe due to the thermal discontinuity across the pipe crosssection.
Thermal stratification has caused leaks in normally stagnant lines connected to the reactor
coolant loop in safety injection systems, reactor coolant drain lines, a residual heat removal
suction line, and an excess let-down line. In addition, stratification has caused cracking ofsteam generator feed-water nozzles and high piping displacements in pressuriser surgelines.

2.3. Thermal cycling
Thermal stratification by itself is generally not a concern for fatigue failure. However, if amechanism for cycling exists, fatigue cracking can occur. There are several sources ofthermal cycling:4.

Locations where there can be leakage past valves separating hot and cold fluids and regions
where there might be intermittent mixing of hot and cold fluids caused by fluid injection are
considered to be susceptible to thermal fatigue. Exceptions are pipe segments where the diameteris 25 mm or less or where the slope of the segment is 45° or more from the horizontal. According
to the EPRI “Fatigue Management Handbook” (TR-104534), when these criteria are exceeded,
additional evaluations can be performed to determine if the maximum temperature difference isgreater than 10°C or Ri > 4.0.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 19•

Changes in interface level. If, for example, stratification is caused by a leaking
valve, if the rate of leakage varies, the hot-to-cold fluid interface level will moveup and down. A point in the pipe cross-section near the interface can alternately be
exposed to tensile and compressive stresses, as well as through-wall thermal
gradient stresses, as the point cycles between cold and hot.

•

Changes in temperature. If either the hot fluid or the cold fluid undergoes
temperature fluctuations, the global bending moments will cycle and the stress
distribution in the pipe cross-section will vary.

•

Turbulence penetration. Turbulence penetration is secondary turbulence in branchlines caused by flow turbulence from high velocity fluid in the run pipe. As thereactor coolant loop flow passes branch lines of connecting systems, some of thehotter reactor coolant periodically enters the branch lines. The hotter fluid caninteract with cold leakage fluid and cycle at a high frequency, depending on the
distance from the run pipe. The length that the turbulence penetrates into the branchvaries with flow conditions, temperature and the randomness of the fluid dynamics.
The phenomena are not yet fully understood, but the maximum extent of turbulence
penetration cycling length in the branch is approximately 25 branch pipes inside
diameters.

Branch piping subject to turbulence penetration can be considered to consist of three
regions. In the section closest to the run pipe, all of the pipe cross-section is near the hottemperature of the run pipe. In the second section, the run pipe temperature permeates into
the branch pipe, but to a lesser extent, such that it stratifies at the top of the pipe. Combinedwith this are flow turbulence effects, which cause local thermal cycling at points in the pipe
cross-section. In the third section of turbulence penetration, further away from the run pipe,
only the effects of the run pipe temperature permeating into the pipe due to naturalconvection effects are present. There is no flow turbulence across the cross-section.
However, cycling can still occur in this region, as the length of turbulence penetration may
fluctuate, causing temperature variations over time in both the axial and radial directions.
Piping that runs downward will behave differently than that which runs horizontally or
upward from the RCS piping.

2.4. Thermal striping
Under certain thermal-hydraulic conditions, the interface between the hot and cold layers
can become turbulent, leading to high frequency cycling over a narrow range in the verticalprofile of the pipe. This phenomenon generally occurs when the relative velocities of the
hot and cold fluids are high. Testing has demonstrated that the temperature amplitude vs.
frequency of cycling can be represented by a spectrum. Fortunately, the higher frequency
cycling occurs with the lower amplitude temperature changes, such that in most cases thecontribution of striping to the overall fatigue usage is not large. In addition, striping stresses
occur only at the pipe surface, so that even though striping can initiate a crack, it cannot
cause the crack to grow through-wall. This phenomenon will generally not occur innominally stagnant lines.

2.5. Valve in-/out-leakage
Two conditions need to be present for valve in-leakage to occur: 1) the isolation valve must
leak, and 2) there needs to be a pressure gradient from the upstream to the downstream sideof the isolation valve. As an example, the high-pressure safety injection system in20 │ NEA/CSNI/R(2019)13
Westinghouse plants is driven by charging pump discharge pressure, up to about 17.6 MPa,
which is higher than reactor coolant system pressure, about 15.5 MPa. The case of concern
is when the leakage is into an unisolable, normally stagnant portion of the piping system,and the leakage interacts with the hot water from the reactor coolant system. The colder
leakage fluid can interact with turbulence penetration from the RCS and cause cycling at ahigh frequency, depending on the distance from the leaking valve to the RCS pipe.
The US NRC Bulletin 88-08 “Thermal Stresses in Piping Connected to Reactor Coolant
Systems” 5 was issued in response to a leak that occurred at the Farley plant in 1987. The
Farley leak was located in the heat-affected zone of the weld between the first elbow andthe horizontal run upstream of the RCS cold leg nozzle. The line rises vertically from thecold leg, and then turns horizontally to the check valve. The crack location was about 1 mfrom the RCS cold leg nozzle. The orientation of the crack was circumferential, in thebottom third of the pipe. Farley is a 3-loop Westinghouse plant. The cause of the crack was
the interaction of hot RCS fluid with the cold valve leakage fluid that had stratified at thebottom of the pipe. Temperature monitoring performed subsequent to cracking indicatedthat downstream of the check valve, the difference in temperature between the top andbottom of the pipe was 102°C. The bottom temperature was not constant; the amplitude of
the cycling was 21°C with a cycle period between two and 20 minutes. Other leaks of asimilar nature have occurred in the safety injection system due to valve in-leakage at
Tihange unit one (Belgium) in 1988 and Dampierre unit 2 in 1992.
With respect to valve out-leakage phenomena, the leak path is from the reactor coolant
loop, through an unisolable branch line, to a lower pressure system. The valve leak could
be either past the isolation valve seat or past the stem packing and out the leak-off line, if
one is present. Out-leakage stratification is generally steady with a limited number ofcycles. Bulletin 88-08 Supplement 3 described an unisolable piping system leak that wascaused by valve out-leakage.
In 1988, a leak occurred in the RHR suction line at Genkai unit one (a 2-loop WE/MHI
pressurised water reactor [PWR] design). The crack location was in the weld at the firstelbow downstream of the reactor coolant loop hot leg connection, between the vertical drop
and the horizontal run. The crack orientation was circumferential and originated in the heataffected zone of the weld. The crack was located at the top of the pipe. A second crack wasfound at the top of the weld between the horizontal pipe and the isolation valve. An
evaluation of the fatigue striations indicated that the stress levels had been about 10 to 15
ksi and the number of cycles to failure was 104 to 105.
The RHR pipe was eight inches in diameter, Type 316 stainless steel. The distance fromthe hot leg to the leak location was nine feet (L/D = 13.5) and the distance to the isolation
valve along the horizontal run was 11 feet (L/D = 17). Examination of the fracture surface
and other root cause analyses indicated that the cause of failure was TF. The crack initiatedat a location with a stress concentration factor of about three, and had been subjected tothermal stratification cycling. The isolation valve did not show any evidence of leakage
past the seat. However, there was evidence of leakage at the connection of the valve leakoff line to the lantern ring. The purpose of the lantern ring is to provide a means to conduct
leakage out from the stem packing.
The NSSS vendor, Mitsubishi Heavy Industries (MHI), conducted tests that demonstrated
that even a small amount of leakage through the stem packing gland could cause thermal
5.

www.nrc.gov/reading-rm/doc-collections/gen-comm/bulletins/1988/bl88008s3.html.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 21
stratification in the horizontal pipe. However, such stratification was steady, without
cycling. Further testing discovered that the disc of the isolation valve normally rests against
the downstream side of the seat, leaving a small gap on the upstream side. When thepacking gland leaks, hot water from the RCS passes by the disc and the thermal expansionof the disc closes the gap, which stops the leak. After a while, the disc cools, the gap reforms
and leakage begins again. This intermittent leakage mechanism was concluded to be thesource of the thermal stratification cycling. The tests demonstrated that the upper pipe
temperature fluctuated by 40°C, and the period of the cycling was 20 minutes. There had
been a temperature sensor installed on the leak-off line to detect leakage but because theline was not insulated, it did not detect the hotter flow.
As a corrective action, the valve disc position was raised so that there would always be a
gap between the disc and the upstream side of the seat. Thus, if there is packing leakage, itwill be constant with no thermal cycling. Although there was no evidence of it, the thermal
stratification cycling could possibly have also been caused by turbulence penetration fromthe RCS hot leg. The vertical distance from the hot leg to the crack of 13.5 inch pipe
diameters is in the turbulence penetration range, and variations in the turbulence
penetration length could have intermittently stratified the fluid in the horizontal section [1].

2.6. Thermal sleeve mixing
PWR plants of the Babcock & Wilcox (B&W) design have combined make-up/highpressure injection lines. Most plants use one line for normal make-up, with one plant using
two. The make-up does not pass through a regenerative heat exchanger, so the flow
temperature is much colder than the RCS temperature. A thermal sleeve is provided in thenozzle to protect the immediate area from TF. However, if there is any problem with thesleeve, turbulent mixing of fluids with a large temperature difference can result.
In 1982, a crack in the make-up nozzle occurred at the Crystal River plant [11]. The nozzle
is of carbon steel 2.5 inches in diameter and is shop welded to a safe-end made of type 316
stainless steel. The safe-end is then field welded to the 2.5-inch, type 316 stainless steel
make-up line. A thermal sleeve of type 316 stainless steel was installed in the nozzle by amechanical press fit.
The crack was in the safe-end to pipe weld. The orientation was circumferential, 140°
around on the outside surface. There were actually two separate cracks: one initiated on theinside due to TF caused by turbulent mixing of the hot reactor coolant and the cold makeup water; the other originated on the outside and was believed to have been caused bymechanical vibration. Other cracks were found in the safe-end, check valve and the thermal
sleeve.
The thermal sleeve was found to have been loosened, and there was a gap between thesleeve and the safe-end. The gap allowed cold make-up flow to mix with the RCS fluid on
the nozzle surface, and the resulting flow-induced vibrations caused additional wear of thethermal sleeve. However, there was no thermal stratification. Five of the other B&W plants
also found cracking at this location although it was not through-wall. The US NRC issued
the Generic Letter 85-20, Resolution of Generic Issue 69: High-pressure injection/makeup nozzle cracking in Babcock and Wilcox Plants and the B&W Owners Group developeda corrective action programme to address the issue. An improved thermal sleeve design
was subsequently developed. It included a hard roll and longer contact area to assure a tight
fit of the nozzle and prevent flow-induced vibrations. An augmented, periodic ISI
programme to detect gaps behind the sleeve was also recommended.

22 │ NEA/CSNI/R(2019)13
Another through-wall crack occurred at Oconee in 1997 in the pipe to safe-end weld [11].
Similar to Crystal River, the cause of the cracking was turbulent mixing of the cold makeup flow with the hot RCS flow. Again, a gap behind the sleeve allowed the RCS fluid topenetrate back past the safe-end. Variations in the make-up flow rate and high frequency
turbulence effects resulted in thermal cycling at the crack location. It was postulated thatthe sleeve had been loosened by differential thermal expansion cycling due to back flow
with partial pump operation and evidence suggested that further wear occurred due to flowinduced vibrations. For corrective actions, the new thermal sleeve design was installed, andwarming line flow was increased to minimise flow isolations. Prior to the weld leak, thecombination stop-check valves were replaced by separate stop and check valves, thus
removing a known condition for back-leakage.

2.7. Thermal transient (TT) fatigue
The number of TT fatigue cycles that will cause fatigue damage depends on the magnitudeof the alternating stress (or strain) incurred during each transient loading cycle. The
endurance limit represents the cyclic stress amplitude that can be applied indefinitly
without causing fatigue damage. From a practical standpoint, the endurance limit istypically assigned to the alternating stress amplitude associated with 106 or 107 cycles on
the material stress-cycle curve. A specific operational thermal and pressure loading
condition identified in the TT degradation mechanism assessment is not considered fatigue
significant if the allowable number of cycles associted with the maximum alternatingtransient stress is greater 106. The CODAP event database includes two recent (2018)
examples of recordable circumferential weld flaws attributed to TT fatigue. The affectedareas were in the charging pump discharge lines.
Areas considered susceptible to TT fatigue include pipe segments where there is relatively
rapid cold (hot) water injection with delta temperature greater than 65°C for carbon steel
pipes and 90°C for austenitic stainless steel pipes. When these temperature changes are
exceeded, additional evaluations can be performed to determine if delta temperature isgreater than delta temperature allowable.

2.8. Significant historical events
Five TF events of historical significance are summarised below. The information is
extracted from the CODAP event database. The term “significance” is a reflection of
combinations of operational impact, and the extent to which industry and regulatory
agencies responded in terms of finding acceptable solutions to the long-term fatigue
management.
•

Olkiluoto unit one (OL-1) major pressure boundary failure. On 29 August 1979 a
DN150 pipe in the reactor water clean-up (RWCU) system fractured, resulting in aspill of 5 000 kg of primary water outside containment. The through-wall crack in
the pipe was 150 mm long and 2 mm wide. The unit was in the commissioning
phase of operation, and the pipe fracture resulted in reactor trip and containment
isolation. The rupture occurred in a mixing tee between the RWCU and residual
heat removal (RHR) systems. The apparent cause of the failure was thermal
stratification, but the underlying cause was an operator error due a deficiency in thesystem operating procedure. The fatigue crack evolved because of plant operation,with an intermediate basic state of RWCU valve 331-V19 leaving two flows withconsiderably different temperatures to mix incompletely. OL-1 is a BWR designed

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 23
by Asea-atom (AA). Additional details on the AA NSSS design are included in
Section 4.2.5.
•

Through-wall TF cracking of make-up nozzle safe-end weld at Crystal River 3. On
21 January 1982, an unidentified 3.8-L/min leak was found while the plant was inoperation. Visual inspection revealed that the leak was associated with the MU/HPI
line. As the leak was unisolable, the plant promptly proceeded to cold shutdown.
Inspection of the safe-end revealed that a through-wall, circumferential crack waspresent in the safe-end-to-check valve weld. The circumferential extent of the crack
at the outside surface was 140 degrees. The crack consisted of two separate cracks:
one initiated at the outside diameter and another initiated at the inside diameter.
The one on the outside surface was initiated and propagated by mechanical fatigue
caused by pipe vibrations. The one on the inside surface was initiated andpropagated by TF caused mainly by turbulent mixing of hot reactor coolant andcold make-up water. Thermal shocks during periodic make-up water additions
could have played a role in causing the fatigue damage.

•Through-wall cracking in the safety injection line at Farley-2 and Tihange-1. In
1987, a leak occurred inside the containment of Farley-2 during normal power
operation. The leak was found in the unisolable location of the safety injection line.
The crack was on the inside surface of the weld and extended approximately
120 degrees circumferentially around the underside of the pipe. About 25 mm ofthis crack was through-wall. The crack was caused by TF and had developed
slowly. The leak rate was 2.7 L/min. The monitoring of circumferential temperature
distribution at the failed weld at Farley-2 carried out after the leak event showedspatial and temporal fluctuations in the temperature. The circumferential
temperature difference at the weld varied from 3°C to as high as 120°C. Based onthese measurements, it was assumed that the temporal variations resulted fromintermittent action of the check valve. There were, however, no test results
supporting this assumption. Experiments performed in Japan simulating the Farley
event showed that the temperature fluctuation in the safety injection line was notcaused by the intermittent action of the check valve (i.e. fluctuation in the flow rate)
but by the mixing of low temperature leakage flow with high temperature turbulent
flow in the pipe downstream of the check valves. The Japanese test results also
concluded that the thermal cycling is severe enough to cause HCF failure of thepiping material when the leak flow rate is equal to or larger than 100 kg/hr, as
measured at Farley-2; see pp 229-240 in reference [1]. A similar occurrence caused
cracking in both base metal and welds of Tihange-1 on 18 June 1988. Both Farley2 and Tihange-1 are similarly designed Westinghouse-3 loop plants.

•

Three Mile Island 1 RCS cold leg drain line failure. On 9 September 1995, a leakoccurred in a cold leg drain line. The through-wall crack was located in the weld
between the first elbow downstream of the reactor coolant loop nozzle and thehorizontal pipe run. At the time of the leak, the plant had been operating for 21
years and was at 0% power, beginning a cool-down. The leak rate was 20 “drops”
per second. The pipe routing is down 14 inches vertically from the cold leg, then
7.3 feet horizontally to the first valve. The vertical run is 1.5 inches diameter pipe,
the horizontal run is 2 inches diameter, and the elbow between them is a reducing
elbow. The elbow and horizontal run are type 316 stainless steel, and the verticalrun included an inconel safe-end. The location of the crack was in the weld between
the elbow and the horizontal pipe, near the top of the pipe. The distance from thecrack to the cold leg inside diameter was 14 inches, or about 10 pipe diameters. The

24 │ NEA/CSNI/R(2019)13
crack was circumferential, 2 inches (50.8 mm) long on the inside and 0.55 inches(14 mm) long on the outside surface, centred at the 11 o’clock position. The drainline was not insulated. The cause of the cracking was TF, attributed to turbulence
penetration of the hot RCS fluid extending into the horizontal pipe. The horizontal
pipe, being uninsulated, allowed heat to escape to the surroundings, and when the
turbulence penetration reached the horizontal run, this produced thermal
stratification in the line. Fluctuations in the extent of turbulence penetration caused
local thermal cycling at the elbow weld, a point of stress concentration. There wereother contributing causes as well. Two improperly installed pipe support U-bolts
restricted the free thermal expansion of the pipe and produced a 37 ksi stress at the
elbow. In addition, the placement of the pipe supports caused the horizontal pipe toslope upward away from the elbow, which facilitated thermal stratification andcycling in this pipe run. In addition, the toe of the cracked weld had a pre-existing
notch.
•

Large leak due to a longitudinal crack in a mixing zone of the RHRS of Civaux unit
1. On 12 May 1998, with the unit in intermediate shutdown, the pressuriser level
decreased and a fire alarm was triggered in the reactor building. A large primaryleak (leak rate of about 8.3 kg/s) had developed, and it was compensated by the
charging system. After rapid cool-down of the reactor, the leak was identified tocome from RHR train “A”; additional details on this event are provided in Section
4.2.4.

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3. Safety significance of TF failures
A pipe failure caused by TF can have a direct or indirect impact on plant operation.
Examples of direct impact include flow diversion and loss of the affected train or systemor an initiating event such as a loss-of-coolant-accident (LOCA). A direct impact may resultin automatic safety system actuation and reactor trip or operator action may be needed toinitiate a controlled plant shutdown. An example of an indirect impact is degraded pipe
condition that is subjected to an operability determination and without need for immediate
operator intervention.

3.1. Consequences of TF failures
Figure 3.1 provides a high-level summary of the total failure population of 253 events,organised by operational impact and consequence of a below ground pipe failure. The
operating experience data is grouped into nine consequence categories 6. While it is possible
for some events to result in more than one consequence, the dominant consequence isassigned in the database.

6.

•

Flooding/spraying of equipment area.

•

Unplanned outage work. Applies to situations involving scheduled below ground
piping system inspections performed during a maintenance or refuelling outage andthe discovery of a degraded pipe condition that requires prompt remediation.

•

Manual/controlled shutdown from full power. This typically means that a manualreactor shutdown is performed when a limiting condition for operation (LCO)
cannot be met; see below.

•

Cool-down from hot shutdown or cool-down from hot standby.

•

ESF/RPS actuation. In response to postulated accidents, the engineered safety
features (ESF) actuation signals actuate or realign safety-related systems,
equipment, and/or components and isolate non-safety-related systems. Various
parameters are monitored by plant instrumentation to determine whether an
accident or other condition requiring protective action has occurred. In the reactor
protection system (RPS), multiple parameters are monitored and compared. Whenenough signals in an appropriate combination have exceeded their respective set
points, the RPS system is designed to safely shut down the reactor and start cooldown actions. In many boiling water reactor (BWR) designs, RPS also includes
containment isolation and radioactive materials release functions. In PWR designs,
these functions are usually situated in ESF systems.

In CODAP, the event classification is done in multiple ways. It is done by “event type”(e.g. small leak, large leak, rupture), “isolable/non-isolable pressure boundary breach” (yes/no),“collateral damage” (e.g. flooding/spraying of equipment area), and “direct impact on plant
operation” (e.g. automatic turbine trip/reactor trip, manual shutdown, power reduction).

26 │ NEA/CSNI/R(2019)13•

Expanded outage work. Applies to situations involving scheduled inspections
performed during a maintenance or refuelling outage and the discovery of adegraded pipe condition that prompts further non-destructive examinations but
without impacting the overall outage schedule.

•

No operational impact. A pipe failure that can be corrected without impacting anyother routine plant operation activities.
Figure 3.1. Operational impact of TF failures•

LCO entry 7. Discovery of a degraded condition (e.g. a through-wall leak) results in
a LCO action statement. When an LCO cannot be met, the reactor must be shut
down or the licensee must follow any remedial action permitted by the licensing
basis (e.g. technical specifications) until the condition can be met. In CODAP, thisconsequence category is applied to pipe failure events for which corrective action
is completed within a time frame as specified in the technical specifications.

3.2. Considerations of TF in probabilistic safety assessment
The history of LOCA frequency model development is as old as the commercial nuclear
industry. There have been numerous efforts to develop plant and location-specific LOCA
frequencies that reflect the unique piping configurations and susceptibilities toenvironmental degradation. The analytical approaches have ranged from top-down
apportionment of fixed LOCA frequencies to specific locations within a reactor coolant
pressure boundary (RCPB), to a bottom-up approach where location-specific LOCA
frequencies are aggregated to a plant-wide LOCA frequency. An example of high-levelguidelines for LOCA frequency assessment are contained in the ASME/ANS “Standard forLevel one/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear
7.

For a definition, refer to the “Glossary of technical terms” in Annex B.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 27
Power Plant Applications”. 8 As another example, the US Nuclear Regulatory Commission
has issued a draft regulatory guide 1.229 [32], Section C (Staff regulatory guidance) of
which addresses “acceptable technical approaches for how to estimate location-specific
LOCA frequencies.”
The “bottom-up” approach to LOCA frequency assessment consists of multiple analysis
steps, beginning with a detailed consideration of the constituent elements of the evaluation
boundary. As one example, for a PWR plant, the evaluation boundary includes thefollowing:•

Reactor coolant system (RCS) hot leg; i.e. the large-diameter piping between thereactor pressure vessel and steam generator.

•

RCS cold leg; i.e. the large-diameter piping from the steam generator to the reactor
pressure vessel.

•

Pressuriser surge line; piping between the pressuriser and RCS hot leg.

•Multiple branch connections, including RCS cold leg and hot leg drain lines, highpressure injection, and chemical and volume control (RCS make-up), pressuriser
spray, accumulator injection, decay heat removal.

For each element of the RCPB a formal degradation mechanism analysis is performed todetermine the potential and observed degradation mechanisms. Next follows an evaluationof the relevant operating experience, including the level of defence-in-depth provided
through the implementation of component degradation mitigation strategies. Finally, thespecific calculation cases to consider (i.e. combinations of material, pipe size anddegradation susceptibility) are defined, including the associated input parameters.
Overviews of LOCA frequency models and the different calculation strategies are
summarised in references [12][13][14][15].
An example of the contribution of TF to the calculated, plant-specific LOCA frequency isgiven in Figure 3.2. It is a composite of results obtained from several relatively recent (2011to 2015) plant-specific LOCA frequency calculations. According to this example,
approximately 20% of the total LOCA frequency is attributed to TF.

8.

https://adams.nrc.gov/, Accession No. ML15335A179.

28 │ NEA/CSNI/R(2019)13
Figure 3.2. An example of the contribution of TF to PWR LOCA frequency

Note: In this example the “SB-LOCA” category considers all pressure boundary break sizes between 12.7 mm
equivalent break size (EBS) and 50.8 mm EBS (12.5 mm < EBS ≤ 50.8 mm). The “MB-LOCA” category
considers the continuum EBS range of (50.8 mm < EBS ≤ 152.4 mm), and the “LB-LOCA” category considers
the continuum EBS range of (152.4 mm < EBS ≤ 1114 mm). LCF = low-cycle fatigue; HCF = high-cycle
fatigue (vibratory fatigue); PWSCC = primary water stress corrosion cracking in dissimilar metal weld
locations; TGSCC = transgranular stress corrosion cracking.

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4. Operating experience data review
Summarised in this section is the TF operating experience as recorded in the CODAP event
database. For each country or economy, a discussion is included of selected significant
events as well as the particular approaches to TF management. The term “significant” refers
to operational impact as well as the impact of an event on regulation and/or industry
initiatives to address below ground piping ageing issues.

4.1. High-level database summary
The CODAP event database content with respect to TF failures is summarised in
Figures 4.1 through 4.4. Of the total data subset, crack-part failures represent around 50%of the US operating experience and 18% of the Swedish experience with TF (this isaddressed in further detail in Section 4.4.5). The operating experience is delineated asfollows:•

Figure 4.1. TF operating experience by period and plant type (i.e. BWR, PHWR
and PWR).

•

Figure 4.2. In this chart, the operating experience is organised by country and modeof failure. It should be noted that as a founding database project member, Sweden
remained an active participant in data exchange through the end of the calendar
year 2014.

•

Figure 4.3. In this chart, the database content is organised by the plant systemaffected by TF as well by the event dates as being pre-1999 or post-1998.

•

Figure 4.4. In this chart, the operating experience is organised by nominal pipe sizeand structural integrity impact.

The majority of the TF events involve non-through-wall cracks discovered duringscheduled ISI using NDE technologies and procedures. To date, there have been two events
resulting in a significant through-wall peak flow rate > 8 kg/s (or circa. 127 gpm, US),corresponding to about 8.3% of the total event population. Various high-level, qualitative
insights may be derived from the charts. Trends in the operating experience data are
attributed to:•material ageing;•changes in reporting practices and requirements;•changes in inspection practices and requirements;•

permanent piping system design modifications.

More in-depth evaluation of the events is needed to distinguish the significance of each ofthese four causal factors in the trends exhibited in Figures 4.1 through 4.4.

30 │ NEA/CSNI/R(2019)13

Figure 4.1. Number of TF failures by time period and plant type

Note: Information on thermal fatigue events is still being collected for the period 2010-2020.

Figure 4.2. TF failures by country

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Figure 4.3. TF operating experience by plant type, system and time period

Figure 4.4. TF failures by structural integrity impact

Note: The two events with “significant through-wall flow rate” are Civaux-1 (May 1998, ca. 8.3 kg/s peak flowrate) and Olkiluoto-1 (August 1979, ca. 20 kg/s peak flow rate).

32 │ NEA/CSNI/R(2019)13

4.2. Specific TF operating experiences
This section includes selected summaries of the post-1998 operating experience with TF in
commercial nuclear power plants. In addition, this section presents the different approaches
among participants to TF management and mitigation.

4.2.1.

CanadaA recent study by the CANDU owners group (COG) [16] concluded that TF via thermal
stratification, cycling and stripping is not a major concern for CANDU nuclear power
plants due to the lack of sources and locations promoting direct contact and mixing of hot
and cold fluids. The study was performed using an adaption and modification of the EPRI
risk-informed in-service inspection (RI-ISI) methodology [17], known as “COG best-fit
RI-ISI” methodology [18]. The CANDU reactors have some specific design features thatdifferentiate them from generic PWR designs; however, supplementary research andsurveys on CANDU operating experience with TF is valuable to compare to and learn fromoperating experience available for the PWR fleet. These features and/or differences indesign imply that certain CANDU systems and components are not as susceptible to TF as
those found in a typical PWR.
The CANDU primary heat transport system (PHTS) operates at pressures and temperatureshigher than those of the interfacing systems do and there are very few possible sources for
hot-to-cold fluid mixing where isolation valves exit. In addition, branch configuration of
the drain and suction lines are mostly contained in the hot feeder cabinet and do notexperience thermal cycling. A significant amount of evidence indicates that certain steam
generator (SG) feed-water nozzles and adjacent piping of PWRs exhibit TF susceptibility.
Since the steam generators of newer CANDU six stations have an internal preheater(instead of a feed-water distribution ring like in most PWR steam generators), the potential
for high-frequency thermal stripping, as seen in a typical PWR, has been eliminated [16].
Nevertheless, due to the inherent temperature differences between incoming feed-water and
SG internal recirculating flow, the potential for TF still exists at the feed-water entrance
nozzles. This could be a problem in the CANDU SGs, as there has been a case where a
CANDU station had a steam generator with cracks found in the weld between the feedwater nozzle and the feed-water thermal sleeve [16].
In PWRs, surge lines are typically instrumented on for temperature and deflection
measurements to monitor for the existence of thermal stratification. The in-surge/out-surge
flow also has an impact on the pressuriser and may cause additional thermal stresses in thepressuriser bowl. Even though there is ISI evidence of TF in CANDUs in Canada, someindirect measurements taken from the commissioning of one CANDU plant showedindications of thermal stratification in the horizontal run of the surge line [16]. To address
this, the station adjusted its warm-up and start-up procedures to mitigate the impact ofthermal stratification in the surge line. There have since been no recorded disturbances
related to thermal stratification for this station. In addition, the potential for TF caused by
the in-surge/out-surge flow is greatly reduced with the use of a thermal sleeve in thepressuriser. However, the weld connecting the surge line to the pressuriser is left exposedand could be susceptible to thermal stresses. Consequently, the pressuriser bowl of
CANDUs has been identified as a potential fatigue location because of differential thermal
expansion of the thermal sleeve and pressuriser bowl during in-surge events [16].
Over the years, there have been few cases potentially related to the appearance of TF at
CANDU nuclear power plants. All of these cases have been related to class one piping
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failures occurring either at or downstream of the thermal mixing tees connecting the D2O
feed pump outlet lines to the main PHTS piping [19]. The most probable cause of these
events was cyclic thermal loading that was not considered in the original CANDU design.
Table 4.1 includes a summary of TF failures in safety class one heat transport systems.
Table 4.1. Examples of safety class one piping failures at CANDU plants
Year
1991
1993
1999
2001
2003

Description
Thermal mixing tee connecting to purification return line to the HTS
Thermal mixing tee connecting the feed line to the HTS
Thermal mixing tee connecting the feed line to the HTS
Mixing tee weld failure at the tee connecting the purification return line to the HTS
The weldolet discovered in 2001 was found cracked again

In Canada, requirements for fatigue management are specified under the CSA standards.
Fatigue evaluations required for all class one components can be found under CSA
Standard N285.0, which contains general requirements for pressure-retaining systems andcomponents. This consists of selection of code classification, design fabrication, inspection
and installation of pressure-retaining systems and components, regulatory requirements
and applies ASME Section 3 or “non-nuclear” standards. CSA Standard N285.4 specifies
the requirements for periodic inspection programmes for the nuclear pressure boundary
components of CANDU nuclear power plants. It requires assessments of components forsusceptibility to fatigue and environmentally assisted cracking arising from normal
operation at select candidate locations for inspection. Additionally, CSA standard N285.7,
first issued in 2015, establishes inspection programme requirements for safety significant
non-nuclear (balance-of-plant) pressure boundary components and requires an assessmentof TF susceptibility.
To date, there has been a single case in which TF was thoroughly investigated at an
operating CANDU station. This was done with the use of EPRI/MRP-132 TF screening
techniques. Results of the inspection showed no significant signs of TF and it is reasonable
to expect that other operating CANDU stations would have similar results [20]. However,
it is worth noting that the screening criteria depend on design and operational parameters,
so detailed screening and evaluations of each operating plant are necessary to come toconclusive findings. A review of the CANDU plant using the EPRI/MRP-146 andEPRI/MRP-132 identified different locations that have potential for thermal [21]. The
identification of components at risk presents an opportunity to proactively assess theseverity of fatigue and take corrective actions to mitigate the effects of TF on these systems
and components, if needed. Automatic monitoring software can be used to efficiently
monitor key operating parameters within systems susceptible to thermal fatigue [22].
In recent years, recommendations and potential improvements to TF management havebeen introduced for CANDU stations. The following monitoring techniques have been
introduced to help address and mitigate the effects of thermal fatigue [21]:•cycle-based fatigue (CBF) monitoring;•

stress-based fatigue (SBF) monitoring;•

flaw tolerance approach.

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CBF monitoring computes fatigue directly from transients and determined parameters. In
reference to TF, CBF algorithms have been developed for three limiting HTS components
and SG lugs [21]. SBF monitoring uses directly measured or computed real-time stressdata. The flaw tolerance approach demonstrates that a postulated (or an actual crack) will
not grow through-wall in a specified operating period [21].
In summary, there is limited information on operating experience with TF associated with
the CANDU design. However, specific screening techniques have identified components
and/or systems of CANDU stations that are susceptible to TF. Fatigue-related failures at
CANDU stations have been observed on class one piping at or downstream of the thermal
mixing tees connecting the D2O feed pump outlet lines to the main PHTS piping. With thesystems and components identified through screening techniques and patterns in previous
events, the integration of various monitoring techniques and software presents anopportunity to gain more insight into TF associated with the CANDU design and to mitigate
its effects.

4.2.2. Czech Republic
The Czech Regulatory Body (SÚJB) in 2017 issued a new regulatory requirement specific
to TF management: C28 [Rozhodnutí č. j. SÚJB/OSKŘaE/12142/2017]. According to thisregulatory requirement, evaluations are to be performed to identify the locations where TF
can occur. This requirement applies to all safety class one and two systems and high-energy
safety class three piping as well as to certain non-safety-related system where a structural
failure can affect safety-related components.
In 2000, a thermal fatigue-monitoring system was installed at the Dukovany nuclear power
plant. At each reactor unit, 103 locations are monitored using 242 thermocouples. A similarsystem was subsequently installed at the two reactor units at the Temelin nuclear power
plant; see Figure 4.5.
Figure 4.5. Schematic of thermal fatigue-monitoring locations at the Temelin nuclear power plant

Source: Reproduced from Žaloudek, J. et al, “Innovation of Temelin Diagnostic and Monitoring System,” Proc.
Int. Topical Meeting VVER – 2004 Experience and Perspectives, Prague, Czech Republic, 19-22 October 2004.

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Fatigue life calculations of the main circulation piping, adjacent high-pressure, lowpressure and passive emergency piping systems of safety class one, SG and pressuriser are
performed on all units in the Czech Republic. Fatigue life calculations are performed using
real operating modes.
Medium temperature variations entering into the calculations are adjusted according tomeasured temperature values obtained from both standard measurements of basic primary
medium parameters and diagnostic measurements of temperatures performed on the outersurfaces of pipes and primary circuit and secondary circuit components. Detailed analyses
are performed using the finite element method.
Temperature measurement was used to assess fatigue life time in cases where a lack of
information about temperature distribution in piping and components is supposed, inparticular in all cases when changes of medium temperature are not correctly determinedby the actions of pumps or valves, or in cases where the medium temperature is notmeasured directly. Piping and component temperature measurements are performed toobtain information on:•

general temperature distribution on whole component;•

thermal expansion of piping systems;•

quick temperature transients;•

stratification – special temperature distribution;•

combinations;•internal tightness of valves;•monitoring of operational modes;•

precise warming up of pressure vessels.

4.2.3. France
On 12 May 1998, with the Civaux unit in intermediate shutdown, the pressuriser level
decreased and a fire alarm was triggered in the Reactor Building [23]. A large (ca. 8.3 kg/s)
primary leak had developed, and it was compensated by the charging system. After rapid
cool-down of the reactor, the leak was identified to have come from RHR train “A”. About
ten hours after the event occurred, the RHR train “A” was isolated and the leak was stopped.
The source of the leakage was a cracked weld on an elbow downstream of a control valve.
The weld was longitudinal, while the elbow was fabricated from two elbow halves andwelded together. The crack length was approximately 180 mm. The event was classified aslevel two on the International Nuclear and Radiological Event Scale (INES). The observed
cracking was immediately downstream of the mixing tee in a DN250 Schedule 40 (9.3 mm
wall thickness) 304 L stainless steel seam-welded elbow. The worst cracking that led to thethrough-wall leak was at the extrados welded seam, and had an outside surface length of
180 mm with an inside surface length of 350 mm. The surrounding area of the elbow andlocations between the elbow and the mixing tee exhibited widespread crazing-type cracking
characteristic of TF. Examination of this elbow and similar ones from other plants
confirmed that high-cycle TF was the mechanism. No fabrication defects were found. In
general, the deep cracks (one through-wall at Civaux, and others up to 80% of the wall
thickness) occurred along the longitudinal and circumferential welds, starting at the root of36 │ NEA/CSNI/R(2019)13
the weld on the inner surface. There were many small cracks, mainly in the weld counterbore regions and where there was evidence of grinding.
Électricité de France (EDF) initiated a major assessment, analysis and research programme
to better understand the phenomenon that caused the incident. In 2013, at the end of the
programme, which ran for more than ten years, the Institut de Radioprotection et de Sûreté
Nucléaire (IRSN) examined EDF’s conclusions and delivered its own analysis. The IRSN
highlighted the fact that EDF’s analysis and research had provided a more detailed
understanding of the phenomenon, but it also said that an accurate record should be kept of
the operating times of circuits vulnerable to TF where there were important differences intemperature, to enable appropriate checks to be introduced.

4.2.3.1 Action taken as a result of the 1998 incident at Civaux
In the 1998 incident at Civeaux Nuclear Power Plant, reactor one was undergoing a normal
shutdown when a 30 m3/hour leak appeared in the Reactor Heat Removal System (RHRS).
This system is used to remove the residual power from the reactor core. This major leak
was caused by a 180 mm long crack through the pipe at an elbow immediately below amixing tee linking a pipe (or line) carrying hot water bypassing the heat exchanger with apipe carrying cold water coming out of the heat exchanger (Figure 4.6). ln the mixing area,
the hot water was at a maximum temperature of 180°C and the cold water at a temperatureof around 20°C.
Figure 4.6. Area where the leak occurred

Inspections were carried out of the other 1 450 MW reactors (Civaux 2, Chooz B1 and B2)
and then of all other reactors. The inspections were then extended to all mixing areas in the
RHRS circuits of French reactors. The ultrasonic testing revealed damage to all the areas
checked, prompting EDF to replace the relevant areas of the RHRS circuits in all reactors
within three years. Examinations of the RHRS piping sections that had been removed
revealed evidence of TF that had not been taken into account at the design stage. Thisphenomenon causes multiple shallow cracks, crazing (cracking of the surface) or isolated
cracks starting at the base of a weld bead (Figure 4.7).

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Figure 4.7. Cross-sectional view and internal view of weld bead of the cracked elbow

From 2000, on IRSN’s recommendation, EDF introduced a policy of regularly checking
the replaced RHRS sections every 450 hours of operation at wide temperature differences.
The 450-hour frequency was based on the first cracks observed in the RHRS circuits
removed from reactor 2 at the Civaux power plant. The plant operator’s initial analyses
could not fully explain the location and extent of the damage observed. Furthermore, thetraditional method of analysing (mechanical) fatigue could neither predict nor explain thedamage observed. According to the reports on numerous inspections carried out since, TF
in the mixing areas of RHRS circuits is a phenomenon that depends on many factors: theoperating time at wide temperature differences, thermal and hydraulic factors (the speed
and temperature of the fluid), the condition of the surface and the mechanical stresses inthe different components. Because there are so many factors, understanding this type of
fatigue is particularly complex.

4.2.3.2 Vulnerability of mixing areas to TF
When France’s nuclear reactors were designed, TF was not taken into account when
designing mixing areas. The Civaux incident prompted EDF to introduce a method forevaluating the risk of TF in mixing areas in pipes.
EDF’s decision was based on an indicator commonly used for fatigue, the usage factor.
This is defined as the ratio between the number of stresses applied to a particular component
and the maximum stress values indicated by the mechanical fatigue curve for the material
that the component is made from. Mixing areas with a usage factor of more than one wereconsidered “vulnerable to thermal fatigue”. This method was applied to all circuits
containing mixing areas, particularly certain pipes connected to the RCS, in order tointroduce appropriate monitoring for the vulnerable areas identified.

4.2.3.3 Research and development on TF
At the end of 1999, EDF and AREVA began a major research and development
programme. The aim of the programme was to improve the knowledge and analysis of TF,
particularly in mixing areas, and to identify more clearly the conditions under which thedegradation mode observed would arise, through an exhaustive survey of the “vulnerable”
areas.
In parallel, to bolster its own expertise in this area, the IRSN also carried out some research.
In partnership with the Commissariat à l'énergie atomique et aux énergies alternatives(CEA), the IRSN conducted a study aimed at understanding the specific nature of the38 │ NEA/CSNI/R(2019)13
stresses involved in TF, compared with the stresses commonly taken into consideration formechanical fatigue. The study was conducted using an experimental device known as
FAT3D (Figure 4.8), in which a pipe was subjected to thermal stresses: a network of cracks
rapidly appeared and spread, with some spreading right through the material. In particular,
it was established that the number of cycles it took to make the cracks appear in the test
specimens was always less than the number predicted by the calculation made using theusual analysis methods for mechanical fatigue. The presence of the welded joint may resultin the first cracks appearing more quickly.
Figure 4.8. FAT3D TF testing device

4.2.3.4 Tests on models representative of a mixing area
EDF had some results of tests that had been performed since 1976 on models that were ageometric representation of the mixing areas and reproduced the flow in these areas. Thesemodel-based tests were primarily designed to provide the loads (average temperature
ranges, heat transfer coefficients) for studies of mechanical behaviour related to operatingtransients. Following the TF incident at Civaux, EDF completed its database by carryingout tests that were not restricted to the RHRS circuit but also covered other mixing areas in
the RCS feed line (Figure 4.9). For the different mixing areas, these later tests allowed:•

The temperature ranges and the coefficients for heat transfer between the fluid andthe internal wall of the pipes to be evaluated.

•

Information to be gained about the sites of temperature fluctuations due to themixing of a hot fluid with a cold fluid, based on the geometry and the ratios of theflow rates of the two fluids.

The model-based tests and the associated computational models gave a realistic assessmentof the thermal loading affecting the pipes in the mixing areas. Taking account of the resultsof the research and development and the model-based tests, the inspection method for
mixing areas “vulnerable to TF” used from the 2000s involved the scheduling of periodic
checks of the identified areas, usually at the time of the ten-yearly inspections.

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Figure 4.9. Diagram of the feed line connected to the RCS

4.2.3.5 Change to the method for assessing mixing areas
Since 2008, EDF has felt that the method and means of monitoring it has introduced address
the problem of TF in mixing areas. EDF has adopted a more representative modelling ofthermal transients based on the results of the model-based tests, and has replaced theconventional values used for heat transfer coefficients with values based on reality, taking
account of the thermal and hydraulic conditions that actually exist in mixing areas.
However, the results of the research have not made it possible to upgrade thethermomechanical method used for estimating mechanical stresses.
Lastly, the research has led to the adoption of a lower threshold than before for thevulnerability of stainless steels to TF: mixing areas where the temperature difference of thefluids is 50°C or more (rather than 80°C or more) are now considered to be “vulnerable”.
This change has prompted EDF to review its list of mixing areas that might be “vulnerable
to TF”.

4.2.3.6 Representativeness of the usage factor of the mixing areas
Between 1999 and 2002, EDF appraised the sections of the RHRS circuits removed after
the Civaux incident. The usage factors of these sections were also calculated on the date ofremoval; the usage factors were greater than one. This confirms their vulnerability to TF.
However, no correlation could be established between the usage factor values and thedimensions of the cracks. ln 2001, EDF decided to investigate the possibility that there
were defects in a mixing area considered “vulnerable to TF”: the branch pipe of the RCS
feed line (Figure 4.9).
For appraisal purposes, EDF took the opportunity provided by the replacement of the steam
generators in reactor one of the Fessenheim power plant after 24 years of operation, toremove a section of piping from the RCS, including the connection joint of the feed line.
The inspection of this connection joint, which had a usage factor calculated at more thanone, did not reveal the beginnings of any cracking due to fatigue. Furthermore, ultrasonic
testing of feed line connection joints carried out since 2004 on around 20 reactors has notdetected any evidence of TF.
In 2013, the IRSN examined the evidence from around a decade of inspections of mixing
areas, including the sections from the RHRS circuit and the branch connections of the feedline. The IRSN felt that the usage factor was an indicator of the risk of cracking, but that it40 │ NEA/CSNI/R(2019)13
was not suitable for assessing the damage that could occur because of TF. It was therefore
important that, in addition to the ten-yearly inspections, the frequency of inspection of all
mixing areas with a usage factor greater than one should be determined based on theoperating time at wide temperature differences. This was implemented by EDF in the caseof the mixing tee in the RHRS circuit.

4.2.3.7 Counting operating time at wide temperature differencesln 2008, EDF introduced a system for metering on a daily basis the amount of time spent
operating at wide temperature differences, for all mixing areas “vulnerable to TF”, and set
thresholds for this. However, it was not until 2013 that EDF specified what action should
be taken when these thresholds were reached. If the operating time of a mixing area at wide
temperature differences is greater than expected, EDF will either bring forward the checkof the area normally scheduled to take place during the ten-yearly inspection or will justify
the fact that its condition will not have changed using calculations. In the opinion of the
IRSN, however, if a threshold for the time spent operating at wide temperature differences
is exceeded, EDF should inspect the mixing area concerned without delay.

4.2.3.8 Conclusion
Following the 1998 incident at Civaux, the IRSN analysed the substantial amount of work
done by EDF to establish the risks to its reactors from TF. A better understanding of local
thermal and hydraulic phenomena, developed because of numerous model-based tests andsome computational modelling, has brought improvements to the method of identifying
mixing areas “vulnerable to TF.” However, the IRSN still has reservations about basing anassessment of the risk of damage to sections of the mixing areas solely on a calculation of
the usage factor. In the IRSN’s view, priority should be given to checking these sections
on the basis of the time spent operating at wide temperature differences.

4.2.4. Finland
Finland has four operating commercial nuclear power plants at two sites; two BWR units(OL1 and OL2) at the Olkiluoto site and two PWR units at the Loviisa site. The fifth power
plant unit (OL3) is under construction. The OL1 and OL2 BWR units are of type AA-IV
BWR-2500, which were designed by AA of Sweden. OL3 is an EPR (European PWR)
plant unit. The two PWR units (LO1 and LO2) at the Loviisa site are of type VVER-440,
which is based on a Soviet concept (Atomenergoeksport). However, the Loviisa units aredesigned and constructed to meet the Finnish safety requirements and conditions. The basic
AA Nuclear Steam Supply System (NSSS) design of the OL1 and OL2 BWR units differs
in many respects from the NSSS design by General Electric. OL1 and OL2 are essentially
identical to Forsmark-1 and Forsmark-2 in Sweden. The AA-specific thermal fatigueinduced damage experience is summarised in Figure 4.10.

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Figure 4.10. TF failures in AA reactors

Source: AA-I (Oskarshamn-1), AA-II (Ringhals-1), AA-III (Barsebäck-1/2 and Oskarshamn-2), AA-IV(Forsmark-1/2 and OL-1/2), AA-V (Forsmark-3 and Oskarshamn-3); AA-I through AA-III units have external
recirculation pumps and AA-IV and AA-V units have internal recirculation pumps.

An example of recent TF experience in OL1 and OL2 involves the discovery of several 2
to 5 mm deep cracks on the inner surface of the main feed-water piping. The cracks weredetected during respective refuelling outages in 2014. These cracks were located in the basemetal of the FW/RHR mixing tees. The feed-water system of the AA-IV design consists of
two main header pipes, each with a RHR branch connection. The RHR system is basicallyalways in operation. When the reactor is in hot standby, there is no preheating of the feedwater. To keep the RPV water level within the normal allowable range, smaller amounts ofcold water are added through one of the FW lines. Thus, the ~20°C feed-water and 280°C
RHR water mix and create cyclic temperature fluctuations in the large-diameter FW pipe
upstream the tee (Figure 4.11).
Figure 4.11. Location of the crack indications in FW/RHR mixing tees

Cracks in the mixing tees in OL1 and OL2 were first discovered in the early 1980s and in
response to these discoveries the mixing tees in both units were replaced in 1986. To
minimise thermal fluctuations and in response to this operating experience, the original42 │ NEA/CSNI/R(2019)13
design of the mixing tee was improved by installing sparges (or static mixers) to improvethe mixing of the fluids and to create a more uniform temperature distribution. 9
Although this 1986 design of the mixing tee did not last until End of Life (EOL), it lasted
about 25 years. The mixing tee is actually loaded by most severe thermal transients in thewhole power plant. Therefore, the tee was considered relatively reliable, especially as thecrack growth was slow, allowing for time to design the replacement of the tees. For thisreason, the design of the mixing tee was not changed, except for one minor change. The
inspection hatch was removed to simplify the geometry. The hatch was not necessary
anymore as it had not been used for a long time and endoscope inspection from further
away proved enough to assess the internal condition of the sparge. Furthermore, thecondition-monitoring of the run pipe (main feed-water pipe) is in the ISI programme andinspected every third year with ultrasonic.
There are two mixing tees per unit. They were placed under closer monitoring after thecrack indications. The preliminary safety analysis concluded that no repair was needed forthe ongoing operating cycle. The three most cracked mixing tees were replaced in the 2015
outage and the last in the 2016 outage; see IRS Report 8439 & CODAP Record No. 4758.
The original design of the VVER-440 units in Loviisa did not consider damage mechanisms
like thermal mixing in T-junctions and thermal stratification in horizontal piping because
their existence was not realised at that time. The importance of these mechanisms is wellunderstood nowadays and corrective actions have been made accordingly. An example ofleakage caused by thermal stratification occurred in the pressuriser auxiliary spray line
control valve on 16 May 1994. Leakage was localised in a valve body, where a throughwall axial crack was found (Figure 4.12). The control valve is supposed to be closed whenplant is on operation; it is used when primary pressure needs to be decreased during normal
unit shutdowns. However, investigations revealed that the control valve had not been
completely leak-tight, leading to a case where hot steam 325°C could have entered fromthe pressuriser into the horizontal part of the auxiliary spray line while colder water 250°C
could have gone through the control valve from the cold leg side. This could have led to asituation where hot steam stratified in the top part of the pipe with cold water in the bottom(stratification was verified to have occurred only in the LO1 unit). A material study also
revealed that the crack initiated from minor internal defects, i.e. inclusions containing
titanium that favoured the crack nucleation. It was also noticed that during shutdown, andespecially during start-ups, this valve body confronts temperature differences of up to
100°C. In early years of operation, this temperature difference was estimated to be as highas 220-270°C on at least one or two occasions [1].

9.

International Reporting System for Operating Experience (IRS), IRS-Report 8439: Thermal
Fatigue Cracks in Feed Water Piping Tee, 2015.

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Figure 4.12. The shape of through-wall axial crack in control valve

A valve of the same design was used to replace the leaking control valves because it waseasily available. After the valves were replaced, both units were restarted on 21 May 1994.
However, in 1996, the valves were replaced with a different type of valve design and thepipelines were modified [1].

4.2.5. Germany
As required by the KTA standards, the management of TF in German nuclear power plants
starts already within the framework of the design, selection of materials, manufacturing
processes and production and installation. These measures are supplemented by monitoring
provisions during the commercial operation phase. As a result, no safety significant events
have been observed in German nuclear power plants so far. The following review includes,
within the context of TF, a brief compilation of the most essential requirements of the KTA
standards, a description of the installed equipment for continuous fatigue-monitoring, anda short summary of the German operation experience of the last 20 years.
According to KTA 3201.2 (“Components of the RCPB of Light-Water Reactors – Part 2:
Design and Analysis”) [25], during the design phase of a nuclear power plant, a fatigue
analysis shall be made independently of the type of component to avoid fatigue failure due
to cyclic loading later in operation. The bases for fatigue evaluation are the design fatigue
curves based on tests carried out in air. KTA 3201.2 presents different procedures forfatigue analysis. In cases where the environment may have a deleterious effect on thefatigue life of the component, a so-called attention value of 0.4 in the cumulated usage
factor is defined, above which certain additional measures are necessary.
KTA 3201.4 (“Components of the RCPB of Light-Water Reactors – Part 4: In-service
Inspections and Operational Monitoring”) [26] includes a requirement to monitor TF loads
by a sufficiently dense net of measurement points of the standard instrumentation 10. The
location of the measurements points shall consider effects of the mode of operation(e.g. little mass flows, indifferent pressure conditions, switching operations, temperature
10.

The standard instrumentation is installed to monitor the parameters relevant for the integrity of
the components within the scope of KTA 3201.4 and comprises measuring equipment to monitor
global loadings and – if required – measuring equipment to monitor local loadings.

44 │ NEA/CSNI/R(2019)13
differentials) and the design (e.g. pipeline installation, isolating function of valves).
Stratification effects shall also be monitored by adequate positioning of the measurement
points such that all relevant loading variables across the pipe cross-section and axially tothe pipe run can be measured. KTA 3201.4 also contains regulations on how to handle andassess the results.
The thermal fatigue-monitoring during operation demanded by the above-mentioned codesand standards is performed using the measuring points of the standard instrumentation as
well as of the special instrumentation. Typical parameters monitored by the standard
instrumentation include coolant pressure and temperature in different areas of the systems
as well as valve positions. The additional special instrumentation allows for monitoring ofthermal transients deviating from the systems specifications and/or unspecifiable transientsas stratification effects, which otherwise would not be measurable, but which may causesignificant fatigue load.
Thermal fatigue monitoring in German nuclear power plants is accomplished by the
installation of a fatigue-monitoring system using the standard and special instrumentation.
Fatigue-monitoring systems of different suppliers are used with the FAtigue-MOnitoring
System (FAMOS) by Framatome GmbH (formerly Siemens/KWU, later AREVA) being
the most commonly used one. Differences that exist between the various systems are mostly
related to the way to bring the thermocouples into contact with the surface, Figure 4.13.
Figure 4.13. FAMOS Thermocouple installation along the circumference of a pipe

Source: [27].

The selection of the measuring points requires a plant-specific analysis of the fatiguerelevant locations. Figure 4.14 gives an example of the distribution of the respective
measurement points of FAMOS in a typical German 4-loop PWR. Typically, around
50 locations are monitored in the primary and secondary circuit. Mostly, locations can befound at the main coolant lines, surge line, pressuriser spray lines and feed-water nozzles.

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Figure 4.14. Typical positions of FAMOS measurement points for a four loop PWR

Source: [27].

The goal of the fatigue-monitoring systems such as FAMOS is to gather real information
about fatigue-relevant temperature changes at locations most prone to TF. The derived data
allows for:•determination of thecomponents/locations;

cumulatedusage

factor

of

the

most

affected•

identification of disadvantageous operation procedures related to TF;•

establishment of a database for fatigue analyses based on realistic fatigue load
parameters.

Temperature changes are usually measured using thermocouples, which are placed on theouter surface of the fatigue-relevant locations. Figure 4.13 shows the thermocouple
installation used in the FAMOS system. The seven thermocouples along the half
circumference of the pipe allow the detection of thermal stratification. At locations wherestratification is not assumed to occur, two thermocouples are installed. The thermocouples
are installed at a certain distance from welds to not disturb the ultrasonic inspection of thewelds. Based on the measured temperature on the outer surface, the temperature on the
inner surface is calculated and, where necessary, transferred to the fatigue-relevant
location.
In German BWRs, thermal fatigue monitoring is installed in the feed-water piping (RPV
nozzles) and at various locations in the RHR system, especially to monitor thermal
stratification. In addition, TF is also monitored at the spray lines of the RPV heads in asimilar way as the spray lines of the pressuriser in PWRs. Safety significant events due to
TF have not occurred in German BWR for the last 20 years.
Table 4.2 lists the German operating experience concerning TF for the last 20 years. The
identified events are grouped depending on component and phenomenon, respectively.
Afterwards, selected events/groups are described in more detail.

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Table 4.2. German reportable events
Event #
1998/099
1998/119
1999/079
2005/036
2005/061
2001/003
2003/018
2009/019
2009/083
2010/026
2011/032
2010/070

Plant
Neckarwestheim-1
Mülheim-Kärlich
Biblis-A
Grafenrheinfeld
Biblis-B
Biblis-A
Biblis-B
Biblis-B
Unterweser
Grafenrheinfeld
GrohndeGrafenrheinfeld

Brief description
TF on injection nozzles of the CVCs into the cold loop due to internal valve leaks
Cracks due to TF loads in the exhaust elbow on an emergency diesel generator
Internal leak in heat exchangers of the reactor coolant clean-up system
Leak at an SG instrumentation line
Crack indications at hard-facing of pressuriser relief and safety valves
Crack indications in the seal housing of the main coolant pump
Indications in the safe-end of the nozzle of the main coolant line connecting to the surgeline

The 1998-99 event in Neckarwestheim-1 is comparable to a number of previous events inGermany and worldwide with internally leaking isolation valves, which cause thermal
striping and/or stratification. It corresponds to the Farley-2 and Tihange-1 event
descriptions presented in Section 2.8. In the case of Neckarwestheim-1, crack indications
were found at tee-pieces of the volume control system (TA). It was known from monitoring
during operation that increased temperature loads may occur at the connection between thecold and the hot TA leg at loop 10 and 30. Therefore, these areas were replaced by
optimised constructions at all loops. The back-up examinations revealed several crack
indications at 2 of 3 dismantled tee-pieces of the volume control system (loops 10 and 30)
at the inner surface of the cold TA feed line before it leads into the hot TA feed line. The
cause of the cracks was attributed to TF. The thermal cyclic loading was caused by creeping
flow due to untighten check valves. The tee-pieces are located in the area, which cannot be
isolated from the reactor pressure vessel.
In Biblis units A and B as well as in Grafenrheinfeld, some internal leakages occurred atheat exchangers in the reactor coolant clean-up system. The leakages were detected either
by activity monitoring of the intermediate cooling water system or by hydrazine input into
the CVCS, depending on operating conditions. The leakages were caused by TF cracks inheat exchanger tubes, which were able to develop due to design deficiencies of the tube
sheet and manufacturing deficiencies in the tube-to-tube sheet connection in the heatexchanger. As a measure, the heat exchangers were replaced by new ones with improved
design.
Due to the findings at a main coolant pump in a Swiss nuclear power plant, the main coolant
pump was opened during the 2009 revision in the Biblis-B Nuclear Power Plant (and later
in Unterweser, Grafenrheinfeld and Grohnde). The seal housing was subjected to a dye
penetrant surface examination. Further investigations revealed transgranular, lowdeformation incipient cracks in the axial direction in the seat of the bearing flange, in thegroove flank of the sealing area, in the groove base of the sealing area, in the cylindricalseal housing underneath the undercut, and in the blind hole threads. The patterns of theaxial cracks and the circumferential cracks were similar and exhibited characteristics offatigue fracture. It was assumed that the fatigue was caused by cyclic thermal stress. The
primary circuit medium with temperatures of about 260ºC and 280°C continuously enters
the pump impeller side chamber. Downstream of the high-pressure cooler, the seal watersupplied by the volume control system has a temperature of about 50ºC. Via oblique holes
in the seal housing, the seal water enters the gap between bearing flange and seal housing
and is discharged again through another hole. This means that the outer surfaces of the seal
housing and the bearing flange are exposed to the temperature of the primary circuit
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medium and the inner surfaces simultaneously to the temperature of the seal water. As acountermeasure, newly manufactured, modified seal housings were installed. An
information notice was filed giving recommendations regarding inspections for indicationsin the affected area as part of the next scheduled pump revision. In case of findings, theseshould be assessed and, where required, the seal housings should be replaced with those of
an improved design.
Within the framework of an outage of the Grafenrheinfeld nuclear power plant from March
to June 2010, an automated ultrasonic in-service inspection was performed on the nozzleof the main coolant line to the surge line, which revealed a circumferential indication in therounded connecting area of the thermal sleeve to the safe-end of the nozzle [27]. The
indication was fully circumferential and was classified as a relevant finding due to its lengthand despite its relatively low echo amplitude. Depending on the method, the indication
depth measured was between 2 and 2.7 mm. Since possible crack growth could not be
excluded, the safe-end was removed, cut and subjected to visual and metallographic
examinations. The visual inspection of the removed part revealed that the actual geometryof the connecting area of the thermal sleeve deviated clearly from the specifications of thedrawing. Instead of having a radius of 2.5 mm, an almost rectangular groove cross-section
was found, superimposed by further geometric discontinuities. The metallographic
examination revealed a minute circumferential crack which had apparently formed throughfatigue because of changing thermal loads. The location of the crack corresponded with theresult of the ultrasonic inspection. The actual maximum depth of this minute crack was,however, clearly below the crack depth determined by means of the ultrasonic inspection.
The maximum depth was approximately 0.3 mm in the 0°position. It was assumed that theunexpected geometry contributed to a higher fatigue load and to the overestimation of thecrack determined by UT. The plant operator replaced the affected safe-end with a new one
during the 2011 plant outage. To prevent recurrence of crack formation, the round area ofthe replaced part was machined with a lathe in accordance with the specifications of thedrawing. The actual safety significance event was limited due to the small depth of theincipient crack. The potential significance of the event is given by the fact that a deviationfrom the specified geometric form of the round area may lead to an active degradation
mechanism with crack growth during operation and such manufacturing defects might not
be identifiable in a reliable manner with the available non-destructive test methods. An
information notice was filed giving recommendations regarding the non-destructive test
sequence and test frequency depending on the review of the documentation of the latest
tests performed.
The monitoring systems used in German plants have proved widely effective. They arehighly developed and optimised continuously. Based on the results of the monitoring
systems, operating procedures have been optimised and technical modifications have been
performed to reduce the fatigue load of the components. Both led to an enhancement insafety and availability of the plants. The continuous monitoring also allowed the very early
detection of damages. The operating experience concerning TF is positive. Over the last 20
years, around 12 events have occurred in German PWRs, but none in BWRs. Regarding
the (generic) safety significance of a TF event, GRS filed information notices withcorresponding recommendations on how to handle such an issue and avoid re-occurrence.
The operating experience, however, also shows that TF damages cannot be fully excluded
despite the installation of a FAMOS, especially in cases where the real component
geometry deviates from the drawing that is used to calculate the fatigue load of certain
locations.

48 │ NEA/CSNI/R(2019)13

4.2.6. Japan
Some Japanese BWR and PWR plants have been affected by TF damage caused bytemperature fluctuations through the mixing of hot and cold streams at tee junctions. Below
are three examples of Japanese operating experience with TF damage:•

Mihama-2 (2-loop PWR), April 1999. The reactor unit was operating at reduced
power when it was determined that the drain flow to the containment sumpincreased. A containment entry revealed a small leakage from the excess let-down
piping off reactor coolant loop B (RCS cold leg). The leak had developed on the
DN80 excess let-down line of the chemical and volume control system (CVCS). It
was subsequently found that there was a through-wall crack of about 7 mm in length
on the outside and of about 24 mm in length on the inner surface near the elbow ofthe excess let-down system piping. The cracked pipe surface showed the featuresof fatigue damage. It became clear that, at the relevant elbow, in addition to theresidual stress generated in the bending process, the repeated thermal stress was
generated due to the change of the temperature boundary surface (thermal
stratification) between the reactor coolant from the RCS cold leg and the stagnant
water in the excess let-down system piping. The corrective action for this event was
to shorten the vertical run to about six pipe diameters, so that the turbulence
penetration boundary was well into the horizontal run, away from the elbow (Figure4.15). This avoided thermal cycling at the elbow, and assured that the horizontal
run would have a continuous source of heat, eliminating stratification cycling. An
elbow with reduced residual stress was used in the replacement.
Figure 4.15. Mihama-2 primary coolant leak

Source: Reproduced from EPRI 10070061 (MRP-25, 2003).

•

Tsuruga-2 (4-loop PWR), July 1999. A primary coolant leak was detected at 6.05
am when fire alarms in the passage outside SG compartments C and D wereactivated. At the same time, another alarm showed a high level in the containment
sump followed by an increase in charging flow and an increase in the reading of
the containment vessel particulate radiation monitor. A controlled shutdown wasstarted at 6.24 am. When cold shutdown conditions had been reached, radiationlevels inside containment began to fall. Engineers entered containment at 18.45 andfound a leak at the outlet of one of the regenerative heat exchangers of the CVCS.
The leak was stopped by closing an isolation valve and upon removal of insulationfrom the pipework a through-wall crack was found on an elbow joint (Figure 4.16).

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Some 51 m3 of primary water had leaked from the crack and the water had
overflowed the containment sump and flooded the floor to a level of about 5cm.
Inspection of the outer surface of the elbow found a longitudinal (running in the
direction of the flow) crack about 44 mm long, on the outer surface of the pipe.
After the elbow had been removed, a 99 mm crack was found in the corresponding
position on the inner surface. The elbow was sent to a laboratory. Visualobservations, and dye penetrant and ultrasonic tests showed 3 longitudinal cracksof 151 mm, 100 mm and 72 mm on the inner surface and a 26 mm crack crossing
the weld on the upstream side. The 151 mm crack had penetrated the wall for alength of 47 mm. In addition, there were five circumferential cracks in the welds.
The fracture surfaces were typical of fatigue cracks with striations (so-called beachmarks) indicating that the cracks had grown from the inside surface of the elbow.
There was no evidence of stress corrosion cracking or other corrosion problems.
Figure 4.16. Location of primary coolant leakage•

Genkai-2 (2-Loop PWR), January 2007. With the unit shut down for its twentiethrefuelling and maintenance outage, ultrasonic examination of the excess let-down
piping revealed a significant flaw in the base metal. The affected pipe section wasremoved for metallurgical examination and flaw characterisation. The nonthrough-wall flaw was about 90 mm in length and about 8.1 mm in depth.

In response to this operating experience, the Japan Society of Mechanical Engineers(JSME) published the “Guideline for Evaluation of High-Cycle Thermal Fatigue of a Pipe
(JSME S017),” in 2003, which provides procedures and methods for evaluating the
integrity of structures with the potential for high-cycle thermal fatigue (HCTF). In JSME
S017, one of the important procedures of TF evaluation is to classify the flow patterns attee junctions, because the degree of TF damage is closely related to the flow pattern
downstream of the mixing junction. The conventional characteristic equations forclassifying flow patterns are only applicable to 90-degree tee junctions (T-junctions).
However, angled tee junctions other than 90 degrees (Y-junctions) are also used inchemical plants and refineries to reduce the pressure drop in the mixing zone and to weaken
the force of the impingement of the branch pipe stream against the main pipe.

4.2.7. Korea
During a planned outage at Wolsong unit 4, a through-wall flaw was found on a welded
connection of the pressuriser steam bleed valve. On 17 June 2014, at around 10:30am,
while the unit was on guaranteed shut down for a planned maintenance outage, a finethrough-wall crack was found at the pressuriser side of the welded connection of thepressuriser steam bleed valve (63332-PCV5); see Figure 4.17. The field investigation

50 │ NEA/CSNI/R(2019)13
revealed that the pipe was locked in because a pipe hanger (PH-29) had not been installed
correctly as designed and the operation of the pressuriser steam bleed valve (63332-PCV5)
generated repeated thermal loading to the restrained pipe, which caused corrosion-fatigue
cracking (CFC) on the weld zone and ultimately through-wall leakage.
Figure 4.17. Pressuriser steam relief valve schematic diagram

The fracture surface and cross-section analysis showed that the flaw morphology was a
typical corrosion-fatigue crack (Figure 4.18). The key damage mechanism was determined
to be circumferential CFC, which grew from the inside to outside surface of the welded
connection along the boundary between the base and welding metals due to extended
exposure to a corrosive environment coupled with the repeated thermal loading fromoperation of the valve as well as the tensile stress caused by the interference with some ofthe pipe hanger.
Figure 4.18. Fracture surface analysis results

The evaluation of the leak rate from the fine through-wall crack showed that the leaked
coolant (about 0.9 kg) was properly recovered as the plant proactively sealed the potential
leak points when the increasing trend of the reactor building tritium concentration wasnoticed. Since there were no meaningful changes in terms of radioactivity level including
the tritium concentration in reactor building and the amount of effluent release from thevent, it was concluded that the leakage from the through-wall crack on the welded

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connection of the pressuriser steam bleed valve did not cause any safety concerns such asradiological leakage to the environment.
The licensee performed a detailed fracture surface and cross-section analysis to identify thecause(s) of the flaw. It also established a repair work plan based on such analysis to conduct
a repair welding onto the flawed location of the steam bleed valve welded connection andcorrected the positions of the pipe hangers that were believed to be the contributors to the
problem. The licensee performed a post-maintenance NDE testing to determine theintegrity of the weld zone. After evaluating the dose received by the workers who
participated in the repair welding work and checking the leak location, the licensee
confirmed that the internal and external exposure dose as well as the annual cumulative
dose of the workers was within the annual dose limit. The investigation team reviewed thelicensee evaluation on this matter and found that the licensee conclusion was acceptable.
The licensee, based on its cause evaluation, took the following short-term actions: 1)
maintenance on the flawed location of the welded connection, 2) inspection andmaintenance on the supports of the reactor coolant pressure and volume control system,and 3) expansion of NDE testing to same type valves and dissimilar weld zones. For longerterm measures, the licensee came up with the following actions: 1) monitoring the
behaviour of steam bleed pipe during plant start-up and shutdown, 2) increasing thefrequency of ISI for similar locations, 3) replacing steam bleed valves with new valves, 4)
periodic monitoring of pipe hangers (PH-28 and PH-29), and 5) establishing a NDE planfor equivalent locations at Wolsong units two and three, as well as inspection of the pipe
hangers. The investigation team deemed these short- and longer-term measures to beadequate to prevent recurrence of the event.

4.2.8. The Netherlands
The Netherlands has only one nuclear power plant in operation: a two-loop PWR of
German design, located in Borssele. The possible ageing mechanisms of the safety-related
components were taken in account in the design phase, with the available knowledge at thetime. During the lifetime of the installation, no incidents related to fatigue occurred. There
are, however, locations in the power plant where high fatigue loading has been observed,
the most relevant being the surge line, affected by TF due to stratification (see Section
4.2.8.1).
In the framework of the LTO assessment project, to extend the lifetime of the installationfrom 40 to 60 years, a fatigue time-limited ageing analysis (TLAA) project was set up. The
study resulted in a list of recommendations:•

Perform further assessments for six locations before the end of 2013 (40th year ofoperation) to prove that adequate safety margins against crack initiation by fatigue
are in place also during LTO. The locations are:
‒
‒
‒
‒
‒
‒

secondary outlet nozzle (steam);
control volume system/primary system nozzle injection;
nozzle of surge line at main coolant line;
nozzle of surge line at pressuriser;
spray nozzles of the auxiliary spray line;
spray nozzles of the spray lines.

52 │ NEA/CSNI/R(2019)13•

Further assess environmental fatigue at the following three locations before the endof 2013:‒‒‒•

control volume system/primary system nozzles;
surge line;
nozzles of the double T-junction.

Update the load catalogue and load specifications after five operating cycles(Borssele has one-year cycles) of monitoring and revalidate the fatigue analyses
that are required by this update.

The operating licence of the Borssele Nuclear Power Plant was changed to grant operation
after 2013. A clause was introduced in the licence regarding the monitoring of fatigue: A
fatigue-monitoring programme must be in place and the operator must report yearly to the
nuclear regulatory authority about the results, news or upcoming trends emerging from theaforementioned programme. According to the licence, the scope of the fatigue-monitoring
programme must comprise all relevant safety-related components and structures.
The monitoring of the fatigue transients is carried out via the FAMOS by Framatome
GmbH, which was installed in 2010. FAMOS consists of rings of thermocouples on theoutside of the pipe measuring the temperature distribution over the circumference of thepipe. The rings are installed on several locations at which temperature differences can besignificant; see Figure 4.19. This monitoring of the FAMOS thermocouples is combinedwith the already existing temperature measurements.
Figure 4.19. The Borssele Nuclear Power Plant FAMOS locations

The fatigue calculations for the identified locations are re-evaluated every year, based on
the monitoring data. New calculations are performed when needed and the cumulative
usage factors (CUFs) are re-evaluated. Environmental fatigue is considered according to
the thresholds prescribed in KTA 3201.2 [25].
The use of an online monitoring system like FAMOS allows for an assessment of thechange of operating procedures on the fatigue loading, to influence the thermal fatigueinitiating transients in the most sensitive locations.
The fatigue monitoring programme is supervised from the nuclear regulatory authority via
the evaluation of the yearly report and taking part in inspections related to the programme.
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An example is the ultrasound inspection of the thermal sleeve at the SG nozzle (Figure4.20). The location where the thermal sleeve connects to the nozzle has a high usage factor.
Even if the thermal sleeve has no pressure-retaining function, a NDE was carried out torule out the presence of possible crack initiation. This examination was carried out during
the outage of 2018.
Figure 4.20. SG feed-water nozzle thermal sleeve design

4.2.8.1 Operating experience with the surge line of Borssele Nuclear Power
Plant
The surge line of the Borssele Nuclear Power Plant is welded onto loop one of the main
coolant line, connecting the main coolant system with the pressuriser. It consists of nine
sections that are welded together. The surge line is made of low-alloy ferrite steel with anaustenitic stainless steel cladding. The design of the main coolant lines and surge line was
generally performed according to the German requirements. Dutch requirements were also
used to supplement these requirements in the design of related pressure-retaining
components.
In the 1980s, it became clear from international experiences that the phenomenon of fatigue
due to stratification was a possible ageing and failure mechanism in the surge line. Thismechanism was not taken in account in the original design of the Borssele Nuclear Power
Plant. In 1991, thermocouples were temporarily installed on the surge line to be able todetect the temperature changes in the line and the occurrence of stratification.
With the available measurements, Siemens/KWU performed a fatigue calculation. It wasassessed that the 40 years of lifetime were still covered from the existing design

54 │ NEA/CSNI/R(2019)13
assumptions. The start-up and shutdown procedures were, however, changed to reduce thetime the line is exposed to high fatigue loading due to stratification.
Thermal stratification may occur in horizontal and slightly sloping parts of the surge line
during start-up and shut down. Cooler, denser water from the primary circuit flows under
the warmer, lighter water from the pressuriser. The operating temperature is 345 C for thepressuriser and 319.4 C for the hot leg of the main coolant line. Therefore, the temperature
difference could be approximately 25 C in the mixing zone during normal power operationin “base load” conditions. During start-up and shutdown, the temperature difference in thisarea will be significantly higher (approximately 180 C).
As previously specified, in the framework of the LTO project, all the relevant locations forfatigue, including the surge line, were recalculated to estimate the number of cycles, theamplitude of these cycles, and the CUF until 60 years of life. Environmental fatigue wastaken in account. The surge line entered the scope of the fatigue-monitoring programme of
the Borssele Nuclear Power Plant.
Every year the occurred cycles are compared with the projected ones and the nature of thecycles is compared with the specified one. This makes it possible to assess if theassumptions made in the fatigue calculations cover the transients that actually take place.
As of today, no unexpected transient has been registered with FAMOS on the surge line(the last available results are from the 2017 scheduled refuelling outage).

4.2.9. Slovak Republic
Four units are in operation in the Slovak Republic. Two are located at the Bohunice site
(units three and four) and two at the Mochovce site (units one and two). Another two units
are under construction at the Mochovce site. All six units are of the VVER-440/213 type.
Initial evaluation of TF was performed in 2000. In Bohunice three and four, analyses werebased on results from FAMOS, which was installed in 1996 on the primary components.
In Mochovce one and two, FAMOS was part of the delivery of the primary circuitcomponents (the system was installed before the start of commercial operation in 1998
[Mochovce 1] and 2000 [Mochovce 2]).
Computational analyses revealed high thermal stratification in the pressuriser surge line
and SG feed-water nozzle. The maximum number of thermal cycles was observed during
the start-up and shutdown of the unit. Mitigation of thermal cycles (stress loads) in SG
feed-water nozzles was achieved by modifying the start-up period. Thermal stratificationin the pressuriser surge line was mitigated by using medium mixing in the pressuriser withmodification of the reactor coolant pump start-up.
In Bohunice, the original FAMOS was replaced with a new system in 2015-2016. The
amount of monitored locations was expanded on the pressuriser surge line (Figure 4.21)
and SG feed-water nozzles. Several new thermocouples were installed on primary
components that had not been monitored before, such as the RHR and emergency corecooling system (ECCS) piping. There are now 118 (367 thermocouples) monitored
locations per unit, compared with the 62 locations (106 thermocouples) of the previous
FAMOS. TF evaluation methodology was updated by the use of: 1) data from the new
FAMOS, 2) new evaluation models with a higher number of computational elements, and3) computational analyses that are now performed for real operating loads.

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Figure 4.21. Thermal fatigue-monitoring of the pressuriser surge line

Results from the new evaluation model demonstrate lower stress loads of the evaluatedcomponents compared with the previous model. The most stressed components are SG
feed-water nozzles. Neither indications of crack development nor any other damage ofprimary components due to TF (1998 to present) have occurred. Based on the evaluationand current CUF numbers, it can be concluded that the limit acceptance criteria value will
not be reached by the end of 60 years of plant long-term operation.

4.2.10. Spain
Spanish nuclear power plants perform TF management through the approved ISI
programmes and the AMP.

4.2.10.1 ISIThe Spanish Nuclear Safety Council (CSN) Instruction IS-23 (November 2009) regulates
the requirements, development, implementation and management of ISI programmes. As
stated in IS-23, Spanish nuclear power plants must carry out their ISI programmes inaccordance with the standards defined in the regulations of the country of origin of thetechnology and accepted in the operating permits, the basic standards applied being section
XI of the code of the American Society of Mechanical Engineers (ASME) and the ASME
operation and maintenance code (ASME-OM), required by the technical specifications.
Consequently, this code is considered an acceptable reference for the drawing up of the ISI
and testing programmes defined for these facilities, which are included in the document
known as manual of ISI (MISI).
TF susceptibility was not originally considered in the design of the Spanish power plants,
all of which stem from the 1970s. A need to manage the TF due to stratification was
subsequently recognised during the 1980s and in response to operating experience as
documented for example in the US NRC Bulletin 88 “Thermal Stresses in Piping
Connected to Reactor Coolant Systems”. Current Spanish PWR nuclear power plant 11 ISI

11.

Thermal fatigue has not been considered relevant for BWR plants until now (EPRI throughthe BWRVIP has just released rev.1 of two documents on this subject, VIP-155 on isolatedbranches and VIP-195 on T's of mixing, but continue to consider susceptibility analysis as a“Good Practice”).

56 │ NEA/CSNI/R(2019)13
programmes include consideration of TF management (Inspection for other regulations
than ASME XI, additional to that required by Section XI of ASME) based on:•

Bulletin 88-08 “Thermal Stresses in Piping Connected to Reactor Coolant Systems& Supplement 1, 2, 3”;•

MRP-192 “Assessment pf RHR Mixing Tee TF in PWR Plants”, Rev. 2.

As a consequence of the appearance of cracks in a line of the safety injection system of
Farley 2 (United States) produced by thermal stratification due to the entrance of relatively
cold water from the leakage through the seat of an isolation valve of the system, and since
it could be a generic problem of the plants of similar design, the US NRC collects in
Bulletin 88-08 a series of inspection and testing requirements in addition to those requiredby ASME XI.
As discussed in Bulletin 88-08, the inspection of pipelines susceptible to thermal
stratification involves volumetric examinations using qualified ultrasonic procedures (UT)
and techniques. The current frequency of inspections is based on the results obtained thefirst time these procedures and techniques were applied. For example, all selected welds
are inspected every five years. Depending on the results obtained, the frequency of future
inspections may be modified. For example, due to the NRC Bulletin 88-08, about 16 newareas of the RCS and 2 new areas of the CVCS are inspected. Also included are isolation
valves whose leakage through the seat, either in the direction of the safety injection or in
the direction of the insulation of the containment, may cause thermal stratification.
Currently, seven high-pressure safety injection system (HPSI) valves and two CVCS valves
are included in the programme.
MRP-192, Revision-2 describes the incident that occurred in the French plant Civaux-1;
see Section 4.2.4 for details. Because of the evaluation of this document, the operators of
the Spanish PWR plants have extended the ISI inspection requirements to the following:•

Welds downstream of the “teas” joining the output lines of the RHR heatexchangers and their respective bypass lines.

•

Welds located in the section of pipe downstream of the mentioned “teas” up to a
distance of four times the nominal diameter of the pipeline.

•

Two base metal bands of one inch each, downstream of the connection of theminimum flow recirculation lines of the waste heat removal pumps.

4.2.10.2 Ageing management
The requirements for ageing management in Spanish nuclear power plants are establishedin safety instruction IS-22, Revision-1 of 15 November 2017 of the nuclear safety council
regarding safety requirements for the management of the ageing and long-term operationof nuclear power plants.
The ageing management activities defined in IS-22 are based on the US regulations as
formulated in 10 CFR 54 “Requirements for Renewal of Operating Licenses for NPP” andthe corresponding implementation guidance documents. Those comprise the “Standard
Review Plan for Review of License Renewal Applications for NPP” (NUREG-1800) andthe “Generic Ageing Lessons Learned” (NUREG-1801). Based on IS-22, the Spanish plantowners have to perform:•

AMP.XI.M35 “One-Time Inspection of ASME Code Class 1 Small-Bore Piping”
based on EPRI 1011955 “Materials Reliability Program: Management of TF in

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Normally Stagnant Non-Isolable RCS Branch Lines (MRP-146)”, 8 June 2005, and
EPRI 1018330 “Materials Reliability Program: Management of TF in Normally
Stagnant Non-Isolable RCS Branch Lines – Supplemental Guidance (MRP-146S)”,
31 December 2008.
•

AMP.XI.M6 (BWR) “BWR Control Rod Drive Return Line Nozzle” is a conditionmonitoring programme for BWR control rod drive return line (CRDRL) nozzles
that is based on the staff’s recommended position in NUREG-0619 for TF.

•

AMP.X.M1 “Fatigue-Monitoring” to monitor and track the number of critical
thermal and pressure transients for the selected components. The programme also
verifies that the severity of the monitored transients is bounded by the designtransient definition for which they are classified.

•

TLAA “Metal Fatigue” (LTO) (Chapter X NUREG-1801).

4.2.10.3 Fatigue monitoring
Since the 1990s, the Spanish plants have tried different technical approaches to monitoring
fatigue. Some plants acquired the “FatiguePro” monitoring system that determines
accumulated fatigue and is able to quantify the current remaining margin. 12 The“FatiguePro” applications were terminated in 2006, however. For a limited time the
Spanish industry also supported the Euratom “THERFAT” project to simulate, measureand quantify the turbulent fluid flow and associated thermal loads in mixing tees. 13
Currently, the Spanish nuclear power plants use the “FatONE” FAMOS, which
automatically detects transients that occur in the plant and that can be identified from theprocess signals.
4.2.10.4 TLAA of metal fatigue
Fatigue is a degradation mechanism by which the failure in metallic components can occur
when they are subject to cyclic loads, even with voltage variation levels lower than the limit
stresses corresponding to a static stress. Due to this fact, some design codes, such as the
ASME boiler and pressure vessel code and ANSI pipe codes, require explicit fatigue
calculations or set design limits that consider this phenomenon.
The origin of the stress cycles on the material of the vessel, pipes and components of thepressure barrier is mainly in the variation of the conditions of pressure and temperature that
occurs during operation of the plant. The fatigue processes are developed mainly under two
mechanisms:•

Fatigue from a high number of cycles: produced generally by the mixture of fluids
at different temperatures, which produces turbulence and therefore thermalvariations of small magnitude, but of high frequency (> 103-104). Also identified
is fatigue from a high number of cycles of mechanical origin in pipes and in the12.

A detailed description of the methodology is found in: Stress-Based Fatigue Monitoring:
Methodology for Fatigue Monitoring of Class 1 Nuclear Components in a Reactor Water
Environment, EPRI, Palo Alto, CA: 2011, 1022876.

13.

For details see NEA/CSNI/R(2007)13: “Assessment of Computational Fluid Dynamics (CFD)
for Nuclear Reactor Safety Problems”, www.oecd-nea.org/nsd/docs/2007/csni-r2007-13.pdf.

58 │ NEA/CSNI/R(2019)13
internals of the reactor. The latter is usually caused by vibrations due to the passageof the flow.
•

Fatigue from a low number of cycles: generally produced by the thermal andpressure variations of the reactor’s refrigerant because of the different modes ofoperation or transients that take place in the plant. These variations have a large
magnitude but occur with a low frequency ( 103 -104).

Both NUREG-1800 Rev. 2 and NUREG-1801 Rev. 2 require the analysis ofenvironmentally assisted fatigue (EAF) in ASME Class 1 components during long-term
operation, based on the use of an environmental factor that modifies the CUFs calculatedwith fatigue design curves obtained based on air tests.

4.2.11. Switzerland
All Swiss nuclear power plants utilise instrumentation for fatigue-monitoring. Three of four
nuclear power plants use integrated fatigue-monitoring software systems (“FatiguePro”,“WESTEMSTM” and “FAMOS”, respectively) and in one plant, the fatigue assessment isdone manually. Enforced by the regulatory guideline ENSI-B01 [28] environmental factors
have to be taken into account. An annual report is required. The Swiss nuclear power plants
have experienced only a few TF events. The majority of the reported fatigue events are
attributed to high-cyclic mechanical vibrations.
In some case high fatigue usage factor have been predicted, which has led to preventive
actions. In the Mühleberg Nuclear Power Plant (KKM) a high fatigue usage factor in a Tjoint of the Reactor Core Isolation Cooling (RCIC) system due to thermal mixing was
identified. In 1997, a double thermo-sleeve was installed. No indications of fatigue cracks
were detected by non-destructive testing and the growth of fatigue usage factors was
minimised.
In the Gösgen Nuclear Power Plant (KKG) thermal stratifications in the pressuriser surge
line were identified. This has caused relatively high fatigue usage factors (lower than100%) on parts of the pressuriser surge line and pressuriser nozzles. The system was
modified in 2005. After the modifications, fatigue loadings were reduced significantly andparts of the surge line with high usage factors were replaced.

4.2.12. Chinese Taipei
TF has been identified as a mechanism that can potentially lead to cracking in lines attached
to PWR primary coolant piping. Taipower has six units at three sites. Two of them,
Maanshan-1 and Maanshan-2, are Westinghouse 3-loop PWR units. For these units, the TF
management program is in full accordance with the EPRI materials reliability programme(MRP) guidelines; refer Table 4.4 to Section 4.2.13 for details.
For investigation and analysis, the geometry, temperature difference between hot and cold
water, maximum effective operating time, and leak rate data were collected. After that, theMaanshan Nuclear Power Plant followed MRP-146 to screen potential TF locations and set
up an inspection programme in normally stagnant, unisolable RCS branch lines. In
addition, MRP-192 was followed to screen locations and set up the same programme for
RHR system mixing tees.
The normally stagnant RCS branch lines where geometry and temperature difference met
the MRP screening criteria were put on the inspection list even if the leak rates were smaller

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 59
than the screening criteria. If the leak rate exceeds the screening criteria, a shortening of
the inspecting period should be evaluated.

4.2.12.1 Research and development
At the end of 2009, research began on the Maanshan Nuclear Power Plant’s TF issue. In
April 2011, hot and cold mixed T pipe thermal fatigue monitoring equipment was set up at
the T pipe at the exit of the RHR heat exchanger at Maanshan unit two. Figure 4.22 shows
the installation positions of the 12 strain gauges and 12 thermometers, which were installedat 4 different positions on 3 sections. Figure 4.23 shows the stress-over-time and Figure
4.24 is a flow chart that uses the measured data to analyse the degree of TF (the maximum
stress amplitude value of > 28.2 ksi is per ASME Section III, Appendix I, Figure A.1. The
maximum stress amplitude threshold of > 13.6 ksi is per ASME Section III, Appendix I,
Figure A.2.).
Figure 4.22. The positions of the strain gauges and thermometers

Figure 4.23. Stress over time of the RHR mixing tee60 │ NEA/CSNI/R(2019)13
Figure 4.24. Measured data to analyse the degree of TF

Amplitude(max)>28.2ksi

No

Amplitude(max)>13.6ksi

Yes

No

Yes
Analysis by Rain Flow Method

No fatigue problem

Calculation of cumulative fatigue factor (CUF)
The strain value measured by the strain gauge is converted into a stress value, thereby
determining the maximum stress amplitude. If the maximum stress amplitude is greaterthan 13.6 ksi, the CUF should be calculated by the “rain flow method” (RFM) as describedin ASME Section three, Appendix I, Figures I-9.2.1 and Figures I-9.2.2 (the RFM is usedin the analysis of fatigue data to reduce a spectrum of varying stress into a set of simple
stress reversals. It allows the application of Miner’s rule to assess the fatigue life of astructure subject to complex loading.) If the maximum stress amplitude is less than 13.6
ksi, there is no fatigue problem. Because maximum stress amplitude at the T pipe at theexit of the RHR heat exchanger at Maanshan unit 2 was lower than 13.6 ksi, it wasdetermined that there was no TF issue in that pipe.

4.2.13. United States
4.2.13.1 Industry and regulator response to TF management
The US operating experience with TF is summarised in Figure 4.25. Thermal stratification
was first identified in 1977 as the root cause for ASME III Code Class 2 feed-water nozzle
cracks which occurred at a number of PWR plants. The next industry-wide thermal
stratification concern arose at Farley unit 2 in late 1987, because of an unisolable primarysystem water leak. Subsequently, reports from foreign reactors (e.g. Tihange unit 1 andGenkai unit 1) described similar events. Table 4.3 summarises the NRC regulatory bulletins
and information notices related to thermal stratification. References [11] [29] and [30]provide more detailed technical information on TF experience for the period from 1970 to
2000.
Figure 4.25. The US TF operating experience

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 61
Table 4.3. List of NRC bulletins and information notices related to thermal stratification
Information notice (IN)/bulletin
Bulletin 79-13
IN 84-87
IN 88-01
Bulletin 88-08
IN 88-80
Bulletin 88-11
IN 91-28
IN 91-38
IN 93-20
IN 93-62
IN 97-46

Subject
Cracking in Feed-water System Piping
Piping Thermal Deflection Induced by Stratified Flow
Safety Injection Pipe Failure
Thermal Stresses in Piping Connected to Reactor Coolant Systems - with two
supplements
Unexpected Piping Movement Attributed to Thermal Stratification
Pressuriser Surge Line Thermal Stratification
Cracking in Feed-water System Piping – Closeout of Bulletin 79-13
Thermal Stratification in Feed-water System Piping
TF Cracking of Feed-water Piping to Steam Generators
Thermal Stratification of Water in BWR Pressure Vessels
Unisolable Crack in High-Pressure Injection Piping

Date ofIssuance
10/16/1979
12/3/1984
1/27/1988
6/22/1988
10/7/1988
12/20/1988
4/15/1991
6/13/1991
3/24/1993
8/10/1993
7/9/1997

The US Nuclear Regulatory Commission issued Bulletin 88-08 “Thermal Stresses in Piping
Connected to Reactor Coolant Systems” in response to three TF failures that occurred inunisolable portions of piping systems attached to the reactor coolant loop piping. The
failure mechanism was thermal stratification and cycling caused by valve leakage and swirl
penetration. In response to Bulletin 88-08, PWR plants inspected susceptible locations,
instrumented piping to measure thermal stratification, and made permanent modifications
to preclude TF failures. Since the issuance of the bulletin, incidents of unisolable leakage
have continued to occur, however. Subsequently, the Electric Power Research Institute(EPRI), via its MRP, developed various TF management guidelines (see Table 4.4). For
example, EPRI published MRP-24 as an interim guideline to manage TF issues in nuclear
plants.
After the issuance of MRP-24, additional testing and evaluation was performed to bettercharacterise TF. In 2003, the Nuclear Energy Institute (NEI) issued NEI 03-08 to provide
guidance on TF management by implementing the requirements in MRP-146 for RCS
branch lines and MRP-192 for RHR system mixing tees. These NEI and MRP documents
provide “good practice” recommendations for evaluating and inspecting regions where
there may be potential for TF cracking. MRP-146 also provides a model for predicting andevaluating thermal cycling for PWR stagnant lines, which has been shown throughbenchmarking results to be effective in predicting the location of thermal cycling in lines
attached to the RCS. In 2009, EPRI issued supplemental guidance to MRP-146 providing
revised evaluation guidance for branch piping.
Since 1998, there have been 21 TF events involving US plants [31], 10 of which occurred
between 2013 and 2015. In response to these recent events, EPRI formed the Thermal
Fatigue Focus Group (TFFG) with the objectives of (1) publishing interim guidance toprevent thermal fatigue degradation from exceeding regulatory limits and (2) identifying
actions for programme improvement [31]. The TFFG evaluated the thermal fatigue events
between 2013 and 2015 and issued interim guidance in MRP 2015- 019, which included
two “good practice” and eight “needed” requirements. Additionally, MRP-146, Revision 2
was issued in 2016 and MRP-192, Revision 3 was issued in 2017, both incorporating theinterim guidance issued in MRP 2015-019.
After the release of this guidance, another seven events occurred between 2016 and 2017.
The evaluation of these events revealed that further modifications to the 2015 guidancemight be needed. Some of the flaws occurred in areas that were not previously considered
to be susceptible to TF. A possible new source of in-leakage from safety injection systemcross flow was identified. In response, EPRI is collecting temperature and strain data fromseveral plants for analysis and may provide further guidance, if needed.

62 │ NEA/CSNI/R(2019)13
Table 4.4. Selected EPRI MRP TF R&D and guidelines
Document
ID
MRP-23
MRP-24
MRP-25
MRP-81
MRP-83
MRP-85

MRP-132
MRP-146
MRP-146S
MRP-192
MRP-275
MRP-409

Title

Year Issued“NDE Technology for Detection of TF Damage in Piping”

2017
Revision 2
2001
2000

“Interim TF Guideline”“Operating Experience Regarding TF of Unisolable Piping Connected to PWR Reactor Coolant
Systems”
Interim Report on Thermal Cycling Model Development for Representative Unisolable Piping
Configurations”“Lessons Learned from PWR TF Management Training”“Operating Experience Regarding TF of Piping Connected to PWR Reactor Coolant Systems”.
The second update includes descriptions of cracking and leakage events as well as related
inspection experience collected through February 2017. Also included are results of a survey
conducted of BWR plants to test the prior assumption that TF cracking is less of a concern than in
PWRs. TF operating experience in CANDU plants has been included as well.
“Thermal Cycling Screening and Evaluation Model for Normally Stagnant Non-Isolable Reactor
Coolant Branch Line Piping With a Generic Application Assessment”“Cyclic Stratification in Non-isolable RCS Branch Lines”“Management of TF in Normally Stagnant Non-Isolable RCS Branch Lines -Supplemental
Guidance”“Assessment of RHR Thermal Mixing Tee TF in PWR Plants”“MRP-146/146S Implementation Survey Summary”“EDF Assessment of US TF Management and Operating Experience and Development ofRecommendations for Guideline Improvement”

2002
2002
2018
Revision 2

2004
2016
Revision 2
2008
2018
Revision 3
2010
2016

4.2.13.2 NRC guidance for addressing TF plant operation beyond 40 years
To address TF for piping and components exposed to LWR environments in the UnitedStates, the US Nuclear Regulatory Commission (NRC) issued two documents:•“Effect of LWR Water Environments on the Fatigue Life of Reactor Materials,”
NUREG/CR-6909 Revision 1, 2017,•“Guidelines for Evaluating The Effects of LWR Water Environments in Fatigue
Analyses of Metal Components”, Regulatory Guide 1.207.

The effects of the environment on fatigue life are addressed by reducing the measured life
in air by an environmental fatigue correction factor (Fen). The Fen factor for carbon andlow-alloy steels is a function of sulfur content, material temperature, dissolved oxygen andstrain rate. The Fen factor for wrought and cast stainless steels and Ni-Cr-Fe alloys is afunction of material temperature, dissolved oxygen and strain rate.
To support licence renewal (i.e. plant operation beyond 40 years), in July 2001, the NRC
published NUREG-1800 “Standard Review Plan for Review of License Renewal
Applications for NPP”, and NUREG-1801, “Generic Ageing Lessons Learned (GALL)
report, Volumes 1 and 2”. To support subsequent licence renewal (i.e. plant operation
beyond 60 years), NUREG-2192 “Standard Review Plan for Review of Subsequent License
Renewal Applications for NPP”, and NUREG-2191 “Generic Ageing Lessons Learned for
Subsequent License Renewal (GALL-SLR), Volumes 1 and 2”, were published in July
2017. These reports discuss TF issues in nuclear power plants and provide guidance
regarding how plant owners should manage TF in piping during the period of extended
operation. NUREG-1800 and NUREG 2191 reference industry guidance on managing TF.
For example, AMP XI.M35 provides guidance for performing one-time inspections on
ASME Class 1 small-bore piping with diameters from NPS-1 to less than NPS-4, and bothreports reference MRP-146 and MRP-146S for guidance on determining the locations that

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 63
are most susceptible to TF. Both documents also reference NUREG/CR-6909 and RG
1.207 for addressing environmental effects in fatigue applications.

4.2.13.3 ASME code activities
The ASME has been working on several code cases related to environmental fatigue thatare used to evaluate TF, as summarised below:•

ASME Code Case N-779 (Ke’-): “alternative rules for simplified elastic-plastic
analysis Class 1, Section III Division 1.” This code case provides a method to
reduce some of the conservatism in calculating the Ke factor in the current
simplified elastic-plastic discontinuity analysis in ASME boiler and pressure vessel
code, Section III sub-sub-paragraph NB-3653.6. This code case utilises two factors:
Kv to account for Poisson’s ratio effects, and Kn to account for plastic strain
redistribution effects at local discontinuities. The evaluation utilises stress terms
that are not directly available from Class 1 stress summary reports.

•

Proposed ASME Code Case N-XXX (Ke*): “Alternative Rules for Simplified
Elastic-Plastic Analysis Section III, Division 1.” This proposed code case is slightly
different from CC N-779 but it also provides a method to reduce some of theconservatism in calculating the Ke factor using ASME boiler and pressure vessel
code, Section III sub-sub-paragraph NB-3653.6. This code case utilises stress terms
that are directly available from Class 1 stress summary reports. The results fromthis proposed code case may be slightly more conservative compared to CC N-779.

•

Proposed ASME Code Case N-884: “Procedure to Determine Strain Rate for Use
with the Environmental Design Fatigue Curve Method and the Environmental
Fatigue Correction Factor (Fen) Method as part of an Environmental Fatigue
Evaluation for Components Analysed per the NB-3200 Rules. Section III, Division
1.” The Fen factor is a function of temperature, dissolved oxygen and strain rate.
This code case provides a methodology to calculate the strain rate for eachapplicable transient when the design by analysis approach in NB-3200 is used forClass 1 components.

•

Proposed ASME Code Case N-X-0: “Thickness and Gradient Factors for Section
III Piping Fatigue Analyses” is under development. This code case addresses two
topics: (1) it accounts for the thickness of real components compared to thinlaboratory specimens upon which the ASME fatigue S-N data is based and (2) itaddresses real stress distributions or gradients through the thickness of realcomponents in addition to the membrane stress component.

•

Proposed ASME Code Case N-XYY: “Procedure to Determine Strain Rate for Use
with the Environmental Design Fatigue Curve Method and the Environmental
Fatigue Correction Factor (Fen) Method as part of an Environmental Fatigue
Evaluation for Piping and Valves Analysed per the NB-3600 and NB-3500 Rules,
respectively. Section III, Division 1.” The Fen factor is a function of temperature,
dissolved oxygen and strain rate. This code case provides a methodology forcalculating the strain rate for each applicable transient.

64 │ NEA/CSNI/R(2019)13

5. Conclusions and recommendations
This sixth CODAP topical report focuses on the operating experience with TF in piping
systems of commercial nuclear power plants. Through an examination of the operating
experience as recorded in the CODAP, the field experience with the post-1998 TF
mechanisms is evaluated in order to draw qualitative and quantitative insights about thedamage and degradation mechanisms and their potential plant operability and safety
impacts.
In the “Specialist Meeting on Experience with TF in LWR Piping by Mixing andStratification” [NEA/CSNI/R(98)8] [1] it was noted that some deterioration mechanisms
related to TF were not considered in the original design of many nuclear power plants.
Therefore, this meeting concluded that effective co-operation between the plant designer,
plant owner and plant operator is necessary to keep the risk of TF under control and toensure safe operation.
In 2005, the CSNI Working Group on Integrity and Ageing of Components and Structures(WGIAGE) finalised the work on a review of operating experience, regulatory framework,countermeasures and current research [5]. The group concluded that thermal cycling
degradation has the potential to be an important safety and economic issue in all reactor
types. It strongly affects the AMP of safety components. The incidents analysed occurredin different systems and countries/economies but fall within three distinct loading modes
connected to thermal cycling: a) stratification; b) dead legs and vortex; c) mixing tees.
Assessment screening criteria and guidelines were developed to identify potential locations
and ISI programmes were modified in many countries/economies to integrate online
monitoring and valve leak tests. In addition, some countries/economies adopted designmodifications in different systems to avoid TF effects. The corrective actions against TF
have been discussed in many international workshops related to the topic.

5.1. Conclusions
It can be seen in Figure 1 that HCF failures form a major part of fatigue events recorded in
CODAP. It should be noted that, based on the number of TF failures, especially HCF
failures, the number of reported TF events by CODAP participants decreased after thenumerous international discussions and led to various projects in this area in the end 1990s.
Based on the corrective actions by CODAP participants as presented in Chapter 4.2 and theresults of the above-mentioned international work related to TF [1-5], the following actions
can be considered effective in preventing TF failures:•

Screening of locations with a higher risk of stratification based on operating
experiences and research results, especially locations with a risk of HCF. For
example, Canada, the Czech Republic, France, Finland, Germany, Japan and theUnited States indicated that identifying most critical locations is an importantcomponent of their corrective actions to mitigate TF.

•

Design modifications to decrease TF risk could be implemented based on screening.
Design changes depend on reactor design and could include, for example, flow

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 65
circulation changes to avoid stratification, material changes to avoid quick
deterioration, and process and procedure changes to avoid high temperature
differences. Canada, France, Finland, the Netherlands and Spain mentioned that
adequate preparations and procedures to mitigate TF had not been included in theoriginal design of the plants. Most participants indicated that optimising operatingprocedures based on operating experiences was an effective method to avoid further
fatigue events.
•

Conducting online monitoring of critical locations during operation based on
detection and screening of critical locations. The monitoring results should becontinuously analysed together with results of periodic inspections. Canada, theCzech Republic, Germany, the Netherlands, the Slovak Republic, Spain andChinese Taipei mentioned the importance of effective online monitoring in theircorrective actions of TF events. Finland, France, Spain and Switzerland provided
specific monitoring considerations to assess the stratification problems and develop
effective corrective actions.

•

Developing more precise NDE methods, techniques and procedures for periodic
inspections. As part of the event descriptions in Chapter 4.2, Canada, France,
Germany, Korea and the United States mentioned improvements in TF crack
detection due to the higher effectiveness of inspection programmes. Korea
specifically described their work on improving NDE techniques.

5.2. Recommendations
The actions of the NEA CODAP participants represent good practices to prevent TF in
nuclear power plants. In addition, the following international actions are recommended:•

Improving the precision of NDE methods, techniques and procedures for periodical
inspections. One possibility could be an international benchmark to establish thereliability of detecting existing cracks. The benchmark could include a comparisonof the detection techniques, the calibration procedures of the equipment for plant
environment, and the ultrasonic examination procedures. This kind of international
benchmark is also recommended in reference [6].

•An international workshop on fatigue could be held and include discussion of how
to ensure that design procedures and monitoring systems adequately take into
account TF and stratification. The workshop could be scheduled during the next
five years to look at the possible need for international actions in the area of TF.

66 │ NEA/CSNI/R(2019)13

References[1]

NEA (1998), “Specialists Meeting on Experience with Thermal Fatigue in LWR Piping Causedby Mixing and Stratification”, NEA/CSNI/R(98)8, OECD Publishing, Paris.

[2]

NEA (2000), “Proceedings of the International Conference on Fatigue of Reactor Components”,
NEA/CSNI/R(2000)24, OECD Publishing, Paris.

[3]

NEA (2002), “Proceedings of the EPRI/USNRC/OECD International Conference on Fatigue ofReactor Components”, NEA/CSNI/R(2003)2, OECD Publishing, Paris.

[4]

NEA (2004), “Proceedings of the Third International Conference on Fatigue of Reactor
Components”, NEA/CSNI/R(2004)21, OECD Publishing, Paris.

[5]

NEA (2005), “Working Group on Integrity and Ageing of Components and Structures (WGIAGE)
Report on Thermal Cycling in LWR Components in OECD/NEA Member Countries”,
NEA/CSNI/R(2005)8, OECD Publishing, Paris.

[6]

NEA (2015), “Proceedings of the Fourth International Conference on Fatigue of Nuclear Reactor
Components”, NEA/CSNI/R(2017)2, OECD Publishing, Paris.

[7]

Dahlberg, M, I. Bretherton, C. Faidy, M. Church, S. Chapuliot, U. Wilke, N. Taylor, J. Solin, K.F.
Nilsson, J. Catalano (2007), Development of a European Procedure for Assessment of High Cycle
Thermal Fatigue in Light Water Reactors: Final Report on the NESC-Thermal Fatigue Project,
EUR 22763 EN, European Commission Joint Research Centre, Institute for Energy, Petten, The
Netherlands.

[8]Wåle, J. and P. Ekström (1995), Crack Characterisation for In-service Inspection Planning - An
Update, SAQ/FoU-Rapport 95/07, SAQ Inspection Ltd., Stockholm, Sweden.
https://inis.iaea.org/collection/NCLCollectionStore/_Public/27/028/27028075.pdf.

[9]

EPRI (2010), MRP-146/MRG-146S Implementation Survey Summary Report (MRP-275),
1020975, Electric Power Research Institute, Palo Alto, CA.

[10]

EPRI (2010), Materials Reliability Program: Proceedings of MRP-146 Supplemental Guidanceand Implementation Insights Training (MRP-299), 1022192, Electric Power Research Institute,
Palo Alto, CA.

[11]

Shah, V.N., M.B. Sattison, C.L. Atwood, A.G. Ware, G.M. Grant and R.S. Hartley (1998),
Assessment of Pressurised Water Reactor Primary System Leaks (NUREG/CR-6582), US Nuclear
Regulatory Commission, Washington, DC, United States.

[12]

Tregoning, R., L. Abramson and P. Scott (2008), Estimating Loss-of-Coolant-Accident (LOCA)
Frequencies Through the Elicitation Process (NUREG-1829), US Nuclear Regulatory
Commission, Washington, DC.

[13]

Fleming, K.N. and B.O.Y. Lydell (2016), “Insights Into Location Dependent Loss-of-coolantAccident (LOCA) Frequency Assessment for GSI-191 Risk-informed Applications”, Nuclear
EngineeringandDesign,Volume305,
pp.
433-450.
https://inis.iaea.org/search/search.aspx?orig_q=RN:48060454

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 67
[14]

Lydell, B. (2017), “A Review of the Progress with Statistical Models of Passive Component
Reliability,” Nuclear Engineering and Technology, Volume 49, pp. 349-359,
https://doi.org/10.1016/j.net.2016.12.008.

[15]

Berg, H.-P., R. Gersinska and J. Sievers (2009), “Improved Approach for Estimating Leak andBreak Frequencies of Piping Systems in Probabilistic Safety Assessment”, pp. 69-81.

[16]

Yetisir, M. (2008), “Thermal Fatigue in CANDU Stations”, Document No. COG-07-4054,
CANDU Owners Group Inc., Toronto, Canada.

[17]

EPRI (1999), Revised Risk-informed Inservice Inspection Procedure, EPRI TR-112657, Revision
B-A, Electric Power Research Institute, Palo Alto, CA.

[18]

Lafreniere, P., P. O’Regan, M. Rezaie-Manesh and S. Chesworth (2013), CANDU Owners Group
Risk-Informed In-Service Inspection (RI-ISI) Project, PVP 2013-97935, Proc. ASME 2013
Pressure Vessels and Piping Conference, Paris, American Society of Mechanical Engineering,
New York, United States.

[19]

Yetisir, M. (2009), “Thermal Fatigue in CANDU Stations” ICONE17-75603, Proc. 17th
International Conference on Nuclear Engineering (ICONE-17), American Society of MechanicalEngineering, New York, United States.

[20]

Yetisir, M., G. Stevens and S. Robertson (2008), “Management of Component Fatigue in CANDU
Stations for Life Extension”, Paper ICONE16-48757 Presented at the 16th International
Conference on Nuclear Engineering (ICONE-16), Orlando, Florida, American Society of
Mechanical Engineering, New York, United States.

[21]

Scovil, A. and D. Gerber (2011), “PLGS Fatigue Management Program Plan Development”,
CNSC CANDU Fatigue Management Programme Development Workshop, 16-17 February 2011,
National Arts Center, Ottawa, Canada, Canadian Nuclear Safety Commission.

[22]

Stevens, G., M. Yetisir, A. Scovil, J. Slade (2008), “Monitoring and Managing Component Fatigue
at Point Lepreau Generating Station”, Proceedings of the 8th International Conference on CANDUMaintenance, Canadian Nuclear Society, Toronto, Canada.

[23]

Chapuliot, S., C. Gourdin, T. Payen, J.P. Magnaud and A. Monavon (2005), “Hydro-ThermalMechanical Analysis of Thermal Fatigue in a Mixing Tee”, Nuclear Engineering and Design,
Volume 235, pp. 575-596, DOI:10.1016/j.nucengdes.2004.09.011[24]

Hakala, J., H. Hänninen and P. Aaltonen (1990), “Stress Corrosion and Thermal Fatigue –Experiences and Countermeasures in Austenitic SS Piping of Finnish BWR Plants”, Nuclear
Engineering and Design, Volume 119, pp. 389-396. https://doi.org/10.1016/0029-5493(90)901792[25]

Safety Standards of the Nuclear Safety Standards Commission (KTA) (2017), “Components of theReactor Coolant Pressure Boundary of Light-Water Reactors: Part 2: Design and Analysis”.
Revision 2017-11, KTA 3201.2, KTA-Geschaeftsstelle c/o Bundesamt fuer kerntechnische
Entsorgungssicherheit (BfE), Germany.

[26]

Safety Standards of the Nuclear Safety Standards Commission (KTA) (2016), “Components of theReactor Coolant Pressure Boundary of Light-Water Reactors, Part 4: In-service Inspections andOperational Monitoring”, Revision 2016-11, KTA 3201.4, KTA-Geschaeftsstelle c/o Bundesamt
fuer kerntechnische Entsorgungssicherheit (BfE), Germany.

68 │ NEA/CSNI/R(2019)13[27]

Elmas, M., U. Jendrich, F. Michel, H. Reck, T. Schimpfke, M. Walter and R. Wenke (2013),
Studies on the Effectiveness of Measures to Maintain the Integrity of Pressurized Components inGerman Nuclear Power Plants: Final Report, GRS-A-3700 (in German), Gesellschaft für Anlagen
und Reaktorsicherheit (GRS) mbH, Cologne, Germany.

[28]

Swiss Federal Nuclear Safety Inspectorate (2011), Ageing Managament Guidelines for Swiss
NuclearPower
Plants,ENSI-B01/d(inGerman),Brugg,
Switzerland.
www.ensi.ch/de/dokumente/richtlinie-ensi-b01-deutsch/[29]

Su, N.T. (1990), Review of Thermal Stratification Operating Experience, AEOD/S902, Office forAnalysis and Evaluation of Operational Data, US Nuclear Regulatory Commission, Washington,
DC, United States.

[30]

Shah, V.N., A.G. Ware and A.M. Porter (1997), Review of Industry Efforts to Manage Pressurized
Water Reactor Feedwater Nozzle, Piping, and Feedring Cracking and Wall Thinning,
NUREG/CR-6456, US Nuclear Regulatory Commission, Washington, DC, United States.

[31]

Crooker, P. (2017), “MRP Thermal Fatigue Program Update,” Materials Research andDevelopment Information Exchange Meeting presentation, 23 May 2017 (NRC ADAMS
accession number ML17142A005).

[32]

US Nuclear Regulatory Commission (2016), “Risk-Informed approach for addressing the effects
of debris on post-accident long-term core cooling” Regulatory Guide 1.229.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 69

Annex A OPDE/CODAP Bibliography
OPDE-1.

Lydell, B., E. Mathet, and K. Gott, (2002) “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, United States.

OPDE-2.

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

OPDE-3.

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

OPDE-4.

Lydell, B., E. Mathet, and K. Gott, (2004) “OECD Pipe Failure Data Exchange Project(OPDE) – 2003 Progress Report,” ICONE12-49217, Proc. ICONE-12: 12th International
Conference on Nuclear Engineering, Arlington, American Society of MechanicalEngineers, New York, United States.

OPDE-5.

Colligan, A., R. Lojk, and J. Riznic, (2004) “The OECD Pipe Failure Data Exchange
Project – Data Validation on Canadian Plants,” Proc. International Conference on Nuclear
Engineering, pp. 65-70, ICONE12-49273, American Society of Mechanical Engineers,
New York, NY.

OPDE-6.

Sun, Y.C., H.C. Young, and K.C. Yeon, (2004) “Application of a Piping Failure Database
to Some Nuclear Safety Issues in Korea,” Proc. International Conference on Nuclear
Engineering, pp. 175-179, ICONE12-49572, American Society of Mechanical Engineers,
New York, NY.

OPDE-7

Fleming, K. N. and B. Lydell, (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, American Society of
Mechanical Engineers, New York, NY.

OPDE-8.

Lydell, B., E. Mathet, and K. Gott, (2004) “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., H.J. Kim, S.Y. Choi, and Y.J. Kim, (2004) “Safety Evaluation of Socket Weld
Integrity in Nuclear Piping,” Key Engineering Materials, 270-273:1725-1730,
OECD/NEA, The OECD/NEA Piping Failure Data Exchange Project, Workshop on
Database Applications, OPDE/SEC(2004)4, Issy-les-Moulineaux, France.

OPDE-10.

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

70 │ NEA/CSNI/R(2019)13
OPDE-11.

NEA (2006a.) OPDE 2006: 2 Coding Guideline (OPDE-CG), OPDE PR01, OECD NEA,
Boulogne-Billancourt, France, France.

OPDE-12.

NEA (2006b.) OPDE Quality assurance program (OPDE-QA), OPDE-PR-02, OECD
NEA, Boulogne-Billancourt, France, France.

OPDE-13.

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

OPDE-14.

Lydell, B., E. Mathet, and K. Gott, (2006) “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-15.

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

OPDE-16.

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

OPDE-17.

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

OPDE-18.

Choi, S.Y., and J-E. Yang, (2007) “Flooding PSA by Considering the Operating
Experience Data of Korean PWRs,” Nuclear Engineering and Technology, 39:215-220.

OPDE-19.

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

OPDE-20.

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

OPDE-21.

Simonen, F.A., S.R. Gosselin, B. Lydell, and D.L. Rudland, (2007) “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, United States.

OPDE-22.

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

OPDE-23.

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

OPDE-24.

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

OPDE-25.

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

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 71
OPDE-26.

Pandey, M. (2008) Piping Failure Frequency Analysis Using OECD/NEA Data,
University of Waterloo, Final Report, CNSC Research and Support Program. Report RSP0236, February.

OPDE-27.

Olsson, A. and B. Lydell, (2008) “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.

OPDE-28.

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

OPDE-29.

Yuan, X.X., M.D. Pandey, and J.R. Riznic, (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-30.

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

OPDE-31.

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

OPDE-32.

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

OPDE-33.

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

OPDE-34.

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

OPDE-35.

Nordic PSA Group (2010) R-Book: Reliability Data Handbook for Piping Components in
Nordic Nuclear Power Plants, Stockholm, Sweden.14

OPDE-36.

Olsson, A., V. Hedtjärn Swaling, and B. Lydell, (2010) “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.

OPDE-37.

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

OPDE-38.

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

OPDE-39.

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

14.

Electronic version on CD (includes methodology description, data input/output files and resultsummaries). For information on the terms and conditions for how to access the restricted RBook, contact the NPSAG Secretariat; www.npsag.org.

72 │ NEA/CSNI/R(2019)13
OPDE-40.

Lydell, B., A. Huerta, and K. Gott, (2011) “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.

OPDE-41.

Olsson A., V. H. Swaling, and B. Lydell, (2011) “Experiences from Implementation ofUpdated 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-42.

Lydell B., A. Mosleh, and D. Chrun, (2011) “Enhanced Piping Reliability Models for Use
in Internal 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-43.

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

OPDE-44.

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

OPDE-45.

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

OPDE-46.

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

CODAP-1.

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

CODAP-2.

Lydell, B., A. Huerta, K. Gott, and J. Riznic, (2013) “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.

CODAP-3.

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

CODAP-4.

Lydell, B. (2013) “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 (2014) 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, Issy-les-Moulineaux, France.

CODAP-6.

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

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 73
CODAP-7.

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

CODAP-8.

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

CODAP-9.

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

CODAP-10.

Lydell, B., and C. Castelao, (2015) “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.

CODAP-11.

Dragea, T., and J. Riznic, (2015) “The Component Operational Experience Degradation
and Ageing Program (CODAP): Review and Lessons Learned (2011-2014)”, 23th
International Conference on Nuclear Engineering, ICONE23-1001, Chiba, Japan, 1721 May 2015.

CODAP-12.

Rivet, M., and J. Riznic, (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 (2015) 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.

CODAP-14.

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

CODAP-15.

Rivet, M., K. Cormier, and J. Riznic, (2016) “The Component Experience, Degradation
and Ageing Programme (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., O. Nevander, and J. Riznic, (2016) “Piping Corrosion Risk Management on
the Basis of the OECD/NEA CODAP Project Database,” RISK16-8327, Proc. Corrosion
Risk Management Conference, Houston, TX, May 23-25, 2016, NACE International,
Houston, TX, 2016.

CODAP-17.

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

74 │ NEA/CSNI/R(2019)13
CODAP-18.

Lydell, B. (2016) “A Review of the Progress with Statistical Models of Passive
Component Reliability,” Paper 063, Proc. 13th International Conference on Probabilistic
Safety Assessment and Management (PSAM13), Seoul, Korea. 15

CODAP-19.

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

CODAP-20.

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

15.

Also published in Nuclear Engineering and Technology (NET), an international journal of the
Korean Nuclear Society (KNS): www.sciencedirect.com/science/article/pii/
S1738573316303254

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 75

Annex B Glossary of technical terms
Backfill. The material used to refill the trench after the pipe and the embedment have beenplaced. 16
Backing ring (BR). Used to align two pipe sections before welding and eliminate the needfor tack welding. BR pins or nubs automatically set the weld gap.
Bayesian reliability analysis. In the Bayesian approach a subject matter expert develops awell-informed estimate of the probability of failure distribution; the prior state ofknowledge. This probability distribution is then updated as more information is collected
about the structural integrity of a certain piping system component.
Below-grade piping. Below ground piping with its location given relative to a reference
point; e.g. ground level per a plant structure elevation.
Below ground piping. Buried piping in contact with soil or concrete and underground
piping that is below grade but is contained within a tunnel or vault such that it is in contact
with air and is located where access for inspection is restricted.
BONNA® pipe. A thin steel pipe embedded in reinforced concrete. It has rebar or a heavy
wire mesh embedded in the reinforced 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) to make the piping behave in a cathodic manner. Using CP, thecorrosion rate is normally reduced to an acceptable level.
CFRP system. A buried piping rehabilitation and repair technique. It is comprised of highstrength carbon fibre fabrics and/or glass fibre fabrics, fully saturated in a 2-part 100%solids epoxy matrix. These laminates are bonded both longitudinally and circumferentiallyto the interior surface of the pipe, forming a structural lining within the pipe. This lining
can be designed to replace the degraded portions of the existing system without reliance on
the degraded piping for the life of the repair, except at the terminal ends of the repair.
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.
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
16.

For additional details on buried piping backfill refer to: “Pipe Bedding and Backfill,”Geotechnical Training Manual No. 7, US Department of the Interior Bureau of Reclamation,
Earth Sciences and Research Laboratory, Denver, CO.

76 │ NEA/CSNI/R(2019)13
between metals and non-metals, 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.
Cured-in-place-pipe (CIPP). A BP temporary repair method. A resin-saturated felt tube
made of polyester, fibreglass 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).
Dealloying (selective leaching). As defined by the Association for Materials Protection and
Performance (AMPP) (formerly NACE International), “dealloying” or “selective leaching”
refers to the selective removal of one element from an alloy by corrosion processes. A
common example is the dezincification of unstabilised brass, whereby a weakened, porous
copper structure is produced. The selective removal of zinc can proceed in a uniform
manner or on a localised (plug-type) scale. It is difficult to rationalise dezincification interms of preferential Zn dissolution out of the brass lattice structure. Rather, it is believedthat brass dissolves with Zn remaining in solution and Cu replating out of the solution.
Graphitic corrosion of gray cast iron, whereby a brittle graphite skeleton remains following
preferential iron dissolution, is a further example of selective leaching. During cast irongraphitic corrosion, the porous graphite network that makes up 4-5% of the total mass of
the alloy is impregnated with insoluble corrosion products. As a result, the cast iron retains
its appearance and shape but is weaker structurally. Testing and identification of graphitic
corrosion is accomplished by scraping through the surface with a knife to reveal thecrumbling of the iron beneath.
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.
Epistemic uncertainty. It is scientific uncertainty in the piping reliability model. It is due tolimited data (or completeness of the database) and knowledge. The epistemic uncertainty
is characterised by alternative models. For discrete random variables, the epistemic
uncertainty is modelled by alternative probability distributions.
Equivalent break size (EBS). The calculated size of a hole in a pipe given a certain throughwall flow rate and for a given pressure.
Frazil ice. A collection of loose, randomly oriented needle-shaped ice crystals in water thatis too turbulent to freeze solid. It resembles slush and has the appearance of being slightly
oily when seen on the surface of water.
Fusion (or heat fusion). In the context of high-density polyethylene (HDPE) piping design
and installation, fusion techniques are used to join pipes together. It is a welding processused to join two different pieces of a thermoplastic. This process involves heating bothpieces simultaneously and pressing them together. The two pieces then cool together andform a permanent bond.
Holiday in pipe coating. A holiday is a hole or void in the coating film, which exposes theburied piping to corrosion.

OECD/NEA CODAP PROJECT TOPICAL REPORT:NEA/CSNI/R(2019)13 │ 77
JRC Operating Experience Clearinghouse (CE-OEF). Located in Petten, Netherlands, theClearinghouse gathers nuclear safety experts performing the following technical tasks insupport to the EU Member States:•“Topical Studies” providing in-depth assessment of particularly significant eventsor families of events. These studies are drafted by experts on the topic and basedon an analysis of usually hundreds of event reports.

•

Trend analysis of events 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 NEA and the International
Atomic Energy Agency (IAEA).

•

Reporting every three months the main events that occurred in nuclear power
plants.

•Developing a European central Operational Experience OE repository to ensure
long-term storage of OE and to facilitate information retrieval.

•Participating in several international co-operation projects on OE, mainly throughthe IAEA and the NEA working groups.

Limiting Conditions for Operation (LCO). According to the technical specifications 17, an
LCO is the lowest functional capability or performance level of a piece of equipment
required for safe operation of a nuclear plant. When an LCO cannot be met, the reactor
must be shut down or the licensee must follow any remedial action permitted by thetechnical specifications until the condition can be met.
Pipeline inspection gauge (PIG). In-line PIGs, or smart PIGs, gather information about thepipeline from within. The type of information gathered by smart PIGs includes the pipeline
diameter, curvature, bends, temperature and pressure, as well as corrosion or metal loss.
PIGs utilise two methods to gather information about the interior condition of the pipeline:
magnetic flux leakage (MFL) and ultrasonics (UT). MFL inspects the pipeline by sending
magnetic flux into the walls of the pipe, detecting leakage, corrosion or flaws in thepipeline. Ultrasonic inspection directly measures the thickness of the pipe wall by using
ultrasonic sounds to measure the amount of time it takes an echo to return to the sensor.
Richardson number (Ri). A dimensionless parameter used in the identification andassessment of locations potentially susceptible to stratification.
Selective leaching. Also referred to as dealloying, demetalification, parting and selective
corrosion, selective leaching is a corrosion type in some solid solution alloys, when insuitable conditions a component of the alloys is preferentially leached from the material.
The less noble metal is removed from the alloy by a microscopic-scale galvanic corrosion
mechanism. The most susceptible alloys are the ones containing metals with high distance
between each other in the galvanic series, e.g. copper and zinc in brass.
Techite pipe. Fibreglass (or Fibre Reinforced Polymer) reinforced mortar pipe. This typeof 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.

17.

“Betriebshandbuch” in German.

78 │ NEA/CSNI/R(2019)13
Thermal striping. Incomplete mixing of high temperature and low temperature fluids near
the surface of structures with subsequent fluid temperature fluctuations giving rise to TF
damage to wall structures.
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. Tritium radiation does not travel
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 is either 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.
Type A sleeve (reinforcing). Used for the repair of non-leaking defects (e.g. pitting, wall
thinning) of below ground pipelines. Such a sleeve provides reinforcement for the defectivearea.
Type B sleeve (pressure containing). A type of steel sleeve used to make pipeline leakrepairs. The ends of a Type B sleeve are fillet welded to the carrier pipe.
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.
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,
VT-2 and VT-3.
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, VT-2 is a visual surface examination to locate evidenceof leakage from pressure-retaining components.
VT-3 examination. A limited visual examination specific to ASME Section XI which isintended to determine the general mechanical and structural condition of components andtheir supports, such as the verification of clearances, settings, physical displacements, loose
or missing parts, debris, corrosion, wear, erosion, or the loss of integrity at bolted or welded
connections. The VT-3 examinations shall include examinations for conditions that could
affect operability of functional adequacy of snubbers, and constant load and spring typesupports. The VT-3 examination is intended to identify individual components withsignificant levels of degradation. As the VT-3 examination is not intended to detect the
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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.
WEKO-SEAL®. A flexible rubber leak clamp that ensures a non-corrodible, 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.
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.
Yoloy pipe. A high-strength low-alloy steel with enhanced corrosion resistance (ASTM A714).